NRC Forms Special San Onofre Review Panel

Victor Dricks
Senior Public Affairs Officer
Region IV

NRC Chairman Allison Macfarlane (second from right) listens as Southern California Edison executive Richard St. Onge (third from right) discusses issues with one of the damaged steam generators at SONGS. The steam generator is in the right foreground.
NRC Chairman Allison Macfarlane (second from right) listens as Southern California Edison executive Richard St. Onge (third from right) discusses issues with one of the damaged steam generators at SONGS. The steam generator is in the right foreground.

The NRC has established a special panel to coordinate the agency’s evaluation of Southern California Edison Co.’s proposed plan for restarting its Unit 2 reactor and ensuring that the root causes of problems with the plant’s steam generators are identified and addressed.

Art Howell, the NRC’s Region IV deputy regional administrator, will serve as co-chairman of the panel along with Dan Dorman, deputy director for engineering and corporate support in the Office of Nuclear Reactor Regulation (NRR). Jim Andersen, chief of NRR’s Electrical Engineering Branch, will serve as deputy team manager of the San Onofre Nuclear Generating Station (SONGS) Oversight Panel.

The panel will ensure that NRC communicates a unified and consistent position in a clear and predictable manner to the licensee, public and other stakeholders, and establishes a record of major regulatory and licensee actions taken and technical issues reviewed, including adequacy of Southern California Edison’s corrective actions.

The panel also will be responsible for conducting periodic public meetings with the utility and providing a recommendation to senior NRC management regarding restart of SONGS Unit 2. In comments to reporters Monday following a tour of the plant, Chairman Allison Macfarlane said Unit 2 will not be permitted to restart unless the NRC has reasonable assurance it can be operated safely.

Other panel members include: 

  • Ed Roach, chief, Mechanical Vendor Inspection Branch, NRO
  • Ryan Lantz, chief, SONGS Project Branch, Region IV
  • Greg Werner, inspection & assessment lead, SONGS Project Branch, Region IV
  • Nick Taylor, senior project engineer, SONGS Project Branch, Region IV
  • Greg Warnick, senior resident inspector, San Onofre Nuclear Generating Station
  • Doug Broaddus, chief, SONGS Special Project Branch, NRR
  • Randy Hall, project manager, SONGS Special Project Branch, NRR
  • Ken Karwoski, senior level advisor, Division of Engineering, NRR
  • Michele Evans, director, Division of Operating Reactor Licensing (alternate is Pat Hiland, director, Division of Engineering)

Author: Moderator

Public Affairs Officer for the U.S. Nuclear Regulatory Commission

136 thoughts on “NRC Forms Special San Onofre Review Panel”

  1. Invaluable analysis , I loved the details – Does anyone know where my assistant could possibly grab a template CA Edison Company Claim Form form to fill in ?

  2. A frightful aspect of this is the extent (generic implications) of the gross incompetence of SCE, MHI, NRC, and INPO.

    NEI fought hard to get enough rope in 10CFR50.59 and SCE, MHI, NRC, and INPO used it to hang themselves.

    The first law of highway engineering is,”Never remove a guardrail that has dents in it.” 50.59, properly applied, had been involved in keeping many plants forming over the cliff. Yet NEI and NRC weakened the rigor. what sense does this make?

  3. Edwin Hackett, Executive Director ACRS
    Thanks for your reply, and staying aware of what is happening at SONGS aka SanO.
    Royale Brodeur

  4. Role and Responsibilities of Agencies in Nuclear Power Generation, Decommissioning and Lessons Learnt From San Onofre – NRC Continuing Education Series

    America is a Democratic Country and NRC Commission Solemn Duty is Public Safety/Licensing, Protection of Workers from Retaliation/Discrimination and Safe Decommissioning of power plants and not vacating ASLB SONGS Ruling under NEI/Industry Pressure for Profits/Production from Unsafe Nuclear Energy. Nuclear Energy Institute Job is Interpretation of NRC Rules For Safe and Reliable Production of Nuclear Energy. Institute of Nuclear Power Operations Job is measuring Operational Excellence of Nuclear Plants.

    SONGS Original Combustion Steam Generators (OSGs) lasted for 28 years at a void fraction of 96%, fluid velocities of 22 feet/sec and 900 psi and did not suffer fluid elastic instability. In 2001, SONGS up-rated the power (steam Flows) of these OSGS from 1705 MWt to 1729 MWt for profits, reduced steam pressures, which increased the flow-induced vibrations. The increased steam flows and reduced steam pressures increased flow-induced vibrations, increased fluid velocities > 22 feet/sec, and increased tube wear and plugging rates in OSGs. Low steam pressures and increased steam flows produce high void fractions for more power, but are detrimental for tube vibrations, tube wear and structural integrity. The high steam flows and high fluid velocities decreased the life of OSGs, which could have lasted for a few years. When increasing the power, SCE Engineers should have foreseen the adverse affects of the power up-rate . But looks like they were focused on profits and getting new replacement steam generators, therefore, they did not care and were not able to foresee the safety consequences of these changes on steam generator tube leaks. No leaks happened, so everything turned out to be ok. Not really. SONGS got new generators in 2010 & 2011 at the cost of $670 Million and Edison Management claimed to be the 21st Century Safest and Most Innovative Steam machine.

    According to SONGS Insider Documents (Useless Now, Since SONGS is being Decommissioned), Edison Specified, “Edison intends to replace the steam generators under the 10 CFR 50.59 rule. Consequently, Edison requests that the RSGs be as close as possible to the existing steam generators in form, fit, and function, subject to additional requirements and limitations stated elsewhere in this [Redacted]. The Supplier shall prepare and submit for Edison’s approval a Licensing Topical Report demonstrating compliance of the RSG design with all SONGS licensing requirements. The report shall include an engineering evaluation, including all necessary analyses and evaluations, justifying that the RSGs can be replaced under the provisions of 10 CFR 50.59 (without prior NRC approval). The report format shall follow the guidelines of in order to facilitate preparation of the 10 CFR 50.59 evaluation. The 10 CFR 50.59 evaluation shall be performed by Edison. Steam Generator Thermal Rating @100% Reactor Power – OSGs -1705 MWt – RSGs – 1729 MWt, Heat Transfer Area, sq. ft (0% tubes plugged) – OSGs 105,000 – RSGs – 116,100, Steam Pressure@100% Power – OSGs -900 psi – RSGs – 833 psi, Circulation ratio – OSGs – 3.2 -RSGs 3.3, RCS Design Flow, gpm (0%Tubes Plugged) OSGs -198, 000 – RSGs – 209,880, Feedwater Flows, gpm, OSGs – 7,414,000 – RSGs – 7,588,000. The service life of the RSGs shall be 40 calendar years from the date of startup following their installation and this duration is to be used as the basis for fatigue analyses and for determining the effects of corrosion, erosion, fretting, wear, and the number of chemical cleanings. Edison desires that from the perspective of performance, the RSGs be designed as large as possible within the dimensional and other limitations imposed.” Edison picked the cheapest contractor to build their impossible dream machines in the shortest time and lowest cost. The steam generator design specification certifying Professional Engineer has to know that whatever he is specifying can be built by using a process known as Technical Bid Evaluation. Complex Steam Generators are not built on a handshake, a faulty design specification or subverting the regulatory process.
    SCE/MHI ignored the Dr. Pettigrews, the world’s most renowned tube vibration expert warning on the effectiveness of MHI flat bar AVBs for prevention of fluid elastic instability, San Onofre Chief Nuclear Office Dwight Nunn\’s warning on high void fractions and SCE/MHI AVB Joint Team recommendations to reduce void fractions. It is the personal opinion of the author since Edison did not comply with the recommendations of Dwight Nunn and SCE/MHI AVB Joint Team to reduce high void fractions, the performance warranties and Liquidated Damages of RSGs will be contested in California Courts for years until resolved mutually and peacefully by SCE/MHI by Arbitration and NRC Assistance.
    The NRC AIT Report is incomplete and erroneous, based on SCE Supplied Faulty Information and needs to be redone to clear the Public Record and Establish NRC Independence. NRC Commission independently needs to interview under oath SONGS Root Cause Team Members and assign a qualified thermal-hydraulic expert to review the complete SG, FW, RCS and other operational records for Units 2 & 3 for Cycle 16 to determine the true causes for Units 2 & 3 FEI, and the role of Mitsubishi Flowering Effect and RSG Manufacturing Differences. Based on interpretation of data published in NRC AIT Reported and SONGS SGM Procedure in 2012, Unit 2 had a steam pressure of 942 PSI and RCS Flows of 74 million Ibs./Hr. and Unit 3 had a steam pressure of 833 PSI and RCS Flows of 76 million Ibs./Hr. All the other Unit 2 Return to service reports had identical operating condition for both Units. Based on Hand calculations, only 5 MWt more in Unit 3 was required to change the flow from nucleate boiling to film boiling (Void fractions 99.0% ~ 99.6%) in 4% area of the Hot-Leg region of high wear. Mr. Ryan Lantz of San Onofre Special SONGS Panel was made aware of these discrepancies, but no action by the AIT Team Leader has been taken to date resolve these discrepancies. Mitsubishi Testing data indicates a tube-to AVB contact force between 10-30N is required to prevent FEI in Unit 2, whereas Unit 2 tube-to AVB contact force was only 2N. This force of 2 N was not sufficient to prevent occurrence of FEI given the benefit of doubt that operational condition were identical in both Units. The tube-to AVB contact force reported in Unit 3 was only 99.6% and fluid velocities 35-50 feet/sec)
    Lessons Learnt for NRC Commission/ from from Defective San Onofre RSG 10CFR 50.59
    Even though SCE, MHI and AREVA claim that operating and thermal-hydraulic conditions were the same in both units, Unit 2 did not experience tube-to-tube wear because of lower reactor thermal power and higher steam generator pressure operation and NOT double tube-to-AVB contact forces and better supports because of inadvertent accidental Unit 2 AVB design as explained by SCE and MHI. FEI did not occur in Unit 2, which is consistent with Westinghouse report. The Unit 2 Return to Service Engineering Reports, Computer Calculations, Testing Data and Tube Inspections prepared by SCE and its World Class Independent Consultants are full of errors, inconclusive and based on invalidated assumptions. The reports are Confusing, Conflicting and full of Smoking Mirrors. SONGS Inside Steam Generator Investigator, DAB Safety Team, Arnie Gundersen, John Large and Dr. Joram Hopenfeld showed ASLB and in other numerous public forums that these reports are incomplete, full of errors and Unit 2 Restart at 70% power is an Unsafe Experiment to keep Edison in Business.

    Edison was afraid of public scrutiny and sworn testimony by its Senior Team Leader Ship Directors certifying the safety and accuracy of these reports. So, afraid of these long legal investigations, Edison Senior Team Leader Ship Directors chose to safely shutdown San Onofre Units 2 & 3 blaming on ASLB ruling, Regulatory Hurdles, MHI delay in building the repaired/replacement generators, opposition from Super Intelligent & Nuclear Trained Safety Specialists, Burden on EIX/SCE Shareholders and Investors, SONGS Anonymous Steam Generator expert, Friends of the Earth and Concerned Rate Payers.

    Despite all foul-crying (No grid and voltage stabilization and summer power outages without San Onofre) EIX/SCE continues to make handsome profits by marking up cheap power bought from other undisclosed vendors and supplying power to customers at its own exorbitant price on its own Transmission and Distribution System. Plus EIX/SCE gets 11% additional rate of return on its Transmission and Distribution System. But the question is, who is going to pay and hire Edison Discriminating, Inept and Retaliating Senior Leader Ship Team Directors with no operating funds coming from Ratepayers. Certainly Not, the EIX/SCE Shareholders and Investors. That is why Senior Leader Ship Team Directors shut down San Onofre to use it as a Safe Haven from Prosecution and now want to enjoy and secure their positions by mismanaging the Huge Decommissioning Funds and Reap Profits, Paychecks, Perks and Bonuses for a Long Time in their Luxury Beach Homes. NRC and CPUC (Both Companies Cozy to Edison) need to break the SCE Brotherhood Tradition and award the San Onofre Decommission contract to a specialized and well-established company like Energy Solutions, so Rate Payers are NOT stuck once again paying for San Onofre Decommissioning Delays and Mistakes. Southern Californians want their beautiful beaches back ASAP in the Safe and Pristine Conditions.

  5. In 2009 songs 2 & 3 decomissioning budget was 3.5 billion in 2013 it exceeds 4. Billion how can nrc and public trust sce to manage decommissioing after rsg catastrotphy give the contact to another vendor sce is only interested how much money they can squeeze out of the decommissioning funds to make profits for themselves shareholders and return to ratrpayers for rsg fiasco sce has no trust left with nrc and public

  6. Sce shutdown songs because of its greed negligence and fear of criminal investigations and not because of aslb nrc arnie gundersen or the anonymous steam generator in the last 36 years sce has shutdown 3 units destroyed eight generators because they do not understand the relationship between high steam flows high void fractions and tube vibrations caused by fei and firv. 18 feet/sec is the limit of fluid velocity and 96% void fraction with a circulation ratio GREATER THAN 4 is the secret to avoid fei & firv but for the last I 6 yrars songs has missed the boat on these operational parameters. The point is sce does not kno what how to operate a power plant. After all this experience how can ratrpayers trust sce with $2B in Decommissioning Funds?

  7. Decontamination, Demolition, Dismantling and Decommissioning San Onofre is a very serious business, which requires the right management, procedures, contractors and strict NRC Oversight. Based on the last 10 years of observations at SONGS for SGRP, all these factors are missing at San Onofre now. Therefore, a 3rd neutral party with competent oversight organization reporting directly to NRC Resident Inspector and with Decontamination, Demolition, Dismantling and Decommissioning experience of a NPP is needed to do the job right first time following the INPO Principles of Excellence. Ratepayers cannot afford by Edison another Multi-Billion Dollar Mess.

  8. By The Times editorial board
    June 8, 2013

    Southern California Edison on Friday made its smartest decision yet about the troubled San Onofre nuclear power plant: It announced it was closing the facility once and for all. A year after the plant was taken off-line as a result of several problem-ridden steam generators, one of which had leaked a small amount of radioactive steam, the company finally decided to cut its losses and move on.

    It didn’t have to be this way. If Edison had gone through full regulatory oversight in 2010 and 2011, when its then-new $670-million steam generators were being designed and built, rather than choosing the cheaper and more expedient route of claiming that the new machinery was virtually the same as the old, there’s a good chance the design errors would have been caught in time. Edison might have a thriving nuclear plant today, well-positioned for license renewal in several years, which would have kept the two reactors operating for decades to come.

  9. REAL REASONS FOR SAN ONOFRE SHUTDOWN, INVESTIGATIONS NOT ASLB OR ECONOMICS

    [Information removed by the moderator]

    There may be lots of questions yet to be answered about Southern California Edison’s permanent shutdown of its San Onofre nuclear plant, but here are a couple about which there’s no doubt.

    Who’s responsible? Edison, 100%. Accept no argument that it did the best it could in overseeing a $700-million generator replacement project, but accidents happen. This wasn’t an accident: It was the failure of what Edison claims was its rigorous and negligent oversight of contractors. MHI was unable to build a steam generator specified by the inexperienced Edison Steam Generator Designers. On top of that Edison Engineers prepared defective 10CFR 50.59, subverted NRC regulatory process, ignored recommendations of SCE/MHI AVB Joint team established by Dwight Nunn, and misdirected MHI, Westinghouse, AREVA and Intertek in preparation of Unit 2 Return to Service Reports. SCE used and abused any body they could find to achieve their end goal, but failed and abandoned the San Onofre Sinking Ship in Panic.

    How much should Edison’s customers pay for the miss-engineering and inept mismanagement that led to mothballing a hugely important generating station? That’s easy. The answer is nothing. Not a dime.

    SONGS Management has been misleading the public since the inception of Steam Generator Replacement Project. Their focus has always been on profits/production and preaching false sermons of their overriding obligation to safety and achieving excellence in operations. They have indulged in systematic retaliation of workers reporting nuclear safety concerns regarding steam generators, cyber security program, fire/safety, discrimination and harassment. SONGS Unit 3 Root cause was rejected in early June 2012 by SONGS Insiders and they were warned about MHI. SONGS Management were warned by many insiders that the Unit 3 Root Cause was a result of design deficiencies and changes as a result of 11% increase in heat transfer area of the tubes due to change of Alloy 600 from to Alloy 690 and evolutionary untested AVB design. Dwight Nunn’s 2004 and 2005 letters warned about high void fractions and the capabilities of MHI to build such massive steam generators and evolutionary AVBs capable of handling high void fractions and tube fractions. Good SONGS SNO’s like Dwight Nunn and Ross Ridenoure were kicked out by SCE and resigned abruptly without explanation.

    It is not the Atomic Safety Board, NRC, MHI, Independent Safety Experts, CPUC or Public, which created significant additional uncertainty regarding SCE’s decision to get to an NRC decision to restart Unit 2 this year. It is the inept, inefficient and cunning SONGS Senior Leadership Team, which was focused on making money and bonuses for themselves, and subverting the regulatory process and not worried about plant safety, workers or the public. Justice Department, NRC Office of Inspector General and Investigations should continue their investigations into allegations of wrong doing by SONGS Senior Leadership Team. In the end, SONGS Senior Leadership Team was so afraid of these investigations, that they decided to abandon the ship by announcing Shutdown of SONGS blaming ASLB, NRC and Economics and coming with a new excuse, “This is not good for our customers, our investors and the region.” SCE was never worried or concerned about the customers, safety and the region. These guys did not have the courtesy of informing NRC, MHI, SONGS Workers, CPUC and SDG&E, their supporters through this crisis before announcing the decision.

    Thought of the Day on Dangers of Unit 2 Restart Experiment issued the day before Sanofre Panicky and Sudden Shutdown Announcement

    Continuous monitoring of primary-to-secondary leaks led to three shutdowns at the Cruas
    NPP: unit 1 in February 2004 and unit 4 in November 2005 and February 2006. Analyses
    carried out by EDF, further to the last two events, resulted in them being attributed to high
    cycle fatigue of steam generator tubes due to flow-induced vibration.

    The results of in situ examination initiated by the Cruas NPP operator showed that the flow
    holes of the uppermost Tube Support Plates (TSPs) were partially or completely blocked by
    corrosion products. This phenomenon is referred to in this paper as TSP “clogging-up” and it
    was considered potentially generic for EDF NPP fleet. For the Cruas leakages, it was
    established that the association of TSP clogging-up and the specificity of the Cruas steam
    generator (central area in the tube bundle where no tubes are installed) were responsible for
    a significant increase in the velocity of the secondary fluid in the tube bundle central area.
    The high velocity of the fluid in this region increases the risk of fluidelastic instability for the
    tubes. Based on this preliminary analysis, EDF has implemented preventive measures
    (stabilizing and plugging of tubes in the central area of the tube bundle deemed sensitive to
    high cycle fatigue risk).

    AREVA states, “Out-of-plane fluid-elastic instability has been observed in nuclear steam generators in the past and has led to tube bursts at normal operating conditions. Given identical designs, Unit 2 must be judged, a priori, as susceptible to the same TTW degradation mechanism as Unit 3 where 8 tubes failed structural integrity requirements after 11 months of operation. Based on the extremely comprehensive evaluation of both Units, supplemented by thermal hydraulic and FIV analysis, assuming, a priori, that TTW via in-plane fluid-elastic instability cannot develop in Unit 2 would be inappropriate. The nominal distance between extrados and intrados locations of neighboring U-bends in the same plane ranges from 0.25 inches to 0.325 inches due to the tube indexing. There are 36 U-bends in Unit 2 SG E-088 and 34 in SG E-089 with a separation less than or equal to 0.050 inches. “

    The circulation ratio of the replacement steam generator secondary-side fluid (ratio of riser mass flow-rate to steam outlet mass flow rate) at 70% power is ~ 4.9. A higher circulation ratio limits concerns regarding heat transfer performance, generator sludge management, corrosion product transfer, and tube dry-out.

    Based on recent Mitsubishi Testing conducted in Japan, tube-to-AVB contact force more than 30N is required to counteract the adverse effects of in-plane fluid-elastic instability. Unit 2 AVBs only have 2N contact force, which cannot stop tube-to-tube wear and tube burst at 70% Unit 2 normal power operations, if in-plane and out-of-plane fluid-elastic instability develops due to abnormal operation occurrences, main steam line breaks, inadvertent equipment errors and other plant transients.

    Let us assume, hope and pray for the benefit of 8.4 Million Southern Californians, IPC, State of California, CPUC, MHI, SCE and NRC, the probability of occurrence of these events is very low and nothing happens. But as stated above, there are 36 U-bends in Unit 2 SG E-088 and 34 in SG E-089 with a separation less than or equal to 0.050 inches. The problem lies that in these U-bends, even at 70% power and a circulation ratio of 4.9, localized areas with very poor circulation ratio and no flow zones (Flow areas blocked by SG debris and corrosion products) can develop resulting in very high void fractions. With no tube damping and insufficient contact forces, in-plane fluid-elastic instability and out-of-plane vibrations can develop, as we witnessed in Unit 3. Just like Unit 3, now, the tubes will start moving in the in-plane direction and hit other worn and plugged/stabilized tubes with low clearances and cause tube-to-tube wear. Also, the tubes in other non in-plane FEI areas will also start moving in the out-of-plane direction, hitting already damaged AVBs with sharp corners (Zero Radius) resulting in the existing incubating cracks in the tubes to grow at a undefined rate.
    Now the tubes are wearing and cracks are growing without the knowledge of the operator, because there is no instrumentation installed in the SGs as a part of the NRC Confirmatory Letter to warn/alarm the operator, as to what is going on about this kind of event. This event can occur at any time and propagate during the Unit 2, 5-month experiment window. Now, one, two or more than 5 tubes can potentially leak and/or rupture and the operator gets sudden warning/alarms through existing radiation monitors and proposed temporary N-16 detectors located on the main steam lines.

    Shift Manager has only 15 minutes to diagnose, trouble shoot, declare the event and notify the Offsite Agencies for activation of the SONGS Emergency Plan. Before, Shift Manager can call for additional help, activate TSC, OSC, EOF, JIC or start taking actions to mitigate the consequences of a nuclear accident in progress, the reactor trips, turbine trips, main steam lines over pressurizes due to sudden turbine load rejection. The main steam lines atmospheric valves and/or main steam line relief valves will instantaneously open to prevent the main steam line from over pressurization and start dumping the un-partitioned radioactive coolant containing iodine with steam into the environment. In less than 15 minutes, 60 tons of radioactive coolant contained in the faulty and un-isolatable steam generator, will leak to the environment, melt the fuel in the reactor and release offsite doses in excess of Control Room limit of 5 Rem TEDE, and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE.

    Based on the NRC Studies, Independent Safety Experts Observations and observation of SONGS Emergency Plan Drills, San Onofre Emergency Plan is not proven to notify, shelter (Plus KI Tablets) and evacuate the transients, disabled residents, affected families and children within the 10 mile zone during rush traffic hours in the event of above described a sudden large early frequency release radiological accident. A nuclear fallout from San Onofre can shutdown completely the business at Los Angeles and Long Beach Harbors and chock the already fragile economy of Southern California.

  10. NRC Job is not done. Now they should start investigating SCE SG cover up, discrimination, intimidation, and retaliation concerns. SCE officers were afraid of testifying under oath and being investigated by US Justice Department, NRC Office of Inspector General and Investigations.

  11. San Onofre Unit 2 Sad Saga Continued – Conclusions and Lingering Questions for the benefit of NRC Commission and San Onofre Panel
    1. It is unanimous decision of unbiased independent safety and SONGS Insider Experts that SONGS 10CFR 50.59 was not performed properly by SCE. If NRC had reviewed these changes in detail through a 50.90 License Amendment like Palo Verde SGRP, then the problems with Unit 3 FEI high void fractions, untested design and unanalyzed operational changes could have been averted. Then this whole embarrassing fiasco for NRC Commission, SCE, MHI and entire nuclear industry could have been averted.
    2. Root Cause for Units 3 and 2 have not been determined because:
    a. Uncertainties with computer modeling for in-plane fluid elastic instability,
    b. Operational differences between Units 2 & 3 have not been analyzed in detail that is why, SCE, MHI, AREVA, Westinghouse, NRC have not arrived at a clear and unanimous conclusions.

    3. SCE, MHI, AREVA, Westinghouse and NRC have not addressed the synergic effects of tube-to-tube wear and high cycle fatigue.

    4. SCE/MHI explanation of tube-to-AVB contact forces and fatigue levels in Unit 2 is incomplete and totally unsatisfactory.

    5. Unit 2 Tubes have not been inspected for internal incubating cracks.

    6. SCE/NRC explanation of new License Amendment to operate at 70% power is incomplete and totally unsatisfactory.

    7. SCE/MHI/NRC still do not understand that you cannot design anti-vibration bars to prevent the adverse effects of high void fractions of 99.6%. Therefore, the best advice is to operate the completely re-built replacement steam generators at void fractions of 96%, steam pressures > 950 psi, 1500 MWt and recirculation ratios >5.

    The former head of the Nuclear Regulatory Commission said Southern California Edison’s plans to restart San Onofre at 70 percent power do not inspire him with confidence. A question specifically about the proposal to restart the San Onofre nuclear power plant elicited a skeptical response from Jaczko. The plant has been offline since January 2012 after a small radiation leak and the NRC is currently considering a proposal to restart one reactor at 70 percent for five months. “Just looking at it as an outsider now,“ Jaczko said, “the approach that is being taken is not one that instills tremendous confidence in me because the approach is for operation at reduced power. In principal, what you should see is design modifications and changes that allow operations at the licensed power levels.” In other words: a fix. Jaczko said the restart proposal creates doubt in his mind that there’s a complete understanding of what’s wrong with the faulty steam generators at San Onofre.

    Arnie Gundersen, said the probability of a nuclear power plant accident is in fact much higher than estimates, becuase five units have melted down in the past 35 years: one at Three Mile Island, one at Chernoble and three at Fukushima.“ We are dealing with a technology that can have 40 great years and one bad day,” Gundersen said, “and that one bad day can destroy a country.”

    Peter Bradford, a commissioner with the NRC at the time of the Three Mile Island accident, said the regulatory agency pledged to become more transparent but in fact the opposite has happened. He pointed out little was done after Fukushima to improve safety at U.S. power plants, whereas there was heightened security after the Boston bombings. Bradford called for appointing nuclear regulatory commissioners who have a record of protecting public safety, as well as a record of technical experience with the industry.

  12. Edison with great confidence and arrogance quotes approval of San Onofre SGRP 10 CFR 50.59 via NRC Safety Evaluations and NRC AIT Report. Edison seeks to minimize damage from Dwight Nunn’s Letters and rejection of recommendation to reduce high void fractions by SCE/AVB Team. The high void fractions of 99.6% to maximize thermal output from RSGs caused a tube leak and unprecedented tube-to-tube wear and high cycle thermal fatigue cracks in Unit 3. The high void fractions of 98.9% to maximize thermal output from RSGs caused high cycle thermal fatigue cracks in thousands of Unit 2 tubes.
    Edison states, “Replacement of the steam generators is a replacement in-kind in terms of an overall fit, form, and function with no, or minimal, permanent modifications to the plant systems, structures, and components (SSCs)· The fact that the RSG were designed, fabricated and examined to the newer edition of the ASME Code is an enhancement over the OSGs.”
    SCE website states, “SCE advised the NRC that the San Onofre steam generators contained a number of different features from the precious design. In fact, safety evaluations prepared by the NRC in connection with amendments to the San Onofre license associated with the steam generator replacements described the most important of those changes in detail. At no time did SCE hide the differences from the NRC, nor did it seek to mislead the NRC concerning the applicability of Section 50.59 to the project. Any suggestion that seeks to draw from the November 2004 letter a contrary conclusion is simply incorrect and relies on the fundamental error of viewing Section 50.59 as applying to identical, or “like for like” replacements.”
    A like-for-like replacement is defined as the replacement of an item with an item that is identical. For example, the replacement item would be identical if it was purchased at the same time from the same vendor as the item it is replacing, or if the user can verify that there have been no changes in the design, materials, or manufacturing process since procurement of the item being replaced. If differences from the original item are identified in the replacement item, then the item is not identical, but similar to the item being replaced, and an evaluation (Such as 10CFR 50.59, Changes, tests or experiments..) is necessary to determine if any changes in design, material, or the manufacturing process could impact the functional characteristics (high void fractions, high steam flows, high fluid velocities, excessive tube vibrations, etc.) and ultimately the component’s ability to perform its required safety function (SG Tube structural Integrity). If the licensee cannot demonstrate that the replacement item is identical and differs in design, or results in a design change, new test or experiment, which adversely affects RSG’s functional characteristics (high Void fractions, high steam flows, Wrong AVBs for FEI) and ultimately the RSGs ability to perform its required safety function (RCS Barrier), then the licensee needs to inform NRC ASAP and proceed safely for a NRC Approved 50.90 License Amendment like Palo Verde.
    Flow-induced Vibrations & Fluid Elastic Instability
    The World’s Foremost Renowned Professeur Titulaire, Michel J. Pettigrew, Ecole Polytechnique de Montreal, on the subject of fluid elastic instability and turbulence-induced vibration states in the 1970’s, “It is concluded that, although there are still areas of uncertainty, most flow-induced vibration problems can be avoided provided that nuclear components are properly analysed at the design stage and that the analyses are supported by adequate testing and development work when required. There has been no case yet where vibration considerations have seriously constrained the designer.” Dr. Pettigrew told the NRC Commissioners again in 2013 that San Onofre Replacement Steam Generators flat anti-vibration bars do not provide a positive restraint against fluid elastic instability.
    Shahab Khushnood, Zaffar M. Khan, M. Afzaal Malik, Zafar Ullah Koreshi and Mahmood Anwar Khan wrote in 2003, “Flow-induced vibration is an important concern to the designers of heat exchangers subjected to high flows of gases or liquids. Two-phase cross-flow occurs in industrial heat exchangers, such as nuclear steam generators, condensers, and boilers, etc. Under certain flow regimes and fluid velocities, the fluid forces result in tube vibration and damage due to fretting and fatigues of tubes. Prediction of these forces requires an understanding of the flow regimes found in heat exchanger tube bundles. Excessive vibrations under normal operating conditions can lead to tube failure. Relatively little information exists on two-phase vibration. This is not surprising as single-phase flow induced vibration; a simpler topic is not yet fully understood. Vibration in two-phase is much more complex because it depends upon two-phase flow regime, i.e. characteristics of two-phase mixture and involves an important consideration, which is the void fraction. The effect of characteristics of two-phase mixture on flow-induced vibration is still largely unknown. Two-phase flow experiments are much more expensive and difficult to carry out as they usually require pressurized loops with the ability to produce two-phase mixtures. Although convenient from an experimental point of view, air–water mixture if used as a simulation fluid, is quite different from high-pressure steam–water. A reasonable compromise between experimental convenience and simulation of steam–water two-phase flow is desired.”
    One Masters Research Student R. Viollette and Dr. Pettigrew state in a 2006 research paper, “Fluid elastic instability is the most important vibration excitation mechanism for heat exchanger, or steam generator type of tube bundles. It is so because of the very high vibrations amplitude that it can induce to the tubes, which can lead to rapid failure by fatigue or wear. Also, unlike vibrations induced by vortex shedding (vortex-induced vibrations), fluid elastic instability is not a self-limiting phenomenon: amplitude of vibrations does continue to increase with velocity past the critical onset of the instability.” Dr. Pettigrew told in his 2006 paper that design of flat anti-vibration bars design need to be verified against fluid elastic instability.
    Quotes from [Redacted] Anonymous Engineers
    One [Redacted] Root cause Leader stated, “ I wish that these [Redacted] Engineers had made these changes one by one and tested them before implementing them in the RSG design.”
    Another [Redacted] Project Manager said, “ I wish that these [Redacted] Engineers had duplicated Palo Verde Generators, went through an independent design review …….”
    Another [Redacted] Retired Manager said, “These whole design changes by these [Redacted] Engineers and [Redacted] Management were geared only to maximize the thermal performance and profits or the new Replacement Generators because of change from Alloy 600 tube material in the OSGs to Alloy 690 in the RSGs.” Even though Alloy 690 has better corrosion resistance than Alloy 600, Alloy 690 has a 10-12 less heat transfer coefficient than Alloy 600.
    10 CFR 50.59 – Changes, tests, and experiments.
    (c) (2) A licensee shall obtain a license amendment pursuant to § 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would:
    (i) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the final safety analysis report (as updated);
    Utility Likely Answer N/A with no explanation
    Comments: The [Redacted] Engineers evasively answered N/A and did not answer the following NRC 50.59 Inspection Manual safety questions required before making the change. (1) Systems and components affected by the change (What is the effect of the change on their capability to perform their specified or intended functions?); (2) Parameters of the accident analysis affected by the change (Are all the relevant design basis accidents and transients identified?); (3) Potential effects of system or component failure (i.e., the question, “what would happen if…” is explored and answered in the evaluation), and (4) How the evaluation criteria are met. The first criterion is if the CTE would result in more than a minimal increase in the frequency of an accident previously evaluated in the FSAR (as updated). The intent of the criterion is to allow changes to be made without approval unless there is a discernible, attributable increase in frequency of an accident. There must be some reason to believe that the CTE would result in a more than minimal impact upon the accident frequency (as because it affects the integrity of the reactor coolant system, or the ability of SSC to remove decay heat, or makes an initiating event more likely to occur). Licensees must still meet applicable regulatory requirements. As noted in NEI 96-07, departures from the design, fabrication, testing and performance standards in the General Design Criteria are not compatible with a “no more than minimal increase” standard.
    Title 10 of the Code of Federal Regulations (10 CFR), “Energy,” establishes the fundamental regulatory requirements for the integrity of the SG tubes. Specifically, the general design criteria (GDC) in Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” state that the RCPB— (1) Shall have “an extremely low probability of abnormal leakage…and gross rupture” (GDC 14, “Reactor Coolant Pressure Boundary”), (2) Shall be designed with sufficient margin” (GDCs 15, “Reactor Coolant System Design,” and 31, “Fracture Prevention of Reactor Coolant Pressure Boundary”), and (3) shall be of “the highest quality standards practical” (GDC 30, “Quality of Reactor Coolant Pressure Boundary”)
    Most of the successful steam generators operate at void fractions less than 98.5%, steam pressures > 900 psi and recirculation ratios >4. This ensures that high dry steam is not produced in localized areas of U-tube bundle and tubes and tube supports are not subject to excessive vibrations/tube contact and tube dry-out due to fluid elastic instability, flow-induced vibrations and excessive fluid hydrodynamic pressures. The original San Onofre steam generators had circulation ratios of 3.3, void fractions of 96.1%, steam pressures of 900 psi and did not experience fluid elastic instability in 28 years of operation. All the four San Onofre replacement units had poor circulation ratios of 3.3, void fractions ranging from 98.9-99.6%, steam pressures ranging from 833-942 psi, RCS flows between 74-76 Mlbs/hour, narrow tube to pitch tube diameter, excessive number of tubes (9,727), extremely tall tubes (average length of heated tube increased by 50 inches, equivalent to the addition of ~650 tubes), 116,000 square feet of tube heat transfer area (increased from 104,000 in the OSGs) and virtually no in-plane restraints. Successful Steam Generator manufacturers control a combination of design and operational features to prevent tube/support damage from fluid elastic instability, flow-induced vibrations and excessive fluid hydrodynamic pressures. Because of the customer’s (SCE) desire to increase thermal output from steam generator and profits, Steam Generator manufacturers do not design and test/experiment anti-vibration bar support systems in an operating steam generator to demonstrate its capability to prevent tube/support damage from high void fractions (fluid elastic instability), flow-induced vibrations and excessive fluid hydrodynamic pressures. This is against the 10CFR 50.59 rules and GDC 14, 15 and 30.
    The changes in void fractions due to numerous unanalyzed and untested design and operational changes not only increased significantly the frequency of occurrence of a steam generator tube leak in Unit 3 but destroyed the Edison’s newly constructed Billion Dollar 21st Century Safest and Innovative Steam Machines in less than 2 years of operation due fluid elastic instability, flow-induced vibrations and excessive fluid hydrodynamic pressures. In addition, 8 tubes in Unit 3 failed structural and leakage integrity criteria during their in-situ pressure main steam line break pressure testing. The high void fractions ultimately destroyed the RCS Barrier protected by the tubes. This was one of the largest steam generators ever built and represented a significant increase in size from those that Mitsubishi Heavy Industries has built in the past. It required Mitsubishi Heavy Industries to evolve a new design beyond that which they currently have available. Such design evolutions required a careful, well thought approach that should have fully evaluated the risks inherent in creating a new and significantly larger steam generator. Such design evolutions challenged the capability of existing models and engineering tools used for proven steam generator designs. Success in developing a new and larger steam generator design required a full understanding of the risks inherent in this process and putting in place measures to manage these risks. Void fraction was an important thermal- hydraulic parameter, related to the probability of tube dry out occurring during power operation (the higher the void fraction, the higher the probability of dry out). Tube dry out is an undesirable phenomenon as it may eventually result in tube cracking. AVB team recognized that the design for SONGS RSGs resulted in higher steam quality (void fraction) and had considered making changes to the design to reduce the void fraction (e.g. using a larger down comer, using larger flow slot design for the tube support plates, and even removing a TSP). But each of the considered changes had unacceptable design consequences and the SCE/MHI AVB design team agreed not to implement them. Among the difficulties associated with the potential changes was the possibility that making them could impede the ability to justify the RSG design under the provisions of 10 CFR 50.59.” In Unit 2, the RSGs were installed and tested in 2009/10 and in Unit 3 in 2010/11. The RSG post-installation test results met acceptance criteria for all specified test parameters except measurements of void fractions to confirm the safety of new design, thus wasting all the money, time and the effort put into their fabrication.
    _X__ Implement the activity per plant procedures without obtaining a License Amendment.
    Likely Wrong Answer by Utility
    __ X__ Request and receive a License Amendment prior to implementation.
    Results in more than a discernible, attributable increase in the frequency of occurrence of steam generator tube leakage/cracking/rupture an accident previously evaluated in the final safety analysis report (as updated). –
    This is the Right Answer

  13. My personal opinion :The insurance companies will not cover damages for a melt down of a NPP when it is in good working order, why should the public take the further added risk with a one of a kind jerry rigged repair job that may have long term flaws that no other plant has experienced, who knows what long term maintenance schedules are needed to run this plant, I have no confidence in the plant owners to go above or beyond the normal maintenance they do with a proven design because of the extra costs, they are determined to work the plant at 70% to brake even with no losses at any risk.

  14. Update From FoE:

    San Onofre: Friends of the Earth to NRC — Operating unsafe reactor as a nuclear experiment is not an option

    Posted May. 24, 2013 / Posted by: Adam Russell

    Motion calls for convening hearing panel on license amendments

    WASHINGTON, D.C. – Friends of the Earth has filed a motion with the Nuclear Regulatory Commission requesting that a Licensing Board be convened to review license amendments that are required for Southern California Edison’s crippled San Onofre reactors, and provide an opportunity for an adjudicatory public hearing before any decision on restart.

    The motion calls on the NRC to implement the decision of its own Atomic Safety Licensing Board, which ruled last week that the current process for evaluating and approving restart of the San Onofre reactor Unit 2 is a de facto license amendment proceeding. The ASLB ruled that without making formal changes in San Onofre’s license to address major safety issues Edison would be in violation of NRC regulations. The licensing board also ruled that a public hearing should be held prior to any decision on a new license.

    Friends of the Earth’s motion, filed jointly with the Natural Resources Defense Council, calls on the NRC to convene an ASLB panel to preside over an adjudicatory license amendment proceeding. If the motion is accepted by the NRC, it would require Edison to submit a new license amendment that addresses unresolved safety issues. It would also consolidate the existing and newly required license amendment into one proceeding.

    “The ASLB was very clear that operating San Onofre in its present uniquely damaged state would be an ‘experiment’ and outside its safety license requirements. Our motion makes clear to the NRC that moving forward on the current plan without further license amendments is not an option for them or Edison,” said Kendra Ulrich, nuclear campaigner with Friends of the Earth.

    “It’s not that complicated,” Ulrich continued. “An unprecedented safety flaw is found at a nuclear reactor located in a highly active earthquake zone, and the agency’s own judicial safety board says it’s not legal to operate. The ASLB’s order confirms that all the major safety issues that Edison and the NRC have failed to acknowledge, never mind resolve, must be considered by a full license amendment process with an adjudicatory public hearing prior to restart.”

    Established by the NRC Commissioners in late 2012 to hear a case brought by Friends of the Earth, the ASLB ruled last week that the current proposal from Edison to restart the reactor was an “experiment” with a reactor that had suffered “unique” phenomenon of steam generator tube damage. Edison has provided no data to prove that it can operate San Onofre without further damage — in fact, its own consultants acknowledge that steam tubes will continue to incur further damage even if the reactor operates at only 70 percent power. Last week nuclear engineers for Friends of the Earth provided evidence to the NRC that operating San Onofre will not only lead to further tube damage and failure but risk a major nuclear accident potentially affecting the lives of millions of people in Southern California.

    Both the NRC and Edison in all of their assessments and proceedings have failed to address and resolve important safety issues. These include the vulnerability of the San Onofre steam generators and the nuclear reactor to seismic impact, which could lead to a significant accident and radiation dosage for the public. San Onofre lies within 50 miles of 8.7 million people in Southern California.

    ###

    Contact:
    Kendra Ulrich, (216) 571-7340, kulrich@foe.org
    Shaun Burnie, (202) 251-1862, sburnie@foe.org

    Categories: Climate And Energy, News Releases, San Onofre News

  15. Road Map for End to San Onofre Sad Saga – NRC/SCE/MHI/CPUC and Public Awareness Series
 – Excuse me for the formatting, misspellings or grammatical errors.
    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    Edison has in excess of $32 billion investment in transmission and distribution system. Edison is guaranteed 10-12 rate of return on this investment, which is added to the Ratepayers monthly electric bills. For Edison to claim this money, SCE needs a base-load plant like San Onofre to produce safe, reliable and affordable nuclear power. In other words, no matter what Ted Cravers says, Edison will opt to replace San Onofre Steam Generators, rather than Decommission and Dismantle San Onofre Units 2 & 3, “when push comes to shove.”

    Now the existing steam generators are unsafe to operate any power even with MHI’s proposed interim repairs. Therefore, a License Amendment with Public Hearings at this time won’t help. MHI does not have the tools, technology, skills, research and testing facilities to rebuild these complex generators @100% power safe operation, as Edison Management wants it. SCE wrote defective specifications, MHI made tall and innovative claims that they can rebuild these generators better than Westinghouse. SCE/MHI AVB team rejected the adverse design changes and avoided informing the NRC. Now NRC, EIX/SCE Shareholders, MHI, CPUC and Ratepayers are holding the $2 Billion Radiological Garbage Bag and suffering the consequences.

    MHI will never be able to rebuild these generators for free as Pete Dietrich wants. MHI says that it will take 5.5 years to rebuild these generators. The new generators for safe operation can be rebuild by MHI to deliver 1500 thermal MWt, so these generators would conform to MHI’s contemporary experience. To avoid fluid elastic instability and flow-induced vibrations, these new generators should have 88% heat transfer area of the MHI replacement generators and operated at 950 psi steam pressure, void fractions of 96% and a recirculation ratio >5. SCE and MHI have to share the cost of new generators and should apply for a 10CFR 50.90 License Amendment with trial like public hearings. MHI can use multiple tube manufacturers and retain AREVA to provide complete independent and detailed design, fabrication and testing oversight for the entire project. This can all be accomplished in less than 2 years with Extended NRC Oversight. SCE should totally keep its hands free of this entire operation.

    In addition, Ted Craver should improve the working environment, maintenance, fire safety, procedure compliance, work control, nuclear training, and configuration control at San Onofre. This can be accomplished by retraining the workers, writing simple procedures and firing the existing inefficient, intimidating, retaliatory, discriminating and self-serving SONGS Senior Leadership Team.

  16. San Onofre Sad Saga Continued – NRC/SCE/MHI/SCE Experts/CPUC and Public Awareness Series
    Excuse me for the formatting, misspellings or grammatical errors.

    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    1. Because of operational differences between Units 2 (Steam Pressure 942 psi, RCS Flows ~ 74 MLbs/Hr.)& 3 (Steam Pressure 833 psi, RCS Flows ~ 76 MLbs/Hr.), FEI did not occur in Unit 2 (out-of-plane vibrations and/or may be in-plane vibrations existed far below the level of FEI to cause tube-to-tube wear). This finding is consistent with Westinghouse OA. Unit 3 FEI occurred in 4% area of the tubes in the Hot leg due to high steam flows (SG Heat Transfer Coefficient Exceeded by 5 MWt, Change from Nucleate Boiling to Film Boiling), high in-plane fluid velocities (35-50 feet/sec), low tube clearances (0.05 – 0.25 inches), extremely tall tubes, low steam pressures, high RCS flows and Mitsubishi Flowering Effects ( increased the tube-to AVB Gaps in Unit 3 compared to Unit 2 as measured by ECT). SONGS Original Combustion Engineering steam generators were operated at a void fraction of 96.1%, fluid velocities of 22 feet/sec and steam pressures of 900 psi, and a circulation ratio of 3.3. That is why FEI did not happen in Original San Onofre Units 2 & 3 for 28 years, but, these generators did suffer from flow-induced random vibrations. According to Dr. Pettigrew, for optimum steam generator operation, operations and design engineers are advised to keep fluid velocities 16 ksi, which exceeds the ASME Limit of 13 ksi. Review of 170,000 San Onofre Tube Inspections indicates that SCE and its vendors have not used the latest technology probes used by the Canadian and Finland Engineers for detection of incubating and circumferential cracks. These cracks can cause instantaneous tube ruptures during SONGS Unit 2 normal 70% steady state power (at any time in the 5 month operation), anticipated operational occurrences, inadvertent equipment manipulations and Design Basis Accidents. Due to the amount of abnormal and unprecedented degradation reported in thousands of Unit 2 tubes and inadequacy of Unit 2 AVBs to prevent FEI, inspections beyond the current NEI Steam Generator Management Program are required to assure adequate protection of health and safety of 8.4 Million Southern Californians and minimize Environmental, Ecological and Economic Damage from potential nuclear accidents. The following types of scenarios are possible to inflict the above damage:
    A. Spontaneous fretting fatigue rupture of a single steam generator tube in the free span with a stuck open relief valve or a broken header
    B. Tube Ruptures from Unplanned closing of an isolation valve.
    C. Seismically –Induced Tube Rupture
    D. Station Blackout, SBO
    E. Main Steam Line Break, MSLB

    From any tube rupture and leakages, concurrent with containment bypass, these events might cause offsite radiation doses in excess of 10 CFR Part 100 as evaluated in the SONGS FSAR. Any of these two events would cause a simultaneous reactor, turbine, feedwater and reactor coolant trips. Due to feedwater pump trip, the RSG U-bundle secondary water level will shrink and tubes will be uncovered for a period of at least 10 minutes and experience a sharp drop in secondary side pressure. The entire sub-cooled feedwater inventory contained in the faulted RSG will instantaneously flash to high dry steam. The combination of resonant, out-of-plane, in-plane vibrations, jet impingement forces, broken tube fragments and RSG debris will cause large axial, bending, dynamic and cyclic loads on all the tubes, tube support plates, retainer bars and anti-vibration structure. The strength of the welded and mechanical connections of these low frequency retainer bars, retaining bars and bridges have not been analyzed for the effects of these cumulative loads to prevent AVB structure displacement, deformation or collapse during anticipated operational transients and main steam line breaks. The displacement, deformation or collapse of AVB structure along with the large axial, bending, dynamic and cyclic loads can potentially cause thousands of worn, cracked, plugged and stabilized tubes to exceed several times the allowed tube ASME Endurance Limit of 13.6 ksi. If this happens, multiple tube ruptures will occur at tube-support plates, mid-spans, free spans and tube-to-anti-vibration bar notched interfaces. Since all the water from the RSG would escape to the environment, the iodine-131 from un-partitioned reactor coolant leaking out the rupture tubes will also escape to the environment in less than 10 minutes with 60 tons of radioactive coolant. Consistent with Fukushima Task Force Lessons Learnt and NRC Commissioner Meeting Transcripts, this event will be considered as a beyond design basis event, and SONGS Operators will be unable to take any timely mitigation actions to stop a severe nuclear accident in progress. If the prevailing winds are towards San Clemente, consistent with NRC Inspector General Reports, NRC Studies and observations of SONGS Emergency Plan Drills for the last six years, SCE and Offsite agencies would not have time to respond, notify, evacuate, shelter or give Potassium Iodide to the affected residents within the 10-mile affected emergency planning zone. The casualties, and short, long-term cancer affects to the affected population will depend upon the iodine spiking factor and the duration of blowdown, but will significantly exceed the NRC approved SONGS Control Room limit of 5 Rem Total Effective Dose Equivalent (TEDE), and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE

    3. MHI tube-to-AVB contact forces to prevent FEI and reduce flow-induced random vibrations based on ECT results, Visual Inspections, Quarter Bundle Model, Statistical Analysis, Manufacturing Dispersions, AVB Twist Forces Testing and New Anti-Vibration Test Data range from 2 N to > 30 N . According to Mitsubishi recent testing data, additional thicker tubes with contact forces in excess of 30N are required in Unit 2 are required to prevent adverse effects of FEI @ @100%RTP. Best on the best available evidence, existing Unit 2 AVBs have a significant smaller contact force (2N) than 30N required to prevent FEI. This data appears to be contradicting and significantly flawed. NRC needs to question MHI and AREVA to determine correct tube-to-AVB contact force number to prevent FEI with tube-bundle uncovered and depressurized during a potential MSLB with Unit 2 at 70% power?

    4. The in-plane critical velocities based on latest 2011 research papers, Dr. Pettigrew’s and Dr. Mureithi Testing and MHI Root Cause Data range between 35-50 feet/sec.

    5. AREVA, Westinghouse, MHI and SCE conclusions on Unit 2 FEI are conflicting, contradicting, smoking mirrors and ambiguous based on a review of SCE Unit 2 return to Service Reports and NRC Commissioners Transcripts.

    6. SCE, NRC AIT, Westinghouse, AREVA, MHI and Intertek have not addressed the combined effects of tube-to-tube wear, circumferential and incubating cracks caused in tubes due to tube-to-tube wear and high cycle metal fatigue caused by fluid elastic instability. One European Nuclear site experienced 3 tube leaks between 2004-2006 due to fluid elastic instability and high cycle fatigue.

    7. Based on Unit 3 tube leak and MSLB in-situ testing, SCE has not addressed the effects of fluid elastic instability on multiple SGTRs concurrent with a MSLB in the Updated UFSAR, 10CFR 50.59 and proposed 10CFR 50.92 No Significant Hazards Analysis License Amendment. Operating Unit 2 degraded RSGs @70% power due to the above described potential accidents results in multiple SGTRs due to FEI and incubating cracks and the consequences are as follows:

    A. The Proposed License Amendment Would Involve a Significant Increase in the Probability or Consequence of an Accident Previously Evaluated.
    B. The Proposed License Amendment Would Involve the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated.
    C. The Proposed License Amendment Would Involve a Significant Reduction in a Margin of Safety.

    In order to issue a finding of no significant hazards considerations, the NRC Staff bears
    the burden of showing that the hazards considerations as a result of the ASLB’s recent decision in the CAL proceeding are insignificant. The Staff cannot make that showing, and consequently the proposed finding must be withdrawn and a hearing on the proposed license amendment held by an ASLB before the amendment may be approved by the NRC. As the ASLB recently held with respect to San Onofre Unit 2:
    We conclude that until the tube degradation mechanism is fully understood, until reasonable assurance of safe operation of the replacement steam generators is demonstrated, and until there has been a rigorous NRC Staff review appropriate for a licensing action, the operation of Unit 2 would be outside the scope of its operating license because the replacement steam generator design must be considered to be inconsistent with the steam generator design specifications assumed in the FSAR and supporting analysis.

    May 23, 2013: California Democratic Sen. Barbara Boxer told Nuclear Regulatory Chairwoman Dr. Allison Macfarlane that she wants two things before the restart of the San Onofre Nuclear Generating Station is considered: completion of an investigation and a public hearing. Boxer made her comments Thursday as part of the Senate reconfirmation hearing for Macfarlane, who is seeking another term as chair of the NRC. Boxer scoffed at Southern California Edison’s (SCE) plan to change its operating license to restart the number two reactor at San Onofre at partial capacity. She repeated the U.S. Atomic Safety Licensing Board’s description of the plan as an “experiment.” Boxer commented on SCE’s plan to operate the reactor at 70 percent. “We’ll see what happens, we’ll see how it goes,” Boxer said during the hearing. “That’s like saying I think I fixed the damaged brakes on your car, but don’t drive it over 40 miles per hour.” Boxer repeatedly brought up the 8 million people living within 50 miles of the nuclear plant, saying if someone came to the NRC today and asked for a license to operate a nuclear power plant at that site, “in a seismic and a tsunami zone, we all know every single commissioner would say, ‘don’t you think you could find a better place for it?’”

    8. SCE has not addressed the True Root Cause of Unit 3 tube-to-tube wear (Untested and unanalyzed design changes, adverse operational and thermal-hydraulic parameters, human performance errors (Avoidance of 10CFR 50.90 by portraying RSGs as “like for like” replacement, rejecting SCE/MHI AVB Team proposed changes to reduce void fractions by improving circulations ratios, failure to review of FEI research papers by Dr. Pettigrew, Dr. Ivan Cotton, Dr. Dhir, etc., failure to benchmark other successful CE replacement generators (Palo Verde)) and actions to prevent tube-to-tube wear in Unit 2 as required by CAL.

    9. Westinghouse tube wear rates calculations are non-conservative and based on old SGs testing, which significantly differ in design compared with San Onofre RSGs.

  17. San Onofre Sad Saga Continued – NRC/SCE/MHI/SCE Experts/CPUC and Public Awareness Series

    SCE is on the road to being Unpopular and Bankrupt without Public Support. EIX/SCE Management and Shareholders will find themselves alone holding the Expensive Bag Full of Radioactive Waste – Holding of Useless Proprietary Information is hurting SCE, its Vendors and NRC – It makes Public more suspicious of wrongdoing by SCE, its Vendors and NRC

    Subject: Review of SONGS 10CFR50.59 and 50.92 Evaluations – SCE Designed and MHI Fabricated 21st Century Safest & Innovative Replacement Steam Generators

    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    QUIZ for NRC/SCE/MHI/Public SCE Experts and Public

    Tube-to-AVB contact forces required to prevent fluid elastic instability in San Onofre Replacement Steam Generators

    A. 2N
    C. ~ 10N
    D. > 30N
    E. None of the above or your best guess based on the information provided below

    Most of the steam generators operate at void fractions below 98.5%, steam pressures > 900 psi and recirculation ratios >4. This ensures operation of the SG in the nucleate boiling regime (damping of hot SG tubes to prevent in-plane vibrations (Fluid Elastic Instability), optimum operation of the SG thermal performance and minimization of tube vibrations. These operational parameters ensure the prevention of adverse effects of FEI (High Dry Steam, high fluid in-plane velocities, Film Boiling), flow-induced random vibrations and excessive dynamic pressures on tube-to-tube wear, tube-to-AVB/TSP wear, high cycle tube thermal fatigue (development of incubating cracks) and retainer bar-to-tube wear. Along with numerous untested and unapproved design changes made under the false pretense of “like for like” to avoid lengthy NRC 10CFR 50.90 Review and Public Hearings, [Redacted]… designed and [Redacted] fabricated 21st Century Safest and Innovative Replacement Steam Generators were operating outside the above operational parameters to maximize the SG thermal output and profits. MHI Root Cause states, “Thus, not using ATHOS, which predicts higher void fractions than FIT-III at the time of design represented, at most, a missed opportunity to take further design steps, not directed at in-plane FEI, that might have resulted in a different design that might have avoided in-plane FEI. However, the AVB Design Team recognized that the design for
    the SONGS RSGs resulted in higher steam quality (void fraction) than previous designs
    and had considered making changes to the design to reduce the void fraction (e.g.,
    using a larger downcomer, using larger flow slot design for the tube support plates,
    and even removing a TSP). But each of the considered changes had unacceptable
    consequences and the AVB Design Team agreed not to implement them. Among the
    difficulties associated with the potential changes was the possibility that making them
    could impede the ability to justify the RSG design under the provisions of 10 C.F.R.
    §50.59. Thus, one cannot say that use of a different code than FIT-III would have
    prevented the occurrence of the in-plane FEI observed in the SONGs RSGs or that any
    feasible design changes arising from the use of a different code would have reduced
    the void fraction sufficiently to avoid tube-to-tube wear. For the same reason, an analysis of the cumulative effects of the design changes including the departures from the OSG’s design and MHI’s previously successful designs would not have resulted in a design change that directly addressed in-plane FEI.” And we saw the end result of that [Redacted missed opportunity. Destruction of $ 1 Billion Dollar Steam Generators and Number 1 US Public Safety Concern/Nuclear Scandal, Controversy and Cover-up involving [Redacted] and others. Dr. Pettigrew told Dr. Macfarlane and the NRC Commissioners that [Redacted] AVBs simply do not provide a positive restraint against FEI.

    Here is the summary of San Onofre Tube-to-AVB Contact Forces and Accident Scenarios for your benefit:
    1. High Void fractions of 99.6%, high steam flows (film boiling), higher thermal reactor power per RSG (RCS Flows ~ 76 Million lbs./hr, 1737 MWt plus), high in-plane fluid velocities (35-50 feet/sec), circulation ratios of 3.3, narrow tube to pitch tube diameter, excessive number of 9,727 tubes, extremely tall tubes (average length of heated tube increased by 50 inches, equivalent addition of 650 new tubes), 116,000 square feet of tube heat transfer area, lack of in-plane restraints, steam generator operation at 833 psi and insufficient tube-to-AVB contact forces ( 2N per (redacted)) and better supports (smaller tube-to-AVB Gaps, Based on ECT Results) caused Flow-Induced Random Vibrations and (Redacted) Flowering Effect in Unit 2 @100%RTP.
    3. [Redacted] companies claim that operating and thermal-hydraulic condition were the same in both units, Unit 2 did not experience tube-to-tube wear because of double tube-to AVB contact forces and better supports because of inadvertent accidental Unit 2 AVB design. FEI did not occur in Unit 2, which is consistent with [Redacted]. The [Redacted] Report noted that the operational differences did not make any difference between Units 2 & 3. Throughout this entire paper, we will review [Redacted] claims: (1) About Unit 2 double tube-to AVB contact forces and better supports because of inadvertent accidental design, and (2) About Unit 3 insufficient tube-to AVB contact forces and loose supports because of intentional precision manufacturing.
    4. According to [Redacted], a Tube-to-AVB Contact Force of 10N is required to prevent FEI@100%RTP. It is noted that Tube-to-AVB clearances are significantly larger than the SONGS steam generator design clearance of 2 mils diametral. For the present, it is sufficient to note that the forces at AVB locations needed to prevent the onset of fluid-elastic instability are low. In contrast, after instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location. Calculation of the probability of the onset of in-plane fluid-elastic instability requires information in three areas: stability ratios, contact forces at AVB locations and a criteria for deciding whether AVB supports are effective or ineffective in terms of in-plane support. Stability ratios need to be known as a function of position in the bundle, number of consecutive ineffective supports and power level. Contact forces at AVB locations cannot be determined deterministically since the dispersion of gaps between tubes and AVB supports is random, and thus probabilistic in nature. The primary source of tube-to-AVB contact forces is the restraint provided by the retaining bars and bridges, reacting against the component dimensional dispersion of the tubes and AVBs. Contact forces are available for both cold and hot conditions. Contact forces significantly increase at normal operating temperature and pressure due to diametric expansion of the tubes and thermal growth of the AVBs. [Redacted] has calculated the response of a large U-bend with AVB supports subjected to turbulence and fluid-elastic excitation forces. Various gap (clearances) conditions were included along with contact forces ranging from 1N to 10N. An equal contact force was applied at all 12 AVB locations. Given the uncertain nature of fluid-elastic excitation forces, a direct application of the selected excitation function to SONGS at 100% power is problematic. However the scale of the contact force that prevented in-plane vibration is highly useful. A contact force of 1N did not resist in-plane motion but a force of 10N was completely effective.

    5. According to [Redacted] recent testing data, additional thicker tubes with contact forces in excess of 30N are required in Unit 2 are required to prevent adverse effects of FEI @ @100%RTP.

    6. Best on the best available evidence, existing Unit 2 AVBs have a significant smaller
    contact force (> 2N) than 30N required to prevent FEI.

    6, During AOO and MSLB events, Unit 2 at 70% power will experience void fractions of 100%, high steam flows (film boiling), high in-plane fluid velocities (35-50 feet/sec) and jet impingement from flashing feedwater. With contact forces of 2N, Unit 2 tube bundle would not be able to prevent the adverse effects of FEI, Flow-Induced Random Vibrations and Mitsubishi Flowering Effect. Multiple tube-ruptures can occur due to tube-tube wear, full circumferential rupture of tubes can occur due to incubating cracks and the entire degraded Anti-vibration structure can collapse.

    Answer . Send email at contact-mnes@mnes-us.com for the correct answer, because all the data is private and proprietary. Release of correct data in the public domain can point to Billion Dollar Mistakes made in the RSG Design, Manufacturing, Testing, Computer Simulation, Mock-up Testing, Statistical Analysis, and Thermal-Hydraulic Operational Analysis.

  18. San Onofre Sad Saga Continued – NRC/SCE/MHI/SCE Experts/CPUC and Public Awareness Series
    SCE is on the road to being Unpopular and Bankrupt without Public Support. EIX/SCE Management and Shareholders will find themselves alone holding the Expensive Bag Full of Radioactive Waste – Holding of Useless Proprietary Information is hurting SCE, its Vendors and NRC – It makes Public more suspicious of wrongdoing by SCE, its Vendors and NRC

    Subject: Review of SONGS 10CFR50.59 and 50.92 Evaluations – SCE Designed and MHI Fabricated 21st Century Safest & Innovative Replacement Steam Generators

    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    Preface: Of particular concern with SONGS Unit 2 restart at reduced power are undetermined and unexamined amount of incubating circumferential cracks located in tubes next to each other caused by fluid-induced random vibrations, high cycle thermal fatigue and in-plane fluid elastic instability. When one circumferentially cracked tube ruptures, the additional stresses can cause multiple or cascading tube ruptures, which can result in a nuclear meltdown. In addition, though the Unit 3 steam generators failed more catastrophically, it appears that there is a much larger pool of tubes out of alignment and in direct contact with support plates in Unit 2. SCE, MHI, AREVA, Intertek, Westinghouse and NRC are ignoring these cracks in their analyses. The difference in management of Steam Generator Tube Rupture between Finland and USA is, that no primary coolant (liquid and steam) release to the environment is allowed in Finland, while in USA, primary steam releases are not forbidden for profits to conduct risky experiments with people’s lives. This situation is unique to San Onofre Steam Generator and the Potential Extent of Condition does not affect any other MHI Steam Generators.

    Conclusions: For SCE to restart “Defectively-Designed and Degraded Unit 2”, in accordance with ASLB’s decision today, a full 50.90 License Amendment with trial like public hearing is required, because the pending license 50.92 amendment, CAL Actions, SCE’s response to NRR RAI’s, SCE Unit 2 Return to Service Reports and MHI Root Cause/Technical Evaluations do not fully satisfy the requirement of the Federal Regulations. SCE prepared a defective 50.59 Replacement Steam Generators (RSGs) evaluation and directed MHI not to inform NRC of the RSGs design deficiencies. NRC region IV and AIT Team did a very poor job of the review of the SCE prepared defective 50.59 evaluation and defended SCE by blaming all the mistakes on the MHI. Now from review of the press reports, one is likely to conclude that NRC Commission and NRR are still leaning towards approving SCE’s permission to Restart Unit 2 in violation of the President of The United States, US Congress, Federal Regulations, NRC ASLB Board and against the safety interests of 8.4 Million Southern Californians.

    NRC News, May 13 (Reuters) – ASLB: San Onofre Confirmatory Action Letter Process Offers Opportunity for Adjudicatory Hearing: The Atomic Safety and Licensing Board (ASLB) has decided partially in favor of Friends of the Earth that petitioned for a hearing on the NRC’s Confirmatory Action Letter process regarding steam generator issues at the San Onofre nuclear power plant in California. The ASLB is a three-member board of administrative judges independent of the NRC staff that conducts adjudicatory hearings on major agency licensing actions. The board’s decision concludes that this particular Confirmatory Action Letter process, in which San Onofre seeks to restart Unit 2, is effectively a license amendment proceeding. Therefore the Atomic Energy Act and NRC rules give the public the opportunity for an adjudicatory hearing. The Board’s decision provided the public interest group, Friends of the Earth, with the relief it requested – namely, the opportunity for a hearing on the license amendment. Accordingly, the Board’s decision terminates the proceeding at the Board level. The Board also offered reasons why this decision applies only to the unusual facts in the San Onofre process and not to the whole category of Confirmatory Action Letters.

    Public Reaction to ASLB Ruling: Damon Moglen of Friends of the Earth called the ruling “a complete rejection of Edison’s plan to restart its damaged nuclear reactors without public review or input.” An SCE spokeswoman said the utility was still reviewing the ruling and declined to comment. Edison’s Chief Executive Ted Craver has said the utility may decide by year end to retire one or both San Onofre reactors if its restart request is denied, citing uncertainty over NRC timing and SCE’s ability to recover costs related to the extended outage. The reactor can only restart if the NRC concludes it can operate safely. Pressure has been growing on the NRC and the utility to agree to a full review of safety issues at San Onofre from elected officials and anti-nuclear groups. The board concluded that SCE’s restart plan, known as the Confirmatory Action Letter process, is effectively a license amendment proceeding that gives the public the right to a hearing with testimony and cross-examination of witnesses.

    CPUC News: Two California Public Utility Commission Judges have banned the media and the public from videotaping the hearings on the broken San Onofre nuclear plant run by SCE. The chair of the California Public Utility Commission is the former CEO of SCE and has taken favors from non-profit corporations funded by the SCE. Governor Brown who represents the utility industry has kept this questionable chair in his position of regulating the utilities in California. One of the judges Melanie Darling literally went out of control at the last hearing and tore down a banner after the hearing was adjourned. Maybe she does not like seeing herself in action so shutdown the cameras. Public Groups are requesting that the Commission provide a good-quality webcast of the entire week of evidentiary hearings currently scheduled for May 13-17, 2013. California Public Utilities Commission is strongly advised to allow citizens to videotape the hearings pursuant to the Bagley-Keene Act, in order to maximize transparency in this case and provide public access, especially for affected people, who live near the San Onofre nuclear plant, 450 miles away from the Commission’s courtrooms.

    Background: There are hundreds of operating steam generators in the world, which have prevented in-plane fluid elastic instability by keeping the void fractions below 98.5% (Ref. AREVA Operational Assessment data for 5 steam generators, NUREG-1841, NRC Approved Power Uprate Applications, etc.) by operating at steam pressures above 900 psi and steam generator circulation ratios above 4. MHI Root Causes states, “SCE/MHI AVB Design Team recognized that the design for the SONGS RSGs resulted in higher steam quality (void fraction) than previous designs and had considered making changes to the design to reduce the void fraction (e.g., using a larger downcomer, using larger flow slot design for the tube support plates, and even removing a TSP).” So, we assume, that Edison Engineers must have foreseen the impact the problem of high void fractions on increased tube vibrations and refused to make the changes, because it could have impeded the ability to justify the RSG design under the provisions of 10 C.F.R. §50.59, delayed the construction schedule, increased the costs and reduced the profit margins. Increasing the circulation ratios meant reducing the void fractions by increasing the steam pressures, reducing pressure losses, reducing moisture content and less thermal output from the generator. High void fractions cause higher tube vibrations, fluid elastic instability and tube-to-tube wear. MHI/SCE AVB Team missed the boat on Academic Research Papers (2003 through 2006), NUREG-1841 Industry Bench Marking (World’s largest CE replacement steam generators installed in 2002 and partly owned by SCE) and ignored the well-established elementary principles of physics, SG tube vibrations, nucleate boiling, heat transfer, void fractions and circulation ratios by refusing to lower the RSG void fractions. The Original Combustion Engineering Steam Generators operated at 900 psi and a void fraction of 96.1%. That is why these steam generators did not suffer fluid elastic instability in 28 years of operation. Increasing the heat transfer area by 11%, addition of 377 new tubes (4% heat transfer area), the average length of heated tubes by 50 inches (Equivalent addition of 650 tubes or 7% heat transfer area), the steam generator thermal output by 24 MWt to make more profits and refusal to reduce the void fractions was a joint decision, which we assume, was known by members of the MHI/SCE AVB Team and SCE Management, which included the Edison Engineers.

    Edison Steam Generator Expert states, “The contract for design, fabrication and delivery of the RSGs was awarded to Mitsubishi Heavy Industries Ltd. (MHI). As specified, the RSGs were supposed to be a replacement in-kind for the OSGs in terms of form, fit and function. At the same time, however, the RSG specification included many new requirements derived from both industry and SONGS operating experience, and the requirement to use the best and most suitable materials of construction. These requirements were aimed at improving the RSG longevity, reliability, performance and maintainability. Also, the specification called for very tight fabrication tolerances of the components and sub-assemblies, especially the tubesheet and the tube U-bend support structure. In addition, SONGS steam generators are one of the largest in the industry, which called for innovative design solutions and improved fabrication processes when working on the RSGs. Conceivably, the MHI and Edison project teams faced many tough challenges throughout the entire project in the design, manufacturing and QC areas, when striving to meet the specification requirements. Both teams jointly tackled all these challenges in an effective and timely manner. At the end, MHI delivered the RSGs, which incorporated all the latest improvements found throughout the industry, as well as innovative solutions specific to the SONGS RSGs. In Unit 2, the RSGs were installed and tested in 2009/10 and in Unit 3 in 2010/11. The RSG post-installation test results met or exceeded the test acceptance criteria for all specified test parameters, thus properly rewarding the effort put into their fabrication.”

    A. Review of SONGS Replacement Steam Generators 10CFR50.59 Evaluation
    SCE states, “Having the OSGs replaced with the RSGs will improve efficiency and reliability of Units 2 & 3 by replacing a large number of plugged or otherwise degraded heat transfer tubes in each OSG with new tubes made from thermally-treated Alloy 690, which is less susceptible to degradation than the mill-annealed Alloy 600 material used for OSG heat transfer tubing. Replacement of the steam generators is a replacement in-kind in terms of an overall fit, form and function with no, or minimal, permanent modifications to the plant systems, structures or components (SSCs). Each RSG is designed to produce 7.588E6 lb/hr (vs. 7.414E6 lb/hr for OSGs) of 833 psia (vs. 900 psia for OSGs) saturated steam with void fraction of 99.6% (vs. 96.1% for OSGs) moisture content when supplied with feedwater at 442oF.
    A.1 – The major physical differences between the RSGs and OSGs are as follows:
    1. The RSGs have a greater number of tubes (9,727 vs. 9,350) and a larger heat transfer surface area than the OSGs (116,100 ft2 vs. ~ 105,000 ft2). The average length of the heated RSG tube is approximately 50 inches more than the average length of the heated OSG tube.
    2. The RSG reactor coolant volume is greater than the OSG volume (2003 ft3 vs. 1895 ft3).
    3.The RSG tube wall thickness is less than the wall thickness of the OSG tubes (0.0429 in. vs. 0.048 in.).
    4. The RSG tubes are Alloy 690 (thermally-treated) while the OSG tubes are Alloy 600 (mill-annealed).
    5. The RSG feedwater ring is fabricated from erosion-corrosion resistant Cr-Mo alloy steel with Alloy 690 TT fittings, whereas the OSG feedwater ring is made of carbon steel (with the exception of the flow distribution box).
    6. All RSG tubes are U-bend shape, whereas the OSG tubes have both U-bend shape (inner rows of the tube bundle) and square-bend shape (outer rows of the tube bundle).
    7. The RSG channel head has a flat bottom, thicker divider plate, as compared to the OSGs, and no stay cylinder.
    8. The RSG tube supports consist of 7 broached tube support plates in the straight-leg region and anti-vibration bars in the U-bend region, while the OSG tube supports consist of the egg-crate type supports in the straight-leg region and batwings and vertical strips in the U-bend region.

    A.2 – Design Function(s) and/or Method(s) of Evaluation: The design functions of steam generators are to:
    1. Function as a part of the reactor coolant pressure boundary (RCPB).
    2. Transfer heat between the RCS and main steam system.
    3. Remove heat from the RCS to achieve and maintain safe shutdown following design basis accidents (except for a large break LOCA) and other UFSAR-described events.

    A.3 – The design functions of the steam generator tubes and tube supports are to:
    1. Limit tube flow-induced vibration and reactor coolant pump-induced vibration to acceptable levels during normal operating conditions.
    2. Withstand blowdown forces from severance of a steam nozzle and ensure that ASME Code allowable stress limits are met.
    3. Maintain acceptable ASME Code stress levels under design basis accident conditions (i.e., to prevent a tube rupture concurrent with other accidents, and to prevent multiple tube ruptures during a postulated single steam generator tube rupture event), and
    4. Function as a part of the RCPB.

    A.4 – State if the proposed activity:
    1. Changes an SSC in a manner that adversely affects the UFSAR/DSAR design function(s) or has an adverse affect on the method of performing or controlling UFSAR/DSAR design function(s).

    Yes. After the Unit 3 Leak, it is clear that the RSGs were designed and fabricated poorly compared with the OSGs. RSGs were not OSGs replacement in-kind in terms of design functions. OSGs lasted for 28 years and RSGs were destroyed in less than 2 years. Let us now examine the other differences between Unit 2 and Unit 3’s Operational Factors, which were significant contributors to the “fluid-elastic instability” in SONGS Unit 3 and the tube-to-tube wear resulting in the tube leak.

    A.4.1 – Adverse Design/Operational Factors responsible for Fluid Elastic Instability: Low steam generator pressures (SONGS RSGs range 800-850 psi, the primary cause of the onset of severe vibrations) caused high dry steam and high fluid velocities conducive for fluid elastic instability and flow-induced vibrations, whereby U-tube bundle tubes started vibrating with very large amplitudes in the in-plane directions. Extremely hot and vibrating tubes need a little amount of water (aka damping, 1.5% water, steam-water mixture vapor fraction 98.5%). When the void fractions exceed 98.5% and are in the range of 99.5-100%, the extremely hot and vibrating tubes cannot dissipate their energy and return to their original in-plane design position. In effect, one unstable tube drives its neighbor to instability through repeated violent and turbulent impact events which causes tube leakage, tube failures at MSLB test conditions and or unprecedented tube-tube wear, Tube-to-AVB/Tube Support Plates wear, as we saw in SONGS Unit 3. So in review, due to narrow tube pitch to tube diameter, low tube frequency, low tube clearances, in certain portions of the RSGs U-tubes bundle, fluid velocities exceeded the critical velocities due to extremely high steam flows (100% SONGS power conditions outside the industry NORM). These high fluid velocities cause U-tubes to vibrate with very large amplitudes in the in-plane direction and literally hit other the tubes with repeated and violent impacts. Due to lower steam operating pressures (required to generate more heat, electricity and profits) and excessive pressure drops due to high flows and velocities, steam saturation temperature drops. This lowering of steam temperature combined with high heat flux in the hot leg side of the U-tube bundle causes steam dry-outs to form (Vapor fraction >99%), known as “NO Effective Thin Tube Film Damping.” Thin film damping refers to the tendency of the steam inside the generators to create a thin film of water between the RSG tubes and the support structures. That film is enough to help keep the tubes from vibrating with large amplitudes, hitting other tubes violently, and protect the Anti-Vibration Bar support structures and maintain the tube-to-AVB gaps and contact forces. These adverse conditions in SONGS at 70% power operation (RTP) with the present defective design and degraded of RSGs known as fluid elastic instability (Tube-to-Tube Wear, or TTW) can lead to rapid U-tubes failure from fatigue or tube-to-tube wear in Unit 2 due to a main steam line break as seen in SONGS Unit 3 RSG’s. In summary, FEI is a phenomenon where due to SONGS RSGs design intended for high steam flows causes the tubes to vibrate with increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit (critical velocity) for a given tube and its supporting conditions and a given thermal hydraulic environment. This occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports. The lack of Nucleate boiling on the tube surface or absence of water is found to have a destabilizing effect on fluid-elastic stability.
    A.4.2 – Unit 2 FEI Conflicting Operational Data
    NRC AIT Report SG Secondary U2/3 Pressure Range 833 – 942 psi
    SCE RCE SG Secondary U2/3 Pressure – 833 psi
    SONGS Unit 3 RCE Team Anonymous Member – Unit 2 SG Secondary Pressure 863 – 942 psi
    SONGS SG System Description Unit 2 SG Pressure Range 892 – 942 psi
    Westinghouse OA SG Secondary U2/3 Pressure ~ 838 psi
    SONGS Plant Daily Briefing Unit 3 Electrical Generation – 1186 MWe
    SONGS Plant Daily Briefing Unit 2 Electrical Generation – 1183 MWe
    A.4.3 -Unit 2 FEI Conclusions
    A.4.3.1 – NRC AIT Report – Operational Differences between U2/3 – The result of the independent NRC thermal-hydraulic analysis indicated that differences in the actual operation between units and/or individual steam generators had an insignificant impact on the results and in fact, the team did not identify any changes in steam velocities or void fractions that could attribute to the differences in tube wear between the units or steam generators.
    A.4.3.2 – SCE Unit 2 Restart Report Enclosure 2 Conclusions – Because of the similarities in design between the Unit 2 and 3 RSGs, it was concluded that FEI in the in-plane direction was also the cause of the TTW in Unit 2.
    A.4.3.3 – SCE U2/3 FEI SONGS RCE Team Member Conclusions – FEI did not occur in Unit 2
    A.4.3.4 – Westinghouse OA Conclusions: (a) An evaluation of the tube-to-tube wear reported in two tubes in SG 2E089 showed that, most likely, the wear did not result from in-plane vibration of the tubes since all available eddy current data clearly support the analytical results that in-plane vibration could not have occurred in these tubes, and (b) Operational data – ATHOS Model shows no differences in Units 2 & 3
    A.4.3.5 – AREVA OA Conclusions – Based on the extremely comprehensive evaluation of both Units, supplemented by thermal hydraulic and FIV analysis, assuming, a priori, that TTW via in-plane fluid-elastic instability cannot develop in Unit 2 would be inappropriate.
    A.4.3.6 – SONGS Insider Investigator Unit 2 FEI Conclusions – Due to higher SG pressure (Range 863 – 942 psi) and lower thermal megawatts compared to Unit 3, FEI did not occur in Unit 2. This is consistent with the position of RCE Team Anonymous Member. NRC AIT Report, SCE, Westinghouse and AREVA conclusions on Unit 2 FEI are inconsistent, confusing and inconclusive.
    A.4.4 – Possible RSG Degradation Causes:
    1. MHI did not benchmark the computer codes for CE steam generators or used 100% mock up for SONGS High Steam Flows and SCE did not check their work.
    2. SONGS Certified Design Specification did not specify the value of FEI or SR and MHI did not design the RSGs for in-plane vibrations.
    3. SONGS Certified Design Specification implicitly implied MHI to avoid the NRC License Amendment Process and make the tube bundle as tall as possible to achieve the maximum heat transfer area.
    4. SCE or MHI did not review NUREG-1841 to see how Westinghouse and BWI were designing CE Replacement Generators AVBs to avoid excessive tube vibrations and areas with high dry steam.
    5. SCE/MHI did not review the research papers published in 2003 by Pakistanis Researchers and by Dr.Pettigrew and Dr. Mureithi published in 2006, which states “In nuclear power plant steam generators, U-tubes are very susceptible to undergo fluid elastic instability because of the high velocity of the two-phase mixture flow in the U-tube region and also because of their low natural frequencies in their out of plane modes. In nuclear power plant steam generator design, flat bar supports have been introduced in order to restrain vibrations of the U-tubes in the out of plane direction. Since those supports are not as effective in restraining the in-plane vibrations of the tubes, there is a clear need to verify if fluid elastic instability can occur for a cluster of cylinders preferentially flexible in the flow direction. Almost all the available data about fluid elastic instability of heat exchanger tube bundles concerns tubes that are axisymmetrically flexible. In those cases, the instability is found to be mostly in the direction transverse to the flow. Thus, the direction parallel to the flow has raised less concern in terms of bundle stability.”
    6. Westinghouse OA ATHOS Analysis shows Unit 2 had 99.6% vapor fraction (FEI) and fluid velocities of 28 feet/sec, but based on results of ECT inspection, Westinghouse concludes that unit 2 did not experience FEI. Westinghouse also states, “Test data shows that the onset of in-plane (IP) vibration requires much higher velocities than the onset of out-of-plane (OP) fluid-elastic excitation. Hence, a tube that may vibrate in-plane (IP) would definitely be unstable OP. A small AVB gap (3 Mil) that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion. Manufacturing Considerations: None were extensively treated in the SCE root cause evaluation.”
    7. AREVA states, “At 100% power, the thermal-hydraulic conditions in the U-bend region of the SONGS replacement steam generators exceeded the past successful operational envelope for U-bend nuclear steam generators based on presently available data. The primary source of tube-to-AVB contact forces is the restraint provided by the retaining bars and bridges, reacting against the component dimensional dispersion of the tubes and AVBs. Contact forces are available for both cold and hot conditions. Contact forces significantly increase at normal operating temperature and pressure due to diametric expansion of the tubes and thermal growth of the AVBs. After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location.”
    8. Average heated length of the tubes is too much (730 inches in RSGs versus 680 inches in OSGs). Unit 3 has historically produced more power than Unit 2 (1186 MWe vs. 1183 MWe, 1178 MWe vs. 1172). Westinghouse states, “In the U-bend region, the gap velocities are a strong function of power level. The steam flow in the bundle is cumulative and increases as a function of the power level and the bundle height which causes high fluid quality, void fraction, and secondary fluid velocities in the upper bundle.”
    9. RSGs were operating at a circulation ratio of 3.3. Most of The CE RSGs are running at a circulation ratio of 5.0 or more.
    A.4.5 – Defects or Deviations:
    The design of San Onofre Replacement Steam generators (RSGs) are identical (Neglecting the impact of Units 3 and Unit 2, Tube-to-AVB contact forces due to manufacturing errors – See Item A.4.6 below). As shown below, SONGS Unit 2 potentially did not suffer in-plane fluid elastic instability due to operation at higher steam pressures and lower RCS flows. SONGS Unit 3 suffered in-plane fluid elastic instability due to operation at lower steam pressures and higher RCS flows. This conclusion is consistent with Westinghouse Operational Assessment, but challenges the SCE, NRC AIT, AREVA and MHI conclusions. NRC AIT Report, SCE, MHI and AREVA conclusions on Unit 3 and Unit 2 FEI are incomplete, inconsistent, confusing and inconclusive and based on faulty computer simulations and hideous testing data (Shielded under the false pretense of Proprietary information). The analysis in these reports does not meet the intent of NRC CAL ACTION 1, which states “Southern California Edison Company (SCE) will determine the causes of the tube-to-tube interactions that resulted in steam generator tube wear in Unit 3, and will implement actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes. SCE will establish a protocol of inspections and/or operational limits for Unit 2, including plans for a mid-cycle shutdown for further inspections.”
    Repeated requests to NRC AIT Leader, NRC SONGS Special Panel and NRC Region IV Allegation Coordinator to examine carefully the operational difference between Units 2 & 3 and determine its impact on the tube-to-tube interactions that resulted in steam generator tube wear in Unit 3, and actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes have not been addressed to date. NRR has not asked SCE in its RAI(s) the impact of operational differences between Units 2 and 3 on Unit 2 and Unit 3 tube-to-tube wear. Honorable NRC Commissioner Mr. Apostolakis was totally confused on Unit 2 FEI inconsistent statements by SCE, Westinghouse and AREVA. The Author tried to tell this information to SCE and MHI Management in June 2012, but of no avail (See copy of attached Emails and SG Nuclear Notifications).
    A.4.6 Contact Force Differences between SONGS Units 2 and 3: NRC AIT, SCE and MHI state that supports were better in Unit 2, so no tube-to-tube wear occurred in Unit 2. Fabrication differences during manufacture of SONGS RSGs caused difference of contact forces in supports between Units 2 & 3. Let us now examine that whether insufficient contact tube-to AVB forces in the Unit 3 upper tube bundle caused “fluid-elastic instability” which was a significant contributor to the tube-to-tube wear resulting in the tube leak.
    A.4.6.1 – MHI states, “By design, U-bend support in the in-plane direction was not provided for the SONGS SG’s”. In the design stage, MHI considered that the tube U-bend support in the out-of-plane direction designed for “zero” tube-to-AVB gap in hot condition was sufficient to prevent the tube from becoming fluid-elastic unstable during operation based on the MHI experiences and contemporary practice. MHI postulated that a “zero” gap in the hot condition does not necessarily ensure that the support is active and that contact force between the tube and the AVB is required for the support to be considered active. The most likely cause of the observed tube-to-tube wear is multiple consecutive AVB supports becoming inactive during operation. This is attributed to redistribution of the tube-to-AVB-gaps under the fluid hydrodynamic pressure exerted on the tubes during operation. This phenomenon is called by MHI, “tube bundle flowering” and is postulated to result in a spreading of the tube U-bends in the out-of-plane direction to varying degrees based on their location in the tube bundle (the hydrodynamic pressure varies within the U bend). This tube U-bend spreading causes an increase of the tube-to-AVB gap sizes and decrease of tube-to-AVB contact forces rendering the AVB supports inactive and potentially significantly contributing to tube FEI. Observations Common to BOTH Unit-2 and Unit-3: The AVBs, end caps, and retainer bars were manufactured according to the design. It was confirmed that there were no significant gaps between the AVBs and tubes, which might have contributed to excessive tube vibration because the AVBs appear to be virtually in contact with tubes. MHI states, “The higher than typical void fraction is a result of a very large and tightly packed tube bundle, particularly in the U-bend, with high heat flux in the hot leg side. Because this high void fraction is a potentially major cause of the tube FEI, and consequently unexpected tube wear (as it affects both the flow velocity and the damping factors).”

    A.4.6.2 – AREVA states – “The primary source of tube-to-AVB contact forces is the restraint provided by the retaining bars and bridges, reacting against the component dimensional dispersion of the tubes and AVBs. Contact forces are available for both cold and hot conditions. Contact forces significantly increase at normal operating temperature and pressure due to diametric expansion of the tubes and thermal growth of the AVBs. After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location.”

    A.4.6.3 – Westinghouse states, “Test data shows that the onset of in-plane (IP) vibration requires much higher velocities than the onset of out-of-plane (OP) fluid-elastic excitation. Hence, a tube that may vibrate in-plane (IP) would definitely be unstable OP. A small AVB gap that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion. Manufacturing Considerations: There are several potential manufacturing considerations associated with review of the design drawings based on Westinghouse experience. The first two are related to increased proximity potential that is likely associated with the ECT evidence for proximity. Two others are associated with the AVB configuration and the additional orthogonal support structure that can interact with the first two during manufacturing. Another relates to AVB fabrication tolerances. These potential issues include: (1) The smaller nominal in-plane spacing between large radius U-bend tubes than comparable Westinghouse experience, (2) The much larger relative shrinkage of two sides (cold leg and hot leg) of each tube that can occur within the tubesheet drilling tolerances. Differences in axial shrinkage of tube legs can change the shape of the U-bends and reduce in-plane clearances between tubes from what was installed prior to hydraulic expansion, (3) The potential for the ends of the lateral sets of AVBs (designated as side narrow and side wide on the Design Anti-Vibration Bar Assembly Drawing that are attached to the AVB support structure on the sides of the tube bundle to become displaced from their intended positions during lower shell assembly rotation, (4) The potential for the 13 orthogonal bridge structure segments that are welded to the ends of AVB end cap extensions to produce reactions inside the bundle due to weld shrinkage and added weight during bundle rotation, and (5) Control of AVB fabrication tolerances sufficient to avoid undesirable interactions within the bundle. If AVBs are not flat with no twist in the unrestrained state they can tend to spread tube columns and introduce unexpected gaps greater than nominal inside the bundle away from the fixed weld spacing. The weight of the additional support structure after installation could accentuate any of the above potential issues. There is insufficient evidence to conclude that any of the listed potential issues are directly responsible for the unexpected tube wear, but these issues could all lead to unexpected tube/AVB fit-up conditions that would support the amplitude limited fluid-elastic vibration mechanism. None were extensively treated in the SCE root cause evaluation.”

    A.4.6.4 – John Large States, “Causes of Tube and Restraint Component Motion and Wear: My study of the various OAs leads me to the following findings and opinion that; (i) degradation of the tube restraint localities (RBs, AVBs and TSPs) occurs in the absence of fluid elastic instability (FEI) activity; (ii) TTW, acknowledged to arise from in-plane FEI activity, generally occurs where the AVB restraint has deteriorated at one or more localities along the length of individual tubes; (iii) the number of tube wear sites or incidences for AVB/TSP locations outstrips the TTW wear site incidences in the tube free-span locations. I find that the ‘zero-gap’ AVB assembly, which features strongly in the onset of TTW, is clearly designed to cope only with out-of-plane tube motion since there is little designed-in resistance to movement in the in-plane direction – because of this, it is just chance (a combination of manufacturing variations, expansion and pressurization, etc) that determines the in-plane effectiveness of the AVB; (iv) Uniquely, the SONGS RSG fluid regimes are characterized by in-plane activity, which is quite contrary to experience of other SGs used in similar nuclear power plants in which out-of-plane fluid phenomena dominate. Moreover, from the remote probe inspections when the replacement steam generator (RSG) is cold and unpressurized, I consider it impossible to reliably predict the effectiveness of the many thousands of AVB contact points for when the tube bundle is in a hot, pressurized operational state., and (5) v) The combination of the omission of the in-plane AVB restraints, the unique in-plane activity levels of the SONGS RSGs, together the very demanding interpretation of the remote probe data from the cold and depressurized tube inspection, render forecasting the wear of the tubes and many thousands of restraint components when in hot and pressurized service very challenging indeed. John Large continues, “Phasing of AVB-TSP Wear -v- TTW: I reason that, overall, the tube wear process comprises two distinct phases: First, the AVB (and TSP) -to-tube contact points wear with the result that whatever level of effectiveness is in play declines. Then, with the U-bend free-span sections increased by loss of intermediate AVB restraint(s), the individual tubes in the U-bend region are rendered very susceptible to FEI induced motion and TTW. Whereas the OAs commissioned by SCE broadly agree that the wear mechanics comprises two phases, there are strong differences over the cause of the first phase comprising in-plane AVB wear: AREVA claim this is caused by in-plane FEI whereas, the contrary, Mitsubishi (and Westinghouse) favor random perturbations in the fluid flow regime to be the tube motion excitation cause. Put simply: (i) if AREVA is correct then reducing the reactor power to 70% will eliminate FEI, AVB effectiveness will cease to decline further and TTW will be arrested; however, to the contrary, (ii) if Mitsubishi is right then, even at the 70% power level, the AVB restraint effectiveness will continue to decline thereby freeing up longer free-span tube sections that are more susceptible to TTW; or that
    (iii) the assertion of neither party is wholly or partly correct. As I have previously stated, I consider that AVB-to-tube wear is not wholly dependent upon FEI activity.
    A.4.6.5 – Violette R., Pettigrew M. J. & Mureithi N. W. state (Ref. 1 – See below), “In nuclear power plant steam generators, U-tubes are very susceptible to undergo fluid elastic instability because of the high velocity of the two-phase mixture flow in the U-tube region and also because of their low natural frequencies in their out of plane modes. In nuclear power plant steam generator design, flat bar supports have been introduced in order to restrain vibrations of the U-tubes in the out of plane direction. Since those supports are not as effective in restraining the in-plane vibrations of the tubes, there is a clear need to verify if fluid elastic instability can occur for a cluster of cylinders preferentially flexible in the flow direction. Almost all the available data about fluid elastic instability of heat exchanger tube bundles concerns tubes that are axisymmetrically flexible. In those cases, the instability is found to be mostly in the direction transverse to the flow. Thus, the direction parallel to the flow has raised less concern in terms of bundle stability.” Reference 1: Fluid-elastic instability of an array of tubes preferentially flexible in the flow direction subjected to two-phase cross flow, Violette R., Pettigrew M. J. & Mureithi N. W., 2006, http://yakari.polytechnique.fr/people/revio/masters_research_subject.html

    A.4.6.6 – Dr. Pettigrew (Presentation to NRC Commission, February 2013): So, you notice the U-bend — the plane of the U-bend is being installed, and on top of the U-bends are bars. They are anti-vibration bars. And so you can see here that from the point of view of out-of-plane motion, the tubes are really very well supported because you have a large number of bars all around; but from the point of view of in-plane motion, there’s really no positive restraint here to prevent the tube to move in the in-plane direction. Essentially, it relies on friction forces to limit the vibration.
    A.4.6.7 – Contact Force Definition: Contact force is the force in which an object comes in contact with another object. Some everyday examples where contact forces are at work are pushing a car up a hill, kicking a ball, or pushing a desk across a room. In the first and third cases the force is continuously applied, while in the second case the force is delivered in a short impulse. The most common instances of contact force include friction, normal force, and tension. Contact force may also be described as the push experienced when two objects are pressed together. The MHI-designed AVBs had zero contact forces in Unit 3 to prevent in-plane fluid elastic instability and subsequently, wear occurred under localized thermal-hydraulic conditions of high steam quality (void fraction) and high flow velocity. Large u-bends were moving with large amplitudes in the in-plane direction without any contact forces imposed by the out-of-plane restraints. The in-plane vibration associated with the wear observed in the Unit 3 RSGs occurred because all of the out-of-plane AVB supports were inactive by design in the in-plane direction. The Unit 3 tube-to-AVB contact force for the tubes with tube-to-tube wear (TTW) was zero. That is why they did not restrain the tubes in the in-plane direction (like a sports car moving with very high speed in freeway express lanes passing by a stalled police car with empty guns and disabled communication systems).

    A.4.6.8 – Contact Force Conclusions: SONGS Unit 3 RSG’s were operating outside SONGS Technical Specification Limits for Reactor Thermal Power and Current Licensing Basis for Design Basis Accident Conditions. I agree with MHI that high steam flows and cross-flow velocities combined with narrow tube pitch-to-diameter ratio caused elastic deformation of the U-tube bundle from the beginning of the Unit 3 cycle, which initiated the process of tube-to-AVB wear and insufficient contact forces between tubes and AVBs. Tube bundle distortion is considered a major contributing cause to the mechanism of tube-to-tube/AVB/TSP wear seen in the Unit 3 SG’s. After 11 months of wear, contact forces were virtually eliminated between the tube and AVBs in the areas of highest area of Unit 3 wear as confirmed by ECT and visual inspections. I conclude that FEI and MHI Flowering effect redistributed the tube-to-AVB gaps in Unit 3 RSG’s. FEI did not occur in Unit 2, because of the absence of high steam dryness and NOT the better supports and/or differences in fabrication, which resulted in substantially increased contact forces (reduced looseness) between tubes and AVBs for Unit 2 and prevented FEI from occurring. My findings on Unit 2 FEI are consistent with the findings of AREVA, Westinghouse, John Large, SONGS RCE Anonymous Root Cause Team Member and latest research performed by Eminent Professor Michel Pettigrew and others in 2006. In-plane fluid elastic instability did not happen in Unit 2 because of operational differences, so therefore double contact forces and better supports is just conjecture in Unit 2 to justify the restart of an Unsafe Unit 2.

    A.4.7 – Dings and Dents Conclusions: A analysis performed by AREVA shows that there are more dents and dings in SG 2E-089 (Unit 2) compared to SG 3E-089 (Unit 3) by a factor of about 13. Overall, analysis found that nearly 12,000 contact indications were found in both Unit 2 steam generators as opposed to just under 4,100 contact indications in both Unit 3 steam generators. Even more alarming is that fact that these indications in Unit 2 were primarily found distributed very distinctly across entire rows of steam generator tubes, much more so than Unit 3. This testing is performed by measuring signals between supports and tubes inside of the steam generators. When they are in contact together a signal will be registered and based on the strength of the contact one can correlate the size and impact of the indications on the tubes. What these results infer is that there is a large discrepancy between the amount of tubes out of place and touching the supports in the Unit 2 and Unit 3 steam generators. Considering the fact that Southern California Edison has repeatedly stated that steam generators are of like design and that no evidence or data has been provided which showed any design deviation in this regard between the two units, it is likely that this accelerated wear seen in Unit 2 occurred within the last cycle of operation. Simply this means that for every one indication found in Unit 3 steam generators, three indications were found in Unit 2 steam generators. Though the Unit 3 steam generators failed more catastrophically, it appears from this analysis that there is a much larger pool of tubes out of alignment and in direct contact with support plates in Unit 2. During any operation, it is presumed that there will be some vibration and movement of all of the tubes in steam generators, but this is offset by supports and spacing between tubes. However in this case, nearly 12,000 tubes in Unit 2 are already in contact with supports, meaning that with any vibration or movement more contact and ergo contact indications will occur in the tubes regardless of operational power rates.
    A.4.8 – RSG 10CFR50.59 Conclusions: The values of the RSG major design parameters are different than the values of the corresponding OSG parameters. The RSG steam flow is slightly higher, the outlet steam pressure is lower and the moisture content is considerably lower than the values for the OSGs. These changes are in a non-conservative direction (increased void fractions) and constitute a significant reduction in margin of safety and increase in probability of cascading tube ruptures over the OSGs.

    The RSG heat transfer area is larger than the OSG area (116,100 ft2 vs. ~105,000 ft2) and the RSG tube bundle is taller than the OSG bundle. The larger and taller RSG tube bundle along with unauthorized and untested design changes provided the mechanism for increased void fractions, fluid velocities, fluid elastic instability, flow-induced random vibrations, high cycle thermal fatigue and Mitsubishi Flowering Effect. These factors indicate that the RSGs performed worse than, the OSGs during the events that credit natural circulation. The RSG primary side volume is larger than the OSG volume (2003 ft3 vs. 1895 ft3). Due to this increase, more radioactivity will be released to the environment during multiple tube ruptures caused by anticipated operational occurrences and main steam line break. The RCS volume increase will also result in a slight increase of the containment flooding level, following a LOCA.
    The RSG tube wall is thinner than the OSG tube wall (0.0429 in. vs. 0.048 in.). The analysis concluded that a tube would have to be plugged if it contained a flaw to a depth lesser than that for the OSGs (35% vs. 44%). This reduction of the tube plugging limit is non-conservative because hundreds of SONGS Unit 2 & 3 RSGs exceeded this limit and were operating beyond their license.

    Based on the above, it is concluded that, the proposed activity significantly and adversely affects the steam generator ability to:
    (1)Function as a part of the RCPB
    (2)Transfer heat between RCS and main steam system and
    (3)Remove heat from the RCS to achieve and maintain safe shutdown following postulated accidents (other than the large break LOCA).

    Therefore, it is concluded, that the replacement adversely affected the ability of performing or controlling these design functions. Based on the above, changing the OSGs to RSGs changed an SSC in a manner that adversely affected UFSAR-described design functions or that had an adverse effect on the method of performing or controlling UFSAR-described design functions.

    B. SONGS Replacement Steam Generators 10CFR50.92 Evaluation
    B.1 Condition of Unit 2 steam Generators: SONGS Unit 2 & 3 RSGs are of the same design. Therefore, the description of unit 3 provided below is also applicable to Unit 2. SONGS Unit 3 RSGs’ unprecedented tube failure and massive tube and AVB/TSP degradation occurred due to fluid elastic instability, flow-induced random vibrations, Mitsubishi Flowering Effect and high cyclic fatigue under the following unique circumstances:
    (1) U-tube bundle areas with high dry steam will experienced double in-plane velocities (> 50 feet/sec, based on review of MHI Root Cause, Dr. Pettigrew and other research papers published between 2006-2011) compared with out-of plane velocities assumed (25 feet/sec) to have been predicted by Outdated Out-of-Plane Westinghouse /NRC /MHI /AREVA ATHOS Computer Models,
    (2) Lack of positive in-plane restraints and zero damping,
    (3) Large number of SONGS Units 2/3 RSG U-bends with tube clearances of only 0.05 inches (Design 0.25 inches, Industry Norm > 0.25 inches),
    (4) Excessive number of tubes with narrow tube pitch to tube diameter,
    (5) Low in-plane frequency tubes and retainer bars compared with MHI SGs’ higher in-plane frequency tubes and retainer bars,
    (6) SONGS’ tubes being much longer than Westinghouse Model 51 steam generators (Average length of heated tube = 730 inches) and other MHI SGs,
    (7) MHI RSGs’ unique floating tube bundle with degraded Retainer Bars can collapse due to 100% tube uncovery for 10 minutes under MSLB SG Depressurization, Multiple SGTR SG over-pressurization and lifting of SG Relief Valves, Combination of MSLB and SGTR Conditions, Release of 100% RCS Iodine to Environment,
    (8) Large amount of uncertainties and unverified assumptions in MHI, AREVA, Westinghouse and Intertek’s contact force (zero for in-plane vibrations), wear rate and tube stress calculations (4.6 ksi versus 16-17 ksi) and computer modeling, and,
    (9) Incomplete tube inspections in SONGS Unit 2. Incubating macroscopic circumferential cracks caused by fluid elastic instability, flow-induced random vibrations and high cycle thermal fatigue are extremely difficult to detect and be accurately sized by nondestructive evaluation techniques including X-ray, ultrasonic, and eddy current based bobbin coil probes, mechanically rotating pancake coil (RPC), etc., which have been used in 170,000 SONGS Tube inspections. State-of-the-art systems: Zetec MIZ-80 iD system, Tecnatom TEDDY+, Circular TE and TM, transmit-receive eddy current array probe C-3 and other specialized radiographic probes capable of detecting sub-surface cracks caused by high cycle thermal fatigue have not been used in the 170,000 SONGS Tube Partial and Limited Inspections as shown below for Unit 2 due to access problems in the most problematic innermost sections of the U-Tube Bundle, the high cost, lack of availability of highly specialized tools and contractors, radiation doses, and time considerations in a rush to start Unit 2. The inspection scope defectively designed and degraded SONGS Unit 2 RSGs should have covered 100% hot leg and cold leg tube inspections, 100% of dents or dings, 100% of tube inspections in the tight radius U-bends, 100% area of the Top of the Tube Sheet and Tube Support Plates.
    B.2 SONGS and Offsite Emergency Plans
    Current SONGS Updated FSAR, Emergency Plans, San Diego County Multi-hazard Regional Emergency Operations Plans, IPC/Orange County & Other Offsite/State of CA Plans and NRC Emergency Rules/Guidance, SONGS Drills and Exercises are based on a slow occurring Steam Generator Tube Leakage/Rupture caused by anticipated operational transients, which are significantly flawed based on the SONGS Unit 2 realistic scenario described below.
    B.3 Main Steam Line Break In Unit 2:
    A potential main steam line break occurs outside Containment in SONGS Unit 2 operating at 70% power. This event causes a simultaneous reactor, turbine, feedwater and reactor coolant trips and MSIVs Close (Conservative assumption for the benefit of SCE). Due to feedwater pump trip and SG U-tube bundle depressurization, the RSG U-bundle secondary water level will shrink and tubes will be uncovered for a period of at least 10 minutes and experience a sharp drop/increase in secondary side pressure. The entire sub-cooled feedwater inventory contained in the faulted RSG will instantaneously flash to high dry steam and over-pressurize the steam generators. Loss of Turbine load will also over-pressurize the steam generator. Main steam safety valves located outside the containment will progressively open to prevent over-pressurizing the steam generators and connect the faulty generators to the environment via open steam safety valves. Now for the next 10-15 minutes, the Control Room is busy in trying to trouble shoot and diagnose the changing plant conditions and flipping through 1000 pages of Emergency and Abnormal operating procedures to determine the correct course of mitigation actions.
    Meanwhile, during the 10 to 15 minutes, the combination of resonant, out-of-plane, in-plane vibrations, jet impingement forces, and RSG debris will cause large axial, bending, dynamic and cyclic loads on all the tubes, tube support plates, retainer bars and anti-vibration structure. The strength of the welded and mechanical connections of these low frequency retainer bars, retaining bars and bridges have not been analyzed for the effects of these cumulative loads to prevent AVB structure displacement, deformation or collapse during loss of offsite power. The displacement, deformation or collapse of AVB structure introduce new and significant axial, bending, dynamic and cyclic loads, which can potentially cause thousands of worn, cracked, plugged and stabilized tubes to exceed their high cycle fatigue stress levels several times than the allowed tube ASME Endurance Limit of 13.6 ksi. If this happens, multiple circumferential tube ruptures will occur at tube-support plates, mid-spans, free spans and tube-to-anti-vibration bar notched interfaces due to macroscopic circumferential cracks caused by tube-to-tube wear and high cycle thermal fatigue. Since all the steam from the RSG would escape to the environment, the iodine-131 from un-partitioned reactor coolant leaking out the rupture tubes will also escape to the environment in less than 10 minutes with 60 tons of radioactive coolant and steam. Consistent with Fukushima Task Force Lessons Learnt and NRC Commissioner Meeting Transcripts, this event will be considered as a beyond design basis event, and SONGS Operators will be unable to take any timely mitigation actions in a radiation/steam environment to stop a severe nuclear accident in progress and notify the Offsite Agencies.
    If the prevailing winds are towards San Clemente, consistent with NRC Inspector General Reports, NRC and Government Studies and observations of SONGS Emergency Plan Drills for the last six years, SCE and Offsite agencies would not have time to respond, notify, evacuate, shelter or give Potassium Iodide to the affected residents within the 10-mile affected emergency planning zone. ODAC, Offsite field monitoring teams, Emergency Vehicles, Helicopters, Orange County Hospitals capabilities will be severely limited or non-functional in a high radiation environment to operate and rescue/transport/shelter disabled, sick, elderly, children, transients and other affected citizens. The casualties, and short, long-term cancer affects to the affected population and ingestion pathway will depend upon the iodine spiking factor and the duration of blowdown, but the offsite releases will significantly exceed the NRC approved SONGS Control Room limit of 5 Rem Total Effective Dose Equivalent (TEDE), and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE.
    NOTE: While this event is occurring, San Diego County, Orange County and Marine Corps Base Camp Pendleton won’t be able to send radiation monitoring teams into areas around the plant due to high radiation levels to locate the plume and take soil and air samples to determine the extent of the release off plant grounds. That offsite field monitoring data, along with the data from the plant wound not be able to sent to the Offsite Dose Assessment Center (ODAC) located in MESA Emergency Operations Facility for making Protective Action Determinations. The offsite plans are recommended to be revised and feasibility demonstrated via an Emergency Plan Drill using Alternate and Parallel Emergency Operation Facilities located in Irvine and San Diego. The Three Mile Island nuclear accident was not as serious as Chernobyl, but was very confusing and chaotic. 40,000 gallons of radioactive waste was released in the Susquehanna River. 140,000 pregnant women and small children were evacuated as a precautionary measure, but cancer risk was not a serious threat.
    If the prevailing winds are towards the Pacific Ocean and San Diego, the Public and SONGS worker casualties will be minimum, and short, long-term cancer affects to the affected human and marine population will depend upon the iodine spiking factor and the duration of blowdown, exceed the NRC approved SONGS Control Room limit of 5 Rem Total Effective Dose Equivalent (TEDE) and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE. The impact on Marine Life and 50 Mile Ingestion Pathway is undetermined.
    B.4 – SCE 50.92 License Amendment
    SCE has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of Amendment”, as discussed below:

    1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
    Response: Yes
    As shown above, the proposed changes affects the probability of multiple SG Tube Ruptures due to a potential main steam line break design basis accident. These changes are in a non-conservative direction (increased void fractions) and constitute a significant reduction in margin of safety and significant increase in probability of cascading tube ruptures over the OSGs. Operation at reduced power is not acceptable under the current licensing basis and operation of the plant will not remain bounded by the assumptions of the analyses of accidents previously evaluated in the UFSAR.
    2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
    Response: Yes, see above
    3. Does the proposed change involve a significant reduction in a margin of safety?
    Response: Yes, see above

  19. HAHN Baba – NRC/SCE/MHI/Independent Experts/Public Awareness Series
    Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog. Please excuse me for my grammatical and computer human performance errors. Ted Craver is more worried about his Investment from Transmission & Distribution system than Public Safety or Repairing San Onofre.

    With San Onoftre Nukes near death, consumers staring at $3B tab, UT-San Deigo, May 4, 2013

    A SCE company spokeswoman said Friday that discussion of fixing the San Onofre plant as “premature.” Yet her CEO is talking openly about permanent shutdown. Here is where it becomes clear that utility executives are not like the rest of us. The incentives that govern regulated monopolies bear no resemblance to those for ordinary businesses.

    Fixing San Onofre and selling the power would bring zero profits to Edison or SDG&E. That’s because utilities simply pass on power costs — with no markup for profits — to customers on our bills, regardless of whether the electricity was purchased under a contract or generated by the utility itself. Instead, utilities make their money based on the cash they invest to buy or build assets, such as power lines, smart meters and power plants. Right now regulators allow SDG&E to bill customers at a rate of 10.3 percent a year of its total assets, and Edison gets 10.45 percent.

    That’s probably 1,000 times more than banks are paying to use your cash in a savings account, by the way. But regulators figure that utility investors need plenty of incentive to build and maintain the power grid. Here’s the rub: Customers are supposed to pay only for assets that help provide electricity.

    Strictly speaking, Edison and SDG&E have no financial interest in selling power from San Onofre. The question for executives is whether regulators will allow them to bill consumers the entire cost of what today is an expensive piece of industrial history on the Camp Pendleton beach.

    That bill is high: $700 million and counting for installing the new steam generators in 2010 and 2011 that broke last year; over $1 billion in unfilled costs for the San Onofre plant itself; more than $1 billion in eventual replacement power costs, if the 2012 spending is any guide.

    Those are just the direct cash costs. In a report set for release this week, state grid managers are expected to predict Southern California may have trouble keeping the lights on this summer.

  20. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog. Please excuse me for my grammatical and computer human performance errors.

    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    In a 2003 paper, Shahab Khushnood, Zaffar M. Khan, M. Afzaal Malik, Zafar Ullah Koreshi, Mahmood Anwar Khan, College of Electrical & Mechanical Engineering, National University of Sciences and Technology, Rawalpindi, state, “There is a strong need for establishing reliable design procedures for two-phase cross-flow tube bundle vibrations. This could be achieved by carrying out modeling and simulation of the system with fluid–structure interaction focusing on void fraction, and reliable experimental data. Test data on high pressure and temperature conditions are insufficient, therefore a potential challenge lies ahead.” Looks like MHI did not do their home work in 2005 and is trying to justifying the adverse safety and public relations consequences for SCE, which did not agree to lower o the void fraction, because it would have delayed the project, triggered a NRC review, increased the costs and cut down the heat output and profits from the RSGs. But, these lame MHI/SCE excuses are not going to work.” MHI needs to be fined by NRC for its ignorance. NRC needs to have SCE license stripped to operate a nuclear power plant and MHI’s license to supply nuclear power plant components for US Nuclear Reactors. SCE has destroyed Unit 1 and Units 2 & 3 twice. It is time for SCE close its shops and decommission San Onofre. It is time for MHI to close its shops to supply nuclear power plant components for US Nuclear Reactors.

  21. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog. Please excuse me for my grammatical and computer human performance errors.

    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    Some Southern Californians demand well-managed, well-maintained, safe, economical, 24/7 reliable nuclear power and grid stability from San Onofre. Others, frustrated with SCE, MHI and NRC refusing to discuss safety issues want to decommission San Onofre. A balance can only be achieved, and the public trust can only be restored by an independent and transparent investigation, conducted by the Office of NRC Inspector General and Senate Committee on Environmental and Public Works. In the end, no matter how long it takes, for SCE to stay in business and earn public trust, San Onofre Leaders have to work very hard and honestly to achieve excellence in Regulatory Compliance, Operational and Public Safety to become an INPO I Plant and make San Onofre work place free of discrimination, retaliation, intimidation and insults to nuclear workers.

  22. HAHN Baba – NRC/SCE/MHI/Independent Experts/Public Awareness Series
    Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog. Please excuse me for my grammatical and computer human performance errors.

    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    Bethann Chambers
    Valley Center, California 92082
    Bethann@vcweb.org
    Governor Edmund G. Brown Jr.
    Office of the Governor
    State Capitol, Suite 1173
    Sacramento, California 95814
    May 2, 2013

    Dear Governor Brown,
    Nearly a year ago I wrote a letter to you voicing my concerns about the San Onofre Nuclear Generating Station (SONGS), and to humbly ask you to use the power of your office to investigate the plant’s ongoing declining performance and equipment conditions. During this past year the facility’s declining performance issues have been lost in the shadow of the much bigger problem with premature tube degradation in the recently installed Replacement Steam Generators, and concerns about the plant’s future operation.

    My husband, James Chambers, is a Nuclear Regulatory Commission (NRC) Licensed Nuclear Reactor Operator for San Onofre Units 2 and 3. He worked at SONGS from 1983 until 2010 when he left his job on medical leave because of work induced health problems. In 2010, my husband came under retaliation by Southern California Edison (SCE) for raising safety concerns and filing allegations of serious violations at the plant with the NRC. When my husband’s medical leave was abruptly terminated, SCE no longer had a job for him; so the company offered him a separation agreement with the stipulation that he not publically disclose any information which might be harmful to SCE or its subsidiaries. SCE is a public utility; the fact that they use their abundant financial resources to actively silence potential critics is a practice which I believe should concern you and all of the members of our state legislature.

    It is my firm belief that there are multiple levels of corruption within SCE, the California Public Utilities Commission, and the Interjurisdictional Planning Committee which needs to be rooted out and exposed. Without complete transparency by our public utilities, and the agencies which oversee them; honest public debate about performance, equipment, and environmental issues pertaining to SONGS cannot occur and places the health and safety of the public in jeopardy. Without honesty and
    transparency, we run the risk of having a significant nuclear event comparable to Chernobyl or Fukushima in southern California. As I am sure you are well aware of, if there were ever a significant radiological release to the atmosphere at SONGS the area surrounding the plant could become uninhabitable for several decades or longer. It is no small task to clean up radioactive contamination from the environment. This is a serious topic which could impact the future of California.

    Currently, SONGS Units 2 and 3 are both shut down because of tube failure and
    premature tube wear in the Replacement Steam Generators, and the situation is being
    investigated by the NRC. Last October, SCE proposed a plan to run Unit 2 at only 70%
    plant power, claiming that this change would eliminate the conditions which led to the
    Steam Generator Tube Rupture in Unit 3, and premature wear in both Units’ Steam
    Generators. However, a reduction in plant power, which is the measurement relating to
    how many megawatts the Turbine Generator produces, can never eliminate the threats
    to a Steam Generator Tube Rupture condition as SCE is claiming it will. SONGS Unit 2 is a
    Pressurized Water Reactor, and the conditions that have caused the failure of the Steam
    Generators are the normal operating pressure of the Reactor, and the flow rate of water
    into the Steam Generators. These are fixed pressures and flows that cannot be changed;
    therefore, any attempt to run the plant will result in exactly the same conditions which
    caused the premature tube wear and tube ruptures in the first year of service of the
    Replacement Steam Generators. Please bear in mind that the new metal alloy that was
    used and the tube failures that resulted from the first year of service was equivalent to
    20-30 years of service in other plants using the original metal alloy. SCE, contrary to the
    NRC’s request for complete transparency, has implied that operating the plant at a
    reduced power level will reduce the threat of further Steam Generator tube ruptures
    and subsequent radioactive release to the environment, when in fact, the threat can
    never be removed because of the design of the plant and the weak alloy which the
    Replacement Steam Generators are constructed of.
    I also believe that SCE’s proposed plan is irresponsible, and shows a serious lack
    of conservative decision making principles. “Conservative decision making” is a nuclear
    fundamental that means the safest decisions should always be made to protect the
    health and safety of the general public, the plant workers, and the environment. If the
    NRC approves SCE’s plan, workers in the Operations Department will be required to
    start up and run the reactor knowing that the Replacement Steam Generators have
    extensive design problems and significant wear which could lead to another tube
    rupture and radioactive release to the environment. As the wife of a reactor operator, I
    lived through many refueling outages and unit start-ups throughout the 1990’s and
    stress that reactor operators experience during normal work conditions. The fact that SCE wants its workers to operate defective equipment shows the flagrant disregard that SCE and SONGS senior management has for the health and safety of nuclear workers at the plant, as well as the people living in the surrounding communities.

    According to the findings of the Institute of Nuclear Power Operations (INPO), San Onofre has been the worst rated nuclear plant in the nation. And according to the World Association of Nuclear Operators (WANO), San Onofre has been the worst or near worst rated nuclear plant in the nation for industrial safety. And according to the NRC, San Onofre has had the longest running cross cutting issues in Human Performance in the history of U. S. nuclear power. “Cross cutting issues in Human Performance” means that in nearly every department significant errors are being made because workers do not follow required procedural steps. The length and breadth of these issues were so egregious it forced the NRC to revise their procedures because San Onofre was actually outside of all postulated conditions set forth in the NRC procedures governing Human Performance failures. This is very condemning evidence which shows that SONGS has been mismanaged for many years. How much more evidence do we need before an adjudicated public hearing is held to investigate the matter?
    The design problems and the conditions which led to the first Steam Generator tube failure in Unit 3 have already been investigated and a root cause analysis has been performed by several industry experts. These analyses confirm that future Steam Generator tube wear and tube ruptures with a resultant radioactive release to the environment are inevitable. Why does SCE need to do a 5 month experimental test run with Unit 2; just to see if the conclusions of the root cause analyses are correct? At what point in time did we decide that doing an experiment with a full scale commercial nuclear reactor was a good idea? What SCE is proposing is unprecedented in the history of U.S. nuclear power. It was an equipment test experiment which led to the nuclear event at Chernobyl in 1986. Didn’t we learn anything from that tragedy?
    Public distrust, and nuclear worker distrust of SCE’s management of SONGS has been growing significantly over the past year, and I believe it has reached a boiling point. The electric ratepayers of southern California do not believe they should be held financially responsible for SCE’s engineering mistakes in the design of the Replacement Steam Generators, or for the cost of running a shutdown nuclear facility which has not produced a single megawatt in over a year. As Governor of the state, I believe you have a responsibility to take action on this matter and not leave it to the sole discretion of the Nuclear Regulatory Commission to decide the fate and future of the San Onofre Nuclear Generating Station. The NRC’s biggest flaw is that they are neither omnipresent nor omniscient; and they cannot regulate the nuclear power industry as everyone assumes they do. In 2010, when my husband filed serious allegations regarding blatant procedural violations and retaliation against himself for raising safety concerns at the plant, he conveyed to me his experience that after all was said and done the NRC would never actually do anything; and nothing has changed in three years.
    I would greatly appreciate a response to this letter. I did not receive a response a year ago when I wrote to your office the first time expressing my concerns about SONGS. Thank you for your time and your dedication to keeping California free from preventable nuclear accidents. In conclusion let me say; the only thing necessary for a nuclear disaster to occur in California is for a good Governor to do nothing.
    Sincerely,
    Mrs. Bethann Chambers Page

  23. HAHN Baba – NRC/SCE/MHI/Independent Experts/Public Awareness Series

    Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog. Please excuse me for my grammatical and computer human performance errors.

    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    SCE website states, “The amendment supports SCE’s plan to initially operate Unit 2 at 70 percent power for five months. SCE asked the NRC to act on the amendment before the end of May to facilitate commencement of the restart process for Unit 2 by June 1 so that the unit will be available to assist in meeting peak summer electricity demand. Following the initial five-month operating period, SCE would shut down Unit 2 for steam generator tube inspections. Based on inspection data, Unit 2 would resume operation at 70 percent power for an appropriate operating period during the remainder of the 18-24 month fuel cycle while SCE updates its analysis to determine the appropriate long-term power level. Operating at 70 percent power prevents conditions that caused the tube-to-tube wear in Unit 3 that resulted in the nuclear plant being shut down since January 2012. SCE and three independent companies with expertise in nuclear generation have confirmed it is safe to operate the Unit 2 steam generators. The NRC has been evaluating these analyses, which are based on exhaustive testing and inspections, since October 2012.”

    NRC, SCE and three independent companies plus MHI with expertise in nuclear generation need to sharpen their pencils with “High-Energy Public Safety Wisdom Knives” to evaluate the unintended and adverse consequences of multiple and instantaneous SG tube ruptures due to FEI, FIRV, tube wall thinning, metal fatigue, incubating cracks, and collapse of AVB Structure caused by AOOs and DBAs with Unit 2, @70% Power Operation. The radiological doses from a Unit 2 Nuclear Meltdown will cause a Trillion Dollar ECO Disaster and undetermined causalities and cancer effects, and will destroy Southern California like Fukushima. Southern Californians do not want to flee their homes, businesses and schools because of SCE’s false pretenses of starting dangerous Unit 2 to meet the summer months peak energy needs. Who wants to leave Southern California for SCE’s profits, the land of one of the most majestic, entertaining, tourist, and pristine places in this World?

  24. HAHN Baba – NRC/SCE/MHI/Independent Experts/Public Awareness Series
    Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog. Please excuse me for my grammatical and computer human performance errors.

    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    SONGS Unit 2 Potential Nuclear Meltdown due to Loss of Offsite Power

    Preface: Of particular concern with SONGS Unit 2 restart at reduced power are incubating circumferential cracks in tubes next to each other. When one circumferentially cracked tube ruptures, the additional stresses can cause multiple or cascading tube ruptures, which can result in a nuclear meltdown. SCE, MHI, AREVA, Intertek, Westinghouse and NRC are ignoring these cracks. The difference in management of Steam Generator Tube Rupture between Finland and France and USA is, that no primary coolant (liquid and steam) release to the environment is allowed in Finland, while in France and USA, primary steam releases are not forbidden for profits to conduct risky experiments with people’s lives.

    Conclusions: As shown below, transformer fires, lightening, offsite grid disturbances, main steam line break or other anticipated operational occurrences can result in automatic reactor, reactor coolant pump, feedwater pump and turbine trips. In these situations, all the nuclear power plants In USA can be safely shutdown using plant’s automatic safety systems and timely operator actions except SONGS Unit 2. However, as shown below, that is not the case with San Onofre Unit 2 operating at reduced power as told by “The San Onofre Insider and Dr. Joram Hopenfeld” to Channel 10 Investigative Team on April 25, 2013. NRC Inspector General Reports show that NRC has allowed safety margins in nuclear power plants to decrease too far. Now, not after an accident in SONGS Unit 2, is the time to consider whether the NRC’s position is prudent (safety overrides production), or political (Forget Barbara Boxer, Representative Ed Markey, Dr. Joram Hopenfeld & 8.4 Million Southern Californians concerns) on Unit 2 restart proposed SCE License Amendment.

    A. Background Events:
    1. March 15, 1994 – Los Angeles Times, Fire Follows Blast at San Onofre, March 15, 1994 – An electrical transformer blew out with a loud blast at the nuclear power plant here but caused no power interruptions, a Southern California Edison Co. spokesman said Monday.
    2. February 05, 2001, Los Angeles Times, Fire Shuts Down San Onofre Unit Reactor – The Fire destroyed Unit 3 switchgear, turbine and the plant was shutdown for 5 months resulting in a loss of 100 Million Dollars.
    3. September 8, 2011, SONGs Unit 2 began the inspection period at essentially full power. On September 8, 2011, the unit tripped due to an offsite electrical grid disturbance. Following restoration of the electrical grid, the unit returned to essentially full power on September 11, 2011, and remained there for the duration of the inspection period. Unit 3 began the inspection period at essentially full power. On August 6, 2011, power was reduced to approximately 65 percent to repair main feedwater pump turbine MK006. Following completion of repairs, the unit returned to full power on August 15, 2011. On September 8, 2011, the unit tripped due to an offsite electrical grid disturbance. Following restoration of the electrical grid, the unit returned to essentially full power on September 15, 2011, and remained there for the duration of the inspection period.
    4. January 10, 2013: Fire in a main transformer resulted in a reactor trip and an unusual event declaration at a reactor near Houston Tuesday. Transformer at unit 2 of the South Texas Project failed at about 4:40 p.m., according to an event report filed with the Nuclear Regulatory Commission. Although the onsite fire brigade extinguished the resulting flames in the switchyard within 16 minutes, the incident took out power to busses supplying reactor systems, which the report indicated resulted in a partial loss of offsite power. The reactor tripped from full power, and two emergency diesel generators energized the affected circuits. According to the report, unit 2 was cooling down under natural circulation because of a loss of power to reactor coolant pumps. Auxiliary feed water systems functioned as needed, and heat was removed using steam generator atmospheric relief valves. A pressurizer power-operated relief valve momentarily opened and reclosed.
    5. January 17, 2003: A transformer fire at a nuclear plant injured a security officer on Wednesday night and led to the automatic shutdown of one of the plant’s two reactors. The fire at the Donald C. Cook Nuclear Plant in southwest Michigan also resulted in a brief activation of the site’s emergency plan, the plant’s owner, the American Electric Power Company, said. The security officer was treated for smoke inhalation. When the transformer, which is adjacent to the plant, failed, the plant’s operating system automatically shut down the Unit 1 reactor, which was operating at full power. All safety systems responded appropriately, and the reactor was not damaged, the company said.
    6. March 31, 2013: Easter Sunday and Arkansas Nuclear One: A 600-ton component was dropped from a crane while being moved out of the turbine building at Unit 1. At the time of the event, Unit 1 was in a refueling outage with all of the fuel still in the reactor vessel, safely cooled. The accident damaged some electrical equipment that supplies off-site power to the plant. The plant’s emergency diesel generators started and power was quickly restored to the decay heat removal systems. Unit 2, which was operating at full power, automatically shut down when power was lost to a reactor coolant pump due to electrical equipment that was damaged when the component fell. At 9:22 a.m. offsite power to one electrical bus was lost because water from a fire main broken by the falling component caused a short circuit. An emergency diesel generator started up and is supplying power to key safety systems. Unit 2 is cooling down using natural circulation.
    7. April 18, 2013 – Shippingport — An Illinois nuclear power plant automatically shut down Wednesday after a lightning strike knocked out its offsite power, Nuclear Regulatory Commission officials said, an unusual event that the Beaver Valley Nuclear Power Station seems to have an added layer of protection against. The LaSalle nuclear power plant in Marseilles, Ill., declared an “unusual event” Wednesday afternoon after it lost offsite power to both of its Mark II reactors. According to an Exelon company statement, power from the switchyard into the site was interrupted during a severe thunderstorm.
    8. April 26, 2013: Illinois — On April 17, a reported lighting strike caused off-site power to fail resulting in a chain of events. Units 1 and 2 were reported to have been vented, only one of which was through a filtered system. At this time it is unknown is radiation was released over the populace. At the time of venting the wind was blowing eastward away from the plant. LaSalle Unit 1 and LaSalle Unit 2 have both experienced an automatic reactor scram, in conjunction with a loss of offsite power. This was caused by an apparent lightning strike in the main 345kV/138kV switchyard during a thunderstorm. 138kV line 0112 has been inspected in the field, and heavy damage has been noted on the insulators on two of the three phases on a line lightning arrestor line side.

    B. Safety Significance of Transformers Fires: Fires represent half of the risk of core meltdowns at nuclear power plants in the United States, according to the Nuclear Regulatory Commission, or NRC. ”In other words, the fire hazard equals all other hazards combined,” said David Lochbaum, a nuclear engineer with the Union of Concerned Scientists. As one of the most watched-over industries in the world, nuclear power generating plants are required to abide by an abundance of regulations and standards to ensure that the facility, its employees, the environment and the local population are protected from potential hazards. One of the most ominous threats that every nuclear power generating facility faces is the risk of a fire developing within the plant and the associated consequences. There is no shortage of hazards within these facilities; the possibility for fires to ignite from sources such as lube oil, fuel oil or general combustibles within a warehouse are genuine concerns. However, one of the most common sources for ignition – and unfortunately one of the most dangerous as well – are the plant’s transformers. It is no surprise that transformers are inherently high-risk, considering the hundreds of thousands of volts that they transfer on a continuous basis. There are a number of events that can trigger transformer fires, from weather-related incidents to failures stemming from equipment operating beyond its intended service life . While lightning and short circuits in electrical equipment can cause transformer failures, breakdowns in the insulation system are frequently found to be the source of failure. As the insulation material protecting the transformer deteriorates over time from exposure to natural elements, it puts the equipment at risk for failure and subsequently, fires.

    C. Safety Significance of Loss of Offsite Power: The offsite power system of a nuclear power plant provides the preferred source of electrical power to all the station auxiliaries. Loss of offsite power condition results in a reactor/turbine/feedwater trip and Main Steam Isolation Valves close accompanied by an immediate shrinking of steam generator inventory below the low-low level setpoint for automatic initiation of auxiliary feed water. The turbine load rejection, the steam generator tube bundle uncovery, and the feed water instantaneous flashing into steam can over-pressurize the steam generators. Main steam atmospheric dump and safety valves will progressively open to prevent over-pressurization of the steam generators and transfer decay heat to the atmosphere. Steam generator tube uncovery is significant because, if the SG tube break locations becomes uncovered for 10 minutes (Westinghouse analysis), a direct path might exist for fission products contained in the primary coolant to be released to the atmosphere without partition with the secondary coolant. Source: NRC Bulletin No. 88-02: Rapidly Propagating Fatigue Cracks In Steam Generator Tubes.

    D. Recent Steam Generator Tube Rupture Events
    Even with the improved SG inspection programs, operating experience provides examples of tube flaws that were not detected by in-service tube inspections. These flaws were later discovered after accidents and did not satisfy the required structural and accident leakage integrity margins as observed in SONGS Unit 3. There have been five such occurrences from 1987 to 2012:
    (1) On July 15, 1987, a steam generator tube rupture event occurred at North Anna, Unit 1 shortly after the unit reached 100% power. The rupture extended circumferentially 360ø around the tube. The cause of the tube rupture has been determined to be high cycle fatigue and flow-induced vibrations.
    (2) Indian Point 2—SGTR event in February 2000. This represented a failure to meet structural and leakage integrity performance criteria.
    (3) Comanche Peak 1—Failure to meet structural and leakage integrity performance criteria in Fall 2002, as determined by in-situ pressure testing during condition monitoring.
    (4) Oconee 2—Failure to meet structural integrity performance criteria in fall 2002, as determined by in-situ pressure testing during condition monitoring.
    (5) SONGS 3 —Failure to meet structural integrity performance criteria in 2012, as determined by in-situ pressure testing during condition monitoring.
    Of these events, only the tube that leaked under normal operating conditions at SONGS 3 likely would had ruptured with 2 additional tubes, if an MSLB event had occurred during a several-month period preceding the SGTR event in January 2012. This experience indicates that the frequency at which SONGS 2 tubes may be vulnerable to rupture (or leakage from multiple tubes) under MSLB may be above the conditional probability value of 0.05 assumed in Westinghouse, AREVA and Intertek Operational Assessments.

    E. Condition of Unit 2 steam Generators: SONGS Unit 2 & 3 RSGs are of the same design. Therefore, the description of unit 3 provided below is also applicable to Unit 2. SONGS Unit 3 RSGs’ unprecedented tube failure and massive tube and AVB/TSP degradation occurred due to fluid elastic instability, flow-induced random vibrations, Mitsubishi Flowering Effect and high cyclic fatigue under the following unique circumstances:
    (1) U-tube bundle areas with high dry steam will experienced double in-plane velocities (> 50 feet/sec, based on review of MHI Root Cause, Dr. Pettigrew and other research papers published between 2006-2011) compared with out-of plane velocities assumed (25 feet/sec) to have been predicted by Outdated Out-of-Plane Westinghouse /NRC /MHI /AREVA ATHOS Computer Models,
    (2) Lack of positive in-plane restraints and zero damping,
    (3) Large number of SONGS Units 2/3 RSG U-bends with tube clearances of only 0.05 inches (Design 0.25 inches, Industry Norm > 0.25 inches),
    (4) Excessive number of tubes with narrow tube pitch to tube diameter,
    (5) Low frequency small diameter retainer bars (56 HZ) installed to fit the excessive number of SCE requested tubes compared with other MHI SGs’ higher frequency and retainer bars (120 HZ to 1200 HZ), which are not prone to vibrations due to fluid-induced vibrations.
    (6) SONGS’ tubes being much longer compared with Westinghouse Model 51 steam generators (Average length of heated tube = 730 inches) and other MHI SGs,
    (7) MHI RSGs’ unique floating tube bundle with degraded Retainer Bars can collapse due to 100% tube uncovery for 10 minutes under MSLB SG Depressurization, Multiple SGTR SG over-pressurization and lifting of SG Relief Valves, Combination of MSLB and SGTR Conditions, Release of 100% RCS Iodine to Environment,
    (8) Large amount of uncertainties and unverified assumptions in MHI, AREVA, Westinghouse and Intertek’s contact force modeling (zero for in-plane vibrations), calculation of impact wear coefficients and tube stress calculations (4.6 ksi versus 16-17 ksi) and computer and statistical modeling, and,
    (9) Incomplete tube inspections in SONGS Unit 2. Incubating macroscopic circumferential cracks caused by fluid elastic instability, flow-induced random vibrations and high cycle thermal fatigue are extremely difficult to detect and be accurately sized by nondestructive evaluation techniques including X-ray, ultrasonic, and eddy current based bobbin coil probes, mechanically rotating pancake coil (RPC), etc., which have been used in 170,000 SONGS Tube inspections. State-of-the-art systems: Zetec MIZ-80 iD system, Tecnatom TEDDY+, Circular TE and TM, transmit-receive eddy current array probe C-3 and other specialized radiographic probes capable of detecting sub-surface cracks caused by high cycle thermal fatigue have not been used in the 170,000 SONGS Tube Partial and Limited Inspections as shown below for Unit 2 due to access problems in the most problematic innermost sections of the U-Tube Bundle, the high cost, lack of availability of highly specialized tools and contractors, radiation doses, and time considerations in a rush to start Unit 2. The inspection scope defectively designed and degraded SONGS Unit 2 RSGs should have covered 100% hot leg and cold leg tube inspections, 100% of dents or dings, 100% of tube inspections in the tight radius U-bends, 100% area of the Top of the Tube Sheet and Tube Support Plates.

    (10) SCE states, “Remote visual inspections were performed to confirm the integrity of the RBs. The results of these visual inspections are summarized below: (1) No cracking or degradation of RBs or RB-to-retaining bar welds was observed, and (2) No cracking or degradation of AVB end caps or end cap-to-RB welds was observed.
    Note: Remote visual inspections do not ensure that retainer bars or RB-to-retaining bar, AVB end caps or end cap-to-RB welds are not cracked. Cracks in welds can only be detected by using advanced Remote Computer Controlled Low-Frequency Ultrasonic Methods.

    F. SONGS and Offsite Emergency Plans
    Current SONGS Updated FSAR, Emergency Plans, San Diego County Multi-hazard Regional Emergency Operations Plans, IPC/Orange County & Other Offsite/State of CA Plans and NRC Emergency Rules/Guidance, SONGS Drills and Exercises are based on a slow occurring Steam Generator Tube Leakage/Rupture caused by anticipated operational transients, which are significantly flawed based on the SONGS Unit 2 realistic scenario described below.
    G. Loss of Offsite Power In Unit 2:

    A potential accident causes loss of offsite power In SONGs Unit 2 operating at 70% power. This event causes a simultaneous reactor, turbine, feedwater and reactor coolant trips. Due to feedwater pump trip, the RSG U-bundle secondary water level will shrink and tubes will be uncovered for a period of at least 10 minutes and experience a sharp drop/increase in secondary side pressure. The entire sub-cooled feedwater inventory contained in the faulted RSG will instantaneously flash to high dry steam and over-pressurize the steam generators. Loss of Turbine load will also over-pressurize the steam generator. Main steam safety valves located outside the containment will progressively open to prevent over-pressurizing the steam generators and connect the faulty generators to the environment via open steam safety valves. In the midst of 100s of alarms and flashing trouble shooting windows, now for the next 10-15 minutes, the San Onofre Control Room is busy trying to trouble shoot and diagnose the changing plant and transient conditions and flipping through 1000 pages of cumbersome, mind boggling and complex Emergency, Abnormal, Post-Trip, Fire and Severe Management Accident Guideline procedures to determine the correct diagnpstic course of mitigation actions.
    Meanwhile, in the plant, during the same 10 to 15 minutes, the combination of resonant, out-of-plane, in-plane vibrations, jet impingement forces, and RSG debris will cause large axial, bending, dynamic and cyclic loads on all the tubes, tube support plates, retainer bars and anti-vibration structure. The strength of the welded and mechanical connections of these low frequency retainer bars, retaining bars and bridges have not been tested and analyzed for the effects of these cumulative loads to prevent AVB structure displacement, deformation or collapse during loss of offsite power. The displacement, deformation or collapse of AVB structure introduces new and significant axial, bending, dynamic and cyclic loads, which can potentially cause thousands of worn, cracked, plugged and stabilized tubes to exceed their high cycle fatigue stress levels several times than the allowed tube ASME Endurance Limit of 13.6 ksi. If this happens, multiple circumferential tube ruptures will occur at tube-support plates, mid-spans, free spans and tube-to-anti-vibration bar notched interfaces due to macroscopic circumferential cracks caused by tube-to-tube wear and high cycle thermal fatigue. Since all the steam from the RSG would escape to the environment, the iodine-131 from un-partitioned reactor coolant leaking out the rupture tubes will also escape to the environment in less than 10 minutes with 60 tons of radioactive coolant and steam. Consistent with Fukushima Task Force Lessons Learnt and NRC Commissioner Meeting Transcripts, this event will be considered as a beyond design basis event, and SONGS Operators will be unable to take any timely mitigation actions in a radiation/steam environment to stop a severe nuclear accident in progress and notify the Offsite Agencies.
    If the prevailing winds are towards San Clemente, consistent with NRC Inspector General Reports, NRC and Government Studies and observations of SONGS Emergency Plan Drills for the last six years, SCE and Offsite agencies would not have time to respond, notify, evacuate, shelter or give Potassium Iodide to the affected residents within the 10-mile affected emergency planning zone. ODAC, Offsite field monitoring teams, Emergency Vehicles, Helicopters, Orange County Hospitals capabilities will be severely limited or non-functional in a high radiation environment to operate and rescue/transport/shelter disabled, sick, elderly, children, transients and other affected citizens. The casualties, and short, long-term cancer affects to the affected population and ingestion pathway will depend upon the iodine spiking factor and the duration of blowdown, but the offsite releases will significantly exceed the NRC approved SONGS Control Room limit of 5 Rem Total Effective Dose Equivalent (TEDE), and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE.
    NOTE: While this event is occurring, San Diego County, Orange County and Marine Corps Base Camp Pendleton would not be able to dispatch radiation monitoring teams into areas around the plant due to high radiation levels to locate the plume and take soil and air samples to determine the extent of the release off plant grounds. That offsite field monitoring data, along with the data from the plant would not be able to be transmitted to the Offsite Dose Assessment Center (ODAC) located in MESA Emergency Operations Facility for making Protective Action Determinations. The offsite plans are recommended to be revised and feasibility demonstrated via an Emergency Plan Drill using Alternate and Parallel Emergency Operation Facilities located in Irvine and San Diego. The Three Mile Island nuclear accident was not as serious as Chernobyl, but was very confusing and chaotic. 40,000 gallons of radioactive waste was released in the Susquehanna River. 140,000 pregnant women and small children were evacuated as a precautionary measure, but the effects of cancer risk are undetermined.
    If the prevailing winds are towards the Pacific Ocean, the total immediate casualties, including SONGS workers, will be at a minimum, although exposures could still exceed the NRC approved SONGS Control Room limit of 5 Rem Total Effective Dose Equivalent (TEDE) and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE, depending on the iodine spiking factor and the duration of blowdown.

    If the prevailing winds are inland towards San Diego County, the immediate and long-term fatalities and cancer affects will significantly exceed the NRC approved SONGS Control Room limit of 5 Rem Total Effective Dose Equivalent (TEDE) and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE. The impact within the 50 Mile Ingestion Pathway is undetermined.

  25. HAHN Baba – NRC/SCE/MHI/Independent Experts/Public Awareness Series

    Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog. Please excuse me for my grammatical and computer human performance errors.

    Reference: Nuclear Regulatory Commission [Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2]

    Defects or Deviations:

    NRC AIT Report, SCE, MHI, Westinghouse, Intertek and AREVA conclusions on Unit 3 and Unit 2 FEI are incomplete, inconsistent, confusing and inconclusive and are based on invalidated assumptions, incorrect benchmarking, faulty computer and statistical simulations, and hideous testing data (Shielded under the false pretense of Proprietary information). As shown below, the causes in these reports does not meet the intent of NRC CAL ACTION 1, which states “Southern California Edison Company (SCE) will determine the causes of the tube-to-tube interactions that resulted in steam generator tube wear in Unit 3, and will implement actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes. SCE will establish a protocol of inspections and/or operational limits for Unit 2, including plans for a mid-cycle shutdown for further inspections.”

    Background:

    1. The World’s Foremost Renowned Professeur Titulaire, Michel J. Pettigrew, Ecole Polytechnique de Montreal, on the subject of fluid elastic instability and turbulence-induced vibration in 1970’s states, “It is concluded that, although there are still areas of uncertainty, most flow-induced vibration problems can be avoided provided that nuclear components are properly analysed at the design stage and that the analyses are supported by adequate testing and development work when required. There has been no case yet where vibration considerations have seriously constrained the designer.”
    2. Violette R., Pettigrew M. J. & Mureithi N. W. state in a 2006 research paper, “In nuclear power plant steam generators, U-tubes are very susceptible to undergo fluid elastic instability because of the high velocity of the two-phase mixture flow in the U-tube region and also because of their low natural frequencies in their out of plane modes. In nuclear power plant steam generator design, flat bar supports have been introduced in order to restrain vibrations of the U-tubes in the out of plane direction. Since those supports are not as effective in restraining the in-plane vibrations of the tubes, there is a clear need to verify if fluid elastic instability can occur for a cluster of cylinders preferentially flexible in the flow direction. Almost all the available data about fluid elastic instability of heat exchanger tube bundles concerns tubes that are axisymmetrically flexible. In those cases, the instability is found to be mostly in the direction transverse to the flow. Thus, the direction parallel to the flow has raised less concern in terms of bundle stability.”
    3. Channel 10News Question to MHI: Edison says that a letter from MHI to the NRC proves that SCE believed the San Onofre nuclear plant’s steam generators were safe when installed and that safety measures were not sacrificed for licensing reasons. Is that true?
    4. MHI Answer: MHI’s top priority is, and always has been, the safe and reliable operation of all the plants and components that it designs, engineers, supplies and supports. In designing steam generators, minimizing tube wear due to tube vibration is always given a high priority, and this was a priority for MHI during the design of the SONGS replacement steam generators (RSGs). The SONGS RSGs were designed according to industry standards and our customer’s specifications. The design went through an extensive review process which included the participation of third-party experts and MHI believed they would operate as expected: safely and successfully. No safety measures were sacrificed in the design.
    Note: MHI Root Cause states, “The forced outage of Unit 3 and the subsequent discovery of thousands of U-bend tube wear indications in both Unit 2 and Unit 3 after such a short operating period was wholly unexpected. Such an outcome should have been prevented by the conservative design and the precision manufacture. The identification of the unexpected tube degradation led to an extensive evaluation as to the causes the degradation and the questioning of the original design assumptions.”

    5. Root Cause: Root Causes” are defined as the basic reasons [e.g., hardware (design deficiency), process (e.g., mechanistic or operational parameters), or human performance errors (e.g., Root Cause: Root Causes” are defined as the basic reasons [e.g., hardware (design deficiency), process (e.g., mechanistic or operational parameters), or human performance errors (e.g., lack of critical questioning & investigative attitude, lack of solid team work and alignment between the designer and manufacturer, lack of academic research and Industry benchmarking, etc.) for a problem, which if corrected, will prevent recurrence of that problem., etc.) for a problem, which if corrected, will prevent recurrence of that problem.

    SONGS SG Root Cause: Negative Safety Culture (Production over Safety)
    Contributing Causes

    5.1 Human Performance Errors
    • Lack of critical questioning & investigative attitude by SCE/MHI
    • Lack of solid team work and alignment between MHI & SCE AVB Design Team
    • Lack of academic research and Industry benchmarking by SCE/MHI
    • Avoidance of 10CFR 50.90 License Amendment Process by SCE/MHI and Defective 10 CFR 50.59 Evaluation/Screen
    • Complacency, Ignorance and Time Pressure by SCE/MHI
    • Lack of Benchmarking of Computer Codes and Full Scale Mock-up Testing by MHI
    • Inexperienced Designer and Low Cost/Inexperienced/Aggressive/Ignorant Manufacturer

    NOTE: Blind Trust by NRC Commission in SCE’s and Independent Experts Error Likely Conservative Assumptions, Unexplained Safety Explanations and False Public Safety Sermons and Assurances can Lead to a SONGS Unit 2 Reactor Meltdown, the unintended and adverse consequences of which will destroy Southern California, the land of one of the most majestic, entertaining, tourist and pristine place in this universe.

    5.2 Design Deficiencies – SCE
    • Increase of the tube bundle heat transfer surface area from 105,000ft2 (OSG) to 116,100 ft2 (an 11% increase) to generate more heat, more heat and more profits
    • Increase in the number of tubes from 9,350 (OSG) to 9,727 (RSG), 7% increase in heat transfer surface area and low tube-to-tube clearances
    • RSG tube bundle being taller than that of the OSG (Average Length of Heated Tube increase from 680- 750 inches – equivalent to 700 tubes, 7% increase in heat transfer surface area
    • Lack of In-Plane AVBs (Incorrect Assumptions)
    • Low-frequency 56 HZ retainer bar to fit excessive number of tubes
    • Reduction in Tube Wall thickness from 0.048 inches to 0.043 inches to pump more RCS Flows
    • Removal of Stay Cylinder

    5.3 Operational Causes – To generate more heat, more power and profits – SCE
    • Operation at Low Steam Pressures
    • Increased Reactor Coolant Flows
    • Poor Circulation Ratios

  26. Will the damage of the combined effects of tube-to-tube wear, wall thinning, metal fatigue and undetec can vted incubating cracks in Unit 2 @70% RTP due to In-plane/Out-of-Plane FEI and Flow-Induced Random Vibrations during AOOs and DBAs., create new unforeseen problems like the one they have now. If tube walls have thinned (is very likely) in vital areas, is it likely to vibrate causing more wear and thinning in the already thin tubes and so far not damaged tubes or pipes, as the tube walls thin the frequency of vibration will change, with more then one tube wearing the combined vibration may drastically increase if the vibrations phase together like pushing a swing it takes little to increase the energy more and more until a tube bursts. The sudden shock may cause new damage and/or more tube wall thinning. Trying to jerry rigg a bad design already built, with potential damages so costly if it melts down is not worth the risks, also is built in a bad location for natural damages.
    I do not believe a plant with thinning wall tubes is safe from any size earthquake that can shake the tubes, have they ever designed for this problem on any existing or proposed plants?

  27. HAHN Baba
    Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    Orange County Weekly – By Nick Schou Wed., May 1 2013 at 10:17 AM

    Southern California Edison, which operates the San Onofre Nuclear Generating Station (SONGS), has apparently acknowledged for the first time that it may have no choice but to shutter the aging plant later this year. SCE is hoping that the Nuclear Regulatory Commission (NRC) will approve its application to restart a reactor at the plant which has been shut down since January 2012 at 70 percent power. The NRC is expected to make a decision about this proposal, which is being strongly opposed by environmental groups, within the next month or so. Now, the company is acknowledging that if the NRC denies its request, it will likely close the plant–and may even be forced to do so regardless of how the NRC proceeds. The news came in the form of a conference call between company officials and analysts, and is bolstered by reports SCE has filed with the NRC, according to a CNS news report yesterday. Officials are blaming hundreds of millions of dollars in costs associated with San Onofre’s 15-month shutdown. Specifically, SCE says, repairs and inspections have so far cost $109 million, with an additional replacement power cost of $444 million. San Onofre is one of the oldest nuclear power plants in the country. Its Unit 1 reactor was shut down after 25 years in 1992. As Southern California’s population has exploded in recent decades, questions have lingered about just how safe it is to have a nuclear power plant in such a highly populated area. The plant also sits atop an earthquake fault. Safety issues have plagued the plant in the last several years, with regulators complaining of everything from failing emergency generators to faulty wiring and falsified data (emphasis added).

  28. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    Jim Messina, Chair, Organizing for Action, for His Excellency, President of the United States, states,” I’ve spent enough time in Washington to know that the way you win a fight with the gun lobby, faced with some of the most powerful special interests, is just to refuse to give up.” Following his example, 8.4 Million Southern Californians will keep questioning NRC and SCE, until they are convinced that SONGS Unit 2 is safe for restart. 8.4 Million Southern Californians pay for SONGS Unit 2, therefore, they are justified in expressing their concerns about their safety.

    So far, all the available evidence indicates that the following major problems have not been addressed:

    Problem Number 1. The design of San Onofre Replacement Steam generators (RSGs) are identical. SONGS Unit 2 potentially did not suffer in-plane fluid elastic instability due to operation at higher steam pressures and lower RCS flows (Rejecting the impact of double Tube-to-AVB contact forces and better supports responsible for prevention of Unit 2 FEI). SONGS Unit 3 suffered in-plane fluid elastic instability due to operation at lower steam pressures and higher RCS flows (Rejecting the impact of insufficient Tube-to-AVB contact forces and loose supports due to manufacturing errors responsible for Unit 3 FEI). This conclusion is consistent with Westinghouse Operational Assessment, but challenges the SCE, NRC AIT, AREVA and MHI conclusions. NRC AIT Report, SCE, MHI and AREVA conclusions on Unit 3 and Unit 2 FEI are incomplete, inconsistent, confusing and inconclusive and based on faulty computer simulations and hideous testing data (Shielded under the false pretense of MHI Proprietary information). The analysis in these reports does not meet the intent of NRC CAL ACTION 1, which states “Southern California Edison Company (SCE) will determine the causes of the tube-to-tube interactions that resulted in steam generator tube wear in Unit 3, and will implement actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes. SCE will establish a protocol of inspections and/or operational limits for Unit 2, including plans for a mid-cycle shutdown for further inspections.” Repeated requests to NRC, SCE and its Independent Experts to examine carefully the operational difference between Units 2 & 3 and determine its impact on CAL Action 1 have not been addressed to date. NRR has not asked SCE in its RAI(s) the impact of operational differences between Units 2 and 3 on Unit 2 and Unit 3 tube-to-tube wear. Honorable NRC Commissioner Mr. Apostolakis was very confused on Unit 2 FEI inconsistent and conflicting statements by SCE, Westinghouse and AREVA.

    Required Action 1: To protect NRC Commission’s Independent Public Safety Charter Mission, Honorable NRC Chairman is humbly requested that NRC Office of Inspector General retain an Independent Thermal-Hydraulic Expert to examine the operational differences between Units 2 & 3 during Cycle 16 and determine its impact on NRC CAL Action 1 by examining the entire SONGS Cycle 16 operational data for Units 2 & 3. Unit 2 Restart Permission at 70% power should be contingent on completion of the corrective actions required by NRC CAL Action 1 and 10CFR 50 Appendix B.

    Problem Number 2. In light of massive amounts of tube damage (wear), fatigue and tube failure in Unit 3, along with incomplete tube inspections for detection of circumferential incubating cracks in Unit 2, NRC is legally required to ask SCE to check MHI Fatigue Calculations and post the results on its website before any approval of SONGS proposed New License Amendment for restart of Unit 2, to demonstrate that the proposed license amendment (1) Would not involve a significant increase in the probability of an accident previously evaluated in the SONGS FSAR; or, (2) Would not create the possibility of a new or different type of accident previously evaluated in the SONGS FSAR; or, (3) Would not involve a significant reduction in the required margin of safety by operating Unit 2 at 70% power.

    Required Action 2: Based on the above review, NRC should ask SCE to provide a calculation justifying the engineering basis of MHI Fatigue Calculations to meet the ASME Code, NRC RG 1.121, the NRC Chairman and its own Standards. The calculation should be performed by a California Licensed Mechanical or Civil Engineer and Independently Verified by a California Licensed Structural Engineer. In addition, SCE and its Independent Experts should address the synergic effects of tube-to-tube wear and high cycle fatigue, which can be caused by in-plane fluid elastic instability in Unit 2 during anticipated operational occurrences and design bases accidents.

  29. HAHN Baba
    Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    I told Ted Craver in December 2012 Public Meeting that even with NRC Permission to restart Unit 2, if an accident happens, the entire burden and responsibility of facing the angry Regulators, Public, News Media and EIX/SCE Board of Directors, Shareholders will be on his shoulders. NRC needs to take its sweet time and address very carefully the combined effects of tube-to-tube wear, wall thinning, metal fatigue and undetected incubating cracks in Unit 2 @70% RTP due to In-plane/Out-of-Plane FEI and Flow-Induced Random Vibrations during AOOs and DBAs. With so many unknowns, No body can predict with 95% probability, 50% confidence that tube ruptures in SONGS Unit 2 won,t result in a nuclear meltdown. Why take more than the significant risk? I believe that the best solution for SONGS is to Re-tube/Repair the RSGs by Westinghouse like PVNGs. This way, both units can be run safely, reliably, economically and @100% RTP till 2042 with a 20-30 Year License Extension.

    Now here is the news:

    U-T San Diego, April 30, 2013 by Morgan Lee – Nuke Plant May Close, If Restart Denied

    Operators of the San Onofre nuclear plant may decide to retire one or both reactors by year-end if regulators deny or delay a request to partially restart the plant. Southern California Edison executives made the announcement Tuesday in regulatory filings and on a conference call with analysts, citing uncertainties about mounting repair and power replacement costs linked to the 15-month outage.

    Edison CEO Ted Craver said a decision likely be made before year-end 2013 on whether to shut down Units 2 and 3 if the Nuclear Regulatory has not approved the company’s proposal to restart the Unit 2 reactor at partial power. Edison also pushed back its goal of restarting Unit 2 by June 1.

    “Without a restart of Unit 2, a decision to retire one or both units would likely be made before year-end 2013,” the company said. Edison, based in Rosemead, has spent $553 million on the plant, which has been shut since January 2012 because of damaged steam generators.

    The company has sought permission from the U.S. Nuclear Regulatory Commission to restart Unit 2 at reduced capacity to avoid shaking damaged pipes. Costs for repairs and inspections at San Onofre climbed to $109 million and replacement power expenses rose to $444 million through March 31, Edison said in an investor slide presentation today.

  30. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors.

    http://www.10news.com/news/investigations/san-onofre-insider-says-nrc-should-not-allow-nuclear-restart-042513

    SAN DIEGO – For the first time, a source from inside the San Onofre nuclear power plant has come forward to warn that restarting the power plant is too dangerous.

    “There is something grossly wrong,” said the inside source, a safety engineer who worked at San Onofre and has 25 years in the nuclear field. The source, who requested anonymity, is not alone in concerns over the safety San Onofre Nuclear Generating Station (SONGS).
    The concerns stem from inside the concrete containment walls, which house steam generators unique to the plant.

    Japanese manufacturer Mitsubishi Heavy Industries (MHI) built replacement generators for the aging nuclear plant in 2010 and 2011.

    “There were many, many changes,” said Dr. Joe Hopenfeld, a former employee of the Nuclear Regulatory Commission (NRC). He described himself as pro-nuclear.

    Hopenfeld spent his entire professional life working with steam generators and nuclear power. Though he lives in Maryland, he is familiar with San Onofre, which is run by Southern California Edison (SCE).

    The new generators were designed to provide low cost power for decades. Instead, they shut it down in just eleven months because of a radiation leak.

    “The manufacturer didn’t have experience in this size unit,” said Hopenfeld. “I have reviewed thousands of pages of assessment and reports that Edison has submitted.”

    He says the 2011 radiation leak that shuttered the plant revealed a potentially catastrophic problem with the tubes that carry scalding water.

    “As far as I’m concerned, it’s a very serious risk,” Hopenfeld said.

    Tubes carry water to and from the reactor core. This creates steam, which turns the turbines and produces energy.

    “The tubes operate under very high pressure,” Hopenfeld said, adding there is no protection provided between the tubes, which are placed in rows, to keep them from hitting each other.

    Our sources said the redesign of the generators had unintended consequences. Tubes began hitting each other, creating cracks.

    “These tubes were hitting each other — that’s dangerous,” said Team 10’s anonymous source.

    He wants to remain anonymous because he told Team 10 he fears for his safety.

    “When they made these changes, they did not look at the academic research nor use critical question and an investigative attitude,” said the source.

    Hopenfeld and the inside source said the tubes’ movement — banging into each other — led to unprecedented tube failures.

    Of 19,400 tubes, a NRC report found more than 17 percent were damaged.

    Hopenfeld said the worst case scenario is a main steam line break, which he says could be caused by tubes cracking, the tube walls thinning or metal fatigue.

    The anonymous insider and Hopenfeld said if there is a main steam line break, there is potential for the reactor core to overheat – which could mean a full or partial meltdown.

    “Many tubes, and I don’t know how many, have exhausted their fatigue life – they have no fatigue life left,” Hopefeld said.

    Just like the airline industry, the effect of fatigue on metal is something of concern in the nuclear industry.

    While metal may not show the effects of fatigue to the naked eye, it is weakened after use.

    According to Hopenfeld, that is what has happened inside SONGS.

    SCE proposed a solution for the restart. The company said out of an abundance of caution, it would operate only Unit Two at 70 percent power if the NRC approves a restart.

    Both Team 10 sources said that may reduce risk, but it is no guarantee of safety.

    “Maybe the vibrations wouldn’t be as severe, but it doesn’t mean they are going away,” Hopenfeld said.

    “If an accident like this happens, (an) emergency plan is not geared to handle such a public safety devastation,” the inside source said. “Those things have never been practiced or demonstrated in a drill scenario.”

    SCE did not agree with the insider’s assessment of its disaster drills. A spokeswoman called late Thursday afternoon and said SCE runs drills four times a year and includes community partners.

    The spokeswoman said the company plans for any issue that can happen at the plant.

    Team 10 obtained an internal safety report that states in part:

    With both units in shutdown due to leaks in the Steam Generator tubes, SONGS Senior Management attention is focused on resolving this problem and seeking NRC’s permission to restart the units. With SONGS under NRC, INPO, NOB, Public and Media scrutiny, Station cannot afford the luxury of dealing with adverse performance and publicity in Emergency Preparedness caused by declining SONGS Drill/Exercise Performance (DEP) indicator metric.

    The inside source said the report refers to the plant’s drill success rate. The NRC’s website states the Exercise Performance Indicator monitors the “timeliness and accuracy of licensees performance in drills and exercises with opportunities for classification of emergencies…”

    Hopenfeld and the inside source said no one can predict what will happen if the plant restarts.

    “I am not trying to scare anybody — you can live there, but you should know what the risk is,” Hopenfeld said.

    The NRC is expected to make a decision about the possible restart of San Onofre within the coming weeks.

    SCE maintains the plant is safe to restart and declined an on-camera interview. SCE did send this statement:

    While Dr. Hopenfeld has an extensive resume, his SONGS analysis is significantly flawed, reflecting his lack of specific expertise in tube vibration analysis provided by the three experts that performed SCE’s analysis, which included more than 170,000 inspections.

    The NRC is the appropriate authority to evaluate steam generator tube integrity and continues in that oversight and regulatory role for SONGS.

    — A fatigue analysis submitted by Dr. Hopenfeld to the CPUC contains many allegations that have been presented before and been refuted; the most obvious example is his criticism of the original initial 50.59 analysis for the Replacement Steam Generator. This issue has been addressed by the NRC in several public venues, and the NRC noted that SCE followed all required regulations in completing the 50.59 analysis.

    — Hopenfeld’s fatigue analysis concerning in-plane tube vibration is significantly flawed in that it applies an unreasonably high stress concentration factor based on solid body geometry rather than the more realistic stress concentration factors for a cylindrical geometry applicable to the SONGS steam generator tubes.

    SCE also responded to Team 10 questions by sending past news releases sent to regional media. Read those statements here.

    Mitsubishi Heavy Industries is based in Japan. Team 10 emailed MHI specific questions. Here are those questions and answers:

    Q: Edison says that a letter from MHI to the NRC proves that SCE believed the San Onofre nuclear plant’s steam generators were safe when installed and that safety measures were not sacrificed for licensing reasons. Is that true?

    A: MHI’s top priority is, and always has been, the safe and reliable operation of all the plants and components that it designs, engineers, supplies and supports. In designing steam generators, minimizing tube wear due to tube vibration is always given a high priority, and this was a priority for MHI during the design of the SONGS replacement steam generators (RSGs). The SONGS RSGs were designed according to industry standards and our customer’s specifications. The design went through an extensive review process which included the participation of third-party experts and MHI believed they would operate as expected: safely and successfully. No safety measures were sacrificed in the design.

    Q: Why is the NRC’s Augmented Inspection Team (AIT) putting all the blame on MHI and not SCE?

    A: It is important to understand that the AIT Report reflected the NRC’s understanding of the issues as of July 2012, and is just one part of an on-going inspection that the NRC has been conducting since the tube leak in one of the SONGS Unit 3 RSGs was detected in January 2012. MHI is committed to cooperating fully with the NRC in its inspection activities and has made and is making available internal MHI documents as they are requested.

    Q: MHI knows how to build steam generators. Were Edison’s design specifications faulty (Demanding 11% Additional Heat transfer area) or MHI did not how to build these San Onofre Replacement Steam Generators?

    A: MHI has built more than 100 steam generators according to customer specifications, industry standards, practices and operating data and experience. MHI worked closely with SCE on the RSG design and fabrication, using the best available technology to meet the customer’s specified requirements and the industry’s high standards.

    The SONGS RSGs experienced an unprecedented condition: in-plane tube vibration resulting in tube-to-tube wear. The NRC and other industry experts have confirmed that the occurrence of in-plane tube vibration causing tube-to-tube wear at SONGS was unexpected and without precedent and that MHI had followed industry practice in its design.

    Q: Did SCE exceed the power limits in Unit 3 the generators could safely produce? Is it possible to anticipate these sort of problems? Does MHI know that Unit 2 steam generators were running at much higher pressures than Unit 3?

    A: The SGs were designed to operate at the licensed power for SONGs, and to our knowledge that licensed power level was not exceeded. The thermo-hydraulic conditions in the RSGs for both SONGS units have been shown to be the same. The in-plane vibration and related tube-to-tube wear discovered at SONGS had never been previously observed in an operating nuclear power plant of this design. The in-plane tube vibration, observed at the steam generators of Unit 3, was caused by the use of smaller, more uniform tube-to-support gaps than Unit 2, which reduced the contact force available to restrain tube movement in the in-plane direction.

    Q: Is there a way to measure fatigue in the tubing you created?

    MHI did analyze the potential for fatigue failure of the RSG tubes under operating conditions and determined that fatigue was not a credible tube failure mechanism because the stresses sustained by the tubes due to in-plane vibration are well below the stresses that would cause fatigue failure. The analysis that supports this conclusion is contained in Appendix 16 to the “Tube wear of Unit-3 RSG – Technical Evaluation Report.” It should be noted that the technical reviews and analysis, both by the NRC and industry experts, have not mentioned fatigue failure of the tubing.

    A: What is the possibility that the “tubes” at issue could be removed from UNITS 2 & 3? What are the problems you might face in pulling this off? Is this a cost effective solution? Has your firm ever done tube removal at other nuclear facilities? Where and when?

    Tubes exhibiting significant wear or which are potentially vulnerable to such wear have already been removed from service at SONGS by plugging. Plugging tubes within the limits set by the plant licensing documents is a standard practice in the industry and poses no safety concern. It has been implemented to some extent or another by all nuclear utilities whose facilities include steam generators.

  31. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors.

    The Dark Past of San Onofre

    The ratio of flow rate of the steam water mixture, which flows through the SG tube bundle, to the flow rate of steam out of the steam nozzle, is called the circulation ratio. It is desirable to maintain a high circulation ratio above 4 (preferably over 5) to reduce the concentration of chemicals, debris, steam blanketing and steam dry-outs, etc., in the SG. San Onofre Unit 3 RSGs had a circulation ratio of around 3.26, which along with high steam flows, narrow tube to pitch diameter, lack-of in-plane restraints, narrow tube clearances, excessive number of extremely tall tubes and low pressure of secondary side (833 psi) caused fluid elastic instability in 4% area of the RSGs. Everything between San Onofre Units 2 and 3 was the same, except Unit 2 was operating with a steam pressure between 892-942 psi (consistent with NRC AIT Report). At low steam pressure and low circulation ratios, a RSG can produce more heat and more thermal megawaats (more profits in the pocket of SCE). But, low steam pressures and low circulation ratio are severe for vibrations, steam blanketing and steam dry-outs. These are basic elementary facts established about the design and operation of nuclear steam generators prior to San Onofre RSG design, which SCE and Mitsubishi should have known.

    Because of the high secondary pressure operation, fluid elastic instability did not occur in Unit 2, which is consistent with Westinghouse Operational Assessment. Contact forces play a role in the out-of plane FEI, but SCE/MHI inadvertently designed better supports and double the contact forces in Unit 2 are not the reason that FEI did not occur In Unit 2. That is just an attempt to mislead the public and NRC and is contested based on an in-depth review of conflicting AREVA, Westinghouse, John Large, SCE and MHI Reports.

    Now MHI says, “Thus, not using ATHOS, which predicts higher void fractions than FIT-III at the time of design represented, at most, a missed opportunity to take further design steps, not directed at in-plane FEI, that might have resulted in a different design that might have avoided in-plane FEI. However, the AVB Design Team recognized that the design for the SONGS RSGs resulted in higher steam quality (void fraction) than previous designs and had considered making changes to the design to reduce the void fraction (e.g., using a larger downcomer, using larger flow slot design for the tube support plates, and even removing a TSP). But each of the considered changes had unacceptable consequences and the AVB Design Team agreed not to implement them. Among the difficulties associated with the potential changes was the possibility that making them could impede the ability to justify the RSG design under the provisions of 10 C.F.R.§50.59. Thus, one cannot say that use of a different code than FIT-III would have prevented the occurrence of the in-plane FEI observed in the SONGs RSGs or that any feasible design changes arising from the use of a different code would have reduced the void fraction sufficiently to avoid tube-to-tube wear. For the same reason, an analysis of the cumulative effects of the design changes including the departures from the OSG’s design and MHI’s previously successful designs would not have resulted in a design change that directly addressed in-plane FEI.”

    The above statement reflects negligence, ignorance, excuses and cover-up both by MHI/SCE. Let us examine what is happening next.

    The Dangerous Future of San Onofre

    San Onofre Unit 2 Retainer Bar and AVB Performance Analysis during Anticipated Operational Transients and MSLB

    San Onofre replacement steam generators (RSGs) consist of about 9,727 extremely tall and very tightly packed inverted U-tubes. The tubes in each RSG, are arranged in a triangular pitch in 142 rows and 177 columns. The tubes form the boundary separating the steam-water mixture in the secondary circuit from the highly pressurized hot radioactive coolant contained in the primary circuit (tubes). After San Onofre Unit 3 Accident, the integrity and the life expectancy of the Unit 2 tubes are therefore of prime concern 8.4 Million Southern Californians.

    The tube bundle top region, known as the U-bend region is supported by a floating Anti-Vibration Bar (AVB) structure consisting of three sets of two V-shaped AVBs between each tube column. The AVBs are made of Type 405 ferritic stainless steel and are equipped with two Alloy 690 end caps. Each AVB end cap is welded to an Alloy 690 retaining bar. The continuous retaining bar wraps around the tube bundle to which is fixed the outboard ends of the AV bars. The retaining bar is pulled in, wrapped around the tube bundle by the hairclip-like retainer bar, this being captured in situ by being threaded through the first two rows of tubes, and held in this position by friction between the retainer bar and the inboard top surfaces of the AV bars. Thirteen Alloy 690 bridges run perpendicular to the retaining bars and hold the entire structure together. A total of 24 Alloy 690 chrome-plated retainer bars welded to the retaining bars is provided to prevent AVB structure displacement during SG fabrication and during anticipated operational transients and main steam line breaks.
    In San Onofre replacement steam generators, to accommodate the increased number of tubes, the retainer bars are relatively long and thin as compared to the retainer bars in other SGs designed by MHI, resulting in their having low natural frequencies (56 Hz). The retainer bars anchor the AVB structure to the tubes, but are designed such as to not contact the tubes under operating conditions. The AVB structure is not attached to any other SG component and under operating conditions is held in place by friction between the AVBs and the tubes.

    In San Onfre replacement steam generators, the relative motion between the tubes and the anti-vibration bars (AVBs), tube support plates, and the retainer bars have resulted in tube wear and fatigue damage in tubes due to fluid elastic instability (FEI), flow-induced random vibrations and hydrodynamic pressures. SCE, Westinghouse, AREVA, MHI and Intertek have not addressed the synergic effects of tube wear and fatigue damage. These adverse phenomena can produce instant (> 10 minutes) multiple tube failures when the stresses generated during vibrations are sufficiently large due to the collapse of unique MHI anti-vibration bar structure and retainer bars during anticipated operational transients and main steam line breaks as discussed below.

    MHI reports, “The Steam Generator tube wear adjacent to the retainer bars was identified as creating a potential safety hazard. The maximum wear depth is 90% of the tube thickness. The cause of the tube wear has been determined to be the retainer bars’ random flow-induced vibration caused by the secondary fluid exiting the tube bundle. Since the retainer bar has a low natural frequency, the bar vibrates with a large amplitude. This type tube wear could have an adverse effect on the structural integrity of the tubes, which are part of the pressure boundary. The plugging of the tubes that are adjacent to the retainer bars was performed. MHI has recommended to the purchaser to remove the retainer bars that would have the possibility of vibration with large amplitude or to perform the plugging and stabilizing for the associated tubes.” Plugging of the at-risk tubes is not a satisfactory solution because it is the retainer bar that vibrates via random fluid flow processes at sub FEI critical velocity levels – these are likely to continue in play or, indeed, exacerbate at the proposed U2 restart at 70% power, leading to through-tube abrasion, the detachment of tube fragments, lodging at other unplugged and in-service tube localities, resulting in the so-called ‘foreign object’ tube wear. MHI’s recommendation that those retainer bars at risk of large-amplitude fluid flow excited vibration should be removed or plugging and stabilizing for the associated tubes is, of course, dependent upon reliable analysis to identify the at-risk assemblies. SCE and MHI have a repeated history of catastrophic design failures and cover-ups with San Onofre RSGs.

    During anticipated operational transients and main steam line breaks, the whole u-tube bundle will be subject to fluid elastic instability (due to formation of 100% void fractions) and would be connected to the outside environment as described below. According to the latest research papers, the in-plane velocity caused by fluid elastic instability is more than double the out-of-plane velocity caused by flow-induced random vibrations. Retainer bar vibrations caused by flow-induced random vibrations was the reason identified by MHI to remove the retainer bars that would have the possibility of vibration with large amplitude or to perform the plugging and stabilizing for the associated tubes. Retainer bars have not been removed, but more than 180 tubes in SONGS Unit 2 RSGs have been plugged and/or stabilized. The problem stems from that none of the SCE consultants and MHI have analyzed what will happen to the structural integrity of the retainer bars and floating Anti-Vibration Bar (AVB) structure, and thousands of worn, cracked, plugged and stabilized tubes during adverse effects of anticipated operational transients and main steam line breaks. Let us examine the scenarios:

    Based on SONGS Unit 3 tube leakage, failure of 8 tubes at main break steam line testing conditions and more than three hundred damaged tubes, the following two potentially risk-significant events have not been considered as beyond-design basis accidents in SONGS NRC Approved FSAR or SCE proposed License Amendment for Unit 2 restart at 70% power with significant adverse consequences:

    (1) Operating experience from SONG Unit 3 and design information of RSGs suggests that the potential exists for a line breach to significantly increase RSG leakage, because of resonant, out-of-plane and in-plane vibrations of RSG tubes from a main steam line break. These events could potentially cause increased tube leakage due to multiple tube ruptures resulting from thousands of worn, cracked, plugged and stabilized tubes.

    (2) Significant RSG tube leakage could lead to secondary system breaches (lifting of main steam line relief valves) from anticipated operational transients (e.g., a loss of offsite power). The resulting SG secondary side blowdown could further increase tube leakage due to resonant, out-of-plane and in-plane vibrations within the affected SG tube bundle.

    From any such leakages, concurrent with containment bypass, these events might cause offsite radiation doses in excess of 10 CFR Part 100 as evaluated in the SONGS FSAR. Any of these two events would cause a simultaneous reactor, turbine, feedwater and reactor coolant trips. Due to feedwater pump trip, the RSG U-bundle secondary water level will shrink and tubes will be uncovered for a period of at least 10 minutes and experience a sharp drop in secondary side pressure. The entire sub-cooled feedwater inventory contained in the faulted RSG will instantaneously flash to high dry steam. The combination of resonant, out-of-plane, in-plane vibrations, jet impingement forces, broken tube fragments and RSG debris will cause large axial, bending, dynamic and cyclic loads on all the tubes, tube support plates, retainer bars and anti-vibration structure. The strength of the welded and mechanical connections of these low frequency retainer bars, retaining bars and bridges have not been analyzed for the effects of these cumulative loads to prevent AVB structure displacement, deformation or collapse during anticipated operational transients and main steam line breaks. The displacement, deformation or collapse of AVB structure along with the large axial, bending, dynamic and cyclic loads can potentially cause thousands of worn, cracked, plugged and stabilized tubes to exceed several times the allowed tube ASME Endurance Limit of 13.6 ksi. If this happens, multiple tube ruptures will occur at tube-support plates, mid-spans, free spans and tube-to-anti-vibration bar notched interfaces. Since all the water from the RSG would escape to the environment, the iodine-131 from un-partitioned reactor coolant leaking out the rupture tubes will also escape to the environment in less than 10 minutes with 60 tons of radioactive coolant. Consistent with Fukushima Task Force Lessons Learnt and NRC Commissioner Meeting Transcripts, this event will be considered as a beyond design basis event, and SONGS Operators will be unable to take any timely mitigation actions to stop a severe nuclear accident in progress. If the prevailing winds are towards San Clemente, consistent with NRC Inspector General Reports, NRC Studies and observations of SONGS Emergency Plan Drills for the last six years, SCE and Offsite agencies would not have time to respond, notify, evacuate, shelter or give Potassium Iodide to the affected residents within the 10-mile affected emergency planning zone. The casualties, and short, long-term cancer affects to the affected population will depend upon the iodine spiking factor and the duration of blowdown, but will significantly exceed the NRC approved SONGS Control Room limit of 5 Rem Total Effective Dose Equivalent (TEDE), and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE.

  32. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors.

    Preface: It is the legal and moral duty of every United States government official and politician to ensure that public safety is not endangered, whether, it is gun violence, terrorist attacks or radiological accidents. But, that does not seem to be case with NRC preliminary approval of San Onofre proposed License Amendment to Operate Unit 2 at 70% reduced power.

    Recommend Action: Office of the Inspector General (OIG), U.S. Nuclear Regulatory Commission is requested to conduct an inquiry concerning the NRC’S handling of issues associated with the San Onofre steam generator tube rupture. This inquiry is required to address concerns raised by 8.4 Million Southern Californians, Numerous Safety Experts and Public Organizations and members of the Congress (Senator Barbara Boxer and Congressman Rd Markey) as a result of some of the issues described below.

    Conclusions: It appears that NRC commission has not given proper time to NRC Brilliant NRR and RES Staff to study and evaluate the significant adverse consequences of San Onofre Unit 2 Restart reduced power experiment. The problems with Unit 2 Restart are intentionally buried in reams of worthless paper submitted by SCE and its consultants. It appears that NRC Commission under pressure from Edison Management and Lobbyists has hastily indicated approval of proposed SCE License Amendment with an attempt to bypass the Public Hearing. NRC and SCE want to subvert the Legal Process for questioning by Safety Experts to challenge the fallacies of the SCE License Amendment documents. From the review of NRC Office of Inspector General, Fukushima Task Force, Davis-Besse, Three Mile Island and SONGS Units 2/3 Reports, it appears that NRC is not following its own rules and Lessons Learnt from operating experience of radiological accidents.

    San Onofre defectively-designed and degraded 21st Century Safest and Magical Unit 2 floating anti-vibration structure is subject to collapse under adverse consequences of fluid elastic instability, flow-induced vibrations, high cyclic fatigue and Mitsubishi Flowering Effect caused by anticipated operational transients and main steam line breaks even with Unit 2 at reduced power. In San Onofre replacement steam generators, the relative motion between the tubes and the anti-vibration bars (AVBs), the tube support plates, and the retainer bars have resulted in unprecedented tube-to-tube wear and fatigue due to adverse phenomena described above. These adverse phenomena can produce relatively quick tube failures when the stresses generated during severe vibrations of tubes are sufficiently large.

    As described in Unit 2 Return to Service Report, Attachment 4, MHI Document L5-04GA564, Appendix 16, page 459 of 474, MHI used a finite element model (“FE”), to conclude that the tubes were subjected to a stress of 4.2 ksi (kilopounds per square inch) . Consequently, MHI concluded that the stress on the tube due to in-plane vibration is under fatigue limit (13.6 ksi) and the structural integrity of the tube is confirmed from the view point of fatigue due to in-plane vibration (page 470 of 474). A review of the MHI report indicates that results are based on two erroneous assumptions. The source of MHI’s error has resulted from how MHI has calculated the increase in the local stress at geometrical discontinuities (notches), which are formed when two metal surfaces come in contact during vibration. Since the worn surfaces of the tubes inside the steam generators cannot be seen, MHI made two incorrect key assumptions, which are inconsistent with the observation that both the tube and the supporting bar are worn into each other. First, MHI assumed that the ASME endurance limit could be applied directly to the notched tube surfaces. Since it is commonly known that surface roughness significantly reduces fatigue life and since the ASME data is for smooth polished surfaces, this assumption would underestimate the amount of fatigue damage. Secondly, when using the Peterson chart, MHI assumed an unrealistically large fillet radius and consequently derived a low concentration stress factor. Large radii would decrease the local stress and cause the tube to fail at a higher level of stress, thereby increasing its fatigue life. Only by using these two, arbitrary, non-conservative assumptions was MHI able to conclude that Unit 2 did not suffer any fatigue damage. When these assumptions are corrected, the opposite conclusion is reached, which is that the tubes will be susceptible to failure from fatigue.

    MHI states, “MHI did analyze the potential for fatigue failure of the RSG tubes under operating conditions and determined that fatigue was not a credible tube failure mechanism because the stresses sustained by the tubes due to in-plane vibration are well below the stresses that would cause fatigue failure. The analysis that supports this conclusion is contained in Appendix 16 to the “Tube wear of Unit-3 RSG – Technical Evaluation Report.” Westinghouse, AREVA, SCE and Intertek have failed to address the synergic effects of tube-to-tube wear and high cycle fatigue induced cracks. NRC Brilliant NRR and RES Staff has completely missed this aspect of tube failure, because they are under political time pressure to approve the Restart of Unit 2. Hence, the NRC Commission has failed to fulfill its exclusive responsibility for enforcing radiological health and safety requirements for 8.4 Million Southern Californians. The proposed SCE license amendment does not meet the qualification criteria under 10 CFR 50.92.

    Let us discuss why the proposed SCE license amendment does not meet the qualification criteria under 10 CFR 50.92 based on a very basic and fundamental understanding of physics, vibrations, stresses and heat transfer.

    Nucleate Boiling Region – In this region of the steam generator, the saturated feedwater picks up energy from the hot reactor coolant tubes and begins to boil. The initial heat transfer process in the tube bundle is called the nucleate boiling. The tubes remain wetted, and small bubbles rapidly form and break away from the surface of the tubes. Nucleate boiling provides a large heat transfer coefficient because of the turbulence resulting from the bubble formation. Most of the primary-to-secondary heat transfer occurs in this region of the tubes – continued below.

    Film Boiling Region – Nucleate boiling continues until enough water is vaporized to allow a blanket of high dry saturated steam to form on the tubes; this condition is known as film boiling. The steam blanket forms gradually as the steam quality reaches higher values. It becomes fully developed within a very short axial distance of the tubes. The steam quality and vapor fraction in the film boiling region are 100%.

    Nucleate boiling (steam bubble formation is at the tube interface) takes place when the surface temperature of the tube is hotter than the saturated steam-water fluid mixture temperature by a certain amount but where the heat flux is below the critical heat flux. Nucleate boiling occurs when the surface temperature is higher than the steam saturation temperature (TS) by between 7.2 °F to 54 °F. The critical heat flux is the peak on the curve between nucleate boiling and transition boiling. In case of Unit 3, the temperature difference was 75 °F (Departure from Nucleate Boiling) and it was film boiling or very high dry devastating steam (in-plane fluid elastic instability, vapor fraction ~ 99.6-100%) in the Unit 3, four percent region of tube-to-tube wear. In case of Unit 2, the temperature difference was 67 °F and it can be described as some state between nucleate boiling and transition boiling and production of moderate dry steam (out-of-plane random vibrations, vapor fraction ~ 98.5 %) in the region of tube-to-AVB wear. It is highly conceivable, that this moderate dry steam in Unit 2 on the way up from the high region of tube-AVB wear exited the tube bundle as very high dry devastating steam due to additional steam flows, but did not do the damage as Unit 3 because of wider tube clearances in the upper most U-bends.

    Steam saturation temperature is a function of steam pressure. Higher the steam pressure, higher the steam saturation temperature and vice-versa. At lower steam pressure, you can produce more heat and more megawatts. This is basic Intermediate College physics and heat transfer knowledge, which SCE and MHI Engineers knew but chose to ignore it by refusing to lower void fraction as described in the MHI Root Cause Analysis. You know why, because it would have produced less heat (less profits for SCE), delayed construction of SGs by modification of SG components, increased the fabrication cost, triggered a lengthy NRC Review, and forced shutdown of old steam Generators costing more money, downtime, inspections and tube plugging.

    Nucleate boiling prevents fluid elastic instability and formation of high dry steam. Now SCE and MHI are blaming each other and are coming out with thousands of pages of faulty and unconvincing analysis and trying to wash their ignorance and sins by blaming faulty computer modeling, contact forces and manufacturing errors. It appears that NRC Commission is watching the whole show with sleepy eyes, and rose-colored glasses with blinders, and telling 8.4 Million Southern Californians not to worry. What is in this Billions of Dollars and Public Safety Poker Game for NRC Commission? Like NRC Brilliant Staff is already not in trouble with 8.4 Million Southern Californians with NRC Commission show of favoritism towards SCE. Based on NRC Inspector General and Congressman Ed Markey’s Reports, one is likely to conclude that NRC Officers, who are favorable to for the low performance of Unsafe and INPO 4 Utilities, like SONGS, expect to find a lucrative consulting assignment with a Utility after retirement.

    Let us get back to Film Boiling and examine what it will do to Unit 2 SG tubes at 70% power operations in case of a main steam line break. Due to main steam line break event with failure of the main steam isolation valve to close, the steam generator u-tube bundle will be depressurized and the pressure will be atmospheric. From High School physics, everybody knows, that at atmospheric pressure, steam saturation temperature is 212°F. With tube surface temperature at 600°F, the temperature difference between the hot tubes and secondary environment would be approximately 400°F. Steam line break event will cause automatic trip of the reactor, turbine, reactor coolant and feedwater pumps. With the feedwater pumps trip, the U-tube bundle will be uncovered for a period of 10 minutes due to shrinking of SG water level. Now the difference between sub-cooled feedwater temperature of 400°F and tubes temperature is conservatively 150°F. Therefore, feedwater will instantly flash to steam and create high energy jet impingement on tubes. In other words, in a matter of less than I minute, the entire SG will develop film boiling region or would be full of 100% high dry steam, and tubes would beginning to experience the adverse consequences of fluid elastic instability, flow-induced vibrations, high cyclic fatigue and Mitsubishi Flowering Effect. Who knows and can predict the short-term and long-term cancerous effects of offsite radiation doses exceeding the Federal Limits caused by a potential radiological accident in Unit 2.

    SONGS Unit 3 RSGs’ unprecedented tube failure and massive tube and AVB/TSP degradation occurred due to fluid elastic instability, flow-induced random vibrations, Mitsubishi Flowering Effect and high cyclic fatigue under the following unique circumstances:
    (1) U-tube bundle areas with high dry steam, double in-plane velocities (> 56 feet/sec, Dr. Pettigrew and others, 2006-2011) compared with out-of plane velocities assumed (28 feet/sec) to have been used in William Krotiuk 2002 Report NUREG-1919 TH calculations and predicted by Outdated Out-of-Plane Westinghouse /NRC /MHI /AREVA ATHOS Computer Models,
    (2) Lack of positive in-plane restraints and zero damping,
    (3) Large number of SONGS Unit 2 RSG U-bends with tube clearances of only 0.05 inches (Design 0.25 inches, Industry Norm > 0.25 inches),
    (4) Excessive number of tubes with narrow tube pitch to tube diameter,
    (5) Low in-plane frequency tubes and retainer bars compared with MHI SGs’ higher in-plane frequency tubes and retainer bars,
    (6) SONGS’ tubes being much longer than Westinghouse Model 51 steam generators (Average length of heated tube = 730 inches) and other MHI SGs,
    (7) MHI RSGs’ unique floating tube bundle with degraded Retainer Bars can collapse due to 100% tube uncovery for 10 minutes under MSLB SG Depressurization, Multiple SGTR SG over-pressurization and lifting of SG Relief Valves, Combination of MSLB and SGTR Conditions, Release of 100% RCS Iodine to Environment,
    (8) Large amount of uncertainties and unverified assumptions in MHI, AREVA, Westinghouse and Intertek’s contact force (zero for in-plane vibrations), wear rate and tube stress calculations (4.6 ksi versus 16-17 ksi) and computer modeling, and,
    (9) Incomplete tube inspections in SONGS Unit 2. Incubating macroscopic circumferential cracks caused by fluid elastic instability, flow-induced random vibrations and high cycle thermal fatigue are extremely difficult to detect and be accurately sized by nondestructive evaluation techniques including X-ray, ultrasonic, and eddy current based bobbin coil probes, mechanically rotating pancake coil (RPC), etc., which have been used in 170,000 SONGS Tube inspections. State-of-the-art systems: Zetec MIZ-80 iD system, Tecnatom TEDDY+, Circular TE and TM, transmit-receive eddy current array probe C-3 and other specialized radiographic probes capable of detecting sub-surface cracks caused by high cycle thermal fatigue have not been used in the 170,000 SONGS Tube Partial and Limited Inspections as shown below for Unit 2 due to access problems in the most problematic innermost sections of the U-Tube Bundle, the high cost, lack of availability of highly specialized tools and contractors, radiation doses, and time considerations in a rush to start Unit 2. The inspection scope defectively designed and degraded SONGS Unit 2 RSGs should have covered 100% hot leg and cold leg tube inspections, 100% of dents or dings, 100% of tube inspections in the tight radius U-bends, 100% area of the Top of the Tube Sheet and Tube Support Plates.

    So from the above basic theory of nucleate and film boiling, we conclude that proposed SCE License Amendment (1) Would involve a significant increase in the probability of an accident previously evaluated in the SONGS FSAR; or, (2) Create the possibility of a new or different type of accident previously evaluated in the SONGS FSAR; or, (3) Would involve a significant reduction in the required margin of safety by operating Unit 2 at 70% power without adequate repairs or replacement.

    Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. As demonstrated above, these barriers would be significantly challenged by operating the Unit 2 at 70% power due to a main steam line break or other anticipated operational transients. The margin of safety will be drastically reduced and NRC cannot ensure that SCE plant operators would be able to protect adequately the health and safety of the public. Therefore, the proposed changes involve a significant reduction in a margin of safety.

    SCE License Amendment has not described the changes in plain English so that it can be understood by 8.4 Million Southern Californians without detailed knowledge of nuclear plant design and operation. SCE License Amendment has not identified and discussed the previously causes of San Onofre Unit 3 accident that are affected by the proposed changes in Unit 2 and requested by NRC in Confirmatory Letter and NRR Request for Additional Information (RAI).

  33. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.
    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors.

    FACT: SONGS Unit 3 SG Tubes leaked and failed due to tube-to tube wear, tube-to-AVB wear, tube-to-TSP wear, Retainer Bar-to-tube wear and high cyclic thermal fatigue induced axial, circumferential, macroscopic and microscopic cracks. Fluid elastic instability, flow-induced vibrations and Mitsubishi Flowering Effect responsible for the above catastrophic effects were caused by exceedingly high steam flows overstretching the thermal performance and reducing significantly the safety margin of SGs tubes to maximize SCE profits, low steam generator pressures, high reactor coolant flows, narrow tube pitch to tube diameter ratio, low tube clearances, extremely tall tube bundle and lack of in-plane restraints.

    Problem Statement: Two Independent Nuclear Experts Certify that MHI SG Tube Fatigue and Stress Calculations Assumptions are erroneous and based on faulty data. SCE is legally required to certify MHI’s calculations to assure that San Onofre Unit 2 does not pose significant radiological risks at 70% normal steady state power operations, and during Anticipated Operational Transients and Design Basis Accidents.

    SCE Preliminary Response: Independent Experts’ Analysis concerning in-plane tube vibration is significantly flawed in that it applies an unreasonably high stress concentration factor based on solid body geometry rather than the more realistic stress concentration factors for a cylindrical geometry applicable to the SONGS steam generator tubes.

    MHI Response: MHI did analyze the potential for fatigue failure of the RSG tubes under operating conditions and determined that fatigue was not a credible tube failure mechanism because the stresses sustained by the tubes due to in-plane vibration are well below the stresses that would cause fatigue failure. The analysis that supports this conclusion is contained in Appendix 16 to the “Tube wear of Unit-3 RSG – Technical Evaluation Report.” It should be noted that the technical reviews and analysis, both by the NRC and industry experts, have not mentioned fatigue failure of the tubing.

    Introduction: The SG functions as a heat exchanger, by means of which the high temperature pressurized radioactive primary water on the inside of the tubes heats up the non-radioactive secondary water on the outside of the tubes, in order to generate the steam that turns the turbine which in turn generates electricity. In addition to providing a barrier (Reactor Coolant Pressure Boundary) to radioactivity and producing steam, a steam generator has many other important functions. It is the major component in the plant that contributes to safety during transients and/or accidents. A steam generator provides the driving force for natural circulation and facilitates heat removal from the reactor core during a wide range of loss of coolant accidents. Proper steam generator operation is of major safety significance and therefore any changes to its design may have significant safety consequences.

    Out-of-plane fluid-elastic instability has been observed in nuclear steam generators in the past and has led to tube bursts at normal operating conditions. However, the observation of in-plane fluid-elastic instability in steam generators of a nuclear power plant is a true paradigm shift. The combined effects of tube-to-tube wear and high cycle thermal fatigue cracks caused by fluid-elastic instability and/or flow-induced random vibrations have been witnessed as sudden tube ruptures in the following nuclear power plants:

    North Ana in 1987- NRC Bulletin No. 88-02: Rapidly Propagating Fatigue Cracks In Steam Generator Tubes:

    MHI SG Japan 1991 – On February 9th, 1991, leakage of about 55 tons of primary coolant occurred due to the failure of one SG tube in a steam generator built by Mitsubishi in the No. 2 pressurized water reactor at the Mihama nuclear power station in Japan. The tube had been severed, causing the massive leakage of contaminated cooling water. At the same time, water pressure in the core had dropped drastically and the ECCS kicked in, flooding the reactor and shutting it down. If the core had been left exposed, a meltdown — an overheating of the fuel that can, if uncontrolled, lead to a large release of radio-activity — could have occurred. Following week an estimated 7 million Becquerels (Bq) had been released into the sea and an estimated 5 billion Bq of radioactive gas had been released into the atmosphere. This tube rupture caused the first INES level 3 nuclear incident in Japan, which ignited social concerns all over Japan because it shattered the nuclear industries myth of 100% safe reactors! The failed tube was removed from the heat exchanger, and the fracture surface was examined by a scanning electron microscope. Striations, which are a characteristic of fatigue failure, were observed on large portions of the fracture surface, and dimples showing tensile fracture were also observed. However, few traces of stress corrosion cracking and corrosion were found on the fracture surface of the tube. Stress amplitude of the failed tube estimated based on the striation spacing was found to be in the range of around 51 to 60 Million Pascal’s (7391-8702 psi > 1.5 X 5,200 psi, SONGS 3 MSLB Test Pressure).

    Indian Point 2, 2000 (See NRC Office of Inspector General Report Below)

    Comanche Peak 1, 2002 – Routine inspections at the Comanche Peak nuclear power plant failed to detect a damaged steam generator tube that later ruptured, forcing a shutdown. The flaw in the tube was “clearly identifiable and missed” about 18 months ago by workers for TXU Energy, the plant’s owner and operator, according to the preliminary findings of a special inspection team of the Nuclear Regulatory Commission.

    Oconee 2, 2002 —Failure to meet structural integrity performance criteria in fall 2002, as determined by in situ pressure testing during condition monitoring.

    Craus NPP: Between 2004 and 2006, three primary-to-secondary leaks occurred at the Cruas NPP: unit 1 in February 2004 and unit 4 in November 2005 and February 2006. The three leaks were all the result of a circumferential crack in the tube at the location where the tube passes through the uppermost tube support plate (TSP #8).

    SONGS 3 in 2012 – See NRC AIT Report.

    Defect or Deviation: In San Onfre replacement steam generators, the relative motion between the tubes and the anti-vibration bars (AVBs), the tube support plates, and the retainer bars have resulted in tube wear and fatigue damage in tubes due to fluid elastic instability, flow-induced random vibrations, excessive fluid hydrodynamic pressures and Mitsubishi Flowering Effect. These adverse phenomena can produce relatively quick tube failures when the stresses generated during vibrations are sufficiently large. As described in Unit 2 Return to Service Report, Attachment 4, MHI Document L5-04GA564, Appendix 16, page 459 of 474, MHI used a finite element model (“FE”), to calculate that the tubes were subjected to a stress of 4.2 ksi (kilopounds per square inch). Consequently, MHI concluded that the stress on the tube due to in-plane vibration is under fatigue limit (13.6 ksi) and the structural integrity of the tube is confirmed from the view point of fatigue due to in-plane vibration (page 470 of 474). MHI results are based on two erroneous assumptions described below.

    The source of MHI’s error described below resulted from how they calculated the increase in the local stress at geometrical discontinuities (notches), which are formed when two metal surfaces come in contact during vibration. Since the worn surfaces of the tubes inside the steam generators U-Tube Bundle region of tube-to-tube wear could not be seen during SONGS Tube Inspections, MHI made two incorrect key assumptions, which are inconsistent with the observation that both the tube and the supporting bar are worn into each other. First, MHI assumed that the ASME endurance limit could be applied directly to the notched tube surfaces. Since it is commonly known in the nuclear and commercial industry that surface roughness significantly reduces fatigue life and since the ASME data is for smooth polished surfaces, this assumption would underestimate the amount of fatigue damage. Secondly, when using the Peterson chart, MHI assumed an unrealistically large fillet radius for sharp corners (Should be zero according to basic knowledge of geometry) and consequently derived a low concentration stress factor. Large radii would decrease the local stress and cause the tube to fail at a higher level of stress, thereby increasing its fatigue life. Only by using these two, arbitrary, non-conservative assumptions was MHI able to conclude that Unit 2 did not suffer any fatigue damage. When these assumptions are corrected, the opposite conclusion is reached, which is that the tubes will be susceptible to failure from fatigue.

    Required Action: The NRC Chairman has publically stated that SCE is responsible for the work of MHI, Westinghouse, AREVA and Intertek. In light of SONGS Units 2 & 3 massive amounts of tube damage (wear) and tube failure in Unit 3, along with incomplete tube inspections for detection of circumferential incubating cracks in Unit 2, based on the review of attached reports and governing standards described below, SCE is legally required to check MHI Fatigue Calculations and post the results on its website before any approval of SONGS proposed New License Amendment for restart of Unit 2, to demonstrate:

    That the proposed amendment (1) Would not involve a significant increase in the probability of an accident previously evaluated in the SONGS FSAR; or, (2) Would not create the possibility of a new or different type of accident previously evaluated in the SONGS FSAR; or, (3) Would not involve a significant reduction in the required margin of safety by operating Unit 2 at 70% power However, because of the wear damage previously sustained by Unit 2, some tubes may now be susceptible to rapid fatigue failure.

    Based on the above review, SCE needs to provide a calculation justifying the engineering basis of the above statement to meet the ASME Code, NRC RG 1.121, the NRC Chairman and its own Standards. The calculation should be performed by a California Licensed Mechanical or Civil Engineer and Independently Verified by a California Licensed Structural Engineer.

  34. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.
    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors.

    NRC OFFICE OF THE INSPECTOR GENERAL EVENT INQUIRY TO NRC’S RESPONSE TO THE FEBRUARY 15, 2000, STEAM GENERATOR TUBE RUPTURE AT INDIANA POINT 2
    FINDINGS – NRC STAFF DID NOT GET TIME TO REVIEW DOCUMENTS – HISTORY REPEATING WITH UNSAFE SCE LICENSE AMENDMENT ON UNIT 2 RESTART. OIG REPORT SUMMARY

    1. NRC’s Oversight of Events Leading Up to the February 15, 2000, Steam Generator Tube Rupture at Indiana Point 2 (IP2)

    NRC Office of Inspector General determined that the NRC and nuclear industry had long-standing concerns about the loss of integrity of steam generator tubes used on PWRs due to a variety of degradation mechanisms. Degradation problems particular to Westinghouse Model 44 steam generators resulted in all plants with this model steam generator replacing their steam generators, except IP2. The NRC has also been long aware of steam generator tube and other problems at IP2. Nevertheless, the NRC did not conduct a technical review of the July 29, 1997, IP2 steam generator inspection report when it was submitted to NRR. However, OIG noted that steam generator inspections are of sufficient importance to be included in plant technical specifications. IP2 technical specifications mandate steam generator inspections to be conducted no less than every 24 months and the inspection report to be submitted to the NRC no later than 45 days after completion of the inspection.

    OIG also found that had NRC staff or contractor’s with technical expertise evaluated the 1997 results of the IP2 steam generator inspection, the NRC could have identified the flaw in the U-bend of row 2, column 5, in steam generator number 24 that was indicated in the inspection report. This flaw, which was recently determined to be nearly 100 percent through the tube wall in 1997, was the cause of the February 15, 2000, IP2 steam generator tube rupture. OIG found that the 1997 IP2 steam generator inspection results were not reviewed by the NRC staff for technical quality or sufficiency because the staff is not required to conduct such a review.

    OIG determined that NRR’s review of a 1999 license amendment request submitted by IP2 was not adequate. The 1999 IP2 license amendment request number 201, asking for a 1 year extension for the steam generator inspection, was approved by NRR based on a safety evaluation completed by a junior engineer with limited experience in steam generator inspection techniques. The safety evaluation review included the junior engineer’s evaluation of the 1997 steam generator inspection results. During the safety evaluation review process the junior engineer was supposed to receive assistance from a senior engineer with extensive steam generator experience. OIG determined that during the amendment review process, the senior engineer did not review the source documents submitted by IP2 nor did he review the 1997 IP2 inspection report. OIG also noted that other technical expertise available to the NRR staff was not employed to review the 1997 inspection report or the amendment request. OIG also found that during the amendment review process, the NRC requested additional information from IP2 in the form of an RAI to clarify outstanding issues relative the steam generator inspection program. Although the junior engineer was not completely satisfied with the response to the RAI, no additional questions were asked by the NRC of IP2.

    OIG found nearly no involvement in the amendment request review by either the NRR Project Manager assigned to IP2 or the EMCB Branch Chief. OIG also found that the NRR staff believed that the level of review given to IP2’s license amendment request 201 was acceptable because steam generator issues at IP2 were not viewed as significant to NRC’s oversight and regulation of the plant.

    2. NRC Oversight of IP2 Emergency Preparedness Issues

    OIG found that the NRC considered IP2 to be a plant that struggled with various challenges in the area of emergency preparedness, but the NRC decided that allowing IP2 time to correct its deficiencies outweighed the benefit of increasing NRC oversight. OIG also found that the NRC inspectors from Region I had concerns about licensee on-site performance during emergency preparedness exercises from 1998 to present. They had identified significant, long-standing weaknesses that had not been corrected. Because of IP2’s poor performance in pre-announced drills of a known scenario, NRC inspectors questioned IP2’s capability to perform during an actual event. OIG learned that recurring weaknesses that had gone uncorrected appeared to play a role in the poor emergency response performance of IP2 during the February 15, 2000, event.

    Concerning off-site emergency preparedness issues, OIG found that communication between county EOCs and the NRC was non-existent. County officials view the NRC as the only independent source they have to provide them credible, objective information. Those officials desire personal interaction on a routine basis with the NRC staff of IP2 to discuss plant activities. OIG also determined that disjointed and misinformation from IP2 during the February 15, 2000, incident adversely impacted the off-site emergency preparedness process.

    3. Adequacy of NRC Staff’s Review of IP2 Restart Proposal

    OIG found that, although there may have been some initial concerns by the NRC staff evaluating IP2’s request for restart, NRC management has not allowed time constraints to impact on the staff’s ability to conduct a thorough review of the data presented by IP2 in its June 2000 restart proposal.

  35. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors.

    If you cannot see or test a component, then you cannot predict its reliability or safety with significant confidence. Fluid elastic instability, flow-induced vibrations and Mitsubishi Flowering Effect caused tube-to-wear, tube-to-AVB wear, Retainer Bar to tube wear, tube-to-TSP wear, and high-cycle thermal fatigue, axial and circumferential cracks, and incubating cracks. Undetected incubating cracks caused by fluid elastic instability, flow-induced vibrations and Mitsubishi Flowering Effect are of the greatest concern, because they can cause tube leakage or ruptures at 70% power in Unit 2 at any time without notice due to inadvertent component manipulation (opening or closing valves wrong valves), anticipated operational occurrence (loss of offsite power, ATWS) and Design Basis Accidents (Main Steam Line Break, Earthquake) and fire caused by a short circuit or electrical fault in an energized system. Public Safety against radiological accidents cannot be risked in terms of profits or peak electrical demands in summer months. SCE cannot get away with answers to three simple questions, with a quick NRC Review and advertisement in Federal Register. Mitsubishi Fatigue calculations are erroneous and based on hideous data, and do not meet ASME and NRC Regulations. SCE, MHI, Westinghouse, AREVA and Intertek have not addressed the combined synergic effects of tube-to-tube wear and incubating/undetected cracks caused due to adverse effects of fluid elastic instability, flow-induced vibrations and Mitsubishi Flowering Effect. SCE and their consultants cannot answer these questions, So NRC has to prove the safety of Unit 2. Who is going to assure the safety of 8.4 million Southern Californians. NRC cannot sit silent and take no action like a Helpless Police Officer sitting in a Stalled Car with Empty Guns and No radio Communications. Nuclear Wisdom requires safe, time proven and reliable actions. Nuclear regulators cannot be pressured by Billion Dollar Corporations and Powerful Politicians. A nuclear accident in Southern California will have immense and significant adverse consequences beyond the nightmares or dream of a Regulator, Corporation, Citizen or Politician.

  36. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors.

    SONGS Unit 3 RSGs massive and unprecedented tube failures and AVB/TSP degradation occurred due to fluid elastic instability, flow-induced random vibrations, Mitsubishi Flowering Effect and high cyclic fatigue under the following unique circumstances:
    (1) U-tube bundle areas with high dry steam, double in-plane velocities (> 56 feet/sec, Dr. Pettigrew and others, 2006-2011) compared with out-of plane velocities assumed (28 feet/sec) to have been used in William Krotiuk 2002 Report TH calculations and predicted by Westinghouse /NRC /MHI /AREVA ATHOS Computer Models,
    (2) Lack of positive in-plane restraints and zero damping,
    (3) Large number of SONGS Unit 2 RSG U-bends with tube clearances of only 0.05 inches (Design 0.25 inches, Industry Norm > 0.25 inches)
    (4) Excessive number of tubes with narrow tube pitch to tube diameter,
    (5) Low in-plane frequency tubes and retainer bars compared with MHI SGs higher in-plane frequency tubes and retainer bars
    (4) SONGS tubes being much longer than Westinghouse Model 51 steam generators (Average length of heated tube = 730 inches) and other MHI SGs,
    (5) MHI RSGs unique floating tube bundle with degraded Retainer Bars, can collapse due to 100% tube uncovery for 10 minutes under MSLB SG Depressurization, Multiple SGTR SG over-pressurization and lifting of SG Relief Valves, Combination of MSLB and SGTR Conditions, Release of 100% RCS Iodine to Environment
    (6) Large amount of uncertainties and unverified assumptions in MHI, AREVA, Westinghouse and Intertek’s contact force (Zero for in-plane vibrations), wear rate and tube stress calculations (4.6 ksi versus 16-17 ksi) and computer modeling, and
    (7) Incomplete tube inspections in SONGS Unit 2. Incubating macroscopic circumferential cracks caused by fluid elastic instability, flow-induced random vibrations and high cycle thermal fatigue are extremely difficult to detect and be accurately sized by nondestructive evaluation techniques like X-ray, ultrasonic, and eddy current based bobbin coil probes, mechanically rotating pancake coil (RPC), etc., which have been used in 17,000 SONGS Tube inspections. State-of-the-art systems: Zetec MIZ-80 iD system, Tecnatom TEDDY+, Circular TE and TM, transmit-receive eddy current array probe C-3 and other specialized radiographic probes capable of detecting sub-surface cracks caused by high cycle thermal fatigue have not been used in the 170,000 SONGS Tube Partial and Limited Inspections as shown below for Unit 2 due to access problems in the most problematic inner-most sections of the U-Tube Bundle, the high-cost, lack of availability of highly specialized tools and contractors, radiation doses, and time considerations in a rush to start Unit 2. The inspection scope defectively designed and degraded SONGS Unit 2 RSGs should have covered 100% hot leg and cold leg tube inspections, 100% of dents or dings, 100% of tube inspections in the tight radius U-bends, 100% area of the Top of the Tube Sheet and Tube Support Plates. The combined effects of tube-to-tube wear and high cycle thermal fatigue cracks have been witnessed by sudden tube ruptures in North Ana in 1987, MHI SG in Mihama, Japan in 1991, three tube leakages in French SGs between 2004 through 2006, 20 tube ruptures/leakages in SGs between 1980-2000 in USA, and SONGS 3 in 2012.

    In light of SONGS Unit 3 massive tube failures and safety concerns of 8.4 Southern Californians, NRC staff needs to reanalyze NUREG 1919, published 2009, “Resolution of Generic Safety Issue 188: Steam Generator Tube Leaks or Ruptures Concurrent with Containment Bypass from Main Steam Line or Feedwater Line Breaches” before approval of SONGS New License Amendment. This is a one-time Public Safety and Knowledge Test for Brilliant NRC Staff and applies only to SONGS Unit 2 Defectively Designed and Degraded Unit 2 steam Generators. Excellent PR move for NRC to restore its degraded image and allegations of a SCE Captured Agency. Dr. Macfarlane will like it…. Thanks

  37. The following note was sent to me anonymously – Revised for Moderation by HAHN Baba
    Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog. San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer, or human performance grammatical or spelling errors.

    Street rumors are getting stronger by the minute that backed by Insensitive and Public Safety Ignorant Anonymous Power Politician(s), the “Captured & Afraid” NRC and “Penny-Wise & Pound-Foolish” SCE plan to perpetrate another act of Profits and Production Experiment upon the people of Southern California to determine the True Root cause of Replacement Steam Generators (RSGs) degradation and wipe out the mistakes of SCE/MHI Team in the $1 billion RSGs debacle. By allowing Southern California Edison to restart the damaged and defective reactor number two, that has not even been repaired, to experimentally restart at 70% power sometime in June or July 2013. The anonymous writer says that Profits and Production over Safety Experiment is not nuclear and financial wisdom, because many people in California are afraid for the health of their children, their property values, and what would happen if a major nuclear accident happened at San Onofre Nuclear Generating Station? By definition, Profits and Production over Safety is a deliberate act to make Profits and Production ignoring Safety of 8.4 Million Southern Californians. Just thinking about the evacuation plan that every Californian knows would not work, and having to shelter in place during a nuclear meltdown at SONGS. Then after the radiation damage to people’s health has been done just like Chernobyl and Fukushima, the government will announce in a month or two (far too late) that in a 10, 20 or 30 mile radius will be an exclusion zone due to high radiation levels, and everyone has to leave their homes and possessions to go live in a refugee camp somewhere in Riverside County. The loss of billions in property values, infrastructure (schools, roads, beaches, farm land and food crops, local governments) personal disruptions and destruction of several millions lives, all put at risk in the name of profits for SCE. What will be the fate of Southern California if this act of Profits and Production over Safety by NRC and SCE are allowed by restart at San Onofre Nuke plant? NRC and SCE need to demonstrate via an EP drill, 100% success of the SONGS/IPC/NRC/FEMA/Offsite Agencies/State of CA Emergency Plan (Classifications, Notifications, Protective Action Determinations, Siren Alert Response, Shelter, Evacuation, Treatment of Sick and Disabled, Viability of Decontamination Centers in Rush Traffic Hours and other considerations following the Scenario described by HAHN Baba on this blog, April 12, 2013 at 2:26 am.

  38. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors.

    Phase 1 – Chernobyl, Fukushima, Three Mile Island, Mihama, North Ana, Davis-Besse, Turkey Point, Robinson and SONGS Unit 3 Nuclear Accidents/Incidents/Events were caused by a combination of hardware, process and human performance errors. Following the example of Arkansas Unit One Event on Easter Sunday, March 31, 2013, water from a fire main broken by a falling component or a fire damages some electrical equipment that supplies off-site power to the San Onofre Unit 2 Transformers. Unit 2, which is operating at 70% power, automatically shuts down the reactor when off-site power is lost to all reactor coolant pumps. Concurrent with reactor trip, Turbine and feedwater pumps trip.

    Phase 2 – Per NRC Information Notice No. 88-31, during the Phase 1 event, the water level on the secondary side could fall below the top of the steam generator tubes for a 10-minute period at the beginning of the event. With tubes uncovered, this condition of ZERO Water in the Unit 2 defectively designed and degraded steam generators would cause fluid elastic instability (FEI), flow-induced random vibrations, excessive hydrodynamic pressures and Mitsubishi Flowering Effect and could conceivably cause the collapse of MHI Anti-vibration structure and failure of retainer bars.

    Phase 3 – The faulted steam generator over-pressurizes due to 100% load rejection and leaking/ruptured tubes, and the main steam safety valves per SONGS procedures progressively open to prevent over-pressurizing the faulted steam generator and start releasing steam to the environment. The force of the flashing steam would create high-energy jets, lift loose parts and debris present in the steam generator. These adverse effects would cut holes into already degraded tubes and create additional loading on tube support plates (TSPs) due to heavy build-up of deposits on trefoil/quadrifoil-shaped holes from SG blowdown and crack high cycle fatigue U-bend tubes not supported by an Anti-Vibration Bars (AVB). Due to lack of in-plane restraints, large U-bends supported without AVBs and with clearances of 0.05 inches start to swing violently with large amplitudes (in-plane velocities > 60 feet/sec.) and cause several tubes to leak and with double-ended ruptures in the mid-span, free span and at the junction of 7th tube support plates in a matter of minutes due to tube-tube wear and thousands of undetected macroscopic circumferential cracks. These cumulative adverse conditions in all likelihood would result in additional massive cascading of RSGs tube failures (tubes would excessively rattle or vibrate, hitting other tubes with violent impacts) due to extremely low tube-to-tube clearances and no in-plane effective anti-vibration bar support protection system. This Titanic and adverse effect would involve hundreds of degraded and active SG tubes along with all the inactive (plugged /unstabilized) tubes causing an undetermined amount of simultaneous tube leaks/ruptures. The iodine in the reactor coolant assumed to be dissolved from allowable operational fuel failure or from an iodine spike produced by the transient conditions during the accident could be significantly larger than that previously approved NRC Limits.

    Phase 4 – The accident would transport with steam 100% of the iodine contained in the 15,000 gallons of reactor coolant to the environment exceeding the federal regulations within 10 minutes during an extremely fast-paced transient beyond the operator control and failure of defense-in-depth actions. Core Damage Probability (CDP) and Large Early Release Probability (LERP) associated with multiple double-ended rupture of steam generator tubes will significantly increase than previously approved NRC SONGS FSAR Limits. The resulting doses would be significantly higher than the dose consequences analyzed in the SONGS UFSAR for the post-trip SLB event with a concurrent iodine spike. The postulated post-trip SLB with multiple tube ruptures and concurrent iodine spike Exclusion Area Boundary, Low Population Zone, and Control Room doses would be significantly higher than the post-trip SLB Control Room limit of 5 Rem TEDE, and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE.

    Phase 5 – SCE DID Actions and unreliable operator actions to detect a leak and to re-pressurize the steam generators as claimed by Edison are not practical to stop a major nuclear accident in Unit 2 in progress in the first 5-15 minutes of the transient during the 5-month trial period.

    Phase 6 – Federal regulators signaled on April 10, 2013 that running California’s San Onofre nuclear power plant at reduced power would not pose a significant safety risk — a key step toward a possible restart of one of the idled reactors. The preliminary ruling from Nuclear Regulatory Commission staff represents a victory for operator Southern California Edison, which is pushing for a restart by June and has argued for months that the Unit 2 reactor is safe to run at lower power. But, Sen. Barbara Boxer, D-Calif., called plans to restart the plant before an investigation is complete “dangerous and premature.”

    Phase 7 – Very Strong Street Rumors are that NRC Brilliant Engineers are being pushed to the side by an adamant and very power politician connected with Edison to ignore public questioning of Edison Mistakes. This is America, not Iran or North Korea, where captured politicians for their selfish motives and hidden agendas can erode NRC’s Regulatory Authority, Safety Mission and Public Trust.

    Phase 8 – Email to Obama Campaign – Please be kind enough to deliver the following message to His Excellency, The President of United States. We need to close this Public Safety NRC Regulatory Loophole for the safety of 8.4 Million Southern Californians, a majority Democratic State.

  39. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors.

    I was in a meeting with Ted Craver few weeks ago. He was surrounded by Powerful Politicians, Bank CEO’s, Mayor’s and other Government Officials. He told the audience that Edison has monopoly franchise agreements with cities in Southern California to supply electric power. Edison owns all the transmission and distribution systems in Southern California with an investment of $20 Billion. According to the press reports, there is no shortage of power or problems with grid stability in Southern California. It appears that Edison controls CPUC and ISO; therefore, Edison will stay in the rate-base and keep supplying its generated or borrowed power to Southern California even with San Onofre Unit 2 Shutdown. Edison will continue to make money even with both San Onofre Units Shutdown, until the defectively designed and degraded RSGs are repaired or replaced. In the end, Edison will recover all the money from ratepayers, Insurance Companies and MHI.

    The question is why NRC Commission is sticking out its neck for San Onofre, which is an INPO 4 Plant, with the worst regulatory, safety, emergency preparedness, fire, cyber security, retaliation, discrimination, harassment, management and maintenance record. I am saying all this based on my firsthand knowledge as an ex-San Onofre employee, public safety expert and my observations/work as a high-energy line break, fire and emergency preparedness engineer and auditor. I am sure, that NRC will not go through so much public opposition defending another utility with such bad nuclear safety record as San Onofre. SCE Management to cover their own mistakes in the design of replacement steam generators are using their Powerful PR Machine, which is ridiculing Honorable Senator Barbara Boxer, Honorable Representative Ed Markey, Friends of the Earth, Dr. Joram Hopenfeld, Arnie Gundersen, John Large, Professor Daniel Hirsch, MHI, pressuring News Reporters not to publish anything adverse and paying Union Leaders to show up at Public Meetings.
    NRC commission has not completed review of SCE Unit 2 Return To Service Reports, SCE Response to NRR RAI’s, DAB Safety Team, Honorable Senator Barbara Boxer, Honorable Representative Ed Markey, Friends of the Earth, Dr. Joram Hopenfeld, Arnie Gundersen, John Large, Professor Daniel Hirsch’ Allegations, NRC san Onofre Special Tube Inspections and yet NRC Commission had indicated approval of SCE’s Special License Amendment designed for subverting NRC Regulatory process, public participation and legal challenges from Public Safety Experts and Attorneys . From NRC’s Commission eagerness and hasty actions, I conclude that NRC is not using its authority properly and using due diligence as an Independent Regulatory Commission in interest of Public Safety. This type of irresponsible NRC behavior directly conflicts and erodes Public Confidence in Public Statements made by you for the NRC Commission’s Safety and Oversight Charter. It gives the clear and undisputed perception that NRC has not learnt any lessons from Three Mile Island, Browns Ferry, Davis-Besse, Fukushima, Chernobyl, Mihama Unit 2, Arkansas Units 1 and 2 Recent Events, SONGS 3 Accident and Dr. Joram Hopenfelds’s concerns. 8.4 Million Southern Californians do not want a Fukushima in their backyards due to blind trust of NRC Commission in the Unsafe Gamble of Restarting of San Onofre Unit 2 for SCE Engineers to sharpen their pencils to learn from their continued mistakes. That is not how the concept of NEI, INPO and NRC well managed, safe, clean and reliable nuclear power works. Please feel free to send me an email, if you need further assistace or have any questions. Sincerely….HAHN Baba

  40. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors

    NRC Proposes No Safety Threat Finding With San Onofre. SCE once again finds shelter in NRC Commission’s favorable Findings. NRC Commission’s Finding is in direct conflict with His Excellency President of the United States, Senator Boxer and Representative Edward Markey Open Doctrine and violates the Trust, Rights and Safety of 8.4 million Southern Californians

    Summary: Edison International (EIX)’s request for a license amendment for a proposed restart of its crippled California nuclear reactor doesn’t pose significant safety risks, federal regulators said in a preliminary finding. Edison’s request to operate its San Onofre reactor at reduced power does not involve an increased risk of an accident or create the possibility of a new or different accident from those previously evaluated for its license, the U.S. Nuclear Regulatory Commission said in an e-mailed document. The NRC may approve Edison’s license amendment before the 60-day waiting period normally required after a notice is published if no hazards are found, the agency said. The public will have 30 days after the finding is published in the federal register to comment on the NRC’s conclusion before the agency makes a final determination. The document will be published next week, said Victor Dricks, a spokesman for the NRC. Without a finding of a significant hazard, the NRC can issue a license amendment before holding a public hearing, Shaun Burnie, director of nuclear campaigns at Friends of the Earth, said in a telephone interview. “It’s a get out of jail free card,” Burnie said. Edison didn’t immediately respond to a request for comment.

    An approval of the amendment won’t allow the plant to restart, the NRC has said. The agency will make a separate determination of the utility’s restart plan and it may be challenged in meeting Edison’s June deadline, Doug Broaddus, chief of the San Onofre special-projects branch for the NRC, said at an April 3 meeting.

    Senator Barbara Boxer, a California Democrat, faulted the agency’s preliminary finding.
    “The NRC staff proposal, which could pave the way for the restart of the San Onofre nuclear power plant before the investigations of the crippled plant are completed, is dangerous and premature,” Boxer, chairman of the Environment and Public Works Committee, said today in a statement. She also said the plant is in an area that’s at risk of an earthquake and tsunami. Representative Edward Markey of Massachusetts, the top Democrat on the House Natural Resources Committee, joined Boxer in criticizing the agency. Sen. Barbara Boxer blasted a preliminary finding by Nuclear Regulatory Commission staff that restarting a unit at the idled San Onofre Nuclear Generating Station would not present significant safety hazards. Boxer and Rep. Edward Markey, D-Mass, sent a letter to the NRC this week demanding that a comprehensive investigation of the plant be completed before any units be permitted to operate. They also said full public hearings should be held before a decision is made. Boxer said in response to the commission staff’s preliminary finding. “It makes absolutely no sense to even consider taking any steps to reopen San Onofre until these investigations look at every aspect of reopening the plant, given the failure of the tubes that carry radioactive water. “In addition, the damaged plant is located in an area at risk of earthquake and tsunami,” she said. “With 8 million people living within 50 miles of this plant, the staff proposal is beyond irresponsible.”
    On Monday, Southern California Edison announced it had formalized a request to amend its operating license to allow it to operate its Unit 2 reactor at 70 percent beginning June 1.
    Based on in-depth review of academic literature, all available San Onofre Reports and Tube Inspection data, and industry operating experience, it is concluded that in SONGS Unit 2 defectively-designed and degraded steam generators, even at 70% power normal power operations with proposed amended SCE License, multiple double-ended rupture of steam generator tubes can occur at any time due to anticipated operational transients and main steam line break accidents. These steam generator tubes ruptures caused by fluid elasticity in-plane tube-to-tube wear and undetected and un-quantified incubating macroscopic high cycle thermal fatigue circumferential cracks rupturing the active, worn and plugged tubes into two pieces at tube-sheets, tube-support plates, mid and free spans not supported by anti-vibration bars can result in large radiation leaks due to 100% tube-bundle uncovery, flashing sub-cooled feedwater into high dry steam and jet impingement. The steam generator with multiple tube ruptures leads to the conclusion that the steam generator could be full of high dry steam for a significant period of time. The amount of iodine released from the ruptured steam generator could be significantly larger than that previously calculated. A potential radiological accident would transport with steam 100% of the iodine contained in the 15,000 gallons of reactor coolant to the environment exceeding the federal regulations within 10 minutes during an extremely fast-paced transient beyond the operator control and failure of defense-in-depth actions. Core Damage Probability (CDP) and Large Early Release Probability (LERP) associated with multiple double-ended rupture of steam generator tubes will significantly increase than previously approved NRC SONGS FSAR Limits. The resulting doses would be significantly higher than the dose consequences analyzed in the SONGS UFSAR for the post-trip SLB event with a concurrent iodine spike. The postulated post-trip SLB with multiple tube ruptures and concurrent iodine spike Exclusion Area Boundary, Low Population Zone, and Control Room doses would be significantly higher than the post-trip SLB Control Room limit of 5 Rem TEDE, and the Exclusion Area Boundary and Low Population Zone limit of 2.5 Rem TEDE.

    Incubating macroscopic circumferential cracks caused by fluid elastic instability, flow-induced random vibrations and high cycle thermal fatigue are extremely difficult to detect and be accurately sized by nondestructive evaluation techniques like X-ray, ultrasonic, and eddy current based bobbin coil probes, mechanically rotating pancake coil (RPC), etc., which have been used in 17,000 SONGS Tube inspections. State-of-the-art systems: Zetec MIZ-80 iD system, Tecnatom TEDDY+, Circular TE and TM, transmit-receive eddy current array probe C-3 and other specialized radiographic probes capable of detecting sub-surface cracks caused by high cycle thermal fatigue have not been used in the 170,000 SONGS Tube Partial and Limited Inspections as shown below for Unit 2 due to access problems in the most problematic inner-most sections of the U-Tube Bundle, the high-cost, lack of availability of highly specialized tools and contractors, radiation doses, and time considerations in a rush to start Unit 2. The inspection scope defectively designed and degraded SONGS Unit 2 RSGs should have covered 100% hot leg and cold leg tube inspections, 100% of dents or dings, 100% of tube inspections in the tight radius U-bends, 100% area of the Top of the Tube Sheet and Tube Support Plates. None of the SCE Global Experts agree amongst themselves with the cause of tube-to-tube wear in Unit 2 and have not addressed the combined synergic effects of tube-to-tube wear and high cycle thermal fatigue cracks in their voluminous 2000 page documents. There are basic errors, invalidated assumptions, incorrect benchmarking of Unit 3 and blunders in Westinghouse, MHI, SCE, Intertek, AREVA and NRC AIT Reports. The combined effects of tube-to-tube wear and high cycle thermal fatigue cracks have been witnessed by sudden tube ruptures in North Ana in 1987 (See Item 7 below), MHI SG in Mihama, Japan in 1991 (See Item 8 below), three tube leakages in French SGs between 2004 through 2006 (See Item 9 below), 20 tube ruptures/leakages in SGs between 1980-2000 in USA, and SONGS 3 in 2012. If SCE is so confident and conservative about safety and prudent actions of their Global Experts, why not provide Independent Experts with the Units 2 & 3 Cycle 16 notarized operational, contact force, tube fatigue analysis and details of tube Inspection data? Independent Experts can be persuaded in interest of public safety to certify beyond NRC validations, whether SCE and their Global Experts are right or wrong about safe claims of restart of Unit 2?

    1. SCE states, “Rosemead, Calif. (Dec. 18, 2012) — The Mitsubishi Heavy Industry testing under review by the Nuclear Regulatory Commission (NRC) was not consulted or relied upon in developing Southern California Edison’s (SCE) proposed restart plan for Unit 2 – a plan which includes preventive tube-plugging and operating the unit at 70 percent power for a five-month period. SCE’s international team of experts conducted more than 170,000 inspections to understand the tube wear problem, and confirmed the effectiveness of the corrective actions we have identified to solve the tube wear problem. This work included three independent operational assessments of tube wear issues conducted by Areva Inc, North America, Westinghouse Electric Company LLC and Intertek/Aptech, none of whom based their review and recommendations on Mitsubishi’s testing. This was confirmed Tuesday by an NRC administrator at an NRC public meeting in Rockville, Md. SCE submitted technical information to the NRC on Oct. 3 in support of a proposed restart of Unit 2, which is safely offline. The unit will not be restarted until all plans have been approved by the NRC. The Unit 3 restart was not included in that regulatory filing and remains shut down.”

    2. SCE SONGS Unit 2 Return to Service Report, Enclosure 2, Section 6.1, “Summary of Inspection Results”, page 22 states, “ This section provides a summary of the different types of tube wear found in the SONGS Unit 2 and 3 SGs. Wear is characterized as a loss of metal on the surface of one or both metallic objects that are in contact during movement. The following types of wear were identified in the SONGS Units 2 and 3 SG tubes:

    • AVB wear – wear of the tubing at the tube-to-AVB intersections
    • TSP wear – wear of the tubing at the tube-to-TSP intersections
    • TTW – wear in the tube free-span sections between the AVBs located in the U-bend region.
    • RB wear – wear of the tubing at a location adjacent to a RB (RBs are not designed as tube supports for normal operation)
    • FO wear – wear of the tubing at a location adjacent to a FO.

    Most of the tube wear identified in the SGs is adjacent to a tube support. Figure 6-1 is a side view of an SG, showing the relationship of the tubes to the two types of tube supports: TSPs in the straight portions and AVBs in the U-bend portions of the tubes. All tubes are adjacent to many of these two types of tube supports. The RB supports are not shown because a very small number of tubes are adjacent to them. TTW indications occurred in the free span sections of the tubes. The “free span” is that section of the tube between support structures (AVBs and TSPs shown in Figure 6-1). TTW occurred almost exclusively in Unit 3 and is located on both the hot and cold leg side of the U-tube. In many cases, the region of the tube with TTW has two separate indications on the extrados and intrados of the tube. The wear indications on neighboring tubes have similar depth and position (ranging from 1.0 to 41 inches long and 4% to 100% through wall) along the U-bend, confirming the tube-to-tube contact. Table 6-1 provides the Wear Depth Summary for each of the four SGs based on eddy current examination results. Detailed results of the examinations performed are provided in the Units 2 and 3 CM reports included as Attachments 2 and 3. Figures 6-2 through 6-5 provide distributions of wear at AVB and TSP supports for all four SGs.”

    3. Attachment 2 by AREVA states, ‘The SONGS Unit 2, 2C17 inspection scope occurred in three distinct phases. The first phase followed the planned shutdown for the 2C17 refueling outage and first SG ISI. The next two inspection phases, performed in April and July 2012, were a direct result of a SG tube leak in Unit 3. The tube leak resulted from tube-to-tube wear (TTW) that was caused by fluid-elastic instability. These subsequent inspections are referred to as 2C17 RTS (Return-to-Service) inspections. The second-phase inspection (April 2012) was a full-length U-bend inspection of tubes deemed most susceptible to tube-to-tube wear based on the degradation identified in Unit 3. The third-phase inspection (July 2012) consisted of eddy current testing to measure the gaps between the AVBs and the tubes. Based on the gap measurements, an additional 104 tubes were examined in the U-bend region with the +PointTM coil. Inspections included the following inspection activities for each of the two replacement steam generators (SG 2E-088 and SG 2E-089) using site validated ECT techniques [7]:

    • Bobbin Coil Examinations – All in-service tubes, full length tube-end to tube-end – 19,454 Inspections
    •The review showed that the +PointTM probe had a slightly improved Probability of Detection (POD) over the bobbin coil. Rotating Coil Examinations – 5492 Inspections
    • Tubesheet periphery and divider lane tubes (from 3” above to 1” below the top of the tubesheet), both legs, approximately 3 tubes in from the periphery and 2 tubes in from the divider lane – 4120 Inspections
    • Hot Leg, Cold Leg, U-Bend Coil 2, Special Interest and Mag Bias – 377
    • Secondary Side Visual Examinations
    – Post sludge lancing FOSAR examination at the top-of-tube-sheet (periphery and the divider lane)
    – Visual inspections of the upper tube bundle at the 7th TSP and AVB / retainer bar regions

    4. Attachment 3 by AREVA is only applicable to unit 3, therefore are not credited.

    5. MHI inspections, although very useful per SCE Claims have not been reviewed and not credited.

    6. It is assumed, that Westinghouse and Intertek are using the inspections performed by AREVA summarized in item 2 above.

    7. SCE has not determined the true cause of Unit 3 Tube-to-Tube wear as required by NRC Confirmation Letter Action 1, which states, “Southern California Edison Company (SCE) will determine the causes of the tube-to- tube interactions that resulted in steam generator tube wear in Unit 3, and will implement actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes. SCE will establish a protocol of inspections and/or operational limits for Unit 2, including plans for a mid-cycle shutdown for further inspections.”
    8. – NRC Information Notice No. 88-31: May 25, 1988, Steam Generator Tube Rupture Analysis Deficiency

    Description of Circumstances: On July 15, 1987, a steam generator tube rupture (600 gallons per minute) event occurred at North Anna Unit 1 shortly after the unit reached 100% power. The cause of the tube rupture has been determined to be high cycle fatigue. The source of the loads is believed to be a combination of a mean stress level in the tube and a superimposed alternating stress (The mean stress is produced by denting of the tube at the uppermost tube support plate, and the alternating stress is the result of out-of-plane deflection of the U-bend portion of the tube above the uppermost support plate, caused by flow-induced vibration). Denting also shifts the maximum tube bending stress to the vicinity of the uppermost tube support plate. The rupture extended circumferentially 360ø around the tube. Based on available information, the staff concludes that the presence of all the following conditions could lead to a rapidly propagating fatigue failure such as occurred at North Anna: (1) Denting at the upper support plate, (2) A fluid-elastic stability ratio approaching that for the tube that ruptured at North Anna, and (3) Absence of effective AVB support.

    Following the steam generator tube rupture at North Anna Unit 1 on July 15, 1987, the Virginia Electric and Power Company (VEPCO) modified the flow resistance of the steam generator downcomers at North Anna by the addition of flow baffle plates. This modification necessitated the reanalysis of certain design basis events including rupture of a steam generator tube. The new analysis utilized a revised Westinghouse method for calculating steam generator water mass and indicated that during the event, the water level on the secondary side could fall below the top of the steam generator tubes for a 10-minute period at the beginning of the event.

    Steam generator tube uncovery is significant because, if the break location becomes uncovered, a direct path might exist for fission products contained in the primary coolant to be released to the atmosphere without partition with the secondary coolant. VEPCO and Westinghouse reanalyzed the design basis steam generator tube rupture accident for Surry using the revised methods and determined that the steam generator tubes at Surry could also become uncovered even though the Surry plants were not modified by the addition of flow baffle plates.

    The licensee further concluded that the offsite dose consequences exceeded those calculated in the Surry Updated Final Safety Analysis Report (UFSAR) because tube uncovery could produce a direct path for fission product release. Based on the Surry results, the analysis of steam generator inventory during a steam generator tube rupture at other plants may show that the steam generator tubes may uncover. Thus, for those plants where the steam generator tubes were thought to remain covered following tube rupture, the previously calculated safety analysis offsite doses might be exceeded and since the primary coolant activity limit in Technical Specifications is based upon the occurrence of this accident, the allowable technical specification limit may be too high.

    Discussion: A postulated steam generator tube rupture is one of the design basis accidents analyzed in plant Safety Analysis Reports (SARs). Using conservative assumptions of single failure and loss of offsite power, it must be shown that the offsite dose consequences will be limited to the guideline doses of 10 CFR 100 or a fraction of the guideline doses depending on the assumptions made for iodine spiking. The iodine in the reactor coolant may be previously dissolved from allowable operational fuel failure or may result from an iodine spike which is the sudden increase in coolant iodine concentration produced by the transient conditions during the accident. Mechanisms for transport of the iodine that exits the reactor system to the atmosphere are discussed in Standard Review Plan (NUREG-0800) Section 15.6.3. In determining the amount of iodine that is transported to the atmosphere, credit may be given for “scrubbing” of iodine contained in the steam phase and in the atomized primary coolant droplets suspended in the steam phase for release points which are below the steam generator water level.

    The Surry UFSAR assumed that the break is always covered with water so that 99% of the iodine would remain within the steam generator coolant and only 1% would be released through the atmospheric relief valves. The break location is assumed to be always covered in the UFSAR calculations because an initial steam generator water mass that may be non-conservatively large was assumed in order to conservatively account for the possibility of overfill and because steam generator tube failures were thought only to occur close to the tube sheet. The North Anna tube rupture demonstrated that steam generator tube failures can occur near the top of the tube bundle. The revised steam generator water mass calculations by Westinghouse with the assumption that the break occurs at the top of the tube bundle led to the conclusion that the break could be uncovered for a significant period of time. Tube uncovery occurs because of the level shrink that accompanies reactor trip/turbine trip during the tube rupture event. The tubes would be recovered by the flow of auxiliary feedwater into the ruptured steam generator and by the reactor coolant which would be added due to the ruptured tube; however, the amount of iodine released from the ruptured steam generator could be larger than that previously calculated.

    9. – MHI SG Tube Rupture Mihama, Japan 1991: On February 9th, 1991, a heat transfer tube (SG tube) in a steam generator of the No. 2 pressurized water reactor at the Mihama nuclear power station of the Kansai Electric Power Company broke off during a rated output operation. As a result, about 55 tons of primary cooling water leaked out from the SG tube into the secondary cooling loop, and the reactor was scrammed by operation of the ECCS (Emergency Core Cooling System). The failure of the SG tube was caused by fretting fatigue resulting from contact of the SG tube with the supporting plate for the SG tubes, because the AVB, which functions to prevent flow-induced vibration, was not inserted deep enough onto the SG tubes in the steam generator. The scale of the accident was ranked “level 3” on the international nuclear events scale (INES). At 13:40, an alarm of a condenser air off take system went off during a rated output operation, warning that the coolant water level in the steam generator was decreasing. At 13:50, an automatic emergency shutdown of the reactor was triggered by the signal of decreasing pressure in the pressurizer. After seven seconds, the ECCS was automatically operated, and coolant water was flooded into the reactor by a high pressure injection pump. However, one main steam isolation valve and one pressurizer relief valve could not be operated by remote control. Therefore, the valve operation was carried out manually. The failed tube was removed from the heat exchanger, and the fracture surface was examined by a scanning electron microscope. Striations, which are a characteristic of fatigue failure, were observed on large portions of the fracture surface, and dimples showing tensile fracture were also observed. However, few traces of stress corrosion cracking and corrosion were found on the fracture surface of the tube. The failure of the tube was, therefore, hypothesized to be due to cyclic loading. The morphology of the fracture surface of the SG tube shows a typical example of the striations formed on the fracture surface of the SG tube. Examination of the other SG tubes near the failed tube showed traces of wearing formed by fretting due to contact between the tubes and the anti-vibration bars on the outer surfaces of the tubes. Stress amplitude of the failed tube estimated based on the striation spacing was found to be in the range of around 51 to 60 MPa. (8.4 Ksi). Occurrence of cyclic loading in the SG tube that had failed was related to the insertion depth of the anti-vibration bar, AVB. The SG tubes were subjected to vibrations due to the flow of secondary coolant outside of the SG tubes. In order to prevent the flow-induced vibration, V-shaped AVBs were installed onto the opposite U-bent SG tubes near the upper part of the steam generator. However, the insertion depth of the AVB for the SG tubes was not enough, because the engineers who installed the AVB did not fully understand the importance of the AVB. In fact, no damage was founding in the SG tubes into which the AVB were inserted to sufficient depth as shown by the design guidelines. Accordingly, the SG tubes were subjected to flow-induced vibration and strongly contacted with the sixth supporting plate, so that the SG tubes incurred damage by fretting fatigue. Inspection of the AVB had not been carried out since installation. In order to provide an opportunity to learn from the accident that resulted in the leakage of primary coolant from the SG tube due to fretting fatigue, the damaged steam generator has been preserved in an exhibition at the Mihama station of the Kansai Electric Power Company. An exhibition is a good way to help everyone to good lessons from an accident. This accident was the first disaster in Japan that resulted in actuation of the ECCS due to leakage of primary coolant in the steam generator. Therefore, the accident caused social concern with nuclear reactors. The international nuclear events scale (INES) is defined by the IAEA to assure coherent reporting of nuclear accident by different official authorities. The INES is characterized from level one to level seven. The level number increases with the scale of the accident. For example, level one is a minor event, and level seven is major accident. The scale of the accident in 1979 resulting in the loss of coolant that occurred in Three Mile Island was ranked level five by the IAEA. The accident reported here was ranked level three.
    10. Cruas NPP: Between 2004 and 2006, three primary-to-secondary leaks occurred at the Cruas NPP: unit 1 in February 2004 and unit 4 in November 2005 and February 2006. The three leaks were all the result of a circumferential crack in the tube at the location where the tube passes through the uppermost tube support plate (TSP #8). Analyses carried out by EDF, further to the last two events, resulted in them being attributed to high cycle fatigue of steam generator tubes due to flow-induced vibration. The results of in situ examination initiated by the Cruas NPP operator showed that the flow holes of the uppermost Tube Support Plates (TSPs) were partially or completely blocked by corrosion products. This phenomenon is referred to in this paper as TSP “clogging-up” and it was considered potentially generic for EDF NPP fleet. For the Cruas leakages, it was established that the association of TSP clogging-up and the specificity of the Cruas steam generator (central area in the tube bundle where no tubes are installed) were responsible for a significant increase in the velocity of the secondary fluid in the tube bundle central area. The high velocity of the fluid in this region increases the risk of fluidelastic instability for the tubes. A new fact caused the leakages observed on the Cruas units: the heavy build-up of deposits on the secondary side of the steam generator which changed the flow conditions in the center of the tube bundle. The deposits reduced or blocked water/steam flow through the quadrifoil-shaped holes in TSPs, forcing more water and steam into the center of the tube bundle, which caused the excessive vibration of the tubes near the center of the tube bundle. This excessive vibration due to fluid-elastic instability resulted in fatigue cracking of the tube.

  41. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.

    San Onofre NRC/SCE/MHI/Public Awareness Series – Please excuse me for any computer or human performance grammatical or spelling errors

    Press Reports and Friends of the Earth state, “Southern California Edison, the majority owner and operator of the idled San Onofre Nuclear Generating Station, announced Monday that it has formally proposed changes to its operating license in an effort to start generating electricity at the plant. SCE is hoping that the amendments, which authorize the plant to operate at 70 percent of generating capacity, will be approved by late May. If the NRC grants the necessary approvals, San Onofre could start generating electricity as soon as June 1. SCE contends that the conditions that cause the vibration don’t occur at 70 percent power. The utility plans to run Unit 2 for five months and then shut it down for inspection of the tubes. After the examination, the reactor would resume operating at 70 percent. Data collected during the inspection will determine a long-term power level, according to Edison. Edison wants the NRC to act as “a rubber stamp. It is a request and a red herring to divert attention from major unresolved safety issues and circumvent meaningful public participation. It would be an outrage and a betrayal of the public’s trust if the NRC were to concede to Edison’s demands.” NRC is paid by the American Tax Payers and American Nuclear Utilities to ensure generation of safe and reliable nuclear power. To side with Southern California Edison and an anticipated permission by NRC Commission to grant permission to restart of Unsafe Unit 2 without resolving the Unit 2 safety issues and a Public Hearing would mean: (1) Pressure by Edison Sympathetic Powerful Politicians, Lobbyists, Financial Institutions, Unknown NRC Commissioners and Appointed Government Officials is working for SCE profits over safety, (2) It would be a betrayal of His Excellency, President of United States, Honorable Senator Barbara Boxer and NRC Chairman’s Policies, Commitments and Standards, and (3) An Insult to 8.4 Million Southern Californians, United States Congress, NRC’s Safety Mission and Transparent Public Policy, US Constitution, American Democratic Principles, and Committed Utilities and Brilliant NRC Engineers dedicated with the charter of ensuring public safety and generation of reliable power.

  42. How can NRC believe any reports from the builders and designers, I hope they are double checking every aspects of the plant and not rely on reports that are not supervised by NRC, reports can be skewed to get closer to what they want it to show to the over seers, will they lie because they think they are right and want to show results that favor the NRC to allow restarting.

  43. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC for posting this blog.
    San Onofre NRC/SCE/MHI/Public Awareness Series – by Hahn Baba

    Steam generator tubes leak and rupture despite inspection and plugging. Question is can the operator detect the leak and shutdown the plant. That is a 95% probability of success and 5 percent chance of failing. There have been more than 20 leaks and tube ruptures in the last 30 years in USA with timely detection and shutdown with no reported offsite releases affecting the public or exceeding federal limits. San Onofre MHI generators, I cannot say.. Even at 70% power, there is a much higher chance of multiple leaks and tube ruptures in Unit 2 due to manufacturing defects, cracked and plugged tubes, operational transients, mother nature’s mood, equipment malfunctions and operator errors. SCE has to either satisfy NRC, Dr. Joram Hopenfeld and the Public about safety of Unit 2, repair or replace the defective generators or decommission the plant. So far SCE/MHI/Intertek, Westinghouse and AREVA response is totally unsatisfactory and unconvincing. Customer service and safety always prevails over SCE profits, if NRC and CPUC rules are followed to the Letter…Let us see what NRC Commission does to satisfy 8.4 Million southern Californians, who pay the bills for San Onofre and SCE Management.

  44. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC posting this blog. San Onofre NRC/SCE/MHI/Public Awareness Series – by Hahn Baba

    Life is a unique opportunity to serve the society. Society needs Energy, which is safe, economical and reliable. Every form of Energy has drawbacks and risks. SCE is responsible for safety, economics and reliability of Unit 2.

    NRC Top SG Expert states (February 2013), “With multiple tube ruptures, you’d have an earlier plant transient and you’d be able to identify the tube or the generator quicker, safety systems would react; so multiple tube ruptures would challenge the operators in a different way. But we have studied that from a risk perspective and we chose not to take regulatory action or regulatory action wasn’t necessary. So in some aspects the operators were benefitted by automatic systems and easier diagnosis, but the timing would create another challenge for them, so..”

    Circumferentially cracked tubes can rupture without notice at any time during Unit 2 reduced steady state 70% power perations, anticipated operational transients and main steam line breaks. The additional stresses and jet impingement loads may cause other tubes to rupture and cut into two pieces in a matter of minutes. NRC Rules governing reactor operations simply do not contemplate cascading tube ruptures. Therefore, San Onofre emergency core cooling systems are not designed to prevent a core meltdown if a number of tubes rupture at the same time. Therefore, In accordance with NRC Fukushima Task Force Lessons Learnt, Dr. Joram Hopenfeld’s Analysis and observation of SONGS Operators Poor Performance/Equipment Maintenance/Reliability for the last six years, SONGS Operators and emergency core cooling systems are not capable of preventing a core meltdown caused by multiple cracked tube ruptures in defectively designed and degraded unit 2 SG caused by fluid elastic tube-to-tube wear and undetected high cycle thermal fatigue cracks.

    Circumferential cracks are more serious than axial cracks because of concerns for double-ended rupture of steam generator tubes and consequent large leaks. In addition circumferential cracks are considered more difficult to detect and accurately size by nondestructive evaluation techniques like X-ray, (local) ultrasonic, and eddy current based bobbin coil probes, mechanically rotating pancake coil (RPC), etc., which have been used in SONGS Tube Inspections for Unit 2. Circular TE and TM, transmit-receive eddy current array probe C-3 and other specialized radiographic probes capable of detecting sub-surface cracks have not been used in the 170,000 SONGS Tube Inspections for Unit 2 due to cost and time considerations. None of the SCE Global Experts have addressed the combined synergic effects of tube-to-tube wear and high cycle thermal fatigue cracks in their voluminous 2000 page documents. There are basic errors and blunders in Westinghouse, MHI, SCE, Intertek, AREVA and NRC AIT Reports. The combined effects of tube-to-tube wear and high cycle thermal fatigue cracks have been witnessed by sudden tube ruptures in North Ana in 1987, MHI SG in Mihama, Japan in 1991, three tube leakages in French SGs between 2004 through 2006, 20 tube ruptures/leakages in SGs between 1980-2000 in USA, and SONGs 3 in 2012. Because of basic SCE/MHI mistakes, continued cover-ups and subverting the regulatory process since 2004, ratepayers have lost several Billion Dollars and Public Safety has been compromised. If SCE is so confident and conservative about safety and prudent actions of their Global Experts, provide our experts with the Units 2 & 3 Cycle 16 notarized operational, contact force, tube fatigue analysis and tube Inspection data and we will certify beyond NRC ,,,,,,,,,,,,,,,whether SCE and their Global Experts are right or wrong about safe restart of Unit 2?

  45. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC posting this blog. San Onofre NRC/SCE/MHI/Public Awareness Series – by Hahn Baba

    NRC, INPO, CPUC, NEI and Scientists expect SCE to supply safe and reliable power at a reasonable cost and not conduct unsafe experiments at the expense of public safety and charge ratepayers for its mistakes.

    San Onofre Restart quoting Albert Einstein, “Insanity: doing the same thing over and over again and expecting different results.”

    Root Causes are defined as the basic reasons (e.g., hardware, process, or human performance) for a problem, which if corrected, will prevent recurrence of that problem.

    MHI Root Cause: Insufficient programmatic requirement to assure effective AVB contact force to prevent in-plane fluid elastic instability and random vibration and subsequent wear under high localized thermal-hydraulic conditions (steam quality (void fraction), flow velocity and hydro-dynamic pressure).

    HAHN Baba Rebuttal: MHI Answer is Incorrect. Contact force is the force in which an object comes in contact with another object. Examples are pushing a car up a hill or kicking a ball or pushing a desk across a room are some of the everyday examples where contact forces are at work. In the first case the force is continuously applied by the person on the car, while in the second case the force is delivered in a short impulse. The most common instances of this include friction, normal force, and tension. According to forces, contact force may also be described as the push experienced when two objects are pressed together. The MHI designed AVBs had zero contact forces in Unit 3 to prevent in-plane fluid elastic instability and subsequent wear under high localized thermal-hydraulic conditions (steam quality (void fraction) and flow velocity). Large u-bends were moving with large amplitudes in the in-plane direction without any contact forces imposed by out-of the plane restraints. The in-plane vibration associated with the wear observed in the Unit 3 RSGs occurred because all of the out-of-plane AVB supports were inactive by design in the in-plane direction. The Unit 3 tube-to-AVB contact forces on the TTW tubes were Zero, that is why did not restrain the tubes in the in-plane direction (Like a Sports Car moving with very high speed in freeway express toll lanes passing by a Stalled Police Car). In-plane fluid elastic instability did not happen in Unit 2 because of operational differences, so therefore double contact forces and better supports is conjecture in Unit 2 and a pre-planned and ill-conceived SCE reason to justify restart of Unsafe Unit 2. The baseless contact force theory based on hideous statistical data and manufacturing simulations capable of stopping super express velocity induced in-plane vibration is refuted based on in-depth review of Speculative and Incomplete SCE Root Cause Evaluation, Dr. Pettigrew’s 2006 Research Paper, Westinghouse, AREVA, John Large and earlier version of MHI Reports.

    NRC AIT Team SCE Root Cause: The combination of unpredicted, adverse thermal hydraulic conditions and insufficient contact forces in the upper tube bundle caused a phenomenon called “fluid-elastic instability” which was a significant contributor to the tube to tube wear resulting in the tube leak. The team concluded that the differences in severity of the tube-to-tube wear between Unit 2 and Unit 3 may be related to the changes to the manufacturing/fabrication of the tubes and other components which may have resulted in increased clearance between the anti-vibration bars and the tubes; (3) Due to modeling errors, the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.

    HAHN Baba Rebuttal: Except contact forces described above, NRC/SCE cause evaluation on SCE created adverse thermal–hydraulic conditions and Mitsubishi faulty computer modeling in Unit 3 is correct. SCE Engineers were running Unit 3 in Test Mode with higher steam flows to check the improvements in Tube-to-AVB support contact forces. But like earthquake and Tsunami in Fukushima and Fire in Chernobyl, the misunderstood experiment destroyed Unit 3. First rule of thumb, like Dr. Pettigrew says, “Avoid in-plane and random vibrations by keeping velocities below 20 feet/sec (SONGS in-plane velocities > 56 feet/sec, out-of-plane velocities > 28 feet/sec). Second rule, “Operate SG at higher pressures (> 900 psi) and circulation ratios > 4 to keep the void fractions less than 98.5%. Third Rule,” Ask Westinghouse/Combustion Engineering how to design anti-vibration bars, which are capable of preventing fluid elastic instability.” Fourth Rule, “ATHOS Models cannot calculate in-plane velocities.”

    HAHN Baba Contributing Causes: Too many adverse Design Changes to produce more megawatts, Adverse Operational Parameters, and Human Performance Errors (Lack of Critical Questioning & Investigative Attitude, Lack of Solid Teamwork & Alignment between SCE/MHI Team, Lack of Adequate research and Industry Benchmarking (e.g., NUREG-1841, Palo Verde, etc.) complacence, time pressure)

    HAHN Baba Root Cause: Nuclear Safety was compromised Lack of Critical Questioning & Investigative Attitude and undermined by Profits Motivations

    Dr. Pettigrew’s Advise*: To prevent the adverse effects of fluid elastic instability and flow-induced random vibrations, need Solid Teamwork & Alignment between Designer & Manufacturer.

    * World’s Foremost Renowned and Canadian Research Scientist Professeur Titulaire, Michel J. Pettigrew advise for the last 40 years in a 1976 address to the Canadian Atomic Energy Commission, “Most flow-induced vibration problems, which can be avoided provided that nuclear components are properly analysed at the design stage and that the analyses are supported by adequate testing and development work when required.” SCE/MHI AVB Design Team in 2005 rejected recommendations to reduce high void fractions, which caused fluid elastic instability in Unit 3. The recommendations were rejected by MHI/SCE Team, because it would have cut down the profits due to less electricity production, cost more money to implement changes discussed in MHI Root Cause Evaluation Report, delayed the fabrication and installation process and Triggered a Lengthy NRC 10CFR 50.90 License Amendment and Public Hearing Process. SCE/MHI subverted intentionally the regulatory process. That is what Barbara Boxer was saying.

    Here are more quotes from SONGS Insiders:

    1. To the best of my recollection, the Root Cause Member told me that Unit 2 was running at higher pressures than Unit 3, that is why Unit 2 did not experience FEI. He had a 2006 paper with him published in 2006 by Dr. Pettigrew in his hand, which warned about the ineffectiveness of the flat bars to prevent fluid elastic instability. He was researching on curved bars, bars with springs, which could be attached to the tubes to prevent in-plane vibrations and repair the RSGs. What the Root Cause Member said matches with SONGS Procedures, Plant Briefing Sheets and NRC AIT Report Data.

    2. The Root cause Team leader told to a friend of mine, “I wish that SCE engineers would have made these design changes one at a time and tested them instead of making all the changes at one time.”

    3. One of the very highly placed SCE Manager and Corporate Emergency Director (Now retired) told me that all these changes were made without much thought and analysis, which consisted of the substitution of Inconel 690 for Inconel 600 as the tube material. Inconel 690 is more resistant to corrosion than Inconel 600. However, Inconel 690 has a thermal conductivity approximately 10% less than that of Inconel 600. The requirement that the SG’s thermal performance be maintained, in conjunction with maintaining a specified tube plugging margin, SCE told MHI for increasing the tube bundle heat transfer surface area from 105,000 ft2 to 116,100 ft2 (an 11% increase).

    4. One of the very highly placed SCE Manager was shaking his head, when he told me, “I wish that SCE Engineers would have duplicated the Palo Verde Replacement Steam Generators and we would not be experiencing this embarrassing day. Combustion Engineering not only designed and replaced six Palo Verde steam generators with considerable improvements and higher thermal megawatt, but solved the problem with the original steam generators.” Please note that San Onofre and Palo Verde Original Steam Generators were designed and fabricated by Combustion Engineering, but Palo Verde steam generators are largest in the world. The Palo Verde Replacement Steam Generators are running fine for the last 10 years without any plugged tubes and San Onofre, everybody knows the story. Now the question is that SCE owns 20% of the share of the Palo Verde and how come SCE Engineers did not contact their counterparts – Answers in the Next Update

  46. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC posting this blog. San Onofre NRC/SCE/MHI/Public Awareness Series – by Hahn Baba

    NRC, INPO, CPUC, NEI and Scientists expect a responsible nuclear utility to supply safe and reliable power at a reasonable cost and not conduct unsafe experiments at the expense of public safety and charge ratepayers for its mistakes.

    San Onofre Restart quoting Albert Einstein, “Insanity: doing the same thing over and over again and expecting different results.”

    NO SIGNIFICANT HAZARDS CONSIDERATION – ANALYSIS OF SCE’S NEW LICENSE AMENDMENT

    Dr. John Hopenfeld states, “A steam generator, in addition to providing a barrier to radioactivity and producing steam, has many other important functions. It is the major component in the plant that contributes to safety during transients and accidents. It provides the driving force for natural circulation and it facilitates heat removal from the reactor core during a wide range of loss of coolant accidents. Proper steam generator operation is of major safety significance and therefore any changes to its design may have potential safety consequences. Southern California Edison(“SCE”) has not identified the root cause for the unusually excessive tube wears in the four steam generators (“SGs”) of units 2 and 3 at San Onofre Nuclear Generating Station, SONGS. Based on my evaluation of the tube wear data and the in situ leak test results it is my opinion that restarting Units 2 and 3 would compromise public safety. The new components in the replacement steam generators (“RSGs”), constituted a major change to the original SGs, this lead to vibrations and the unusual rapid tube wear. The components causing the wear would have to be replaced and the SONGS license amended before the units can be restarted. SCE and MHI did not provide any data to support their contention that the various design change options that were discussed in 2006 by the AVB Design Team would have had no significant effect on flow velocities and steam quality. SCE consultants AREVA, Westinghouse and MHI differed on the root cause of tube vibrations and none of their opinions were based on sound scientific principles. The safety consequences of operating with degraded tubes are more serious than envisioned by the consultants. It is apparent that SCE focused its attention on explaining away errors in the design, fabrication and management of the RSGs. There is no indication that any consideration was given to the long-term safety risk of operating Units 2 and 3, with each containing more than 1500 defects (3401 in all).”

    Let us examine, why Dr. Hopenfeld, Arnie Gundersen, John Large, Professor Daniel Hirsh and many other experts are saying that it is not safe to operate at Unit 2 at 70% power, while SCE, AREVA, Westinghouse, MHI, Intertek and the NRC Augmented Inspection Team say that Unit 2 reactor operation at no more than 70 percent power will limit unusual tube wear and is safe to operate by reviewing the following questions:

    Let us assess the condition of the defectively designed and degraded U2 RSGS, before we answer the following questions:

    • A steam generator, in addition to providing a barrier to radioactivity and producing steam, has many other important functions. It is the major component in the plant that contributes to safety during transients and accidents. It provides the driving force for natural circulation and it facilitates heat removal from the reactor core during a wide range of loss of coolant accidents. Proper steam generator operation is of major safety significance and therefore unanalyzed and untested design changes to SONGS RSGs have created major safety consequences as observed by damages in Units 2 & 3.

    • The Root Cause determined by SCE and MHI for both Units 2 & 3 RSGs does not address the exact reason for RSG design and operational flaws. Root Cause is defined as the exact reason (e.g., hardware, process, or human performance) for a problem, which if corrected, will prevent recurrence of that problem. Therefore, SCE/MHI have not determined the exact root cause of the tube-to- tube wear in Unit 3 per CAL ACTION 1, and have not implemented actions to prevent the loss of tube-to-tube wear and demonstrated via a deterministic safety analysis that the AVB structural integrity in the Unit 2 steam generator will be maintained (e.g., collapse of AVB structure and retainer bars failure due to fluid elatic instability) due to Main Steam Line Break (e.g., Mihama, Turkey Point, Robinson), Station Blackout (Fukushima), SG Tube Ruptures ( Mihama, SONGS 3 & 20 other Incidents in US/Europe in the last 20 years) and other anticipated operational transients at power operation between 70 t0 100% power.

    • Based on the evaluation of the tube wear data and the in situ leak test results, restarting Units 2 and 3 would compromise public safety. The new components in the replacement steam generators constituted a major change to the original SGs, which lead to the unusual and rapid tube wear. The components causing the wear would have to be replaced and the SONGS license amended before the units can be restarted. The above assessment also applies to SCE proposed five-month test of Unit 2 at 70% of licensed power. After correcting an error in the SCE stress calculations, the present analysis shows that because of the wear damage previously sustained by Unit 2 some tubes will be susceptible to rapid fatigue failure. The tubes will exceed their allowable fatigue life by 22 to 29% during the next operating cycle. The risk of tube rupture increases with operating time but the MHI and Intertek analysis is not capable of quantifying it in terms of operating time. Unit 2 should not be permitted to operate until SCE provides a thorough assessment of the fretting fatigue discussed in Dr. Joram Hopenfeld’s Reports.

    • To meet the performance criteria as specified in the plant’s technical specifications (“TS”) it is necessary to relate tube defect geometry to primary/secondary side leakage. When tube degradation is caused only by thinning due to tube-to tube wear, the degree of tube damage can be assessed by modeling, pulling selective tubes for testing or by in-situ pressure testing. In addition to losing strength due to wall thinning, some of the tubes at SONGS have used up a significant fraction of their allowed fatigue life. Such damage cannot be detected by even the NRC Special Tube Inspections due to time, cost, unavailability of high technology probes, contactors, and/or impossible access within the tube bundle or radiation dose limitations. These tubes are significantly susceptible to sudden ruptures without notice and/or early warnings during steady state normal operations at 70% power, Operational Transients (opening or closing of valves, scrams, loss of offsite power, moderate earthquakes, etc.) and under steam line break accidents at other reactors. The stress on the tube [SONGS Unit 3 Tube, which leaked, Row 106 Column 78, 100 percent through wall wear, length of wear – 29 inches] due to in-plane vibration calculated by MSI was 4.2 ksi and shown to be under fatigue limit (13.6 ksi).” In contrast, Dr. Hopenfeld’s calculations in the attached report demonstrate that the stress concentration factor is much higher than calculated by MHI and therefore the fatigue limit of 13.6 ksi will be exceeded by at least 22% (16.7 to 17.5 ksi). Now, the ruptured tube in Mihama experienced out-of-plane FEI and according to the latest research between 2006 -2011 by Dr. Pettigrew and others, the in-plane velocities are double the out-of plane velocities, therefore, during FEI conditions, tubes can realistically experience fatigue of 16.8 ksi. This demonstrates that MHI calculations are under-conservative by a factor of 4, just like the fluid velocities calculated earlier by MHI, which led to the SONGS Unit 3 tube leak, failure of 8 tubes under MSLB testing conditions and loss of more than 35% wall thickness in almost 400 tubes. Fatigue damage has resulted in several tube failures in US, French and Japanese steam generators, yet in spite of its importance, none of the SCE consultants included fatigue in their evaluation of restarting Unit 2.

    • Between 2004 and 2006, three primary-to-secondary leaks occurred at the Cruas NPP: unit 1 in February 2004 and unit 4 in November 2005 and February 2006. The three leaks were all the result of a circumferential crack in the tube at the location where the tube passes through the uppermost tube support plate (TSP #8). Analyses carried out for the last two events, resulted in them being attributed to high cycle fatigue of steam generator tubes due to flow-induced vibration.

    • On February 9th, 1991, a heat transfer tube (SG tube) in a Mitsubishi steam generator of the No. 2 pressurized water reactor at the Mihama nuclear power station of the Kansai Electric Power Company, broke off during a rated output operation. As a result, about 55 tons of radioactive primary cooling water leaked out from the SG tube into the secondary cooling loop, and the reactor was scrammed by operation of the ECCS (Emergency Core Cooling System). The failure of the SG tube was caused by fretting fatigue resulting from contact of the SG tube with the supporting plate for the SG tubes, because the AVB, which functions to prevent flow-induced vibration, was not inserted deep enough onto the SG tubes in the steam generator. One main steam isolation valve and one pressurizer relief valve could not be operated by remote control. Therefore, the valve operation was carried out manually. The amount of steam released from the main steam relief valve to atmosphere was about 1.3 tons. The amounts of radioactive rare gas and iodine discharged to the atmosphere were about 2.3E10 and 3.4E8 becquerels, respectively. The scale of the accident was ranked “level 3″ on the international nuclear events scale (INES). Stress amplitude of the failed tube estimated based on the striation spacing was found to be in the range of around 51 to 60 MPa (8.4 ksi).

    • Unit 2 RSGs have 505 plugged and/or staked tubes. Inspections reveal that there are numerous U-bends in both RSGs with tube-to-tube clearances as small as 0.05 inches (SONGS RSGs Design 0.25 inches, Industry NORM 0.35-0.55 inches). The design distance between tubes on the sides at the intersection with the top TSP should be 0.250 inch plus or minus the small broached hole tolerance. The Nominal Gap between tube and AVB in SONGS RSGs built by Mitsubishi is 0.002″, while in Fort Calhoun RSGs (another US plant built by Mitsubishi) the Nominal Gap is 0.0031”. The tube diameter (d) and pitch (P) Tube Index (P/d) in SONGS RSGs built by Mitsubishi is 1.33-1.433, whereas tube diameter (d) and pitch (P) Tube Index (P/d) in Arkansas Nuclear One Unit 2 RSGs built by Westinghouse is 1.518-1.672. The in-plane tube spacing at Apex in SONGS RSG is (0.298, 0.344, 0.400 inch) whereas at another plant, it is (0.442, 0.502, 0.562 inches). According to Westinghouse, the actual distance may be between 0.040 and 0.120 inch (1-3 mm). According to AREVA, “The nominal distance between extrados and intrados locations of neighboring U-bends in the same plane ranges from 0.25 inches to 0.325 inches due to the tube indexing. There are instances where the closest approach distance is less than this value, based on field measurements using bobbin coil ECT. The bobbin probe on the 140 kHz absolute channel can detect neighboring U-bends if the separation distance is less than approximately 0.15 inches. Using a proximity signal calibration curve, the separation distance between U-bends was measured for all steam generators. The smallest detected U-bend separation distance is close to contact. There are 36 U-bends in Unit 2 SG E-088 and 34 in SG E-089 with a separation less than or equal to 0.050 inches. The separation of the U-bends in Unit 2 with TTW is 0.190 inches as measured by UT. The U-bends with the smaller separation distances are much better candidates for wear by rubbing yet do not exhibit TTW. Intertek states, “There were 4348 indications detected at AVB contact points with a maximum NDE depth of 35%TW during the end-of-cycle (EOC) 16 tube examinations. Wear at tube support plates (TSP) was also detected (364 indications) with a maximum NDE depth of 20%TW.” MHI Tube Plugging Criteria states, “Tubes, which exhibit a potential for losing their integrity during the next operating period due to progressive through-wall wear and/or susceptibility to FEI should be plugged. All tubes with ECT tube-tube wear indications in the free span section should be plugged regardless of the wear depth. Furthermore, tubes with wear indications at the AVB and TSP locations, which are similar to those on the tubes with the wear indication in the free span section, should be preventatively plugged. Tubes with Tube-AVB/TSP wear equal to, or greater than, 35% should be plugged in accordance with Technical Specifications.” Approximately ~ 1600 tubes in Unit 2 steam generators were found with wear indications. Only ~500 tubes out of these were plugged due to wear (Tube-to-Tube/AVB/TSPs). That means ~ 1080 tubes in Unit 2 steam generators with wear indications have not been plugged. These plugged tubes will continue to wear and requires a safety analysis to demonstrate that these tubes will not cause damage to adjacent tubes during accidents and meet the present licensing requirements.

    •The following very important issues have not been addressed in Unit 2 Return to Service reports.
    1. Long term Implications of restarting any of the SONGS units.
    Because of the large number of defects in the tubes, there is a risk that additional tubes will have to be removed from service even if the FEI problem has been solved. Defects such as now exist (more than 3000) are known to form nucleation sites for stress corrosion (SCC) and fatigue cracks. It is important to understand that even though alloy 690 is not as prone to SCC as alloy 600 it is not completely immune to SCC. This problem will become more important as the units age because of crud build up in the tube support plates.
    2. 2. Uncertainties in the analyses of Fluid Elastic Instability Analysis (FEI) Long Term Effects.
    The consultants did not address uncertainties in sufficient details. Because of the complexity of the technical issues a reviewer, therefore, cannot asses of the robustness of the analysis. John Large states, “A difficulty that I have with the AREVA and, generally, with the other OAs is that whereas the results of analyses, particularly relating probability and confidence, are often resolutely stated, very little of the analytical procedures arriving at the results are open to inspection. Because of the uncertainties I very much doubt that in the present circumstances tube structural integrity could be guaranteed to satisfy the 95% probability at 50% confidence criterion but, that said, AREVA presents no substantial data that enables me to explore and possibly resolve these doubts.” Because the uncertainties in predicting how fatigued tubes can propagate failures, it is impossible to assess quantitatively safety risks. It is believed that a main steam line break (MSLB) represents a bounding case. A conservative estimate of the probability of a large early release of radiation with containment bypass would be 10E-4 per year for any operating cycle. Such risk exceeds NRC’s safety goals. Units 2 and 3 at SONGS have the highest risk of tube rupture related core damage than any other nuclear power plant in the USA.

    John Large states, “SCE’s assertion that reducing power to 70% will at the best alleviate, but not eliminate, the TTW and other modes of tube and component wear is little more than hypothesis – the supporting Operational Assessments and analyses have not proven it to be otherwise. I am of the opinion that trialling this hypothesis by putting the SONGS Unit 2 back into service will, because of the uncertainties and unresolved issues involved, embrace a great deal of change, test and experiment. The terms of the Confirmatory Action Letter of March 11 2012, are versed such that to meet compliance the response of SCE via its Return to Service Report,11 must include considerable changes of conditions and procedures that are outside the reference bounds of the present FSAR – this is because the physical condition of the RSGs, and the means by which this is evaluated and projected into future in-service operation, have substantially and irrevocably changed since the current FSAR was approved. The fact that SCE fails to satisfy the requirements of the CAL is neither here nor there, although it illustrates the scope and complexity of the response required. At the time of preparing the CAL, the NRC being well-versed in the failures at the San Onofre nuclear plant, surely must have known that the only satisfactory response to the CAL would indeed require considerable changes, tests and experiments to be implemented.”

    1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

    SCE’s Answer: No, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
    Basis – Please see SCE License Amendment (www. songscommunity.com)
    HAHN Baba Response: SCE, its Independent Experts and NRC Augmented Inspection Team investigations and their answers are incorrect and misleading like all the past actions pertaining to: (1) Inability to determine the true reasons for faulty replacement steam generator design, (2) Inability to determine the true root cause of tube-to-tube wear in Unit 3 and take corrective actions to prevent recurrence of tube-tube wear in Unit 2. Because the true causes have not been determined, that is why both San Onofre Units are shutdown, and NRC/SCE/MHI are facing so much opposition from public and Independent Safety Experts. Southern Californians were lucky, that a nuclear meltdown was averted with the Unit 3 accident, and ratepayers have already lost more than a Billion Dollars. The proposed Unit 2 reactor operation at no more than 70 percent power involves more than significant increase in the probability or consequences of a potential Steam Line Break Accident resulting in multiple steam generator tube leakages and/or ruptures than previously evaluated by SCE, their nuclear industry experts or some at the NRC. This is also a new and different type of accident applicable only to the San Onofre Unit 2 & 3 replacement steam generators because of their faulty design, which has not been previously evaluated. The proposed Unit 2 reactor operation at no more than 70 percent power involves a significant reduction in safety because of the reasons stated below.
    Basis – The Unit 2 reactor operation at no more than 70 percent power affects the probability or consequences of steam generator tube rupture significantly because many of the steam generators tubes can leak/rupture under the following conditions:
    Three different accident scenarios should be considered in attempting to determine accident severity when degraded tubes rupture.

    (i) 70% normal Steady State power operations: Due to random variations in local vibration intensity, tube failure will be initiated in a single tube with relatively slow through the wall crack propagation. Such failures likely would be confined to a single tube leakage lending to a may be timely detection and action by operator, or after the 5 month inspection period and removal from service, if the wall thickness exceeds 35%.
    (ii) Operational Transients @ 70% power operations. Next in severity are tube ruptures from operational transient (opening or closing of valves, scrams, etc). To ensure that the tubes withstand such transients, the ASME code requires that their Cumulative Usage Factor (“CUF”) be less than one. When the RSGs were installed the predicted CUF was a low number because it was calculated for pristine tubes. The fact that the tubes were stressed to above their endurance limit and they vibrated with frequencies on the order of 10 to 50 HZ, means that their CUF exceeded unity. A tube with CUF >1 would be prone to a rapid crack growth and sudden rupture under operational and DBA transients.

    Notes: MHI states, “The stress on the tube [SONGS Unit 3 Tube, which leaked, Row 106 Column 78, 100 percent through wall wear, length of wear – 29 inches] due to in-plane vibration was 4.2 ksi and was under fatigue limit (13.6 ksi).” In contrast, Dr. Hopenfeld’s calculations demonstrate that the stress concentration factor is much higher than calculated by MHI and therefore the fatigue limit of 13.6 ksi will be exceeded by at least 22% (16.7 to 17.5 ksi). Now, the ruptured tube in Mihama experienced out-of-plane FEI and according to the latest research between 2006 -2011 by Dr. Pettigrew and others, the in-plane velocities are double the out-of plane velocities, therefore, during FEI conditions, tubes can realistically experience fatigue of 16.8 ksi. This demonstrates that MHI calculations are under-conservative by a factor of 4, just like the fluid velocities calculated earlier by MHI, which led to the SONGS Unit 3 tube leak, failure of 8 tubes under MSLB testing conditions and loss of more than 35% wall thickness in almost 400 tubes. Now, SONGS Unit 2 RSGs have hundreds of fatigue damaged tubes, which have not been plugged and a number of U-bends with tube-to-tube clearances as low as 0.05 inches, which is almost one fifth of the design clearance of 0.25 inches. In addition, there are some active and pressurized tubes, some of which have lost wall thickness up to 28%, therefore, these tubes are just below the safety margin of 35% of the tube plugging limit. SCE has chosen not to plug these tubes as a conservative measure in their belief that these tubes will not rupture during the next 5 months. In addition to losing strength due to wall thinning, some of the worn tubes at SONGS Unit 2 have used up a significant fraction of their allowed fatigue life. Such damage cannot be detected by even the NRC Special Tube Inspections due to time, cost, unavailability of high technology probes and contactors, and/or impossible access within the tube bundle or radiation dose limitations. These tubes will be significantly susceptible to sudden ruptures without notice or early warnings during steady state normal operations at 70% power due to Operational Transients (opening or closing of valves, scrams, loss of offsite power, lifting of steam safety valves due to SG Over-pressurization, etc.) and can result in accidents like Mihama Unit 2 (Japan, 1991, INES Level 3 nuclear incident) or even leading to a nuclear meltdown.

    (iii) Main Steam Line Break (“MSLB”) @ 70% power operations. An MSLB could lead to the most severe consequence for operations with degraded tubes. SCE’s proposed revision to Technical Specification or current licensing basis (“CLB”) at 70% power would require that the plant accommodate such accidents. The MSLB accident is of particular concern, because of what happened at Unit 3. Its unprecedented tube-to-tube wear, never experienced in the 60 years of nuclear steam generator operational history, was caused in a very small section of the tube bundle due to the unanticipated creation of high dry steam (fluid elastic instability). The question is why was this condition unanticipated and how could high dry steam cause so much devastating damage in less than 11 months of operation to these tubes, which were designed to last between 40 to 60 years. Because, SCE, AREVA, Westinghouse, MHI, Intertek and NRC Augmented Inspection Team have not actually answered this important question, Dr. Hopenfeld, Arnie Gundersen, John Large, Professor Daniel Hirsh and other experts are saying that it is not safe to operate Unit 2 at 70% power.

    Now, I will try my best to answer the question, why it is unsafe to restart Unit 2 at 70% power?

    The most severe design basis accident to meet the SONGS Unit 2 TS 5.5.2.11.b.1 steam generator structural integrity is a MSLB at the first weld outside containment. The outside containment scenario includes the assumption that the main steam isolation valve (MSIV) in the steam line with the least flow resistance fails to close following the main steam isolation signal (MSIS). Super-heating within the SG initiates upon U-tube uncovery as specified in the NRC Information Notice 84-90. The turbine stop valves are assumed to close instantaneously at the time of the reactor trip. This assumption is conservative for a MSLB event because the entire steam inventory at the time of reactor trip is assumed to be forced out the break in 300 seconds or 5 minutes.

    The depressurization of the non-isolable steam generator would result in 100% void fractions in the degraded Unit 2 U-Tube bundle due to flashing of the feedwater into steam. This condition of ZERO Water in the steam generators would cause fluid elastic instability (FEI) flow-induced random vibrations, excessive hydrodynamic pressures and Mitsubishi Flowering Effect. The force of the flashing steam would create high-energy jets, lift loose parts and debris present inside the steam generator. These adverse effects would cut holes into already degraded tubes and create additional loading (See Note A below) on tube support plates (TSPs) due to heavy build-up of deposits on trefoil/quadrifoil-shaped holes from SG blowdown and cracked high cycle fatigue U-bend tubes not supported by an Anti-Vibration Bars (AVB). More than 500 tubes were plugged at Unit 2. Even though these tubes would not be not in service, they will continue to wear at the same rate as before. At a certain time, these tubes will break without being detected because no radioactivity will be released. The broken tubes will have relatively low natural frequency and therefore it would be prone for resonance excitation by external forces. The jets, formed at the ends of a broken tube, would cause it to whip and impact adjacent tubes thereby propagating further ruptures. These cumulative adverse conditions in all likelihood would result in a massive cascading of RSGs tube failures (tubes would excessively rattle or vibrate, hitting other tubes with violent impacts) due to extremely low tube-to-tube clearances and no in-plane effective anti-vibration bar support protection system. This Titanic and adverse effect would involve hundreds of degraded and active SG tubes along with all the inactive (plugged /unstabilized) tubes causing an undetermined amount of simultaneous tube leaks/ruptures. Under this adverse scenario, approximately 60 tons of very hot high-pressure radioactive reactor coolant would leak into the secondary system. The release of this amount of radioactive primary coolant, along with an additional approximately 200 tons of steam in the first five to fifteen minutes from a broken steam line would EXCEED the SONGS NRC approved offsite radiological release doses safety margins. So, in essence, the RSG’s are loaded guns, or a nuclear accident like Fukushima, waiting to happen. Any failure under these conditions would allow significant amounts of radiation to escape to the atmosphere and a major Loss of Coolant Accident (LOCA) could easily result causing much wider radiological consequences and even a potential nuclear meltdown of the reactor.

    SCE States, “A MSLB alone does not generate sufficient differential pressure to cause tube rupture. (See Notes below). The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators prevent RCS re-pressurization in accordance with Emergency Operating Instructions.”

    In Summary: SCE DID Actions and unreliable operator actions to detect a leak and to re-pressurize the steam generators as claimed by Edison are not practical to stop a major nuclear accident in Unit 2 in progress in the first 5-15 minutes of a MSLB during the 5-month trial period.

    Notes: This additional loading would exceed: (2) the safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1 .4 against burst applied to the design basis accident primary-to-secondary pressure differentials, and (3) significantly affect burst or collapse pressures determined and assessed in combination with the loads due to a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads. SCE’s planned “defense-in-depth” actions are insufficient to stop multiple tube ruptures due to a main steam line beak event.

    2. Does the proposed change create the possibility of a new or different kind of an accident from an accident previously evaluated?

    SCE’s Answer: No, the proposed changes do not create the possibility of a new or different kind of an accident from an accident previously evaluated.
    Basis – Please see SCE License Amendment (www. songscommunity.com)
    HAHN Baba Response: The proposed changes create the distinct possibility of a new or different kind of an accident from an accident previously evaluated. Please see Item 1 above for details.

    3. Does the proposed change involve a significant reduction in margin of safety?
    SCE’s Answer: No, the proposed changes do not involve a significant reduction in margin of safety.
    Basis – Please see SCE License Amendment (www. songscommunity.com)
    HAHN Baba Response: The proposed change involves more than a significant reduction in margin of safety. Please see Items 1 and 2 above for details.

    Based on the above, HAHN Baba concludes that the proposed amendment involves more than a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and
    accordingly, a finding of “no significant hazards consideration” claimed by SCE is another attempt to to defame, subvert and erode the NRC Regulatory Process. NRC Brilliant Engineers know better what to do this time. Let us wait for the NRC Intelligent Safety Answers.

  47. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC posting this blog. San Onofre NRC/SCE/MHI/Public Awareness Series – by Hahn Baba

    Researchers have said long ago that nucleate boiling on the tube surfaces has a stabilizing damping effect to preclude fluid-elastic instability. At least 1.5 % water or void fraction less than 98.5% in a steam-water mixture and areas without localized tube dry-out conditions are required in a nuclear steam generator to preclude the onset of fluid elastic instability. A review of NUREG-1841 published during the SCE/MHI design stages of San Onofre Replacement Steam Generators indicates that experienced manufacturers of steam generators (like Westinghouse/CE & BW&I) have used a combination of design and operational features [(high circulation ratios(>4), high steam pressures (> 900 psi) and low friction losses] to keep the void fractions at 98.5% or below and have prevented localized tube dry-out conditions and steam blanketing in more than several hundred operating US Steam Generators. Dr. Pettigrew has emphasized the solid teamwork and alignment between the designer and manufacturers of nuclear power plant components since 1970’s to prevent the adverse effects of fluid elastic instability and flow induced vibrations. Low steam pressures are severe for tube vibrations, but saturated steam also has high enthalpy at low steam pressures. Therefore, from steam generators, by operating at low steam pressures, you can produce more thermal megawatts, and thereby, supply more electricity to the grid.

    Per NRR RAI #13, “SONGS RSGs have installed 377 (~4%) more tubes (9,727 versus 9,350) than the OSGs. RSG tubes have a larger average heated length (729.56 in. versus 680.64 in.) than the OSG tubes. These features resulted in larger values for the RSG for heat transfer area, tube-bundle flow area, and tube-bundle water volume.” MHI Root cause states, “Thus, not using ATHOS, which predicts higher void fractions than FIT-III at the time of design represented, at most, a missed opportunity to take further design steps, not directed at in-plane FEI, that might have resulted in a different design that might have avoided in-plane FEI. However, the AVB Design Team recognized that the design for the SONGS RSGs resulted in higher steam quality (void fraction) than previous designs and had considered making changes to the design to reduce the void fraction (e.g., using a larger downcomer, using larger flow slot design for the tube support plates, and even removing a TSP). But each of the considered changes had unacceptable consequences and the AVB Design Team agreed not to implement them. Among the difficulties associated with the potential changes was the possibility that making them could impede the ability to justify the RSG design under the provisions of 10 C.F.R.§50.59. Thus, one cannot say that use of a different code than FIT-III would have prevented the occurrence of the in-plane FEI observed in the SONGs RSGs or that any feasible design changes arising from the use of a different code would have reduced the void fraction sufficiently to avoid tube-to-tube wear. For the same reason, an analysis of the cumulative effects of the design changes including the departures from the OSG’s design and MHI’s previously successful designs would not have resulted in a design change that directly addressed in-plane FEI.”

    This design change of average tube heated length (729.56 in. versus 680.64 in.) increase is the most damaging design change made without a 50.90 License Amendment both by MHI and SCE. Mitsubishi was instructed by SCE to make the RSG tubes as tall as possible in orders to maximize the tube heat transfer area. This change caused unacceptable FEI consequences in Unit 3, which may be one of the reasons that U.S. Sen. Barbara Boxer and Congressman Edward J. Markey expressed the following concern on February 6, 2013 in their letter to NRC Chairman Allison McFarlane: “Southern California Edison was aware of problems with replacement steam generators at its San Onofre nuclear power plant but chose not to make fixes.” And SCE proudly claims, “These changes result in larger values for the RSG for heat transfer area, tube bundle flow area, and tube bundle water volume. This is beneficial in the short and long term for SB LOCAs, which rely upon the steam generators for RCS heat removal.” In response to Senator Barbara’s letter, SCE states, “The report was submitted to the NRC by MHI months ago as part of the voluminous records, data, information and other materials the NRC has been thoroughly reviewing and inspecting as part of its consideration of SCE’s request to restart Unit 2 safely. It is simply not accurate to suggest, as the letter does, that when they were installed ‘SCE and MHI were aware of serious problems with the design of San Onofre nuclear plant’s steam generators.’ Indeed, MHI, the manufacturer of the steam generators warranted the steam generators to be free from defects for 20 years after installation. SCE would never, and did not install steam generators that it believed would not perform safely. SCE, like other utilities seeking to replace their steam generators, sought to purchase replacement steam generators that would meet or improve upon the safety standards and performance of the original steam generators.”

    Press Reports state, “The operator of the troubled San Onofre Nuclear Generating Station wants to put its reactor restart plan into effect on June 1. Doug Broaddus, the chief of the special project branch that oversees the northern San Diego County power plant for the NRC, responded that such a timeline presented “a challenge.” I’m not going to make any commitments as to whether we can make that date or not and the Nuclear Regulatory Commission staff told Edison officials the company’s June 1 target date might be out of reach, given the complexity of a proposal that comes with reams of technical paperwork. The NRC must stand firm and demand a comprehensive license amendment process that includes all safety issues and the opportunity for a full public hearing, said Kendra Ulrich, nuclear campaigner for Friends of the Earth. Unit 2 is “a severely damaged reactor that is unsafe to operate,” she said.” Press Reports further state, “The future of the heavily damaged Unit 3 reactor, where the radiation leak occurred after a tube break last year, is not clear. Edison has said that because of manufacturing differences, Unit 2’s generators did not suffer the extent of deep tube wear witnessed in its sister. Decaying generator tubes helped push San Onofre’s Unit 1 reactor into retirement in 1992, even though it was designed to run until 2004. The following year, the Trojan nuclear plant, near Portland, Oregon was shuttered because of microscopic cracks in steam generator tubes, cutting years off its expected lifespan. Cracked and corroded generator tubing has vexed the nation’s nuclear industry for years.

    Not in so many words, but NRC Chairman has publically stated “SCE is responsible for the work of all its contractors and sub-contractors including MHI, AREVA, Westinghouse and Intertek. Regulators may need to be “buffered” from political winds, but they need to be fully subjected to the pressure of scientific and engineering truth and cannot be allowed to make decisions or order actions that are “independent” of facts. They cannot be allowed to push rules that are aimed at addressing emotional feelings and reinforcing irrational fears.” “Our safety culture, skills and ability were all insufficient,” said Naomi Hirose, president of Tokyo Electric Power Co. (TEPCO), at a recent news conference for the release of a new report on the Japanese utility company’s response to the Fukushima Daiichi accident in March 2011. The report determined TEPCO’s equipment and safety provisions were insufficient; that the company’s safety measures were inadequate; and that it did a poor job of keeping the public informed.

    Let us examine the operator timing issue of multiple tube ruptures about which NRC Staff has been publicly silent since this issue was raised in 2000.

    By, Emergency Planning Rules, an operator has 15 minutes to diagnose an event, declare the event, notify the offsite agencies and start taking initial actions to stop the nuclear accident in progress and manually trip the reactor. We have already discussed the potential of multiple tube ruptures caused by a main steam line break in Unit 2 with assumed failure of main steam isolation valve to close, discounted the operators of taking any mitigating safety actions within 5-15 minutes, because reactor will already be tripped by the automatic systems. So let us examine, what automatic systems do and what are the consequences of the steam generator ruptured tubes in first 15 minutes without operator action.

    Abnormal operation (Main Steam Line Break) modes of the main steam system consist of rapid (step) reductions in power demand (zero demand) and transients that result in an automatic reactor, turbine, feedwater pump and reactor coolant pump trips. the generic signs of a main steam line break are rapid deterioration of steam generator pressure sudden change in steam generator level, decreasing reactor coolant system temperature and pressure, decreasing pressurizer level, large differential pressure between steam generators of 2250 psi between the tubes and steam generator. SCE states, “MSLB alone does not generate sufficient differential pressure to cause tube rupture. The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators prevent RCS re-pressurization in accordance with Emergency Operating Instructions.”

    So now, the steam generator is isolated with Zero Pressure connected to the environment, with 100% Fluid Elastic Instability, Feedwater Jet Impingement on the damaged and cracked tubes and differential pressure between Steam Generators of 2250 psi between the tubes and steam generator and SCE is hoping that operators would stop the auxiliary feedwater flow to the broken generator. So, now what will happen to the tubes, and I will expand on scenario:

    According to Federal Regulations, steam generator tubes are required to be designed with extremely low probability of leakage and assumed to maintain structural integrity for 40 years. According to NRC rules and Nuclear Energy Institute Tube Management Program Criteria, these tubes are required to be plugged, and taken out of service, if they lose their wall thickness greater than 35%.

    Based on a review of SONGS plant data/procedures, various San Onofre Technical Reports, Academic Research Papers (Dr. Pettigrew and others) and input from a leading professional chemical engineer in the world with Ph.D. from McGill and Post-Doctorate work in two-phase flow, it is concluded: (1) Due to extremely high reactor coolant & steam flows (To generate more power for profits more than the design thermal heat transfer capacity of the steam generators and more thermal reactor power than allowed by NRC), high hot leg heat flux, extremely high in-plane velocities (> 56 feet/sec), very tall tube bundle, narrow tube pitch to tube diameter ratio, low steam pressures, low tube-to-tube clearances, no in-plane restraints (zero contact forces, U-bends free span not supported by effective in-plane anti-vibration bars) and zero water on tube surfaces (film boiling, departure from nucleate boiling), fluid elastic instability, flow induced vibrations and Mitsubishi Flowering Effect in SONGS Unit 3 caused tube-to-tube wear and fatigue cracking causing 1 tube to leak, 8 tubes to fail main steam line break test pressure and 380 tubes lost their wall thickness greater than 35% in only 11 months of operation. Besides the same design defects as Unit 3, SONGS Unit 2 did not experience fluid elastic instability due to lower steam flows and higher steam pressures. Unit 2 has very tall tube bundle, narrow tube pitch to tube diameter ratio, low tube-to-tube clearances (A large number of U-bends with only 0.05 inch clearances, Design clearance = 0.25 inches), no in-plane restraints (zero in-plane contact forces, U-bends free span not supported by effective in-plane anti-vibration bars), low frequency retainer bars, thousands of tubes with undetected tube-to-tube wear and macroscopic cracks, and large number of plugged/stabilized tubes. These tubes would be prone to rapid tube-to-tube wear, crack growth and sudden cascading tube leakages and/or ruptures under operational and DBA transients described below. Once an accident is progress, the situation will be beyond operator control and SONGS Defense–in-Depth Engineered safety features can likely be insufficient to prevent a partial or complete nuclear core meltdown like Fukushima, Three Mile Island and Chernobyl.

    The phenomenon of cracking due to high cycle fatigue of U-bend tubes not supported by an Anti-Vibration Bars (AVB) was highlighted as early as 1987 by NRC (tube rupture in the North Anna 1 plant) and Japanese Society of Mechanical Engineers in 1991 (Mihama MHI SG Tube Rupture). Between 2004 and 2006, three primary-to-secondary leaks occurred at the Cruas NPP: Unit 1 in February 2004 and unit 4 in November 2005 and February 2006. The three leaks were all the result of a circumferential crack in the tube at the location where the tube passes through the uppermost tube support plate (TSP #8). The tubes which leaked at Cruas were all located in raw 8 and were not supported by AVB by design. They were not considered sensitive to high cycle fatigue. A new fact caused the leakages observed on the Cruas units: the heavy build-up of deposits on the secondary side of the steam generator which changed the flow conditions in the center of the tube bundle. The deposits reduced or blocked water/steam flow through the quadrifoil-shaped holes in TSPs, forcing more water and steam into the center of the tube bundle, which caused the excessive vibration of the tubes near the center of the tube bundle. This excessive vibration due to fluid-elastic instability resulted in fatigue cracking of the tube.

    Based on a leading Public Safety Nuclear Research Scientist and numerous Independent Technical Experts, Southern California Edison has not identified the root cause for the unusually excessive tube wears and loss of fatigue life in the four steam generators (RSGs) of units 2 and 3 at San Onofre Nuclear Generating Station. The components causing the wear (Anti-Vibration and retainer Bars, and damaged tube bundle) would have to be replaced and the SONGS license amended before the units can be restarted. The above assessment also applies to SCE proposed five-month test of Unit 2 at 70% reduced power. Based upon identification of an error in the MHI stress calculations, the revised analysis shows that because of the wear damage previously sustained by Unit 2, some tubes will be susceptible to rapid fatigue failure. The tubes will exceed their allowable fatigue life by 22 to 29% during the next operating cycle 17. The risk of tube rupture increases with operating time but the SCE/MHI analysis is not capable of quantifying it in terms of operating time.

    Because of the unprecedented severity and scale of tube degradation at SONGS 3, there is no data to determine the safety consequences of operations with failure prone tubes. The existing performance criteria are of little help because they were derived from accident progression of tubes with small stress corrosion cracks—rather than accident progression with fatigued tubes. These criteria are not sufficiently conservative because primary to secondary leakage is expected to develop more suddenly when tube failure is caused by fluid elastic instability vibration fatigue. Because the uncertainties in predicting how fatigued tubes can rapidly and instantaneous propagate failures, it is impossible to assess quantitatively safety risks. The assessment by SCE Independent experts was therefore limited to a discussion of discrete accident scenarios without assigning a probability to any specific scenario. It is believed that a main steam line break (MSLB) represents a bounding case. A conservative estimate of the probability of a large early release of radiation with containment bypass would be 10E-4 per year for any operating cycle. Such risk exceeds NRC’s safety goals. Units 2 and 3 at SONGS have the highest risk of tube rupture related core damage than any other nuclear power plant in the US.

    Based on review of the AIT report, the March 2013 MHI root cause report, NEI 96-07, and 10 CFR 50.59, it is concluded that SCE did not adhere to the 50.59 guidelines regarding license amendment requirements. A main cause of the RSG failures can be attributed to the fact that the design changes initially were not reviewed by a financially independent and competent engineering organization. The documents produced by SCE following the January 2012 incident contain serious technical flaws and reflect lack of impartiality. SCE and MHI did not provide any data to support their contention that the various design change options discussed in 2006 by the AVB Design Team would have had no significant effect on flow velocities and steam quality. SCE consultants AREVA, Westinghouse and MHI differed on the root cause of tube vibrations and none of their opinions were based on sound scientific principles. The safety consequences of operating with degraded tubes are more serious than envisioned by the consultants. It is apparent that SCE focused its attention on explaining away errors in the design, computer modeling, fabrication and management of the RSGs by MHI. There is no indication that any consideration was given to the long-term safety risk of operating Units 2 and 3, with each containing more than identified 1500 defects (3401 in all) and thousands more unidentified defects.

    Westinghouse obtained the K values in their test model MB-2. Since the MB-2 tubes are shorter in length than their model F, the reported values of K are subject to large uncertainties. It is not at all clear how the MB-2 data is applicable to the SONGS RSGs. To appropriately scale the Westinghouse K values to the RSG, it would first be necessary to apply correction factors to allow for the difference in length and tube support geometries. While reducing power level decreases the probability of tube wear it is not at all clear that the 70% power level represents a conservative value when considering the large uncertainties in the Westinghouse analysis.

    Thermal-Hydraulic computer codes are based on the porous model, proprietary experimental data and out-of-plane correlations. Sensitivity studies are commonly used to highlight the effects of design changes on code predictions, thereby allowing uncertainty estimates. Since the FIT-lll code has not been benchmarked and MHI has not provided any sensitivity results, it is impossible to tell whether MHI’s predictions were in error due to a faulty computer code or a faulty input. When SCE assigns a confidence level of 50% to certain predicted SR value, it is impossible to assess its validity because the uncertainties are unknown. It is therefore difficult to assess the risk associated with the 70% power operations. An independent reviewer has no choice but to use bounding models to predict risk.

    In spite of the synergy between wear and fatigue, SCE did not address it. None of SCE’s consultants apparently considered the synergetic effects of wear and fatigue in their assessment of tube wear in Unit 2 at the reduced power level. For a given tube, the repeated impacts on its surface together with its sliding motion will affect tube wear. The major uncertainty in using the ASME data relates to the incubation period, which depends on the combination of number of impacts and sliding motions. No data is available to clarify this point; physical reasoning would dictate that the ASME plots used by MHI are not sufficiently conservative for the type of wear that has occurred at SONGS.

    Three different accident scenarios should be considered in attempting to determine accident severity when degraded tubes rupture.

    (i) Steady State Operations. Due to random variations in local vibration intensity tube failure will be initiated in a single tube with relatively slow through the wall crack propagation. Such failures likely would be confined to a single tube lending to a timely detection and removal from service.

    (ii) Operational Transients. Next in severity are tube ruptures from operational transient (opening or closing of valves, scrams, etc). To ensure that the tubes withstand such transients, the ASME code requires that their Cumulative Usage Factor (CUF) be less than one. When the RSGs were installed the predicted CUF was a low number because it was calculated for pristine tubes. The fact that the tubes were stressed to above their endurance limit and they vibrated with frequencies on the order of 10 to 50 HZ, means that their CUF exceeded unity. A tube with CUF >1 would be prone to a rapid crack growth and sudden rupture under operational and DBA transients.

    (iii) Main Steam Line Break (“MSLB”). An MSLB could lead to the most severe consequence for operations with degraded tubes. SCE’s current licensing base (“CLB”) requires that the plant accommodate such accidents. The MSLB accident is of particular concern to the type of tube failures that occurred at SONGS because of the high excitation frequency. A plugged tube, that has been removed from service, still continues to vibrate and hit other tubes and erode in strength. Eventually, these tubes will breaking in two half during a main steam line break. Each loose end of the broken tube will act as a whip damaging (cutting) adjacent tubes. The energy for the whips originates with the vibrations of the main steam line, the swelling of two-phase mixture during the initial phase of the transient and the energy from jets of any leaking adjacent tubes. In comparison, SCE and NRC state that only few drops of radioactive coolant will escape to the environment with the operator quickly isolating the leaking steam generator. NRC staff at the NRC Emergency Response Center during the accident would be scratching their heads saying, “We have not seen anything like that before, we only studied what happened to cracked tubes during steam line breaks. Our performance technical specifications were based on the behavior of cracked, but strong tubes not paper-thin walled tubes.”

  48. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC posting this blog. San Onofre NRC/SCE/MHI/Public Awareness Series – by Hahn Baba

    SUMMARY: Unit 2 anti-vibration bar bundle and support structure is not designed for positive in-plane vibrations and it has hundreds of damaged and plugged tubes with unspecified number of undetected and incubation cracks in progress caused by high cycle thermal fatigue. During normal operations at reduced power operations during the 5-month test period, due to anticipated operational transients and main steam line breaks, these damaged tubes due to fluid elastic instability, flow-induced vibrations and Mitsubishi Flowering Effect (Due to the Extremely Tall Tube Bundle compared wit other Mitsubishi Generators) can potentially experience from one tube leakage to cascading tube ruptures beyond operator’s control and can potentially result in a nuclear meltdown. This is consistent with NRC Fukushima Task Force Lessons Learnt. The Reports prepared by NRC AIT Team, SCE, MHI, AREVA, Westinghouse and Intertek are inconsistent, confusing, conflicting, ambiguous, based on invalidated data, faulty computer and statistical modeling and fail to arrive at a clear and concise conclusion. The reports prepared by Dr. Joram Hopenfeld, Arnie Gundersen, John large, Professor Daniel Hirsch and me based on review of plant data/documents, review of research papers and industry benchmarking, and conversations with SONGS insiders (Root Cause Team, Operations Shift Managers, Nuclear Fuels, Hazard Barrier Group, Nuclear Regulatory Affairs, Project Management Organization, etc.) are consistent and clear that Unit 2 is not safe for power at 70% operations.

  49. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC posting this blog. San Onofre NRC/SCE/MHI/Public Awareness Series – by Hahn Baba

    NRC, INPO, CPUC, NEI and Scientists expect a responsible nuclear utility to supply safe and reliable power at a reasonable cost and not conduct unsafe experiments at the expense of public safety and charge ratepayers for its mistakes.

    Life is a unique opportunity to serve the society. Society needs Energy, which is safe, economical and reliable. Every form of Energy has drawbacks and risks. SCE is responsible for safety, economics and reliability of Unit 2.

    These factors will be measured on a scale of 0 to 10, 10 being the best and 0 being the worst rating.

    Definitions and Ranking

    • Highest Safety – Low probability of a tube rupture due to operational transients and main steam line break – 10
    • Lowest Safety – Cascading tube ruptures due to operational transients and main steam line break – 0

    • High Base-Load Reliability – 24/7 uninterrupted 2100 MWt supply to grid with voltage support -10

    • No Base Load Reliability – 730 MWt with frequent shutdown and interrupted supply to grid with only 30% of the time voltage support – 0

    • Best Economics – Nuclear plants are the lowest-cost producer of base- load electricity. The average production cost of 2.19 cents per kilowatt-hour includes the costs of operating and maintaining the plant, purchasing fuel and paying for the management of used fuel. (www.nei.org)) -10
    • Worst Economics – With SONGS Restart, cost is unknown, but will be highest in the nation – 0

    SONGS Unit 2 Restart Statistics
    • Lowest Safety – 0
    • Worst Economics -0
    • No Reliability – 0
    • INPO Rating – 4 (Worst Operating Record of A Nuclear Power Plant)
    • General Safety Record – Worst in the USA according to NRC Data
    • Fire Safety Record – Unit 3 shutdown for 5 months in 2001, $100 Million Loss, falsification of fire watch records for 5 years, 250 ignition sources/welding/grinding/procedure violations between 2010-2012
    • Emergency Preparedness Record – Lowest in the nation between 2006-2012, Some of the best Shift Managers, Station Emergency Directors, Corporate Emergency Directors and Plant Operators have resigned, retired or Laid Off
    • Cyber Security Initial Awareness Training – Audit found 1300 site workers, Cyber Security Program Manager, Chief Nuclear Officer and several other Directors out of SONGS not EIX procedure compliance
    • Management Record – Several of the Chief nuclear Officers have retired or resigned. Some of Present Senior Leadership Directors are inefficient, few are inefficient and retaliating, some are inactive and complacent, and the whole team is production/profit oriented rather than safety oriented
    • Steam Generators – 8 Steam generators destroyed by flawed design and mis-operations at a cost of Several Billion Dollars to Rate Payers, Arrangements are on the way to destroy the remaining 2.
    • Future – Depends upon NRC, affects the safety of 8.4 Million Californians

    • Steam Generator Repair/replacement Project – Estimated Duration – 5 Years – Cost – Unknown – NRC Report: The U.S. Nuclear Regulatory Commission (NRC) inspection team observed various activities associated with the mock-up tests of a portion of the upper tube bundle. The activities being done by Mitsubishi Heavy Industries, Ltd. (MHI) were conducted to determine if a design modification to repair the San Onofre Nuclear Generating Station (SONGS) steam generators was feasible. The design modification testing consisted of anti-vibration bar insertion tests using three different designs. The three different anti-vibration bar designs were:
    (1) Thicker – inserted between and parallel to existing anti-vibration bars
    (2) 30 Degree – inserted at a 30 degree angle to existing anti-vibration bars, forming intersections with existing anti-vibration bars
    (3) Comb – shaped like a comb and will be inserted into the bundle on every other row and then rotated 90 degrees, locking tubes into place between the “teeth” of the comb.
    In discussions with MHI personnel, they indicated that the thicker anti-vibration bar will likely be the least difficult to insert and the comb anti-vibration bar the most difficult to insert due to slight differences in the gaps and arrangement of the tubes.

    The last option (3) is beneficial for fluid elastic stability (FEI), but also increases significantly the risk of cascading tube ruptures tube rupture @100%, 1729 MWt per RSG @ MSLB Conditions. The first two options (1) (2) are not beneficial for fluid elastic stability. MHI has does not have the tools, technology or skills to repair/replace these flawed and degraded generators, unless MHI works out a deal with Westinghouse to acquire the required tools and skills. If the deal does not work, SCE is only left wit two options, which were recommended a year ago, but were ignored:
    Option 1 – Give a turnkey contact to repair/replace the replacement Steam Generators to Westinghouse/Bechtel, Fire/Retire/Lay Of the Inefficient and retaliating Leaders, and Hire Capable and Human Managers, or
    Option 2 – Dismantle and Decommission San Onofre Units 2 & 3.

  50. Sincere Thanks to NRC Chairman, Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC posting this blog. San Onofre NRC/SCE/MHI/Public Awareness Series – by Hahn Baba
    NRC, INPO, CPUC, NEI and Scientists expect a responsible nuclear utility to supply safe and reliable power at a reasonable cost and not conduct unsafe experiments at the expense of public safety and charge ratepayers for its mistakes.
    San Onofre Restart quoting Albert Einstein, “Insanity: doing the same thing over and over again and expecting different results.”
    First Strike – 1992 – Decaying generator tubes helped push San Onofre’s Unit 1 reactor into retirement in 1992, even though it was designed to run until 2004.
    Second Strike – 2001Power Uprate – To generate more power, Edison Engineers increased the steam flows and lowered the steam generator pressures, which increased vibrations, and shortened the life of Rate Payers Paid Original Steam generators.
    Third Strike – 2011 – SONGS Unit 3 – To generate more power, believing that Unit 3 anti-vibration structure was built better than Unit 2, Edison Engineers tested the new supports by increasing the reactor coolant flows, steam flows and lowered the steam generator pressures, which increased vibrations, and destroyed the Rate Payers Paid brand New Replacement Steam generators. Edison has said that because of manufacturing differences, Unit 2′s generators did not suffer the extent of deep tube wear witnessed in its sister plant. Unit 2 was not operating in the test mode and did not experience fluid elastic instability because of lower reactor coolant flows, lower steam flows and higher steam generator pressures. Unit 2 better supports and double the contact forces unproven theory is just a conjecture on the part of SCE/MHI based on hideous data, faulty computer simulations and an excuse to start defectively designed and degraded Unit 2. Unit 2 better supports and double the contact forces unproven theory is just a cheap SCE scheme to charge insurance money from MHI and more money from ratepayers. This bogus and unconvincing theory is contested and challenged based on the available plant data and review of Dr. Pettigrew’s research papers and testimony, John Large, MHI, Westinghouse and AREVA Reports.
    Fourth Strike – 2013 – Edison officials are also preparing long-range plans under which the plant might run for years, even though some of Edison’s own research has suggested tube damage could cut short its life span. Precise projections about the future are dependent on a restart — Edison engineers need to study how the reactor behaves at 70 percent power before being able to sharpen longer-range calculations. The plant could be started then shut down, as many as five times during a trial run to assess its operation and safety. “To propose an experiment in which the damaged reactor is repeatedly turned on and off shows a disgraceful contempt for public safety,” said Kendra Ulrich, a spokeswoman for the Friends of the Earth. Unit 2 restart without complete and thorough review by NRC Brilliant Engineers and Public Hearings on the basis of meeting peak summer electricity demands is an unapproved experiment and just a cheap SCE scheme to charge more money from ratepayers.
    Last Strike – Albert Einstein, “Any intelligent …. can make things bigger, more complex, and more ….. It takes a touch of genius — and a lot of courage — to move in the opposite direction.” Ted Craver needs to tell Ron, Pete, Tom, Rich, John, Mike, Vic, Doug and others to stop wasting NRC and Public’s time and money and award a turnkey contract to Westinghouse and Bechtel to repair or replace both Units Steam generators. This will be expensive, but wise long term and an excellent PR move for Ted, Ron, Pete and EIX/SCE shareholders, and will be in the best interests of NRC, INPO, NEI, Nuclear Industry, CPUC and the Concerned Public. Of course, On SCE Federally Leased Territory, SCE can retaliate, discriminate, harass, intimidate, fire, lay off, demote shut up an employee in violation of federal regulation and get away with it due to their power, attorneys, money and political connections, but my friends, this is America, The Greatest Democratic Country in the world, and outside the Federally Leased SCE Controlled Secured Territory, SCE/NRC cannot ignore and shut up the paying and concerned public in violation of the Open Government Doctrine of His Honorable Excellency, The President of the United States, US Justice Department, US Congress and the US Constitution.

  51. Special Thanks to NRC and Moderator Mr. Victor Dricks for posting this blog
    SPECIAL Public/NRC/SCE Awareness Series by HAHN Baba
    AREVA, The number 2 French manufacturer of nuclear plants in the world states, “Out-of-plane fluid-elastic instability has been observed in nuclear steam generators in the past and has led to tube bursts at normal operating conditions.” Recent history of steam generator tube ruptures is as follows:
    • At around 13: 50, on February 9th, 1991, leakage of about 55 tons of primary coolant occurred due to the failure of one SG tube in a steam generator built by Mitsubishi Heavy Industries in the No. 2 pressurized water reactor at the Mihama nuclear power station in Japan. At the same time, water pressure in the core had dropped drastically and the ECCS kicked in, flooding the reactor and shutting it down. If the core had been left exposed, a meltdown — an overheating of the fuel that can, if uncontrolled, lead to a large release of radioactivity — could have occurred. One main steam isolation valve and one pressurizer relief valve could not be operated by remote control. Therefore, the valve operation was carried out manually. In the following week an estimated 7 million Bq was released into the sea and an estimated 5 billion Bq of radioactive gas was released into the atmosphere. This tube rupture caused the first INES level 3 nuclear incident in Japan, which raised social concerns. In order to provide an opportunity to learn from the accident that resulted in the leakage of primary coolant from the SG tube due to fretting fatigue, the damaged steam generator has been preserved in an exhibition at the Mihama station of the Kansai Electric Power Company. An exhibition is a good way to help everyone to good lessons from an accident. http://www.sozogaku.com/fkd/en/cfen/CB1061010.html
    • On 9 August 2004, an accident occurred in a building housing turbines for the Mihama 3 reactor. Hot water and steam leaking from a broken pipe killed four workers and resulted in seven others being injured. The accident had been called Japan’s worst nuclear power accident before the crisis at Fukushima Nuclear Power Plant. http://en.wikipedia.org/wiki/Mihama_Nuclear_Power_Plant
    • Between 2004 and 2006, three primary-to-secondary leaks occurred at the Cruas Nuclear Power Plant in France: unit 1 in February 2004 and unit 4 in November 2005 and February 2006.
    http://www.eurosafe-forum.org/files/Presentations2008/Seminar%201/Abstracts/1.5%20Tubes%20support%20plates%20clogging%20up%20of%20French%20PWR%20steam%20generators%20V5%20correct%20figures%20Bodineau_20080923.pdf
    The point is that from the above real accident information, you can see a tube rupture can occur in SONGS unit 2 at 70% power due to numerous undetected internal cracks in tubes opening up at any time because of Mother Nature’s Mood, adverse operational conditions and/or equipment/operator failures. There are hundreds of damaged tubes with unknown amount of internal cracks and 100% tube inspection of these damaged tubes with computer controlled high technology laser, radiographic probes and remote video cameras is an extremely difficult, very time consuming and expensive process. SCE, MHI and AREVA have inspected only those areas in Unit 2 steam generators, which were previously damaged, and areas easily accessible with average technology probes. That is why NRC is doing special tube inspections, and nobody really knows or will ever know because of proprietary information, what percent of the steam generator areas, NRC Inspectors with the assistance of their highly specialized and qualified contactors will be able to access and inspect. FYI, under the NRC and NEI steam generator tube inspection and management rules, a tube is required to be plugged, if its looses more than 35% wall thickness. Tubes are only 0.043 inches thick and pressurized with 2200 psi highly radioactive coolant.
    Here is what the NRC AIT report states after the Unit 3 tube leak inspection, “At 4:10 p.m., operations personnel evaluated that the primary-to-secondary leak rate exceeded 75 gallons per day on steam generator 3E0-88 and that the leak was increasing at greater than 30 gallons per day per hour, and consequently, initiated a rapid power reduction to be ≤ 50 percent power in one hour and in Mode 3 within the next two hours per Abnormal Operating Instruction SO23-13-14. In accordance with Abnormal Operating Instruction SO23-13-14, when reactor power was less than 35 percent, operations personnel tripped the reactor at 5:31 p.m. to enter Mode 3. These examinations identified over 160 tubes in each steam generator with long free-span indications similar to that found on the leaking tube. More than half of the free-span indications in each steam generator had maximum measured depths exceeding the 35 percent plugging limit in the technical specifications, and ranged to as much as 99 percent (for the non-leaking tubes).”
    So now, according to the emergency planning rules, a Shift Manager has only 15 minutes to diagnose the leak and take protective actions to stop the leak. The last time leak was small and growing, but it took operators more than one hour and 20 minutes to trip the reactor. There were also 2 more tubes with loss of 99% wall thickness besides the leaking tube and could have leaked and operators would not have been able to control the plant. Now, if any of these tubes would have busted, steam generator would have pressurized and we would have seen a reoccurrence of 1991 Mihama accident described previously. Southern Californians were lucky, because the SONGS best Shift Manager was on duty. In 2001, Unit 3 was shutdown due to a fire for 5 months due to a combination of faulty alarms and human errors resulting in a loss of 100 Million dollars.
    SONGS Emergency Planning drill record is one of the lowest in the nation for the last six years, an average between 94-96%. That means, operator can fail 5% of the time in detecting and diagnosing the problem and bringing the reactor under control in time, if it is a full blown tube rupture. If the reactor is not brought under control in time and if it is one or more than one tube rupture and radioactivity gets into the environment, then it depends upon the wind direction. If the wind is blowing towards the ocean, damage will be minimum to the residents. If the wind blows towards San Clemente, and it is rush-hour traffic, nobody really knows and can estimate the immediate and long-term damage to the human life with untested computer programs and models.
    Therefore, there are lot of unknowns with Mother Nature’s Mood, condition of Unit 2 tubes, SONGS equipment failures and SONGS Operators Skills and Intuitions. Is it worth the risks, NRC has to decide. Running at 70% power does not make any economic sense. Can Southern Californians live with the risks and without SONGS in operation, until both units are completely repaired or replaced? Of course, they have lived for 15 months without SONGS and can live for two to three more years by conserving more. I am sure, Southern Californians rather be safe than sorry. SCE can buy surplus power from North and Palo Verde. All the Hospitals, Emergency Rooms have Emergency Generators. Can SCE, MHI or NRC guarantee safety of Southern Californians, ask them? In the end, it is a game of money, politics, power and a chance. Human life is priceless compared to any amount of SCE Profits or temporary inconvenience of living without electricity.

  52. Thanks to NRC Special San Onofre Review Panel for posting this blog.
    San Onofre Billion Dollar Debacle SCE/MHI/NRC Fukushima Lessons Learnt and Public Awareness Series – HAHN BABA
    SCE 10CFR 50. 92 (No Significant Hazards Consideration Analysis) License Amendment for San Onofre Nuclear Plant – SCE has to provide NO answer to one of the following 3 questions:
    (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or
    (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or
    (3) Involve a significant reduction in a margin of safety.

    It appears from a review of all the NRC Regulations and NRC Staff reports/ correspondence that NRC/Industry/NEI/EPRI purposely does not believe in the Ex-NRC Highly Educated, Brilliant Researcher, Eminent Scientist, Public Safety Expert [NAME REDACTED] theory, actual observations and concerns of cracked tubes, cascading tube leaks/ruptures caused by steam line breaks at several power plants (SONGS Unit 2 Scenario -100% void fractions resulting in uncovery of the tube bundle, jet impingement loads on tubes caused by flashing feedwater and stored energy in SG, loose metallic objects and parts of of broken tubes cutting other tubes, 36 U-bends with clearances of 0.05 inches, failure of retainer bars due to FEI velocities causing collapse of Anti-vibration bar bundle and rupturing other pressurized tubes, resonance vibrations and re-pressurization of SG resulting in loss of more coolant, etc.) and operator errors even after what happened with SCE/MHI designed Magical out-of-plane anti-FEI vibration bar bundle in SONGS 3, then the readymade cookbook SCE answers based on knowledge of SONGS 50.59/FSAR Culture are as follows

    1) Yes, Operating SONGS Unit 2 between 70-95% power involves more than a minimal increase per SONGS 50.59 Readymade Cookbook, but does NOT Really Involve a significant increase in the probability or consequences of an accident previously evaluated;
    (2) Operating SONGS Unit 2 between 70-95% power does NOT Really create the possibility of a new or different kind of accident from any accident previously evaluated; and
    (3) Yes, Operating SONGS Unit 2 between 70-95% power involves more than a minimal increase, but does NOT Really involve a significant reduction in a margin of safety.

    Then the question is if SCE is so sure of readymade cookbook answers like the Replacement Steam Generator 50.59 Evaluation Conservative NO Answers, why does not SCE want to apply for a License Amendment to operate Unit 2 at 95% power and afraid to go through a questioning public hearing. If SCE applies for a License Amendment to operate Unit 2 at 95% power and goes through a satisfactory public hearing, it makes good economic and public relations sense for SCE. Satisfying the concerned public in a democratic country like USA is key to the future of SCE, NRC, MHI and Nuclear Industry and intentionally avoiding the worried and restless public is an admission of guilt, negligence of duty, and is an indication of SCE using money and political clout to subvert the NRC regulatory process and violation of Federal Regulations. NRC has very brilliant and dedicated safety staff and I hope they do not get trapped and tainted like the Japanese regulators with Fukushima….Very Strong Rumors on the street from several sources that some elected and appointed officials with authority over Senator Barbara Boxer and Dr. Macfarlane are adamant and pushing restart of Unit 2 despite very strong objections from public, technical experts, scientists, Ex-NRC Chiefs, Friends of the Earth, Media and several Safety Organizations. In the End, Southern Californians need safe, reliable, well managed/maintained nuclear plants free of worker retaliation, discrimination and harassment for expressing nuclear safety concerns and Not a SONGS Fukushima in their Backyards.

  53. PROBLEM STATEMENT: Based on review of power data supplied by Southern California Edison for Unit 3 during January 2012, one concludes that SONGS most likely exceeded Unit 3 Reactor Thermal Power allowed Upper NRC Limit of 3478 MWt (Includes Reactor Coolant Pumps contribution of 20 MWt ±0.58% allowed NRC Crossflow UFM instrument Error) ~ 23 times out of 31. NRC needs to review Unit 3 Operational data for 2011 to determine if it was a procedural problem of calibrating the Crossflow UFM instrument system, and if so, why it was not detected and corrected by SCE.

    Background: Maximum Power Level Southern California Edison Company (SCE) is authorized to operate the facility at reactor core power levels not in excess of full power (3438 megawatts thermal). Based on its review of the information provided by the SCE regarding the Crossflow UFM system measurement uncertainty and plant power calorimetric measurement uncertainty, the NRC staff finds that the SONGS Units 2 and 3 thermal power measurement uncertainty using the Crossflow UFM is limited to ±0.58 percent of reactor thermal power and can support the proposed increase in licensed reactor power.

  54. February 26, 2013, 89.3 KPCC, Southern California Public Radio, “Boxer says documents from a whistle blower show SoCal Edison was trying to avoid having to reapply for a permit and was “aware” the repairs made to the plant aren’t the ones that should have been done.” I support the Honorable Senator Barbara Boxer 100%. I wish she had taken action earlier.
    1. Dr. Pettigrew: So, you notice the U-bend — the plane of the U-bend is being installed, and on top of the U-bends are bars. They are anti-vibration bars. And so you can see here that from the point of view of out-of-plane motion, the tubes are really very well supported because you have a large number of bars all around; but from the point of view of in-plane motion, there’s really no positive restraint here to prevent the tube to move in the in-plane direction. Essentially, it relies on friction forces to limit the vibration.
    2. Westinghouse states, “Test data shows that the onset of in-plane (IP) vibration requires much higher velocities than the onset of out-of-plane (OP) fluid-elastic excitation. Hence, a tube that may vibrate in-plane (IP) would definitely be unstable OP. A small AVB gap (3 Mil) that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion. Manufacturing Considerations: None were extensively treated in the SCE root cause evaluation.”
    3. AREVA states, “At 100% power, the thermal-hydraulic conditions in the U-bend region of the SONGS replacement steam generators exceeded the past successful operational envelope for U-bend nuclear steam generators based on presently available data. The primary source of tube-to-AVB contact forces is the restraint provided by the retaining bars and bridges, reacting against the component dimensional dispersion of the tubes and AVBs. Contact forces are available for both cold and hot conditions. Contact forces significantly increase at normal operating temperature and pressure due to diametric expansion of the tubes and thermal growth of the AVBs. After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location. There are 36 U-bends in Unit 2 SG E-088 and 34 in SG E-089 with a separation less than or equal to 0.050 inches. These tubes are the first ones to break/rupture in 5 months or during an accident.”
    4. John Large States, “Causes of Tube and Restraint Component Motion and Wear: My study of the various OAs leads me to the following findings and opinion that; (i) degradation of the tube restraint localities (RBs, AVBs and TSPs) occurs in the absence of fluid elastic instability (FEI) activity; (ii) TTW, acknowledged to arise from in-plane FEI activity, generally occurs where the AVB restraint has deteriorated at one or more localities along the length of individual tubes; (iii) the number of tube wear sites or incidences for AVB/TSP locations outstrips the TTW wear site incidences in the tube free-span locations. I find that the ‘zero-gap’ AVB assembly, which features strongly in the onset of TTW, is clearly designed to cope only with out-of-plane tube motion since there is little designed-in resistance to movement in the in-plane direction – because of this, it is just chance (a combination of manufacturing variations, expansion and pressurization, etc) that determines the in-plane effectiveness of the AVB; (iv) Uniquely, the SONGS RSG fluid regimes are characterized by in-plane activity, which is quite contrary to experience of other SGs used in similar nuclear power plants in which out-of-plane fluid phenomena dominate. Moreover, from the remote probe inspections when the replacement steam generator (RSG) is cold and unpressurized, I consider it impossible to reliably predict the effectiveness of the many thousands of AVB contact points for when the tube bundle is in a hot, pressurized operational state., and (5) v) The combination of the omission of the in-plane AVB restraints, the unique in-plane activity levels of the SONGS RSGs, together the very demanding interpretation of the remote probe data from the cold and depressurized tube inspection, render forecasting the wear of the tubes and many thousands of restraint components when in hot and pressurized service very challenging indeed.
    5. John Large continues, “Phasing of AVB-TSP Wear -v- TTW: I reason that, overall, the tube wear process comprises two distinct phases: First, the AVB (and TSP) -to-tube contact points wear with the result that whatever level of effectiveness is in play declines. Then, with the U-bend free-span sections increased by loss of intermediate AVB restraint(s), the individual tubes in the U-bend region are rendered very susceptible to FEI induced motion and TTW. Whereas the OAs commissioned by SCE broadly agree that the wear mechanics comprises two phases, there are strong differences over the cause of the first phase comprising in-plane AVB wear: AREVA claim this is caused by in-plane FEI whereas, the contrary, Mitsubishi (and Westinghouse) favor random perturbations in the fluid flow regime to be the tube motion excitation cause. Put simply: (i) if AREVA is correct then reducing the reactor power to 70% will eliminate FEI, AVB effectiveness will cease to decline further and TTW will be arrested; however, to the contrary, (ii) if Mitsubishi is right then, even at the 70% power level, the AVB restraint effectiveness will continue to decline thereby freeing up longer free-span tube sections that are more susceptible to TTW; or that
    (iii) the assertion of neither party is wholly or partly correct. As I have previously stated, I consider that AVB-to-tube wear is not wholly dependent upon FEI activity.
    6. John Large continues, “Tube Wear Rates – Predicting the In-Service Period: SCE presents the findings of its commissioned OAs in a positive light, claiming that at 70% power the restarted Unit 2 plant will maintain RSG tube integrity for 16 to 18 months of continuous running, that is considerably longer than the proposed 150 day inspection interval. However, closer study of the OAs reveals that the reasoning behind important aspects of the deterioration period for the AVB effectiveness in Unit 2 is flawed, being overly dependent upon a number of uncertainties that I identify and expand upon in my affidavit. Some account of these uncertainties has been taken by AREVA in revising the TTW time-to-burst period down to 2.5 months which is well below the 150 days inspection interval but, without much justification, it determines and front-ends the time-to burst with a further 3.5 month AVB wear-in period, thereby delaying the onset of TTW and the unacceptable level of risk of tube burst to about 1 month longer than the proposed inspection period. I have little confidence in the outcome of the AREVA and other OAs projection of the time period through which the Unit 2 nuclear plant could be reliably expected to operate without a) incurring a tube failure or b) running at a greater risk of a tube failure occurring. This is primarily because (i) it is generally accepted that Unit 2 is following along the same path of deterioration as Unit 3 (AVB wear and loss of effectiveness preceding TTW), although the reasons why it lags so much behind are not at all understood by SCE and, indeed, subject to disagreement between the OA consultants; (ii) moreover, the pattern of AVB breakdown is not clear from the more advanced TTW degradation of Unit 3, thus the extrapolation to Unit 2 is not robust – again, there is disagreement between the OAs on this; so, it follows, (iii) there is very little justification in adding to the time-to-burst for Unit 2 tubes a 3.5 month AVB wear-in period, this is particularly so because so there is no certainty of just where Unit 2 is presently at along the path towards TTW wear. In account of these uncertainties, together with the uniqueness of the in-plane FEI in the SONGS RSGs that I will touch upon later, I consider that restarting Unit 2 to continuous running, even at 70%, will incur a great deal of change, test and experiment.
    7. John Large continues, “Plugging of the at-risk tubes is not a satisfactory solution because it is the retainer bar that vibrates via random fluid flow processes at sub FEI critical velocity levels – these are likely to continue in play or, indeed, exacerbate at the proposed U2 restart at 70% power, leading to through-tube abrasion, the detachment of tube fragments, lodging at other unplugged and in-service tube localities, resulting in the so-called ‘foreign object’ tube wear.”
    8. Comments from Mel Silberberg [NRC-RES, Retired [Chief, Severe Accident Research Branch; Waste Management Branch] to Region IV: I am disappointed in the composition of the special panel! Where is the representation from NRC-RES? The issues at SONGS involve thermal hydraulics and material science. The NRC-RES and its contractors are experts in these areas. The Office of Research was created by the Congress for such situations. Two RES staff covering these disciplines and one or two consultants, serving as peer-reviewers. Perhaps there needs to be a separate peer review. Public confidence can only be gained using logical, informed measures as I described above. Inspection Reports are only one facet of the problem, no question. However, understanding the reasons for the fluid instability, possible cavitation corrosion effects, etc. are phenomena which require evaluation by T/H as well as materials experts, with appropriate oversight by the ACRS. The SCE, the nuclear industry, the NRC and the public need assurance, not educated guesses. I have not seen a bona fide attempt to understand resolve the issue such that all can be alert to potential problems. I still remain puzzled as to why the ACRS [at least one of the Subcommittees]. I am trying to reach the ACRS Exec. Director to discuss this point. Thank you.”
    9. According to NRC Insiders, “NRC does not really have experts in T-H, Materials and QA. You may find one or two at the ACRS and none at the ASLB.”
    10. ATHOS Modeling Limitations: NRC AIT Report states, “The result of the independent NRC thermal-hydraulic analysis indicated that differences in the actual operation between units and/or individual steam generators had an insignificant impact on the results and in fact, the team did not identify any changes in steam velocities or void fractions that could attribute to the differences in tube wear between the units or steam generators. The above analyses apply equally to Units 2 and 3, so it does not explain why the accelerated fluid-elastic instability wear damage was significantly greater in Unit 3 steam generators. The ATHOS thermal-hydraulic model predicts bulk fluid behavior based on first principals and empirical correlations and as a result it is not able to evaluate mechanical, fabrication, or structural material differences or other phenomena that may be unique to each steam generator. Therefore this analysis cannot account for these mechanical factors and differences which could very likely also be contributing to the tube degradation.” Based on comments from Dr. Pettigrew and other researchers, the results of ATHOS Models for FEI are very time-consuming and expensive to conduct and the results can vary from 20 to 100%. Westinghouse states, “We’ve performed similar tests in the past, and that’s what our analytical codes are based on, the data from the tests we’ve performed in the past. They may not be as extravagant as Dr. Pettigrew’s, but yeah, we perform tests, and that’s what our models are based on only out of plane. Basically, we’re using the same tools that we’ve used. We’re just staying within our comfort levels. We’re not pushing our design limits. We’re staying with what we know, what’s been proven to work in the past.” AREVA Staes, “And our analysis codes also are based upon testing that was performed in mockups and boilers in France, which is where most of the design work occurs for the replacements. But they — those tests were performed in the late ‘80s and early ‘90s to validate some of the design changes that they were making to the components. In the same context, we are in France developing a new thermal hydraulic code. That’s been underway, but as you might imagine, the development of a code that has to handle so many variables and these conditions that are very uncertain, it’s time — you know, you have to vet the process and make sure that, again, you’re staying in the bounds of what you’ve known and your technology that you’ve used, and continually use that to benchmark anything new that you’re working on or that you’re developing. But as far as the pinnacle of the replacement market or the replacement design, I would say that most of it is fairly standard, you know, at this point. I don’t think there’s anything outside of the norm that anyone is looking at.” John Large states, “The input energy is the dynamic velocity (~ ) of the two-phase fluid impinging on the tube. The energy dissipation is via damping which is strongly related to the two-phase mix of the fluid, here water and steam as described by the void fraction. Increase in steam content, a greater void fraction, reduces the damping and, correspondingly, the increased volume results in an increase of the impinging velocity. AREVA states, “At 100% power, the thermal-hydraulic conditions in the u-bend region of the SONGS replacement steam generators exceed the past successful operational envelope for U-bend nuclear steam generators based on presently available data.” The inference here is that AREVA is comparing like-with-like, but that would require AREVA having undertaken an ATHOS flow analysis48 for each of the comparative SGs. This I consider unlikely because for this AREVA would have required access to very detailed information on the design geometry and flow paths throughout the comparative SG tube bundles – being a designer/manufacturer of steam generators itself, I very much doubt that AREVA would have had access to such proprietary information from competitor manufacturers. So since it is unlikely that AREVA would have carried out an ATHOS computer simulation for each of the five (A to F) comparative nuclear plants, then analysis is unlikely to be directly comparing two-phase fluid flow velocity distribution in the critical FEI regions of the SONGS and comparative plant SG tube bundles. I can only surmise that the analysis comparison is between the mean or average velocity within the overall tube bundle for SONGs and each of the comparative plants. Moreover, since the velocity distributions within each of the comparative plants, because of different design geometries, flow areas, etc, will not be identical, it is very unlikely that the mean or average velocity provides even a crude basis of comparison of the FEI potential of the SONGS RSGs.” The question is as Arnie Gundersen asks, “How similar to the SONGS S/Gs are these other S/Gs? Do the other steam generators, for example, use alloy 670 tubes and have similar spacing, similar support structures, etc.? To the best of my knowledge and belief, no other steam generator in the nation is as large as those at San Onofre with broached tube supports, a tight Combustion Engineering tube pitch, and no stay cylinder. Therefore, comparing San Onofre to “several other successfully operating large S/G’s” is simply not a valid engineering or scientific comparison.”
    11. Controversy regarding Removal of Central Stay Cylinder: Palo Verde and ANO Unit One 2 replaced the RSGs without removing the central stay cylinder, added more tubes and made the tubes taller (Read NUREG-1841) with NRC Region IV blessings and SONGS avoided that Blessing because it did not believe in the NRC Blessing Process in 2004 and 2013 as demonstrated in Yesterday’s SCE/NRC Meeting. . NRC AIT Report states, “The licensee’s bid specification required that the stay cylinder feature of the original steam generators be eliminated to maximize the number of tubes that could be installed in the replacement steam generators and to mitigate past problems with tube wear at tube supports caused by relatively cool water and high flow velocities in the central part of the tube bundle.” Elimination of the stay cylinder to increase the added 377 tubes and increase the average length of the heated tubes for increasing the heat transfer area by ~ 11% in The RSGs caused the following problems: (a) John Large states, “Indeed, this need to increase the heat transfer area (ie putting more tubes into the RSGs) and, with this, reducing the steamside flow area, may have been a strong contributory factor to the enhanced FEI activity in the SONGS FSGs. Moreover, the location of the additional tubing, particularly in what I would describe as the lower swirl space immediately above the tube support sheet, may have contributed to and/or determined the unique in-plane flow characteristics of the SONGS RSGs.” (b) Arnie Gundersen states, “The center section of the original San Onofre steam generators contained a key structural element called a “stay cylinder” and no steam generator tubes. In 2005 or early 2006, Edison made a management decision to eliminate this vital support pillar and add additional tubes in its place. In the original steam generator design, there was no heat input in this central area of the steam generator, because there were no tubes to add the heat. When Edison added almost 400 tubes (4% of the tubes) to the center of the tube bundle in the San Onofre Replacement Steam Generators, Edison effectively increased the power distribution to the center of the steam generator. This radical and unanalyzed design change moved 4% of the heat to the inside of the tube bundle while reducing the heat by 4% to the outside of the tube bundle. Adding this heat to the center of the bundle was then exacerbated by removing the egg crate tube supports and replacing them with a broached tube support plate design that further reduced flow to the center of the steam generator. As the NRC confirmed in its AIT report, a large steam void has developed near where the additional tubes were added in the Replacement Steam Generators (called fluid elastic instability) that allows many types of excess vibrations to occur. Fairewinds review of Figure 1 below from Edison’s Condition Report clearly shows that the location within the steam generators where the steam “fluid elastic instability” has developed is precisely the region where the extra heat created by the 400 new tubes would create an excess of steam and various vibrational modes. While 4% may seem like a small change, it is not. Each San Onofre reactor generates a total thermal output of approximately 3400 megawatts of heat. If one mathematically converts 4% of 3400 megawatts of heat, it equals 135 megawatts, or to illustrate it differently: 180,000 horsepower of thermal heat that was transferred from the outside of the tube bundles to the center, (c) Unit 3 has historically produced more power than Unit 2 (1186 MWe vs. 1183 MWe, 1178 MWe vs. 1172 MWe). Westinghouse states, “In the U-bend region, the gap velocities are a strong function of power level. The steam flow in the bundle is cumulative and increases as a function of the power level and the bundle height which causes high fluid quality, void fraction, and secondary fluid velocities in the upper bundle.” According to the Plant Procedures, Unit 3 RCS flow is 79.79 Million Ibs/hour and the delta T between Hot leg and Cold Leg is 58 degrees Fahrenheit. According to the SONGS Plant Procedures, Unit 2 RCS flow is 75.76 Million Ibs/hour and the delta T between Hot leg and Cold Leg is 57 degrees Fahrenheit. I go along with Westinghouse that higher power of 79.79 Million Ibs/hour caused FEI in Unit 3, because steam saturation temperature was achieved due to lower secondary side pressures of 833 psi and poor circulation ratios of 3.3 earlier in the U-tube bundle than anticipated due to increased average length of the tubes from 680 inches to 730 inches. The critical heat flux was achieved in Unit 3 area of wear ( ~ Hot Leg side Columns 75 -90, Rows 90-120, vertically located z-axis cut at about 20 inches above the 7th TSP) due to increased height of the bundle and narrow tube to pitch diameter. This caused high fluid quality, void fraction, and secondary fluid velocities in Unit 3 area of TTW. Unit 2 did not experience FEI, because power levels were low and critical heat flux was lower at the same point of wear as Unit 3, and by the time steam-water mixture achieved high fluid quality, void fraction, and secondary fluid velocities, it exited the u-tube bundle. It is also my observation, that FEI is an intermittent phenomena controlled by the varying circulation ratios and pressures in the steam generator. When this happens, the RCS return flow temperature in the cold leg is higher than usual, say by 4 degree Farenheit, even though the power is at a steady state level subject to CROSSFLOW uncertainty calculation of 0.5%. FEI causes movement of U-Tubes with large amplitudes starting from the tube sheet to the top of the U-bend and this could have conceivably caused tube-to-tube violent impact at the bottom of the tube sheet registering VLPMS alarms in Unit 3. Since Unit 2 did not experience FEI, no VLPMS alarms occurred in Unit 2.
    12. Palo Verde RSG Design: The tube supports have three basic configurationsl-() horizontal grids (eggcrates/lattice) that provide support to the vertical run of the tubes, (2) vertical grids that provide vertical and horizontal support to the horizontal run of the tubes in the upper bend region, and (3) diagonal strips (batwings) that provide out-of-plane support to the 90-degree bends. The upper tube bundle support system (1) supports the horizontal tube spans against high velocity, two-phase cross flow, (2) permits an expanded vertical tube pitch (from 1.0 inch to 1.75 inches) so as to promote free flow through the bend region and prevent low-flow dryout regions, and (3) supports the upper tube bundle via structural beams against postulated accident condition loads, seismic loads, transportation loads, and dead weight. The U-bend support structure for the replacement steam generator differs from the original design in that it includes welded connections between the vertical grids and the diagonal (batwing) supports. Other features of the U-bend support system are that the batwings bisect the 90-degree bends, the bend region supports are perforated and narrower than the original design, and the bend region supports have ventilation holes. These changes in design improve the thermal/hydraulic conditions in the upper bundle region, preventing crevice dryout and reducing secondary-side fouling, as well as addressing tube-wear phenomena observed in the original steam generator. The diagonal strips (batwings) are located at every row and are designed to prevent out-of-plane deflection and thus preclude the deflection amplitude required for fatigue. The replacement steam generator design has an increased circulation ratio when compared to the original steam generator.
    13. ANO Unit 2 RSG Design: Strict ovality control was implemented during the manufacture of the tubes to limit dimensional variability in the U-bend region. The thickness of the AVBs was also tightly controlled. To limit the potential for U-bend vibration and wear, AVBs support the U-bends. The AVBs provide sufficient support to the U-bend so that all the tubes remain elastically stable even if it is assumed that some of the support points are inactive. The AVBs in adjacent columns are inserted to different depths (i.e., staggered) to limit the U-bend pressure drop and to discourage the formation of flow stagnation regions. The AVBs are nearly perpendicular to the centerline of the tubes at all locations in the U-bend region to provide support without unnecessary tube contact. These features provide margin against flow stagnation, corrosion, and tube vibration.
    14. The NRC AIT Report states, “The team identified that the design of the replacement steam generators did not expect any potential vibration concerns in the area of the tube bundle where the retainer bars were located. The basis for Mitsubishi’s design philosophy relied on the following factors: (a) Based on the calculated natural frequency of the retainer bar, Mitsubishi considered that there would not be a resonant vibration condition relative to the flow conditions in the location of retainer bars, and (b) The vibration analysis of the tube bundle only considered out-of-plane vibration because in-plane vibration was not expected to be an operational concern for the retainer bars. The outermost tubes were considered the least susceptible to flow-elastic instability; therefore retainer bar locations were not included in the vibration analysis. Retainer Bars in other MHI SGs range with a frequency between 120-1180. Because of the excessive number of tubes due to change of Alloy 600MA to Alloy 690TT, the tube-to-tube clearance tightened towards the apex of the U-bend. Therefore, the restraint assemblies required a smaller diameter retainer bar with 56 HZ frequency in order to fit between the tube rows.
    15. SCE SNO states: As a plant operator, I have operational control over two of the three components needed for in-plane fluid elastic instability. Specifically, reducing power can reduce steam velocity and reduce steam dryness sufficiently to preclude in-plane fluid elastic instability since the conditions for fluid elastic instability no longer occur concurrently.
    16. SCE In Enclosure 2 states, “A Probabilistic Risk Assessment (PRA) was performed to analyze the risk impact of the degraded SG tubes on SONGS Unit 3 SG 3E-088 with respect to two cases: (1) any increased likelihood of an independent SG tube rupture (SGTR) at normal operating differential pressure (NODP), or (2) due to a SGTR induced by an excess steam demand event, also referred to as a main steam line break (MSLB). The SONGS PRA model was used to calculate the increases in Core Damage Probability (CDP) and Large Early Release Probability (LERP) associated with each case. In both cases, all postulated core damage sequences are assumed to result in a large early release since the containment will be bypassed due to the SGTR; therefore, the calculated CDP and LERP are equal. The total Incremental LERP (ILERP) due to the degraded SG tubes (i.e., the sum of the two analyzed cases) was determined to be less than 2×10-7. This small increase in risk is attributed to two factors. First, the exposure time for the postulated increased independent SGTR initiating event frequency case was very short (0.1 Effective Full Power Month (EFPM)). Second, a MSLB alone does not generate sufficient differential pressure to cause tube rupture in Case 2. The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators prevent RCS re-pressurization in accordance with Emergency Operating Instructions.
    17. So from the above Information, I conclude that the “As-designed and Degraded Unit 2 RSG Tube-Bundle is not capable of preventing the FEI caused multiple tube ruptures from a MSLB with failure of a MSIV to close. We saw that in Unit 3 with failure of 8 tubes at MSLB test conditions. SCE SNO says, he has control of the plant operations, but SCE Enclosure 2 contradicts SNO with an “If statement” by stating,” The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators prevent RCS re-pressurization in accordance with Emergency Operating Instructions.”
    18. Operator Action as claimed by Edison to detect the leak by N-16 and other radiation monitors and the ability to re-pressurize the steam generators are not practical to stop a major nuclear accident in Unit 2 in progress in the first 15 minutes of a MSLB due to the following factors:
    • The operator action will not work due to the short duration of the initial and devastating event, the radiation/steam environment, communication errors between the control room and field operators due to sonic booms and hissing steam noises (sound-powered phones, pagers, cell phones and radios will not work in such an environment), darkness, difficult terrain and other unknown equipment failures/troubles [e.g. San Onofre’s auxiliary feed-water steam supply piping, which would provide water to the steam generators, if their main supply was lost is vulnerable to a big flood; A big fire in auxiliary feed-water pump room would knock 2 out of three pump’s electrical circuits, etc.], and other contingencies.
    • During a partial walk down of the Unit 2 high-pressure safety injection system in August 2011, NRC inspectors found a drain valve partially open, when it was required to be closed. “Operations personnel failed to implement instructions for filling, venting, draining, startup, shutdown, and changing modes of operation for emergency core cooling systems as written,” the NRC said. “seismic class I valves continue to be miss positioned, safety-related plant systems may be unable to accomplish their safety functions after an accident”.
    • One of the well-known SONGS Shift Managers told SCE Management that he was not going to put his “License on the line” by operating a “Defective Unit.” Several other shift managers have retired rather than work for SONGS’ Profit-Motivated and Retaliating Management.
    • The Operator Union has warned the SCE Management that with the proposed operator reductions, it will not be safe to restart Unit 2.
    • There have been 10 SGTRs (or significant leaks) in U.S. PWRs from 1975 to 2000. Human performance weaknesses, such as misdiagnosis, substantial delays in isolating the faulted steam generator, and delayed initiation of the residual heat removal system, have been identified in these events. The events also involved unnecessary radiation releases; lack of RCS subcooled margin, excessive RCS cooldown rates, and overfilling the SG because of human or procedural problems.
    • Additional complications would add to operator burdens. These include high noise levels preventing normal communications; RCS cooldown with potential recriticality; actions to recover RWST inventory; many radiation alarms, unexpected high radiation areas in the turbine building, and atmospheric releases; fire alarms and fires from steam and shrapnel from the break; and emergency communications with local, state, and Federal governments diverting operations personnel before the technical support center is manned or additional operations personnel arrive on site. The Halden Control Room Staffing study found poor operator performance in one of two simulations of a SG leak with a failed open SG safety relief valve, as well as simulations where crew size was decreased to attend to other duties.
    • Below are some of the weaknesses witnessed during review and/or observation of the Simulator Evaluations, Emergency Planning Drills and discussions with the Shift Managers during 2012. Each weakness may be attributed to one or the other Drills/Exercise Performance (DEP) Miss-classifications:
    A. Unclear and confusing Emergency Action Levels (EALs) and less than adequate Basis Documents.
    B. Too many Priority Reading Assignments to clarify the EALs and Basis Document.
    C. Lack of solid teamwork between the Operating Crew, Control Room Supervisor (CRS), Station Technical Advisor (STA) and Emergency Coordinator/Shift Manager (EC/SM).
    D. Crew members confused and concerned about their roles and responsibilities. Crewmembers held back or failed to provide information, which resulted in SM and CRS to trip the reactor.
    E. Poor communications between the Operating Crew, CRS, STA and EC. Briefs were ineffective at focusing on the crew priorities. Three way communication not used for direction or when providing information relative to plant status.
    F. Poor diagnostics/interpretation of the transient events by the Operating Crew, CRS, STA and EC. Serious omissions, delays, or errors made in interpreting indications resulting in degraded plant conditions. Failed to use, or misused, or misinterpreted indications that resulted in improper diagnosis.
    G. Procedures were not followed correctly which impeded plant recovery or caused unnecessary degradation of plant conditions. Crews did not recognize EOI Entry Conditions.
    H. Repeat failures of the STA to provide consistent & independent check of the EAL by EC.
    I. Lack of Stringent Operations Department/NTD Evaluation and Remediation Criteria for SM/STA/ Operations Crew to achieve excellence and eliminate above shortcomings to prevent DEP Failures.
    J. Lack of practice by the Operating Crews, CRS, STA and EC following the coaching/critique provided by the OPS SM Supervisor and NTD Evaluators.
    • During the April, 2012 Fire Notification of Unusual Event, it took 40 minutes between the Control Room and Electricians to find the drawings to determine the location of the breaker to de-energize the power to the electrical panel in the Unit 2 turbine building to extinguish the fire and terminate the event. Luckily, the Unit 2 was in shutdown and the SONGS Fire Department was present at the scene to extinguish the fire, if it got out of the control. Later it was determined, that the Fire Department and Control room did not take timely action to extinguish the fire due to an over-conservative fire procedure.
    • In 2001, Ratepayers lost 100 million dollars in a Unit 3 Switchgear fire due to faulty alarms and miscommunication between the SONGS Fire Department and Control Room. Unit 3 was in shutdown for 5 months due to Main Turbine repairs, which was damaged in the fire event.
    • During the 2011 NRC/FFEMA Evaluated Exercise, the General Emergency Declaration was missed by 29 Minutes due to a communication error between the Emergency Offsite Facility Health Physics Supervisor and Technical Support Center Station Emergency Director. If this was a real event, the public would have been potentially subject to offsite radiation releases unnecessarily for 29 minutes.
    THANKS To NRC FOR POSTING THIS BLOG

  55. Special Thanks to NRC and Moderator Mr. Victor Dricks for posting this blog
    SPECIAL Public/NRC/SCE Awareness Series by HAHN Baba
    Subject: 2-6-2013: Senator Barbara Boxer Letter To The Honorable Allison M. Macfarlane (Continued)
    So what is new. SCE and MHI have been avoiding regulatory process since 2004 under the pretense of, “like for like.” NRC Region IV has not done anything to stop that. Now Cat is out of the bag. Senator Barbara was advised in August 2012, and a recommendation was made to form a Joint Task Force of Justice Department, Senate Committee on Environment and Public Works and NRC to investigate both SCE/MHI. But, that recommendation was swept under the rug for reasons unknown.
    Email To President Obama’s Campaign – info@2013pic.org
    Attention: Your Excellency Honorable President Barack Obama, Greatest People’s President in the Modern History of United States
    Continued cover-up by NRC Region IV, Southern California Edison and Mitsubishi Heavy Industries in San Onofre Nuclear Generating Station Replacement Generators has become a nightmare for 8.4 Million Southern Californians. Here is a copy of a press release FYI.
    On Wednesday February 6, 2013 , Sen. Barbara Boxer pressed federal regulators to open an investigation at the plant after uncovering documents that she said suggest that Southern California Edison took engineering shortcuts and compromised safety. The Democratic senator said in a letter to Nuclear Regulatory Commission Chair Allison Macfarlane that a confidential report obtained by her office shows Edison and Mitsubishi Heavy Industries, the Japan-based company that built the plant’s steam generators, were aware of design problems before the equipment was installed in 2009 and 2010. Boxer, who chairs the Environment and Public Works Committee, said the report written by Mitsubishi raises concerns that Edison and its contractor rejected safety modifications and sidestepped a more rigorous safety review. “Safety, not regulatory short cuts, must be the driving factor in the design of nuclear facilities, as well as NRC’s determination on whether (San Onofre) can be restarted,” Boxer said in a letter co-signed by Rep. Edward Markey, D-Mass.
    Billion Dollar Safety Question is who is telling the truth? It is time now that a Joint Task Force of Justice Department, Senate Committee on Environment and Public Works and NRC ASLB & San Onofre Special Panel be created to investigate both SCE/MHI Engineers, NRC Region IV and NRC AIT Engineers involved with San Onofre and let them testify under oath. I will testify under oath to help everybody with whatever I know on the Joint Task force. 8.4 Million Southern Californians will certainly appreciate their President’s help in resolving this matter of Great National Interest.
    FEI did not occur in Unit 2, because, (1) It was the absence of high steam dryness ALONE in Unit 2 that FEI did not occur in Unit 2 (emphasis added), and (2) Not because of the better supports (Dietrich is stating backwards and claiming incorrectly) and/or differences in fabrication, which resulted in substantially increased contact forces (reduced looseness) between tubes and AVBs for Unit 2 and prevented FEI from occurring. These supports are designed for out-of plane protection and not for-in plane protection. MHI stated, “In the design stage, MHI assumed that the tube support in the out-of-plane direction with “zero” tube-to-AVB gap in hot condition was sufficient to prevent tube from becoming fluid-elastic unstable during operation. But, the recent SONGS experience shows that the flat bar AVBs does not provide friction forces required to prevent tubes from vibrating in the in-plane direction and eventually becoming fluid-elastic unstable under high local secondary thermal-hydraulic conditions such as in the SONGS RSGs. In addition, MHI concludes that in the Unit-3 RSGs low tube and AVB fabrication dimensional dispersion causes that the tube-to-AVB contact forces are not sufficient to prevent the in-plane motion of tubes. Because of pressure from NRC and SCE and to continue Business in United States, MHI has reverted its stand from its original design position and contemporary experience. Honorable Senator Barbara Boxer is absolutely right, when she contends, “Mitsubishi and Edison were aware of safety problems with steam generators before the equipment was installed starting in 2009 but rejected some enhancements to avoid a more rigorous regulatory review. SCE and MHI accepted some adjustments to the replacement steam generators, further safety modifications were found to have “unacceptable consequences” and were rejected: “Among the difficulties associated with the potential changes was the possibility that making them could impede the ability to justify the RSG [replacement steam generator] design” without the requirement for a license amendment. The Report also indicates that SCE’s and MHI’s decision to reject additional safety modifications contributed to the faulty steam generators and the shutdown of reactor Units 2 and 3.” DAB Safety Team member warned Senator Barbara Boxer of similar concerns in August 2012, but the warning were swept under the rug for reasons unknown. DAB Safety Team findings on Unit 2 FEI and supports are consistent with the findings of AREVA, Westinghouse, John Large, SONGS RCE Anonymous Root Cause Team Member and latest research performed by Eminent Professor Michel Pettigrew and others in 2006. Therefore, SCE claims that insufficient contact forces in Unit 3 Tube-to-AVB Gaps ALONE caused tube “to” tube wear are misleading, erroneous and designed to put the blame on MHI for purposes of making SCE look good in the public’s eyes and for collecting insurance money from MHI’s manufacturing so called defects….. Thanks HAHN Baba

  56. Special Thanks to NRC and Moderator Mr. Victor Dricks for posting this blog
    SPECIAL Public/NRC/SCE Awareness Series by HAHN Baba
    Subject: 2-6-2013: Senator Barbara Boxer Letter To The Honorable Allison M. Macfarlane
    So what is new. SCE and MHI have been avoiding regulatory process since 2004 under the pretense of, “like for like.” NRC Region IV has not done anything to stop that. Now Cat is out of the bag. Senator Barbara was advised in August 2012, and a recommendation was made to form a Joint Task Force of Justice Department, Senate Committee on Environment and Public Works and NRC to investigate both SCE/MHI. But, that recommendation was swept under the rug for reasons unknown.
    Billion Dollar Safety Question is who is telling the truth? It is time now that Joint Task Force of Justice Department, Senate Committee on Environment and Public Works and NRC ASLB & San Onofre Special Panel be created to investigate both SCE/MHI Engineers, NRC Region IV and NRC AIT Engineers involved with San Onofre and let them testify under oath. I will testify under oath to help everybody with whatever I know on the Joint Task force.
    On Wednesday, Sen. Barbara Boxer pressed federal regulators to open an investigation at the plant after uncovering documents that she said suggest that Southern California Edison took engineering shortcuts and compromised safety. The Democratic senator said in a letter to Nuclear Regulatory Commission Chair Allison Macfarlane that a confidential report obtained by her office shows Edison and Mitsubishi Heavy Industries, the Japan-based company that built the plant’s steam generators, were aware of design problems before the equipment was installed in 2009 and 2010. Boxer, who chairs the Environment and Public Works Committee, said the report written by Mitsubishi raises concerns that Edison and its contractor rejected safety modifications and sidestepped a more rigorous safety review. “Safety, not regulatory short cuts, must be the driving factor in the design of nuclear facilities, as well as NRC’s determination on whether (San Onofre) can be restarted,” Boxer said in a letter co-signed by Rep. Edward Markey, D-Mass.
    But Edison said in a statement “it is simply not accurate” to suggest the company was aware of design problems, and pointed out the equipment carried a 20-year warranty against defects. “SCE would never, and did not, install steam generators that it believed would not perform safely,” the company said. Edison “sought to purchase replacement steam generators that would meet or
    Mitsubishi said design decisions were made “in accordance with well-established and accepted industry standards” along with a wealth of operating experience. “Nothing is more important to us than the safe design and manufacturing of nuclear-energy facilities,” a company statement said. “A thorough investigation has been ongoing and will continue. We will continue cooperating fully.”
    In a statement, the NRC said it received the letter and “will review all available information in making a judgment as to whether the plant would meet our safety standards if restart were permitted.”
    Special Thanks to NRC and Moderator Mr. Victor Dricks for posting this blog

  57. Special Thanks to NRC and Moderator Mr. Victor Dricks for posting this blog
    SPECIAL Public/NRC/SCE Awareness Series by HAHN Baba
    Subject: 2-6-2013: Senator Barbara Boxer Letter To The Honorable Allison M. Macfarlane
    We have become aware of new information contained in a 2012 Mitsubishi Heavy Industries (MHI) document entitled “Root Cause Analysis Report for tube wear identified in the Unit 2 and Unit 3 Steam Generators of San Onofre Generating Station” (Report). The Report indicates that Southern California Edison (SCE) and MHI were aware of serious problems with the design of San Onofre nuclear power plant’s replacement steam generators before they were installed. Further, the Report asserts that SCE and MHI rejected enhanced safety modifications and avoided triggering a more rigorous license amendment and safety review process.
    For example, the Report states that although SCE and MHI accepted some adjustments to the replacement steam generators, further safety modifications were found to have “unacceptable consequences” and were rejected: “Among the difficulties associated with the potential changes was the possibility that making them could impede the ability to justify the RSG [replacement steam generator] design” without the requirement for a license amendment. The Report also indicates that SCE’s and MHI’s decision to reject additional safety modifications contributed to the faulty steam generators and the shutdown of reactor Units 2 and 3.
    This newly-obtained information concerns us greatly, and we urge the NRC to immediately conduct a thorough investigation into whether SCE and MHI did in fact fail to make needed safety enhancements to avoid the license amendment process. All people in our nation, including the 8.7 million people who live within 50 miles of the San Onofre plant, must have confidence in the NRC’s commitment to put safety before any other concern. We believe this alarming Report raises serious concerns about SCE’s and MHI’s past actions. Safety, not regulatory short cuts, must be the driving factor in the design of nuclear facilities, as well as NRC’s determination on whether Units 2 and 3 can be restarted.
    So what is new. SCE and MHI have been avoiding regulatory process since 2004 under the pretense of, “like for like.” NRC Region IV has not done anything to stop that. Now Cat is out of the bag. Senator Barbara was advised in August 2012, and a recommendation was made to form a Joint Task Force of Justice Department, Senate Committee on Environment and Public Works and NRC to investigate both SCE/MHI. But, that recommendation was swept under the rug for reasons unknown. Anyhow, the bottom line is neither, SCE nor MHI have the knowhow to design/build a CE Replacement Steam Generator. That being said, only Westinghouse has the skills and technology to design and build a CE Replacement Steam Generator. SCE was told that in June of 2012. But the problem is nobody listens, until it is too late. For example, Westinghouse/Combustion Engineering designed several CE Replacement Generators in 2000-2005 (e.g., PVNGS 1, 2, 3, ANO-2, etc.), which are running successfully. NRC Region IV licensed these generators with assistance from NRC Commission under a 50.90 process with a “Critical Questioning and Investigative Attitude.” During San Onofre Replacement Generator Design, Installation & Accident Investigation Process, NRC Region IV has been acting as a silent observer going along with SCE instead of a strict regulator for reasons unknown.
    AREVA says in its Operational Assessment, ” Weaver and Schneider [16], in 1983, examined the flow induced response of heat exchanger U-tubes with flat bar supports. It is worth quoting the first
    conclusion of their paper: “The effect of flat bar supports with small clearance is to act as apparent nodal points for flow-induced tube response. They not only prevented the out-of-plane mode as expected but also the in-plane modes. No in-plane instabilities were observed, even when the flow velocity was increased to three times that expected to cause instability in the apparently unsupported first in-plane mode.” Additionally, in an effort to encourage the development of in-plane instability, Weaver and Schneider substantially increased the clearances between flat bar supports and U-tubes, but no in-plane instability was observed. Other investigators, notably Westinghouse, have deliberately searched for in-plane instability with only support from flat bars and have not detected the phenomena. However in 2005, Janzen, Hagburg, Pettigrew and Taylor reported in-plane instability. The abstract to their paper states, “For the first time in a U-bend tube bundle with liquid or two-phase flow, instability was observed in both the out-of-plane and in-plane direction.”
    A document published in 2006, “Fluid-elastic instability of an array of tubes preferentially flexible in the flow direction subjected to two-phase cross flow. (http://yakari.polytechnique.fr/people/revio/masters_research_subject.html) by Violette R., Pettigrew M. J. & Mureithi N. W. stated, “In nuclear power plant steam generators, U-tubes are very susceptible to undergo fluid elastic instability because of the high velocity of the two-phase mixture flow in the U-tube region and also because of their low natural frequencies in their out of plane modes. In nuclear power plant steam generator design, flat bar supports have been introduced in order to restrain vibrations of the U-tubes in the out of plane direction. Since those supports are not as effective in restraining the in-plane vibrations of the tubes, there is a clear need to verify if fluid elastic instability can occur for a cluster of cylinders preferentially flexible in the flow direction.” Retainer bars also suffer with similar problems. SONGS Root cause team members told SCE Management about the problems with ineffective flat bars, but no body listed and they kept repeating only one phrase, “insufficient tube-AVB contact forces caused FEI. Westinghouse and AREVA said, Guys, Once FEI starts, contact forces do not count.
    The point is when you are designing and building such a complex steam generator as a San Onofre replacement steam generator, whether it is a designer or manufacturer, one is supposed to keep up with the latest university research and industry benchmarking. It is absolutely clear that SCE and MHI did not do that. They both broke the federal regulations, public trust, wasted 1 Billion Dollars of Rate Payer’s Money and almost created twin accidents. What is the use of crying now. It appears, now NRC is going on behalf of SCE to Japan to supervise MHI’s quality assurance activities associated with the mock-up and testing of re-designed anti-vibration bars that may be used as a long-term repair of both Unit 2 and Unit 3 San Onofre Nuclear Generating Station (SONGS) steam generators. As I have said with John Large, Arnie Gundersen, Professor Daniel Hirsch, David Lochbaum, and dozen other anonymous steam generator experts and DAB Safety Team before, these defectively-designed and degraded generators will not withstand a MSLB or other anticipated transients due to FEI even at 70% power.
    The bottom line is that MHI was a subcontractor to Westinghouse and they make huge claims, but really do not how to design a CE Replacement Generator. SCE also makes huge claims and preaches safety sermons, but is not a steam generator designer. So NRC is wasting their valuable time with SCE and MHI and wasting money of Southern California’s Rate Payers and putting their safety on line. So I request NRC Chairman, NRC ASLB & San Onofre Special Panel Members, Please tell Westinghouse to build replacement steam generators for San Onofre and tell Ted Craver and Pete Dietrich to fire the SCE Retaliating and inefficient management, work on producing safe electricity and build public trust and respect both for SCE and NRC. Thanks once again to Mr. Victor Dricks for posting this blog.

  58. The Research by the World’s Number 1 Expert in 2006 shows that flat bars are not effective to protect the SG tubes from the adverse effects of Fluid Elastic instability and Low Frequency reatainer bars can damage tubes from turbulence induced random vibrations. According to SONGS insiders, SONGS RCE Team had access to this information. But the results were both ignored by SCE and MHi. Transparenecy and accountabilty is the rule of law for a licensee and its contractor for nuclear safety. if this information true, it is a blatant violation of federal regulations and public trust by SCE and MHI. Senator Barbara Boxer was warned of these types of cover ups in August 2012 and investigation by Justice Department was requested. These recommendations were swept under the rug for reasons unknown. Thanks to Mr. Victor Dricks for posting this blog.

  59. Special Thanks to NRC and Moderator Mr. Victor Dricks for posting this blog
    SPECIAL EIX CEO/Chairman Awareness Series by HAHN Baba
    Promoting “Critical Questioning & Investigative Attitude”
    EIX CEO/Chairman Ted Craver should follow the example of wise decision taken by Brilliant Duke Energy CEO Jim Rogers and shutdown the Terminally Sick San Onofre just like Terminally Sick Crystal River. The EIX/SCE should review alternatives to replace the power produced by San Onofre by construction of new, state-of-the-art, natural gas-fueled/solar 50-100 MW plants installed throughout the grid. The decision to retire the Terminally Sick San Onofre nuclear plant would be the best in overall interests of Southern California Customers/Public, EIX Investors/Shareholders, the State of California, CPUC and NRC Region IV. The decision would be very difficult, but it would be the right economical and safety choice and politically popular.

  60. Special Thanks to NRC and Moderator Mr. Victor Dricks for posting this blog
    SPECIAL Public/NRC/SCE Awareness Series by HAHN Baba
    Courtesy of The DAB Safety Team
    A total of 24 Alloy 690, chrome-plated retainer bars welded to the retaining bars are provided to prevent AVB structure displacement during SG fabrication and during a limiting design basis accident such as a main steam line break. The retainer bars anchor the AVB structure to the tubes, but are designed not to contact the tubes under operating conditions. As shown in Section 5.5, Edison response to NRR RAI #15, SCE states, “The limited vibration amplitude of the tubes and retainer bar, combined with stabilizer development, prevents developing wear displacement /wear geometry that could severe any of the tubes adjacent to the retainer bars, either in the short term or long term.” This statement is unacceptable, because the conclusions appear to be drawn without any publicly available SCE auditable scientific/testing data and structural, materials engineering and thermal-hydraulic calculations. The structural integrity of SONGS Unit 2 replacement steam generators degraded retainer bar system welds, retainer bars, stabilized and non-stabilized plugged tubes to withstand combined loads that result from postulated accident conditions events as assumed in the RSG Design/FSAR Analysis has not been demonstrated. This includes a design basis earthquake (DBE) in combination with a LOCA (multiple SG tube leak and/or rupture events due to FEI caused by U-tube bundle uncovery) and MSLB (high energy flashing feedwater jet impingement and loose parts causing multiple tube leak and/or rupture events due to SG depressurization). Like John Large says,”Put another way, the extensive and rapid rates of tube wear experience at the SONGS Unit 2 and Unit 3 RSGs, have necessitated an extensive raft of analysis, assessments and projections to qualify, or otherwise, that Unit 2 is fit for purpose. Not only is this prequalifying work unique to the San Onofre nuclear plant, much of it has never been undertaken before so, it follows, its inclusion in safety considerations must be a new and hitherto unconsidered component now required to be incorporated into an updated version of the FSAR.”
    San Onofre NRC AIT Report, SCE Unit 3 Cause Evaluation, SCE Unit 2 Return to Service Reports, SCE Response to NRR RAIs, San Onofre Special Tube Inspection Reports and 10 CFR 50.59/FSAR Justifications need to be thoroughly reviewed and a GAP Analysis prepared by NRC NRR, Civil, Mechanical, Chemical, Materials, Structural, Electrical/I&C, T/H Engineers, Computer Modeling and San Onofre Special NRC Panel Members. San Onofre Special NRC Panel Members need to make accurate and precise engineering decisions based on validated and auditable facts in accordance with Honorable and Respected Dr. McFarlane’s High Standards. These decisions have to be made without any political/financial/time pressures from EIX/SCE Officers, CPUC Chairman, NRC Commissioners, Pro-SCE Politicians, Attorneys or Industry Lobbyists. Ex NRC Branch Chiefs (Dr. Joram Hopenfeld, Mel Silberberg, etc.), Anonymous “Critical Questioning & Investigative Attitude” Genius NRC Branch Chief, US Public, San Onofre Workers and Southern Californians would appreciate San Onofre Special NRC Panel Members “Critical Questioning & Investigative Attitude”, True, Unbiased and Diligent Public Safety Efforts.

  61. Special Thanks to NRC and Moderator Mr. Victor Dricks for posting this blog
    SPECIAL Public/NRC/SCE Awareness Series by HAHN Baba
    Courtesy of The DAB Safety Team
    Causes of SONGS Unit 3 Replacement Steam Generators Fluid Elastic Instability
    Subject: Promoting Human Performance Tool “Critical Questioning & Investigative Attitude”
    SONGS Unit 3 FEI Root Cause: Lack of “Critical Questioning & Investigative Attitude” by SCE/MHI
    Design Causes: Narrow tube pitch/diameter ratio, too many tall tubes and lack-of in-plane supports
    Operational Causes: High Steam Flows, High Steam Velocities and Low Steam Pressures
    Primary Mechanistic Cause: Fluid Elastic Instability AKA Vapor Fraction >99.6% AKA Steam Dry-outs, Lack of Tube Damping (No Thin Water Film on Tube to Release Heat)
    Secondary Mechanistic Causes: Flow-induced Random Vibrations, Excessive Hydrodynamic Pressures (Mitsubishi Flowering effect)
    Human Performance Errors:
    1. Lack of Solid Team Work, Alignment and Design Reviews
    2. Avoidance of 10CFR 50.90 Amendment Process
    3. Financial and Time Pressures
    4. Lack of Academic and Industry Benchmarking
    5. Complacence and Negligence
    6. Lack of NRC Region IV Strict Oversight
    7. Lack of skills, experience and technology to design & build a CE Replacement Steam Generators

  62. San Onofre Special Public/NRC/SCE Awareness Series by HAHN Baba
    Promoting Critical Questioning & Investigative Attitude
    Edison has said exhaustive research by its team of global experts demonstrates the safety of what it calls a conservative plan to re-open SONGS Unit 2. NRC AIT Team/Edison and its team of global experts are not sure among themselves, whether fluid elastic instability occurred in Unit 2. San Onofre NRC AIT Report, SCE Unit 3 Cause Evaluation, SCE Unit 2 Return to Service Reports, SCE Response to NRR RAIs, San Onofre Special Tube Inspection Reports and 10 CFR 50.59/FSAR Justifications need to be thoroughly reviewed and a GAP Analysis prepared by brilliant NRC NRR, Civil, Mechanical, Chemical, Materials, Structural, Electrical/I&C, T/H Engineers, Computer Modeling and San Onofre Special NRC Panel Members. San Onofre Special NRC Panel Members need to make accurate and precise engineering decisions based on validated and auditable facts in accordance with Honorable and Respected Dr. McFarlane’s High Standards. These decisions have to be made without any political/financial/time pressures from EIX/SCE Officers, CPUC Chairman, NRC Commissioners, Pro-SCE Politicians, Attorneys or Industry Lobbyists. Ex NRC Branch Chiefs (Dr. Joram Hopenfeld, Mel Silberberg , etc.), Anonymous “Critical Questioning & Investigative Attitude” Genius NRC Branch Chief, US Public, San Onofre Workers and Southern Californians would appreciate San Onofre Special NRC Panel Members “Critical Questioning & Investigative Attitude”, True, Unbiased and Diligent Public Safety Efforts. More and more Southern Californians, Cities, Businesses, School Districts are joining every day the chorus to press Federal Regulators to hold a trial-like hearing before deciding whether the San Onofre nuclear plant is safe to reopen. Newly elected San Diego Congressman Vargas said, “When he was an assemblyman he questioned industry executives under oath during the California’s energy crisis. Vargas said that process could be effective for San Onofre as long as those questioning majority owner Southern California Edison executives know what they’re talking about. “Get experts in there,” Vargas said. “To ask them true questions: is it really safe do you really have this under control if not why are you firing it up? Makes no sense. The only reason they’re doing this is they want to get some money and if it sits vacant for a long time they actually can’t recoup their investment.” Federal Regulators have no choice to abide by the wishes of Southern Californians, because they pay for the cost of San Onofre and their safety is at risk. SCE, its global experts and NRC have nothing at risk in this unapproved and potentially lethal experiment. Thanks to Mr. Victor Dricks for Posting this Blog.

  63. I think that California does not NEED any nuclear power plants since they have plenty of spare capacity without relying on nuclear generation! Add in the RISK of an Earthquake and/or a Tsunami and the fact that California has plenty of sunshine, not to mention the possibility of off shore (out of sight of those on land) wind generation and you realize that California could become an Energy exporter, all without any nuclear generation or the massive amount of waste they create! The only thing keeping California from going Non-Nuclear is the “Public Utilities” which now have a strangle-hold on the states Political Leadership and their Utility Regulators…

  64. San Onofre Special Public/NRC/SCE Awareness Series by HAHN Baba
    Critical Questioning & Investigative Attitude Quiz
    An unnamed US Nuclear Power Plant increased its power by 7.5 % and increased the heat transfer surface area of its Replacement steam Generators by 25%. The RSG design addressed Fluid Elastic Instability, Flow -induced Random Vibrations and compared the results with Several Operating Reactors. It will be interesting to find out if NRC region IV can answer what was the name of that US Plant and how the RSG heat transfer surface area was increased. Did this plant apply for a 50.90 License Amendment and which NRC Branch checked the results of Thermal-Hydraulic Modeling/FEI/FIRV Vibrations? Thanks to Mr. Victor Dricks for Posting this Blog.

  65. San Onofre Special Public/NRC/SCE Awareness Series by HAHN Baba
    Courtesy of DAB Safety Team
    Press Release – The DAB Safety Team: January 31, 2013
    Four More Statements From NRC Region IV Augmented Inspection Team (AIT) That Require A Nuclear Reactor Regulation (NRR) Investigation And Resolution.
    The DAB Safety Team Has Transmitted The Following Request To The Offices Of Chairman Of The NRC, The California Attorney General and Senator Barbara Boxer’s Committee on Environment and Public Works (EPW).
    1. NRC AIT in its report dated November 09, 2012 (Re: NRC ADAMS Library Accession Number ML 2012010 – Unresolved Item 05000362/2012007-03, “Evaluation of Unit 3 Vibration and Loose Parts Monitoring System Alarms (V&LPM)”) closed the referenced item by stating that, “The inspectors determined that the licensee properly responded to and evaluated the alarms and followed the applicable station alarm procedures and vendor recommendations. Subsequently, the licensee requested from the vendor an in-depth evaluation of the available acoustical data, which was documented in Nuclear Notification NN 201818719. This evaluation established the likely source of the alarms. The results were inconclusive because of limitations with the monitoring system. Specifically, because of sensor locations (lower portion of the steam generator below the tube sheet in the support structure) and sensitivity, it was not possible to determine the exact source of the Unit 3 alarms. Westinghouse engineering personnel performed an evaluation (Evaluation 201818719-SPT-2) of acoustical data and determined from the shape and intensity of the particular responses that the acoustic source was not likely from the upper bundle of the replacement steam generator or related to the tube-to-tube wear. The licensee (SCE) is considering additional sensor locations which are not required, but may help with monitoring the upper bundle region of the steam generator during power operation. The results of this additional monitoring and increased sensor sensitivity may provide the licensee with a potential means to monitor for tube-to-tube degradation.” (The wonders of of this improved version of V&LPM system related to of tube-tube wear as claimed by AIT Team and SCE and questioned by NRR as NO detection capability below). According to the December 18, 2012 SCE NRC Public meeting Press and Webcast Reports, Edison officials came under sharp questioning about the Vibration and Loose Parts Monitoring System monitors at a U.S. Nuclear Regulatory Commission panel meeting in Maryland. Richard Stattel of the NRC’s Nuclear Reactor Regulation (NRR) Instrumentation Branch told the Edison Officials in a roaring and loud voice on an international live web cast, “The equipment could not do the job described by the company or provide additional safety if the plant is restarted. The instrumentation that you’re proposing … does not appear to be capable of detecting the conditions that would lead to actual tube wear.” Edison depicted the equipment in its restart plan as an important safety measure “but it doesn’t appear to do that.” See the DAB Safety Team’s Press Release + 12-12-28 Thirty Alarms Demonstrates SONGS Unsafe for details on this subject.
    DAB Safety Team Comments: The NRR is saying loud and clear that both NRC AIT and SCE Engineers need to understand the basic functions of “Safety-Grade” Instrumentation and the concept of “tube-to-tube” wear (Fluid Elastic Instability). Since there are no means of monitoring tube wall thinning while the plant is in service, the risk of tube burst is wholly dependent upon the accuracy and reliability of SCE’s “Safety-Grade” Instrumentation. The DAB Safety Team has stated earlier that NRC AIT Report is just a replication of SCE Root Cause Evaluation and not a true assessment by an Independent Regulator tasked with ensuring Public Safety. On December 21, 2012, the US Nuclear Regulatory Commission (NRC) blog posted a letter from Chairman Macfarlane titled, “A Visit to Japan: Reflections from the Chairman.” She said, “Regulators may need to be “buffered” from political winds, but they need to be fully subjected to the pressure of scientific and engineering truth and cannot be allowed to make decisions or order actions that are “independent” of facts.” According to the March 16, 2012 Press reports, Senators Barbara Boxer (D-CA), Chairman of the Senate Environment and Public Works Committee (EPW), and Dianne Feinstein (D-CA) sent a letter to the Chairman of the Nuclear Regulatory Commission (NRC), Dr. Gregory Jaczko, calling on the NRC to perform a thorough inspection at the San Onofre plant, located in San Clemente. The collusion and casual relationship between NRC AIT Team and SCE requires an Investigation by the Offices of NRC Chairman and Honorable and respected Senator Barbara Boxer to determine the impact on both future US reactor operations and emergency preparedness planning. This investigation by the AIT does not meet the Honorable and Respected NRC Chairman’s Standards.
    2. NRC AIT in its report dated November 11, 2012 (Re: Unresolved Item 05000362/2012007-03, “Evaluation of Retainer Bars Vibration during the Original Design of the Replacement Steam Generators”) closed the referenced item by stating that, “The inspectors determined that the licensee’s failure to verify the adequacy of the retainer bar design as required by SONGS Procedure SO123-XXIV-37.8.26 was of very low safety significance (Green) based on NRC Inspection Manual Chapter 0609.04, “Phase 1 – Initial Screening and Characterization of Findings,” and Inspection Manual Chapter 0609, Appendix A, “The Significance Determination Process (SDP) for Findings At-Power,” because the finding did not involve a degraded steam generator tube condition where one tube could not sustain 3 times the differential pressure across a tube during normal full power, steady state operation and none of the replacement steam generators violated the “accident leakage” performance criterion in plant Technical Specifications as a result of the retainer bar vibrations. The licensee also implemented actions to inspect all affected tubes in Unit 2 and 3 and remove from service all those tubes surrounding the smaller retainer bars that could wear due to vibration of the retainer bar. Because this violation has been determined to be of very low safety significance (Green) and has been entered in the licensee’s corrective action program as SONGS Nuclear Notification (NN) 201843216, it will be dispositioned as a non-cited violation in accordance with Section 2.3.2 of the NRC’s Enforcement Policy.”
    John Large, internationally known Consulting Engineer, Chartered Engineer, Fellow of the Institution of Mechanical Engineers, Graduate Member of the Institution Civil Engineers, Learned Member of the Nuclear Institute and a Fellow of the Royal Society of Arts states concerning SONGS Restart Unit 2 in his testimony to the Atomic Safety Licensing Board, “In October 2012 MHI reported directly to the NRC safety concerns about the retainer bars: The Steam Generator tube wear adjacent to the retainer bars was identified as creating a potential safety hazard. The maximum wear depth is 90% of the tube thickness. The cause of the tube wear has been determined to be the retainer bars’ random flow-induced vibration caused by the secondary fluid exiting the tube bundle. Since the retainer bar has a low natural frequency, the bar vibrates with a large amplitude. This type tube wear could have an adverse effect on the structural integrity of the tubes, which are part of the pressure boundary. The plugging of the tubes that are adjacent to the retainer bars was performed. MHI has recommended to the purchaser [SCE] to remove the retainer bars that would have the possibility of vibration with large amplitude or to perform the plugging and stabilizing for the associated tubes. According to MHI, it is the lower resonance frequency of the smaller diameter retainer bars that is susceptible to turbulent two-phase flow exciting the bar into its prime resonance or some harmonic frequency thereof [p10, item 3].14 Whatever, a number of the tubes capturing the retainer bar had sustained abraded wear from interaction with it. These tubes comprised six tubes in U2 and four tubes in U3, with seven tubes in total showing wear greater than the 35% limit of the tube wall thickness for which isolation from service is required by plugging with, as previously noted, an incidence site in one of U2 RSGs having worn through 90% of its wall thickness. I agree with the findings of MHI that the tube wear at the retainer bar localities arises because of random flow induced (not FEI) vibration of the retainer bar itself, it being entirely independent of any tube motion excited from other sources. However, MHI’s advice to either plug the local tubes and/or remove the retainer bars at risk raises two issues unique to the retainer bar and its sub-assembly: (i) Plugging of the at-risk tubes is not a satisfactory solution because it is the retainer bar that vibrates via random fluid flow processes at sub FEI critical velocity levels – these are likely to continue in play or, indeed, exacerbate at the proposed U2 restart at 70% power, leading to through-tube abrasion, the detachment of tube fragments, lodging at other unplugged and in-service tube localities, resulting in the so-called ‘foreign object’ tube wear; (ii) MHI’s recommendation that those retainer bars at risk of large-amplitude fluid flow excited vibration should be removed is, of course, dependent upon reliable analysis to identify the at-risk assemblies; and, importantly, and (iii) this restraint system probably also serves to contain the tube bundle geometry during a main line steam break (MSLB) design basis event, so any change or removal of the retaining bar assemblage would require a full safety justification.”
    Westinghouse states, “For most of the straight leg section of the tube, the gap velocities at lower power levels and at 100% power are similar. The recirculating fluid flow rate is relatively constant at all power levels. However, in the U-bend region, the gap velocities are a strong function of power level. The steam flow in the bundle is cumulative and increases as a function of the power level and the bundle height which causes high fluid quality, void fraction, and secondary fluid velocities in the upper bundle.”
    SCE in its November 30, 2012, NRC Presentation stated, “Four tubes with retainer bars wear above 35% limit in Unit 2 were plugged.” The NRC website states, “The severity of one of the wear indications at a Unit 2 retainer bar was significant enough (90 percent thru-wall) to warrant in-situ pressure testing. This pressure test confirmed the structural integrity of this tube (there was no leakage).”
    DAB Safety Team Comments: Let us summarize what John Large and Westinghouse are saying: (1) Plugging of the at-risk tubes is not a satisfactory solution because it is the retainer bar that vibrates via random fluid flow processes at sub FEI critical velocity levels – these are likely to continue to vibrate or, indeed, exacerbate at the proposed U2 restart at 70% power, leading to through-tube abrasion, the detachment of tube fragments, lodging at other unplugged and/or in-service tube localities, resulting in the so-called ‘foreign object’ tube wear, (2) For most of the straight leg section of the tube, the gap velocities at lower power levels and at 100% power are similar. Therefore, even at 70% power, the tube-to-retainer bar wear will continue at the same rate as 100% power and plugging the tubes is not a satisfactory solution in terms of reducing the active tubes rupture safety risks. SCE is not stating the facts either in its Root Cause Evaluation nor in its NRC Presentation. Two better questions are, “How many tubes in Unit 2 have what amounts of fatigue cracks and why has SCE not used state-of-the-art technology to visually examine all RSG tubes at San Onofre?” What this really means is that Southern Californians were lucky once again, that Unit 2 just happened to be shutdown for refueling! Otherwise, one or more worn tubes could have leaked or failed due to a design bases accident and/or any unanticipated transients. Almost 180 tubes had to be plugged and stabilized in Unit 2 Replacement Steam Generators due to retainer bar design mistakes. In addition, no reports are available to determine the extent of tube fatigue damage or damage to the small retainer bars caused by the worn tubes and whether the damaged retaining bars are strong enough to restrain the movement of the anti-vibration bar assembly during a main steam line break design basis event (Ref: NRR RAI #32). The design of the retainer bars approved by SCE and manufactured by MHI clearly violated the Code of Federal Regulations, 10 CFR Part 50, GDC 14, “RCPB—shall have “an extremely low probability of abnormal leakage…and gross rupture” and Appendix B, Criterion III, “Design Control.” The DAB Safety Team’s opinion is that NRC AIT is treating the retainer bar mistakes and its design approval by SCE just as a routine matter like “No big deal, nothing happened, so who cares” instead of performing the strict enforcement required of an Independent Regulator tasked with ensuring Public Safety. This investigation by the AIT does not meet the Honorable and Respected NRC Chairman’s Standards.
    3. NRC AIT in its report dated November 11, 2012 (Re: Unresolved Item 05000362/2012007-08, “Non-Conservative Thermal-Hydraulic Model Results”) states that, “The licensee and Mitsubishi continued to evaluate this unresolved item and no final conclusions were reached at the time of the inspection. The NRC is continuing to perform independent reviews of existing information, and will conduct additional reviews as new information becomes available.” In the original Report in July 2012, the NRC AIT concluded that, “Due to modeling errors, the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.”
    John Large states, “I identify a number of issues with the … AREVA Tube-to-Tube Report, including: (i) it is not exactly clear which properties are being represented on the spider diagram for comparison with the other operational SGs; even so (ii) since it is most unlikely that AREVA has undertaken a comprehensive (ATHOS) simulation of each of the five nominated SGs, the comparisons drawn are likely to be between aggregate or bulk flows within the entire tube bundle of each SG; (iii) as acknowledged by AREVA, the SONGS RSGs are dominated by in-plane flow regimes whereas all other SGs are characterized by out-of-plane flow regimes; and (iv) none of the comparative SGs has been identified. In other words, … I cannot reason how, are making a direct comparison of the complex two-phase fluid cross-flow situation in the SONGS and other five comparative plant steam generators, then these figures only provide the bases of a somewhat meaningless comparisons. A complete understanding of the causation of the in-plane FEI is essential to ensure that the SONGS Unit 2 plant is acceptably safe to restart and, once restarted, predictably safe to continue in operation over the proposed 150 day inspection interval. To the contrary, the understanding presented by SCE is neither comprehensive nor convincing. In my opinion, simply sweeping the FEI issue under the carpet on the basis of (in- or out-of-plane) FEI will not reoccur at 70% power is not only disingenuous but foolhardy.”
    Arnie Gundersen states, “The AIT report indicated that the change to the FIT-III evaluation methodology was not discussed as part of Edison’s 50.59 screening because the details of thermal hydraulic models used for the design of the OSG were not discussed in the original FSAR. It should have been obvious to Edison that FIT-III has not been benchmarked and had not been previously used in licensing procedures showing that the use of FIT-III might have an adverse effect on the FSAR safety analysis thus necessitating the entire license amendment review and public hearing process. As noted by the AIT, Edison approved the use of FIT-III code even though the code was not benchmarked nor identified as acceptable in the FSAR. Consequently, Edison operated San Onofre without knowing the uncertainties in the Replacement Steam Generators’ performance characteristics. Predicted liquid levels, pressure drops, vibrations, and temperatures at both Units 2 and 3 were all subject to unknown uncertainties during both normal and abnormal operations. In my opinion, by approving the use of an un-benchmarked and untested design tool like FIT-III, Edison did not meet the requirements expected from a nuclear licensee. Use of an un-benchmarked computer code that is not included in the FSAR protocol demands a formal FSAR license amendment process including the requisite public hearings.”
    Arnie Gundersen further states, “The AIT reported that FIT-III predictions differed considerably in comparison to an Electric Power Research Institute developed code named ATHOS. FIT-III predicted lower flow velocities and void fractions that were not conservative compared to ATHOS. The AIT Report neglected an analysis of the root cause of the critical differences between FIT-III and ATHOS, and the negative impact such lax calculational modeling had on the design, fabrication, and successful operation of the San Onofre RSGs. Had Edison sought the required FSAR license amendment, differences between FIT-III and ATHOS would have been identified six years ago. The AIT did not address the possibility that the lack of conservatism in FIT-III predictions, in addition to causing tube vibrations, could also result in non-conservative predictions of the behavior of the steam generator pressure vessel and associated main steam piping during accident conditions that are required to be analyzed in the FSAR. The AIT noted that the non-conservatisms in FIT-III are a contributor to the failure by Edison to adequately calculate the San Onofre RSG tube vibrations. But equally important, the AIT failed to address that FIT-III could also create non-conservative predictions of the behavior of the steam generator pressure vessel and associated main steam piping during accident conditions that are required to be analyzed in the FSAR. Such a conclusion implies that damage to the steam generator pressure vessel itself, and not just the tubes, might have occurred at San Onofre and remains unanalyzed by either Edison or the NRC. The probability of an accident exceeding the plant’s Current Design Basis is increased by the radically different Edison Replacement Steam Generators. Hence, the risks involved in operating the San Onofre RSGs should have been addressed as part of an FSAR license amendment and hearing process. It is my professional opinion that Edison should have applied for the 50.59 process so that the FSAR license amendment evaluation and public hearings would have occurred six years ago, prior to creating an accident scenario and facing losses that by the end of this process will easily total more than $1 Billion. The seriousness of the licensing and safety impact of the damaged RSGs at San Onofre cannot be overstated or underestimated. Any Design Basis Accident (DBA) as defined in the FSAR needs to be accurately modeled in order to protect public health and safety. The FSAR’s DBA analysis including the extent of tube leakage in the event of a Main Steam Line Break significantly impacts the design and implementation of Emergency Evacuation Plans. In the event of a steam line break accident in the San Onofre Replacement Steam Generators with the degraded condition of the tubes, an accident would have occurred that is more severe than any design basis accident scenario previously analyzed by Edison in the FSAR. Such a DBA steam line break accident would render the San Onofre emergency plan totally inadequate and most likely cause a permanent evacuation of a large portion of Southern California.”
    DAB Safety Team Comments: After the June 18, 2012 public Meeting, the NRC AIT Team Chief announced to the world, “The computer simulation used by Mitsubishi during the design of the steam generators had under-predicted velocities of steam and water inside the steam generators by factors of three to four times.” Now, six months later, the AIT Team is saying the matter is unresolved. The AIT Team is just repeating what SCE says or is not sure what they said four months ago. ATHOS Modeling results are not reliable, because the results by NRC AIT Team, Westinghouse, MHI, AREVA and Independent Experts show that fluid elastic instability occurred both in Units 3 and 2. The investigations in the Root cause of SONGS Unit 3 FEI regarding computer modeling have not been completed by NRC AIT Team, SCE and MHI. FEI did not occur in Unit 2 according to DAB Safety Team and Westinghouse. As also shown in other DAB Safety Team reports, FEI was not caused in Unit 3 by tube-to AVB gaps as claimed by NRC AIT Team and SCE. This is consistent with the findings of Westinghouse, AREVA, MHI, John Large and SONGS Anonymous Insiders. The AIT Team is hurting its own credibility by issuing contradicting and conflicting statements. This investigation by the AIT does not meet the NRC Chairman’s Standards.
    4. NRC AIT report dated November 11, 2012 (Re: Unresolved Item 05000362/2012007-10, “Evaluation of Departure of Methods of Evaluation for 10 CFR 50.59 Processes”) closed the referenced item by stating: (a) The change from ANSYS to ABAQUS did not require a license amendment prior to implementing the change, so with respect to section 2.10.D.6 of the NRC Enforcement Manual, there is no reasonable likelihood that the change from ANSYS to ABAQUS would ever require NRC approval. Therefore, in accordance with the NRC Enforcement Manual, the inspectors determined that the licensee’s change from ANSYS to ABAQUS was a minor violation of 10 CFR 50.59(d)(1), and (b)n Based on this, the inspectors determined that the licensee had changed from using ANSYS and STRUDL to analyze several events for the original steam generators, to using only ANSYS to analyze a single limiting event for the replacement steam generators. Therefore, because the licensee did not change the method described in the Updated Final Safety Analysis Report, the inspectors concluded that the licensee did not need to obtain a license amendment prior to implementing that change. In the original Report in July 2012, the NRR technical specialist reviewed SCE’s 10 CFR 50.59 evaluation and found two instances that failed to adequately address whether the change involved a departure of the method of evaluation described in the updated final safety analysis report: (a) Use of ABAQUS instead of ANSYS: The SCE’s 50.59 evaluation incorrectly determined that using the ABAQUS instead of ANSYS was a change to an element of the method described in the updated final safety analysis report did not constitute changing from a method described in the updated final safety analysis report to another method, and as such, did not mention whether ABAQUS has been approved by the NRC for this application.
    (b) Use of ANSYS instead of STRUDL and ANSYS: While SCE’s 50.59 evaluation correctly considered this a change from a method described in the FSAR to another method, the 50.59 evaluation did not mention whether the method has been approved by NRC for this application.
    NRC AIT Report states, “For the Unit 2 and Unit 3 replacement steam generators, the licensee determined that the proposed activity did not adversely affect a design function, or the method of performing or controlling a design function described in the updated final safety analysis report. The licensee evaluated the following updated final safety analysis report design functions in the 50.59 screening: Steam Generator Design Functions. Let us examine the effect of these changes on Steam Generator Design Functions: The design functions of the steam generators tubes and tube supports are to: (1.) Limit tube flow-induced vibration to acceptable levels during normal operating conditions, and (2) Prevent a tube rupture concurrent with other accidents.
    Change Number 1: 105,000 square feet tube heat transfer area in OSGs; 116,100 square feet tube heat transfer area in RSGs; 11.1% increase in heat transfer area, which is more than a minimal change of 10% in the non-conservative direction. Change accomplished by addition of 377 tubes in the central region by removal of stay cylinder and increasing the length of 9727 tubes by > 7 inches in each of the four RSGs.
    Change Number 2: Operating Secondary Pressure – OSGs: 900 psi, RSG: 833 psi ~ 10% change
    – A catastrophic change for onset and ongoing exponential fluid elastic instability
    Change Number 3: Tube wall thickness was reduced from 0.048 inches to 0.043 to pump more reactor coolant through the tubes > 11.6% change – The latest academic research indicates that the tube vibrations become large as T/D decreases and L/D increases, because the in-plane tube vibrations strongly depend on the dynamic characteristics of tubes such as the natural frequency and the damping ability.
    Four other changes: Moisture content was reduced from 0.2% to 0.1% to improve SG performance, RCS Volume was increased from 1895 cubic feet to 2003 cubic feet, RCS Flow was increased from 198,000 gpm to 209,000 gpm, feedwater flow was increased from 7.4 million pound per hour to 7.6 million pound per hour and AVBs were not designed to prevent against adverse effects of fluid elastic instability (In-plane vibrations, Tube-to-Tube wear, steam dry-outs). These unapproved and unanalyzed changes were claimed to be a conservative decision and improvements in the RSGs from OSGs were presented as a “like for Like” change. No mixing baffles were added in the SONGS RSGs to improve the T/H Performance in the RSGs. FEI and SR Values were not provided by SCE in the RSG Design Specifications. SCE told MHI to avoid the NRC Approval… MHI neither provided in-plane supports, nor provided the operational criteria to prevent FEI in one of the largest steam generators with such high steam flows. MHI did not benchmark CE SG Computer codes or design details, neither did SCE, nor did SCE check the work of MHI. And Honorable and Respected Dr. McFarlane says, “SCE is responsible for the work of its vendors and contractors. Look at Palo Verde RSGs, a Success Story and SONGS RSGs, a $ Billion Blunder….
    NRC AIT Report states, “The licensee’s bid specification required that the stay cylinder feature of the original steam generators be eliminated to maximize the number of tubes that could be installed in the replacement steam generators and to mitigate past problems with tube wear at tube supports caused by relatively cool water and high flow velocities in the central part of the tube bundle. Mitsubishi employed a broached trefoil tube support plates instead of the egg crate supports in the original design. In addition to providing for better control of tube to support plate gaps and easier assembly, the broached tube support plates were intended to address past problems with the egg crate supports by providing less line of contact and faster flow between the tubes and support plates, reducing the potential for deposit buildup and corrosion.”
    Arnie Gundersen states, “As the NRC confirmed in its AIT report, a large steam void has developed near where the additional tubes were added in the Replacement Steam Generators (called fluid elastic instability) that allows many types of excess vibrations to occur. Fairewinds review of Edison’s Condition Report clearly shows that the location within the steam generators where the steam “fluid elastic instability” has developed is precisely the region where the extra heat created by the 400 new tubes would create an excess of steam and various vibrational modes.”
    NRC AIT report states, “Mitsubishi’s preliminary explanation of the failure mechanism started with the combination of two factors: (1) a relatively small tube pitch to tube diameter ratio (P/D), and (2) high void fraction in the tube bundle area where the tube-to-tube wear was identified. The small pitch to diameter ratio was a fixed parameter in the replacement steam generators established by the nominal center-to-center distance between adjacent tubes (P) and the nominal outside diameter of the tubes (D). The high void fraction was identified from the results of Mitsubishi’s thermal-hydraulic model for the secondary side of the replacement steam generators. Mitsubishi considered that the combination of these two factors may have resulted in favorable conditions for in-plane tube vibration based, in part, on the results of recent studies in fluid-elastic instability.” Mitsubishi also states, “Low secondary pressures are severe for vibration.”
    John Large states, “Referring to the short section of the FSAR provided to me by SCE, which I understand is not to be amended for the Unit 2 restart: (a) there is no account of the changes that have been made in the evaluation of the tube structural and leakage integrity, that is from the stage of predicting those tubes at risk of TTW and other forms of wear, the tube thinning wear rates, through to the nature of the tube failure being unique to the type and extent of the wear pattern and tube thinning; and (b) the methods of deducing, mainly by unproven inference, from the probe inspection results particularly to determine the in-plane AVB effectiveness, includes unacceptably large elements of test and experimentation that are inconsistent with the analyses and descriptions of the FSAR.”
    John Large states, “SCE’s assertion that reducing power to 70% will at the best alleviate, but not eliminate, the TTW and other modes of tube and component wear is little more than hypothesis – the supporting Operational Assessments and analyses have not proven it to be otherwise. I am of the opinion that trialling this hypothesis by putting the SONGS Unit 2 back into service will, because of the uncertainties and unresolved issues involved, embrace a great deal of change, test and experiment. The terms of the Confirmatory Action Letter of March 11 2012, are versed such that to meet compliance the response of SCE via its Return to Service Report,11 must include considerable changes of conditions and procedures that are outside the reference bounds of the present FSAR – this is because the physical condition of the RSGs, and the means by which this is evaluated and projected into future in-service operation, have substantially and irrevocably changed since the current FSAR was approved. The fact that SCE fails to satisfy the requirements of the CAL is neither here nor there, although it illustrates the scope and complexity of the response required. At the time of preparing the CAL, the NRC being well-versed in the failures at the San Onofre nuclear plant, surely must have known that the only satisfactory response to the CAL would indeed require considerable changes, tests and experiments to be implemented.”
    DAB Safety Team Comments: Therefore, the DAB Safety Team concludes that the changes in design functions of the RSGs tubes and tube supports described above definitely: a) did not limit tube flow-induced vibration to acceptable levels during normal operating conditions and, b) involved a significant reduction in a margin of safety – Failure of 8 Unit 3 SG Tubes under MSLB test conditions and significant TTW > 35% of ~381 tubes in Unit 3 RSGs. A multiple tube failure event, if actually would have occurred during a MSLB would have resulted in a significant increase in the off-site radiological consequences over the single tube burst event, if currently considered in the SONGS approved FSAR by NRC Region IV. The Replacement Steam Generator (RSG) modifications at San Onofre increased both the likelihood of equipment failure and the radiological consequence of such failure and therefore directly affect the FSAR Current Design Basis. The AIT has no business contradicting conclusions made earlier by the NRR technical specialist. This investigation by the AIT does not meet the NRC Chairman’s Standards.
    NRC Region IV Response to DAB Safety Team Analysis of SONGS 10 CFR 50.59 Evaluation Comments: The NRC has already conducted several reviews of the 10 CFR 50.59 documents associated with the replacement of the steam generators at SONGS. These reviews involved NRC inspectors from multiple offices including Region IV, Region II and the Office of Nuclear Reactor Regulation at NRC headquarters. The results of these reviews are contained in NRC two inspection reports that are available at http://www.nrc.gov/info-finder/reactor/songs/tube-degradation.html. [see the Augmented Inspection Team Report dated July 18, 2012, and the Augmented Inspection Team Follow-Up Report dated November 9, 2012]. It is worthy of note that the NRC staff is currently reviewing 10 CFR 50.59 documents associated with the licensee’s proposed restart activities. The results of the ongoing review will be documented in a future inspection report.
    Comments from Mel Silberberg [NRC-RES, Retired [Chief, Severe Accident Research Branch; Waste Management Branch] to Region IV: I am disappointed in the composition of the special panel! Where is the representation from NRC-RES? The issues at SONGS involve thermal hydraulics and material science. The NRC-RES and its contractors are experts in these areas. The Office of Research was created by the Congress for such situations. Two RES staff covering these disciplines and one or two consultants, serving as peer-reviewers. Perhaps there needs to be a separate peer review. Public confidence can only be gained using logical, informed measures as I described above. Inspection Reports are only one facet of the problem, no question. However, understanding the reasons for the fluid instability, possible cavitation corrosion effects, etc. are phenomena which require evaluation by T/H as well as materials experts, with appropriate oversight by the ACRS. The SCE, the nuclear industry, the NRC and the public need assurance, not educated guesses. I have not seen a bona fide attempt to understand resolve the issue such that all can be alert to potential problems. I still remain puzzled as to why the ACRS [at least one of the Subcommittees]. I am trying to reach the ACRS Exec. Director to discuss this point. Thank you.”
    DAB Safety Team Further Comments: NRC Region IV Inspectors need to be re-trained in interpretation of significance of 10 CFR 50.59 Evaluation rules and meaning of changes in design function on safety evaluations. Simply sweeping the 10 CFR 50.59 mistakes under the carpet on the basis of meaningless statements, “The NRC has already conducted several reviews of the 10 CFR 50.59 documents associated with the replacement of the steam generators at SONGS. The present SONGS NRC approved for the total S/G tube leakage assumes a limit of 1 gpm for all S/Gs, which ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10CFR100 limits in the event of either a S/G tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents.
    The 0.5 gpm (720 gpd) leakage limit per S/G ensures that S/G tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.” These reviews of 10 CFR 50.59 and SONGS FSAR S/G tube rupture limits from NRC inspectors from multiple offices including Region IV, Region II and the Office of Nuclear Reactor Regulation at NRC headquarters” are not only disingenuous but foolhardy. A single tube leakage and/or rupture could result in a nuclear incident or accident with tube leakages assumed in the current SONGS FSAR as shown in the Table below. A multiple tube failure event for all phases of the reactor in-core fuel cycle, would result in a significant increase in the off-site radiological consequences (e.g., Fukushima, Chernobyl, etc.) over the single tube burst event currently considered in the FSAR. The rapid and extraordinarily severe wear that resulted in the 2012 failures of all of Edison’s San Onofre Replacement Steam Generators was the result of Edison’s 2005 decision to radically change the RSG design and to claim that the Part 50.59 licensing process did not apply. Arnie Gundersen and DAB Safety Team have stated consistently that San Onofre Replacement Steam Generator tube damage discovered in 2012 was so severe and extensive that both reactors have been operating in violation of their NRC FSAR license design basis as defined in their Technical Specifications. While the NRC Augmented Inspection Team (AIT) briefly described how Edison addressed its 50.59 requirements, the evidence shows that Edison did not comply with the NEI guidelines for implementing 50.59. Published reports indicate that the strategic decision made by Edison that the 50.59 process would not be applied to the RSGs was made by corporate officials before any engineering personnel had actually performed the 50.59 engineering analysis. Consequently, Edison made a management decision to claim that the 50.59 process did not apply and therefore San Onofre was not required to seek NRC approval for the proposed changes at San Onofre Units 2 and 3. These unlicensed unapproved design changes to the containment boundary violated Federal Regulations and therefore the FSAR must be amended prior Unit 2 Restart to reflect multiple steam generator tube ruptures with MSLB plus DBE due to Edison’s significant untested and unanalyzed modifications.The DAB Safety Team: Don, Ace and a BATTERY of safety-conscious San Onofre insiders plus industry experts from around the world who wish to remain anonymous. These volunteers assist the DAB Safety Team by sharing knowledge, opinions and insight but are not responsible for the contents of the DAB Safety Team’s reports. We continue to work together as a Safety Team to prepare additional DAB Safety Team Documents, which explain in detail why a SONGS restart is unsafe at any power level without a Full/Thorough/Transparent NRC 50.90 License Amendment and Evidentiary Public Hearings.
    Our Mission: To prevent a Trillion Dollar Eco-Disaster like Fukushima, from happening in the USA.
    Copyright January 31, 2013 by The DAB Safety Team. All rights reserved. This material may not be published, broadcast or redistributed without crediting the DAB Safety Team. The contents cannot be altered without the Written Permission of the DAB Safety Team Leader and/or the DAB Safety Team’s Attorney… Thanks to NRC Moderator for Posting this Blog. HAHN Baba

  66. Southern Californians need safe, affordable, reliable, well managed, well maintained and excellently operated nuclear power plants, where workers are free for raising nuclear safety and personnel concerns and Rate Payers, Regulators, Politicians and News Media are Proud. San Onofre does not meet any of the above listed criteria. With that said, SCE has to meet all of the above criteria or Decommission San Onofre. Thanks To the NRC Moderator for posting this blog… HAHN Baba

  67. Happy Anniversary!

    I predict that time will show that a nuclear accident (not a nuclear incident) was narrowly avoided at SanO on January 31, 2012 only because of shear luck, due to the timing of the discovery of Edison’s poorly in-house designed replacement steam generators (RSG). Had that Unit 3 tube been just a tiny bit stronger and not leaked when it did; then with both Unit 2 & 3 back online, if a main steam line break or something similar occurred, we now know that it would have probably resulted in the complete venting of the core coolant within minutes, and we all know what that means…

    SanO is now a 1.5 Billion Dollar RED FLAG that illustrates how easy NRC regulations can be gamed (without ANY enforcement penalties) which allow Utilities/Operators to make changes that have enormous implications to safety and the Public Health, with little to N☢ actual oversight, until it is to late!

    The two basic problems at Fukushima, Japan were that:
    (1) TEPCO’s regulator pushed too much paper instead of being “hands on”.
    (2) TEPCO had total control over what data the public had access to, which prevented any real oversight by the public.

    The USA cannot afford a Trillion Dollar Eco-Disaster like Fukushima, that is why the NRC needs to “overhaul” how it enforces its current regulations and develop new regulations ASAP to patch all the regulatory holes that now exist!

    The first step is to really open up the entire NRC process to the public, so that true public oversight can take place, instead of the flawed system we now have, as SanO illustrates all too well! As it is now, the public does not have enough access to NRC documents, reports and/or data which prevents all knowledgable people from providing true input into the decision making process.

    Or said another way, we cannot afford to have a Trillion Dollar Eco-Disaster in the USA for any reason and that includes GREED…

    What we don’t know can indeed hurt US, especially if it is radioactive!

  68. If a picture is worth a thousand words, then a video is worth an AIT Report…

    The latest video from Friends of the Earth US:
    “No way out”

  69. Salute to Mel Silberberg, If you do reach the Exec. Director of ACRS, please tell him to contact the DAB Safety Team, we have posted more factual data/information* about San Onofre’s FEI problems than anybody else! The DAB Safety Team’s documents explain in detail why a SONGS restart is unsafe at any power level, especially without a Full/Thorough/Transparent NRC 50.90 License Amendment and Evidentiary Public Hearings. For much more from the DAB Safety Team, please visit the link* below.

    * https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?pli=1

  70. BIG TIP: The Attorney General of CA has now requested Party Status in the CPUC investigation of Edison’s San Onofre Debacle!

  71. Note from the Moderator: Information about the NRC Inspector General’s HotLine is located here: http://www.nrc.gov/insp-gen/oighotline.html . The Office of the Inspector General at NRC established the Hotline (1-800-233-3497) program to provide the NRC employee, other government employee, licensee/utility employee, contractor employee, and the public with a confidential means of reporting incidences of suspicious activity to the OIG concerning fraud, waste, abuse, and employee or management misconduct. Mismanagement of agency programs or danger to public health and safety may also be reported through the Hotline.

  72. Hi Mr. Mel Silberberg Can you please comment on the following….. Thanks HAHN Baba
    Subject: FYI – Exchange of Notes
    To: Victor.Dricks@nrc.govHelpAllHurtNeverBaba January 29, 2013 at 1:04 am
    Request for independent re-review of SONGS 50.59 Screen/Evaluation by NRC Region II – Please send me an email after you complete the review ASAP. These guys who performed the screen and evaluations are very close friends of mine and I want to make sure they were on the right track. Trying to help my friends and NRC Region IV. Thanks… HAHN Baba

    Reply
    Moderator January 29, 2013 at 2:21 pm

    The NRC has already conducted several reviews of the 10 CFR 50.59 documents associated with the replacement of the steam generators at SONGS. These reviews involved NRC inspectors from multiple offices including Region IV, Region II and the Office of Nuclear Reactor Regulation at NRC headquarters. The results of these reviews are contained in NRC two inspection reports that are available at http://www.nrc.gov/info-finder/reactor/songs/tube-degradation.html. [see the Augmented Inspection Team Report dated July 18, 2012, and the Augmented Inspection Team Follow-Up Report dated November 9, 2012]. It is worthy of note that the NRC staff is currently reviewing 10 CFR 50.59 documents associated with the licensee’s proposed restart activities. The results of the ongoing review will be documented in a future inspection report.

    Victor Dricks

    Reply

    HelpAllHurtNeverBaba January 29, 2013 at 8:34 pm Your comment is awaiting moderation.
    Mr. Dricks, Respectfully, Along with Arnie Gundersen and John Large, I totally disagree with the NRC assessments on SONGS 10 CFR 50.59 RSG Evaluations. I was qualified SONGS 50.59 Screener/Evaluator for a decade besides being qualified at several other nuclear power plants. I have performed numerous 50.59 changes and reviews at SONGS. The changes shown below were claimed by Edison to be in the conservative direction and improvements.

    NRC AIT Report states, “For the Unit 2 and Unit 3 replacement steam generators, the licensee determined that the proposed activity did not adversely affect a design function, or the method of performing or controlling a design function described in the updated final safety analysis report. The licensee evaluated the following updated final safety analysis report design functions in the 50.59 screening: Steam Generator Design Functions….

    Let us examine the effect of these changes on Steam Generator Design Functions and then you go back to your peers for more soul searching/research and provide more arguments and we will go from there:

    The design functions of the steam generators tubes and tube supports are to: (1.) Limit tube flow-induced vibration to acceptable levels during normal operating conditions, and (2) Prevent a tube rupture concurrent with other accidents.

    Change Number 1: 105,000 square feet tube heat transfer area in OSGs; 116,100 square feet tube heat transfer area in RSGs; 11.1% increase in heat transfer area, which is more than a minimal change of 10% in the non-conservative direction. Change accomplished by addition of 377 tubes in the central region by removal of stay cylinder and increasing the length of 9727 tubes by > 7 inches.

    Change Number 2: Operating Secondary Pressure – OSGs: 900 psi, RSG: 833 psi ~ 10% change

    Change Number 3: Tube wall thickness was reduced from 0.048 inches to 0.043 to pump more reactor coolant through the tubes > 10% change

    Other changes: Moisture content was reduced from 0.2% to 0.1% to improve SG performance, RCS Volume was increased from 1895 cubic feet to 2003 cubic feet, RCS Flow was increased from 198,000 gpm to 209,000 gpm, feedwater flow was increased from 7.4 million pound per hour to 7.6 million pound per hour and AVBs were not designed to prevent against adverse effects of fluid elastic instability (In-plane vibrations, Tube-to-Tube wear, steam dry-outs). These unapproved and unanalyzed changes were claimed to be a conservative decision and improvements in the RSGs from OSGs were presented as a “like for Like” change. No mixing baffles were added in the SONGS RSGs to improve the T/H Performance in the RSGs. FEI and SR Values were not provided by SCE in the RSG Design Specifications. SCE told MHI to avoid the NRC Approval…… MHI did not either provided in-plane supports, or provided the operational criteria to prevent FEI in one of the largest steam generators with such high steam flows. MHI did not benchmark CE SG Computer codes or design details, neither did SCE, nor did SCE check the work of MHI. And Dr. McFarlane says, “SCE is responsible for the work of its vendors and contractors. Look at Palo Verde RSGs, a Success Story and SONGS RSGs, a $ Billion Blunder….

    NOTE: ATHOS Modeling results are not reliable, because the results by NRC AIT Team, Westinghouse, MHI, AREVA and Independent Experts show that fluid elastic instability occurred both in Units 3 and 2. The investigations in the Root cause of SONGS Unit 3 FEI regarding computer modeling have not been completed by NRC AIT Team, SCE and MHI. FEI did not occur in Unit 2 according to DAB Safety Team and Westinghouse. As also shown in other DAB Safety Team reports, FEI was not caused in Unit 3 by tube-to AVB gaps as claimed by NRC AIT Team and SCE. This is consistent with the findings of Westinghouse, AREVA, MHI, John Large and SONGS Anonymous Insiders.

    NRC AIT Report states, “The licensee’s bid specification required that the stay cylinder feature of the original steam generators be eliminated to maximize the number of tubes that could be installed in the replacement steam generators and to mitigate past problems with tube wear at tube supports caused by relatively cool water and high flow velocities in the central part of the tube bundle. Mitsubishi employed a broached trefoil tube support plates instead of the egg crate supports in the original design. In addition to providing for better control of tube to support plate gaps and easier assembly, the broached tube support plates were intended to address past problems with the egg crate supports by providing less line of contact and faster flow between the tubes and support plates, reducing the potential for deposit buildup and corrosion.”

    Problems in SONGS Original CE Steam Generators: In the Original 2001 Power Uprate Application (NRC ADAMS Accession Number ML010950020), “Proposed Change Number NPF-10115-514 Increase in Reactor Power to 3438 MWt San Onofre Nuclear Generating Station Units 2 and 3”, SCE stated “ By the above reference Southern California Edison (SCE) submitted Amendment Application Numbers 207 and 192 to the facility operating licenses for the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, respectively, to increase the licensed reactor thermal power level to 3438 MWt. At 100% power operation, steam generator pressures typically vary between 800 psia and 815 psia, compared to the original nominal design operating pressure of 900 psia. Wear at tube support structures is a known degradation mechanism at SONGS. At SONGS, rapid wear was observed on tubes surrounding the stay cylinder in the center of the steam generator during the first cycle of operation. Many tubes in the most susceptible region around the stay cylinder have been preventively plugged. The first preventive plugging was done after 0.7 EFPY of operation. The preventively plugged region was expanded during the Cycle 3 outage. Typical active wear in CE designed steam generators has occurred at the support structures in the upper bundle region of the steam generator. These supports consist of diagonal straps (frequently called bat wings) and vertical strap supports. This currently active wear mechanism is influenced by both flow velocities and tube to support gap wear. The variable influenced by the proposed uprate is the inner bundle flow velocities. The hydrodynamic stability of a steam generator is characterized by the damping factor. A negative value of this parameter indicates a stable unit, i.e., small perturbations of steam pressure or circulation ratio will diminish rather than grow in amplitude. The damping factors remain highly negative, at a level comparable to the current design, for all cases. Thus, the steam generators remain hydrodynamically stable for all uprate cases. Based on a projected increase of 2.3% in the secondary side fluid velocity, normal operation flow induced vibration analysis is impacted by the velocity increase. Current analysis considered that tubes with more than one consecutive inactive eggcrate were staked and plugged, and two nonconsecutive inactive eggcrates are acceptable. The Stability Ratio (SR) is defined as: SR = Veff/Vcr, where, Veff= effective velocity, Vcr = critical velocity; and Values of SR 35% of ~381 tubes in Unit 3 RSGs.

    Palo Verde made similar changes to their RSGs under a 50.90 License Amendment. PVNGS Generators are running after 10 years with very little tube plugging whereas the above changes in SONGS RSGs destroyed Unit 3 and crippled Unit 2 RSGs. Because of these adverse design changes, everybody is on the run: NRC Region IV, SCE, Mitsubishi, California Public Utilities Commission, Senator Barbara Boxer and Senator Dianne Feinstein. NRC Region IV, Westinghouse, AREVA, MHI, World’s Experts, SCE (Except DAB Safety Team SONGS Anonymous Insiders) are not sure whether fluid elastic instability in Unit 2 occurred or not. Southern Californians Ratepayers have lost $1 Billion in this game without electricity and now are faced with the trauma of restart of defectively-designed and degraded Unit 2 due to SCE’s continued mistakes. I am just trying to help, so please, wake up NRC Region IV and San Onofre Special Panel, Your charter is public safety and not whether SCE looses or makes money. I guarantee that SCE will make more money by admitting their mistakes and win NRC/Public Confidence by correcting their mistakes and using “Critical Questioning & Investigative Attitude” in the future. Remember, Mr. Dricks, Truth always prevails….. HAHN BABA

  73. Allegation – NRC Region IV Violating Presidential Directive and the Public Trust
    ttps://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?docId=1S3tUKp-sV-rS2OFMY0Z-A1H4BJ4Ha1ftiow7577tsdY
    snip:

    SONGS UNIT 3 RSG ROOT CAUSE: It appears that Complacent SCE and Inexperienced MHI Engineers did not perform proper academic research and industry benchmarking about the potential adverse consequences of the reduction of original CE steam generator pressures from 900 psi to say, 800 psi on fluid elastic instability and flow-induced vibrations. These lower secondary steam operating pressures (800-833 psia) are the primary cause for shortening the life of SONGS Original Combustion Engineering Generators due to increased tube wear and plugging caused by flow-induced random vibrations and destruction of SONGS Unit 3 Replacement Steam Generators due to flow-induced random vibrations, Mitsubishi flowering effects and steam voids or steam dry-outs (AKA fluid elastic instability). In addition, SCE Engineers prepared a defective 10 CFR 50.59 Evaluation and design specifications, which were not challenged by MHI, and/or adequately reviewed by NRC Region IV. MHI at the direction of SCE Engineers made numerous untested and unanalyzed design changes to the steam generators under the pretense of “like for like”, and even the NRC’s Region IV administrator Elmo Collins said, “The guts of the machinery look …. Different.”

  74. Special Public Awareness Series – SONGS $1Billion Dollar Radiation Steaming Crucibles
    Unbiased and Factual Information provided for the benefit of NRC San Onofre Special Panel
    Addressed To: Ryan Lantz, Brian Benney, Randy Hall, Edwin Hackett, Dan Dorman, Victor Dricks
    Good Moring Mr. Dricks, SONGS Insider Information from Anonymous Sources and From DAB Safety team for SONGS Special Onofre Team – Response appreciated from the San Onofre Special Panel
    A NRC Branch Chief gifted with MIT Intelligence, Intuition and a Sixth Sense, who is an acquaintance of mine, told me at an Industry Conference, “Sir, to resolve any complex technical problem and understand unclear regulations, you have to, ‘Read and reread in between the lines’, use, ‘Critical questioning and an investigative attitude’ and ‘Solid teamwork & alignment.”
    Allegation – NRC AIT Report Incomplete, Inconclusive, Inconsistent and Unacceptable
    SONGS UNIT 3 RSG REAL ROOT CAUSE: Lack of “Critical Questioning & Investigative Attitude” by SCE, MHI, NRC Region IV and the AIT Members.
    NOTE: ATHOS Modeling results are not reliable, because the results by NRC AIT Team, Westinghouse, MHI, AREVA and Independent Experts show that fluid elastic instability occurred both in Units 3 and 2. The investigations in the Root cause of SONGS Unit 3 FEI regarding computer modeling have not been completed by NRC AIT Team, SCE and MHI. As shown in item 3 below, FEI did not occur in Unit 2 according to DAB Safety Team and Westinghouse. As also shown in other DAB Safety Team reports, FEI was not caused in Unit 3 by tube-to AVB gaps as bogusly claimed by NRC AIT Team and SCE. This is consistent with the findings of Westinghouse, AREVA, MHI, John Large and SONGS Anonymous Insiders.
    The concerns raised by Dr. Hopenfeld are extremely important safety issues. As the ACRS stated:
    • Steam generators constitute more than 50% of the surface area of the primary pressure boundary in a pressurized water reactor.
    • Unlike other parts of the reactor pressure boundary, the barrier to fission product release provided by the steam generator tubes is not reinforced by the reactor containment as an additional barrier.”
    • Leakage of primary coolant through openings in the steam generator tubes could deplete the inventory of water available for the long-term cooling of the core in the event of an accident. In the decade since Dr. Hopenfeld first raised his safety concerns, the NRC has allowed many nuclear plants to continue operating nuclear power plants with literally thousands of steam generator tubes that are known to be fatigue cracked! The ACRS concluded that the NRC staff made these regulatory decisions using incomplete and inaccurate information. After receiving the ACRS’s report, the NRC staff considered Hopenfeld’s concerns “resolved” even though it had taken no action to address the numerous recommendations in the ACRS report.
    Mel Silberberg January 21, 2013 at 6:31 pm US NRC Blog
    I am disappointed in the composition of the special panel! Where is the representation from NRC-RES? The issues at SONGS involve thermal hydraulics and material science. The NRC-RES and its contractors are experts in these areas. The Office of Research was created by the Congress for such situations. Two RES staff covering these disciplines and one or two consultants, serving as peer-reviewers. Perhaps there needs to be a separate peer review. Public confidence can only be gained using logical, informed measures as I described above.
    Mel Silberberg, NRC-RES, Retired [Chief, Severe Accident Research Branch; Waste Management Branch.
    1. Changes in SONGS RSGs from Original CE OSGs In the SONGS RSGs: the number of tubes were increased by 377 and made > 7 inches taller to achieve 11% increase in Heat transfer Area of Tubes to increase 24MWt per RSG, tube wall thickness was reduced from 0.048 inches to 0.043 to pump more reactor coolant through the tubes, moisture content was reduced from 0.2% to 0.1% to improve SG performance , secondary pressure was reduced from 900 psi to 833 psi to push more heat from the reactor coolant to the feedwater, RCS Volume was increased from 1895 cubic feet to 2003 cubic feet, RCS Flow was increased from 198,000 gpm to 209,000 gpm, feedwater flow was increased from 7.4 million pound per hour to 7.6 million pound per hour and AVBs were not designed to prevent against adverse effects of fluid elastic instability (In-plane vibrations, Tube-to-Tube wear, steam dry-outs). These unapproved and unanalyzed changes were claimed to be a conservative move and improvements in the RSGs from OSGs under a “like for Like” change. No mixing baffles were added in the SONGS RSGs to improve the T/H Performance and eliminate dead zones in the RSGs. Palo Verde made similar changes to their RSGs under a 50.90 License Amendment. PVNGS Generators are running after 10 years with very little tube plugging whereas the above changes in SONGS RSGs destroyed Unit 3 and crippled Unit 2 RSGs. These fatal changes definitely: a) Caused a significant increase in the probability or consequences of an accident previously evaluated (SGTR) and, b) involve a significant reduction in a margin of safety – Failure of 8 Unit 3 SG Tubes under MSLB test conditions and significant TTW > 35% of ~381 tubes in Unit 3 RSGs.
    2. Problems in SONGS Original CE Steam Generators: In the Original 2001 Power Uprate Application (NRC ADAMS Accession Number ML010950020), “Proposed Change Number NPF-10115-514 Increase in Reactor Power to 3438 MWt San Onofre Nuclear Generating Station Units 2 and 3”, SCE stated “ By the above reference Southern California Edison (SCE) submitted Amendment Application Numbers 207 and 192 to the facility operating licenses for the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, respectively, to increase the licensed reactor thermal power level to 3438 MWt. At 100% power operation, steam generator pressures typically vary between 800 psia and 815 psia, compared to the original nominal design operating pressure of 900 psia. Wear at tube support structures is a known degradation mechanism at SONGS. At SONGS, rapid wear was observed on tubes surrounding the stay cylinder in the center of the steam generator during the first cycle of operation. Many tubes in the most susceptible region around the stay cylinder have been preventively plugged. The first preventive plugging was done after 0.7 EFPY of operation. The preventively plugged region was expanded during the Cycle 3 outage. Typical active wear in CE designed steam generators has occurred at the support structures in the upper bundle region of the steam generator. These supports consist of diagonal straps (frequently called bat wings) and vertical strap supports. This currently active wear mechanism is influenced by both flow velocities and tube to support gap wear. The variable influenced by the proposed uprate is the inner bundle flow velocities. The hydrodynamic stability of a steam generator is characterized by the damping factor. A negative value of this parameter indicates a stable unit, i.e., small perturbations of steam pressure or circulation ratio will diminish rather than grow in amplitude. The damping factors remain highly negative, at a level comparable to the current design, for all cases. Thus, the steam generators remain hydrodynamically stable for all uprate cases.Based on a projected increase of 2.3% in the secondary side fluid velocity, normal operation flow induced vibration analysis is impacted by the velocity increase. Current analysis considered that tubes with more than one consecutive inactive eggcrate were staked and plugged, and two nonconsecutive inactive eggcrates are acceptable. The Stability Ratio (SR) is defined as: SR = Veff/Vcr, where, Veff= effective velocity, Vcr = critical velocity; and Values of SR 99%), known as “NO Effective Thin Tube Film Damping.” Thin film damping refers to the tendency of the steam inside the generators to create a thin film of water between the RSG tubes and the support structures and each other. That film is enough to help keep the tubes from vibrating with large amplitudes, hitting other tubes violently, and to protect the Anti-Vibration Bar support structures and maintain the tube-to-AVB gaps and contact forces. These adverse conditions in Unit 2 at 70% power operation (RTP) with the present defective design and degraded RSGs, known as fluid elastic instability (Tube-to-Tube Wear, or TTW) can lead to rapid U-tube failure from fatigue or tube-to-tube wear in Unit 2 due to a main steam line break as seen in Unit 3’s RSG’s. In summary, FEI is a phenomenon where due to San Onofre RSGs design intended for high steam flows causes the tubes to vibrate with increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit (critical velocity) for a given tube and its supporting conditions and a given thermal hydraulic environment. This occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports. The lack of Nucleate boiling on the tube surface or absence of water is found to have a destabilizing effect on fluid-elastic stability.
    C.2 – Unit 2 FEI Conflicting Operational Data
    • NRC AIT Report SG Secondary U2/3 Pressure Range 833 – 942 psi
    • SCE RCE SG Secondary U2/3 Pressure – 833 psi
    • RCE Team Anonymous Member – Unit 2 SG Secondary Pressure 863 psi
    • SONGS SG System Description Unit 2 SG Pressure Range 892 – 942 psi
    • Westinghouse OA SG Secondary U2/3 Pressure ~ 838 psi, Void Fraction 99.55%
    • SCE Enclosure 2, MHI ATHOS results – U2/3 Void Fraction 99.6%
    • SCE Enclosure 2, Independent Expert results – ATHOS U2/3 Void Fraction 99.4%
    • DAB Safety Team SG Secondary U2 Pressure 863 -942 psi, Void Fraction 96-98%
    • SONGS Plant Daily Briefing Unit 3 Electrical Generation – 1186 MWe
    • SONGS Plant Daily Briefing Unit 2 Electrical Generation – 1183 MWe
    C.3 – Unit 2 FEI Conclusions
    C.3.1 – NRC AIT Report – Operational Differences between U2/3 – The NRC analysis indicated a correlation with the tube-to-tube wear based on a combination of high void fraction and high steam velocities. It should be noted that the traditional forcing function, fluid velocity squared times density, does not show good agreement with the tube-to-tube wear patterns. This indicated that the high quality steam fluid velocities and high void fraction may be sufficiently high to cause conditions in the generators conducive for onset of fluid-elastic instability.
    The ATHOS code predicted regions of high void fraction and high steam velocities are
    super-imposed with tube-to-tube wear indications from Unit 3 steam generator 3E0-88
    The above analyses apply equally to Units 2 and 3, so it does not explain why the accelerated fluid-elastic instability wear damage was significantly greater in Unit 3steam generators. The result of the independent NRC thermal-hydraulic analysis indicated that differences in the actual operation between units and/or individual steam generators had an insignificant impact on the results and in fact, the team did not identify any changes in steam velocities or void fractions that could attribute to the differences in tube wear between the units or steam generators.
    C.3.2 – SCE Unit 2 Restart Report Enclosure 2 Conclusions – Because of the similarities in design between the Unit 2 and 3 RSGs, it was concluded that FEI in the in-plane direction was also the cause of the TTW in Unit 2.
    C.3.3 – SCE U2 FEI SONGS RCE Team Anonymous Member Conclusions – FEI did not occur in Unit 2.
    C.3.4 – Westinghouse OA Conclusions: (a) An evaluation of the tube-to-tube wear reported in two tubes in SG 2E089 showed that, most likely, the wear did not result from in-plane vibration of the tubes since all available eddy current data clearly support the analytical results that in-plane vibration could not have occurred in these tubes, and (b) Operational data – Westinghouse ATHOS Model shows no operational differences in Units 2 & 3 (void fraction ~99.6%) and then Westinghouse says in (a) above that FEI did not occur in Unit 2. Westinghouse is contradicting its own statement.
    C.3.5 – AREVA OA Conclusions – Based on the extremely comprehensive evaluation of both Units, supplemented by thermal hydraulic and FIV analysis, assuming, a priori, that TTW via in-plane fluid-elastic instability cannot develop in Unit 2 would be inappropriate.
    C.3.6 – John Large States, “I note here that there are three clear conflicts of findings between the OAs: From AREVA that AVB-to-tube and TTW result from in-plane FEI, contrasted to Westinghouse that there is no in-plane FEI but most probably it was out-of-plane FEI, and from MHI that certain AVB-to-tube wear results in the absence of in-plane FEI from just turbulent flow. My opinion is that such conflicting disagreement over the cause of TTW reflects poorly on the depth of understanding of the crucially important FEI issue by each of these SCE consultants and the designer/manufacturer of the RSGs.”
    C.3.7 – DAB Safety Team Conclusions – Due to higher SG pressure (Range 863 – 942 psi) and lower thermal megawatts as compared to Unit 3, FEI did not occur in Unit 2. This is consistent with the position of RCE Team Anonymous Member. The NRC AIT Report, SCE, Westinghouse, MHI, Independent Expert and AREVA conclusions on Unit 2 FEI are Contradicting, Confusing, Inconclusive, Full of Smoking Mirrors, Inconsistent and Unacceptable
    PROBBABLE ROOT CAUSE: Lack of “Critical Questioning & Investigative Attitude” of SCE Supplied Operational Data by Westinghouse, AREVA, MHI and Other World’s Leading Experts

  75. Victor: Inspection Reports are only one facet of the problem, no question. However,understanding the reasons for the fluid instability, possible cavitation corrosion effects, etc.are phenomena which require evaluation by T/H as well as materials experts, with appropriate oversight by the ACRS. The SCE, the nuclear industry, the NRC and the public need assurance, not educated guesses. I have not seen a bona fide attempt to understand resolve the issue such that all can be alert to potential problems. I still remain puzzled as to why the ACRS [ at least one of the Subcommittees]. i am trying to reach the ACRS Exec. Director to discuss this point. Thank you.
    Mel Silberberg

  76. The NRC has already conducted several reviews of the 10 CFR 50.59 documents associated with the replacement of the steam generators at SONGS. These reviews involved NRC inspectors from multiple offices including Region IV, Region II and the Office of Nuclear Reactor Regulation at NRC headquarters. The results of these reviews are contained in NRC two inspection reports that are available at http://www.nrc.gov/info-finder/reactor/songs/tube-degradation.html. [see the Augmented Inspection Team Report dated July 18, 2012, and the Augmented Inspection Team Follow-Up Report dated November 9, 2012]. It is worthy of note that the NRC staff is currently reviewing 10 CFR 50.59 documents associated with the licensee’s proposed restart activities. The results of the ongoing review will be documented in a future inspection report.

    Victor Dricks

  77. To: NRC Moderator Mr. Victor Dricks, Senior Public Affairs Officer, NRC Region IV
    Request for independent re-review of SONGS 50.59 Screen/Evaluation by NRC Region II – Please send me an email after you complete the review ASAP. These guys who performed the screen and evaluations are very close friends of mine and I want to make sure they were on the right track. Trying to help my friends and NRC Region IV. Thanks… HAHN Baba

  78. Special Thanks to NRC Moderator Mr. Victor Dricks, Senior Public Affairs Officer, NRC Region IV for Posting this Blog
    Special Public Awareness Series – US #1 Nuclear Safety Concern
    Contrary to what the PUC news release led the public to believe the PUC issued a “scoping” memorandum today limiting the review of San Onofre issues to those helpful to SCE and hurtful The scoping memo makes a mockery of the PUC “investigation” because it allows only a very limited review of the issues: (1) assessing the reasonableness of SCE’s actions and expenditures after the outage; (2) whether SCE’s 2012 expenditures for SONGs was reasonable; (3) the reasonableness of SCE’s expenditures for community outreach; and (4) whether SCE should refund any money they were allowed to keep under the General Rate Case issued in December 2012.
    Here is what will not be allowed: (1) whether SCE was imprudent and unreasonable in spending $800 million for the 4 new generators to replace the previous generators which tube problems, when the new generators had tube problems worse than those replaced; (2) whether the 4 generators should be taken out of the rate base. The Scoping Order does not address the first question and pushes off the second to some undetermined time in the future. The PUC has mislead the People of California by issuing a news release announcing an investigation while issuing an order that does not permit a reasonable investigation.
    It is clear that the PUC has decided to get San Onofre back in operation as soon as possible. The PUC “investigation” is nothing more than a cynical public relations stunt.

  79. Special Thanks to NRC Moderator Mr. Victor Dricks, Senior Public Affairs Officer, NRC Region IV for Posting this Blog – Special Public Awareness Series – US #1 Nuclear Safety Concern

    Portions of the following information have been extracted from the DAB Safety Team Reports (Search Google Drive for DAB Safety Team & Related Info). It is the DAB Safety Team’s goal to help educate both the NRC and the Public by providing unbiased, logical and factual information in order to help assess the real dangers of any San Onofre Unit 2 restart. According to Press Reports and San Onofre Insiders, Unit 2 permission for restart by the NRC is imminent yet the REAL Root Cause for the $1 Billion destruction of Units 2 and 3 RSGs (Including equipment cost and expenses) has not yet even been determined. The Public does not know the status of SCE’s ongoing cause evaluations, SCE’s response to 32 NRR’s RAI’s and NRC’s Special San Onofre Inspections. We like to remind NRC San Onofre Special Panel, what NRC Chairman Macfarlane said during her recent Fukushima Trip, “Regulators may need to be ‘buffered’ from political winds, but they need to be fully subjected to the pressure of scientific and engineering truth and cannot be allowed to make decisions or order actions that are ‘independent’ of facts.” The NRC rush to a faulty judgment cannot be allowed to compromise Public Safety just to please SCE, as this conflicts with President Obama’s Policy, the new NRC Chairman’s Standards and the advice of NRC retired Branch Chiefs who have spoken out.

    Comments – SONGS Unit 2 Restart Reports Contradicting, Confusing, Inconclusive, Smoking Mirrors, Inconsistent and Unacceptable

    PROBBABLE ROOT CAUSE: Lack of “Critical Questioning & Investigative Attitude” of SCE Supplied Operational Data by Westinghouse, AREVA, MHI and Other World’s Leading Experts –

    Public to Judge for themselves
    .C. Let us now examine the other differences between Unit 2 and Unit 3’s Operational Factors, which were significant contributors to the “fluid-elastic instability” in San Onofre Unit 3 and the tube-to-tube wear resulting in the tube leak.
    C.1 – Adverse Design/Operational Factors responsible for Fluid Elastic Instability: Low steam generator pressures (SONGS RSGs range 800-850 psi, the primary cause of the onset of severe vibrations) allow the onset of FEI, whereby U-tube bundle tubes start vibrating with very large amplitudes in the in-plane directions. Extremely hot and vibrating tubes need a little amount of water (aka damping, 1.5% water, steam-water mixture vapor Fraction 99.5%). Without the water, the extremely hot and vibrating tubes cannot dissipate their energy. In effect, one unstable tube drives its neighbor to instability through repeated violent impact events which causes tube leakage, tube failures at MSLB test conditions and/or unprecedented tube-tube wear, Tube-to-AVB/Tube Support Plates wear, as we saw in San Onofre Unit 3. So in review, due to narrow tube pitch to tube diameter, tube natural frequency, low tube clearances, in certain portions of the RSGs U-tubes bundle, fluid velocities exceed the critical velocities due to extremely high steam flows (100% power conditions). These high fluid velocities cause U-tubes to vibrate with very large amplitudes in the in-plane direction and literally hit other tubes with repeated and violent impacts. Due to lower secondary steam operating pressures (required to generate more heat, electricity and profits) and excessive pressure drops due to high flows and velocities, steam saturation temperature drops. This lowering of steam saturation temperature combined with high heat flux in the hot leg side of the U-tube bundle causes steam dry-outs to form (Vapor fraction >99%), known as “NO Effective Thin Tube Film Damping.” Thin film damping refers to the tendency of the steam inside the generators to create a thin film of water between the RSG tubes and the support structures and each other. That film is enough to help keep the tubes from vibrating with large amplitudes, hitting other tubes violently, and to protect the Anti-Vibration Bar support structures and maintain the tube-to-AVB gaps and contact forces. These adverse conditions in Unit 2 at 70% power operation (RTP) with the present defective design and degraded RSGs, known as fluid elastic instability (Tube-to-Tube Wear, or TTW) can lead to rapid U-tube failure from fatigue or tube-to-tube wear in Unit 2 due to a main steam line break as seen in Unit 3’s RSG’s. In summary, FEI is a phenomenon where due to San Onofre RSGs design intended for high steam flows causes the tubes to vibrate with increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit (critical velocity) for a given tube and its supporting conditions and a given thermal hydraulic environment. This occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports. The lack of Nucleate boiling on the tube surface or absence of water is found to have a destabilizing effect on fluid-elastic stability.
    C.2 – Unit 2 FEI Conflicting Operational Data
    • NRC AIT Report SG Secondary U2/3 Pressure Range 833 – 942 psi
    • SCE RCE SG Secondary U2/3 Pressure – 833 psi
    • RCE Team Anonymous Member – Unit 2 SG Secondary Pressure 863 psi
    • SONGS SG System Description Unit 2 SG Pressure Range 892 – 942 psi
    • Westinghouse OA SG Secondary U2/3 Pressure ~ 838 psi, Void Fraction 99.55%
    • SCE Enclosure 2, MHI ATHOS results – U2/3 Void Fraction 99.6%
    • SCE Enclosure 2, Independent Expert results – ATHOS U2/3 Void Fraction 99.4%
    • DAB Safety Team SG Secondary U2 Pressure 863 -942 psi, Void Fraction 96-98%
    • SONGS Plant Daily Briefing Unit 3 Electrical Generation – 1186 MWe
    • SONGS Plant Daily Briefing Unit 2 Electrical Generation – 1183 MWe
    C.3 – Unit 2 FEI Conclusions
    C.3.1 – NRC AIT Report – Operational Differences between U2/3 – The NRC analysis indicated a correlation with the tube-to-tube wear based on a combination of high void fraction and high steam velocities. It should be noted that the traditional forcing function, fluid velocity squared times density, does not show good agreement with the tube-to-tube wear patterns. This indicated that the high quality steam fluid velocities and high void fraction may be sufficiently high to cause conditions in the generators conducive for onset of fluid-elastic instability.

    The ATHOS code predicted regions of high void fraction and high steam velocities are
    super-imposed with tube-to-tube wear indications from Unit 3 steam generator 3E0-88
    The above analyses apply equally to Units 2 and 3, so it does not explain why the accelerated fluid-elastic instability wear damage was significantly greater in Unit 3steam generators. The result of the independent NRC thermal-hydraulic analysis indicated that differences in the actual operation between units and/or individual steam generators had an insignificant impact on the results and in fact, the team did not identify any changes in steam velocities or void fractions that could attribute to the differences in tube wear between the units or steam generators.
    C.3.2 – SCE Unit 2 Restart Report Enclosure 2 Conclusions – Because of the similarities in design between the Unit 2 and 3 RSGs, it was concluded that FEI in the in-plane direction was also the cause of the TTW in Unit 2.
    C.3.3 – SCE U2 FEI SONGS RCE Team Anonymous Member Conclusions – FEI did not occur in Unit 2.
    C.3.4 – Westinghouse OA Conclusions: (a) An evaluation of the tube-to-tube wear reported in two tubes in SG 2E089 showed that, most likely, the wear did not result from in-plane vibration of the tubes since all available eddy current data clearly support the analytical results that in-plane vibration could not have occurred in these tubes, and (b) Operational data – Westinghouse ATHOS Model shows no operational differences in Units 2 & 3 (void fraction ~99.6%) and then Westinghouse says in (a) above that FEI did not occur in Unit 2. Westinghouse is contradicting its own statement.
    C.3.5 – AREVA OA Conclusions – Based on the extremely comprehensive evaluation of both Units, supplemented by thermal hydraulic and FIV analysis, assuming, a priori, that TTW via in-plane fluid-elastic instability cannot develop in Unit 2 would be inappropriate.
    C.3.6 – John Large States, “I note here that there are three clear conflicts of findings between the OAs: From AREVA that AVB-to-tube and TTW result from in-plane FEI, contrasted to Westinghouse that there is no in-plane FEI but most probably it was out-of-plane FEI, and from MHI that certain AVB-to-tube wear results in the absence of in-plane FEI from just turbulent flow. My opinion is that such conflicting disagreement over the cause of TTW reflects poorly on the depth of understanding of the crucially important FEI issue by each of these SCE consultants and the designer/manufacturer of the RSGs.”
    C.3.7 – DAB Safety Team Conclusions – Due to higher SG pressure (Range 863 – 942 psi) and lower thermal megawatts as compared to Unit 3, FEI did not occur in Unit 2. This is consistent with the position of RCE Team Anonymous Member. The NRC AIT Report, SCE, Westinghouse, MHI, Independent Expert and AREVA conclusions on Unit 2 FEI are Contradicting, Confusing, Inconclusive, Full of Smoking Mirrors, Inconsistent and Unacceptable

    C.3.8 – The NRC San Onofre Special Review Panel should direct other branches within the NRC (NRC-RES and/or the ACRS) to review the above data without any prior “turf” bias and present their findings to the public for review and comment prior to any restart decision being made by the NRC.

  80. Hi Mr. Steinberg, Please See
    DAB Safety Team Media Alert 13-01-28
    Allegations
    1. NRC AIT Report Incomplete, Inconclusive, Inconsistent and Unacceptable
    2. SONGS UNIT 3 RSG REAL ROOT CAUSE: Lack of “Critical Questioning & Investigative Attitude” by SCE, MHI and NRC Region IV and AIT Team.
    Google Drive – DAB Safety Team & Related Info Share …
    docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4…

  81. Hi Mr. Silberberg, Brilliant Question And Great Recommendation… My Salute … HAHN BABA
    SONGS RSG Failure Root Cause – Lack of “Critical questioning and an investigative attitude” by SCE, MHI and NRC Region IV
    A NRC Branch Chief gifted with MIT Intelligence, Intuition and a Sixth Sense, who is an acquantaince of mine, told me at an Industry Conference, “Sir to resolve any complex technical problem and understand unclear regulations, you have to, ‘Read and reread in between the lines’, use, ‘Critical questioning and an investigative attitude’ and ‘Solid teamwork & alignment.”
    Thanks to NRC for posting this comment.. HAHN BABA

  82. Portions of the following information has been extracted from the DAB Safety Team Reports (Search Google Drive for DAB Safety Team & Related Info). DAB Safety Team is a group of Public Service Oriented Southern Californians and Anonymous San Onofre Insiders trying to help the NRC and Public by providing unbiased, logical and factual information to assess the real dangers of San Onofre Unit 2. Unit 2 permission for restart by NRC is imminent and REAL Root Cause for destruction of $1 Billion Units 2 and 3 RSGs (Includes equipment cost and expenses) has not even been determined. Public Safety by NRC in a rush to judgment cannot be compromised due to please profit-motivated SCE.

    Commenting on the NRC Augmented Inspection Team San Onofre Report… Just trying to help NRC Augmented Inspection Team Chief and NRC San Onofre Special Panel.. Thanking to the Moderator for posting this comment HAHN Baba

    NOTE: Highly recommend that NRC Augmented Inspection Team and NRC San Onofre Special Panel thoroughly review SONGS Unit 2 Return to Service MHI, AREVA, Westinghouse, DAB Safety Team and John Large Reports and carefully examine the operational differences between Unit 2 and 3 and then update the NRC AIT report with real Root cause for FEI in Unit 3 and NO FEI in Unit 2.

    The AIT inspection concluded that: (1) SCE was adequately pursuing the causes of the
    unexpected steam generator tube-to-tube degradation. In an effort to identify the causes, SCE retained a significant number of outside industry experts, consultants, and steam generator manufacturers, including Westinghouse and AREVA to perform thermal-hydraulic and flow induced vibration modeling and analysis; (2) The combination of unpredicted, adverse thermal hydraulic conditions and insufficient contact forces in the upper tube bundle caused a phenomenon called “fluid-elastic instability” which was a significant contributor to the tube to tube wear resulting in the tube leak. The team concluded that the differences in severity of the tube-to-tube wear between Unit 2 and Unit 3 may be related to the changes to the manufacturing/fabrication of the tubes and other components which may have resulted in increased clearance between the anti-vibration bars and the tubes; (3) Due to modeling errors, the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability. Unless changes are made to the operation or configuration of the steam generators, high fluid velocities and high void fractions in localized regions in the u-bend will continue to cause excessive tube wear and accelerated wear that could result in tube leakage and/or tube rupture; (4) The thermal hydraulic phenomena contributing to the fluid-elastic instability is present in both Unit 2 and 3 steam generators; (5) Based on the updated final safety analysis report description of the original steam generators, the steam generators major design changes were appropriately reviewed in accordance with the 10 CFR 50.59 requirements.

    So based on a review of the AIT Report and World’s Experts, the potential causes, which were significant contributors to the “fluid-elastic instability” in SONGS Unit 3 and the tube-to-tube wear resulting in the tube leak are as follows:

    A. Insufficient contact tube-to AVB forces and differences in manufacturing/fabrication of the tubes and other components between Units 2 & 3

    B. Due to modeling errors, the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.

    C. Operational Factors

    A. Let us now examine that whether insufficient contact tube-to AVB forces in the Unit 3 upper tube bundle caused “fluid-elastic instability” which was a significant contributor to the tube-to-tube wear resulting in the tube leak.

    A.1- MHI states, “By design, U-bend support in the in-plane direction was not provided for the SONGS SG’s”. In the design stage, MHI considered that the tube U-bend support in the out-of-plane direction designed for “zero” tube-to-AVB gap in hot condition was sufficient to prevent the tube from becoming fluid-elastic unstable during operation based on the MHI experiences and contemporary practice. MHI postulated that a “zero” gap in the hot condition does not necessarily ensure that the support is active and that contact force between the tube and the AVB is required for the support to be considered active. The most likely cause of the observed tube-to-tube wear is multiple consecutive AVB supports becoming inactive during operation. This is attributed to redistribution of the tube-to-AVB-gaps under the fluid hydrodynamic pressure exerted on the tubes during operation. This phenomenon is called by MHI, “tube bundle flowering” and is postulated to result in a spreading of the tube U-bends in the out-of-plane direction to varying degrees based on their location in the tube bundle (the hydrodynamic pressure varies within the U bend). This tube U-bend spreading causes an increase of the tube-to-AVB gap sizes and decrease of tube-to-AVB contact forces rendering the AVB supports inactive and potentially significantly contributing to tube FEI. Observations Common to BOTH Unit-2 and Unit-3: The AVBs, end caps, and retainer bars were manufactured according to the design. It was confirmed that there were no significant gaps between the AVBs and tubes, which might have contributed to excessive tube vibration because the AVBs appear to be virtually in contact with tubes. MHI states, “The higher than typical void fraction is a result of a very large and tightly packed tube bundle, particularly in the U-bend, with high heat flux in the hot leg side. Because this high void fraction is a potentially major cause of the tube FEI, and consequently unexpected tube wear (as it affects both the flow velocity and the damping factors).”

    A.2 – AREVA states, “At 100% power, the thermal-hydraulic conditions in the U-bend region of the SONGS replacement steam generators exceeded the past successful operational envelope for U-bend nuclear steam generators based on presently available data. The primary source of tube-to-AVB contact forces is the restraint provided by the retaining bars and bridges, reacting against the component dimensional dispersion of the tubes and AVBs. Contact forces are available for both cold and hot conditions. Contact forces significantly increase at normal operating temperature and pressure due to diametric expansion of the tubes and thermal growth of the AVBs. After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location.”

    A.3 – Westinghouse states, “Test data shows that the onset of in-plane (IP) vibration requires much higher velocities than the onset of out-of-plane (OP) fluid-elastic excitation. Hence, a tube that may vibrate in-plane (IP) would definitely be unstable OP. A small AVB gap that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion. Manufacturing Considerations: There are several potential manufacturing considerations associated with review of the design drawings based on Westinghouse experience. The first two are related to increased proximity potential that is likely associated with the ECT evidence for proximity. Two others are associated with the AVB configuration and the additional orthogonal support structure that can interact with the first two during manufacturing. Another relates to AVB fabrication tolerances. These potential issues include: (1) The smaller nominal in-plane spacing between large radius U-bend tubes than comparable Westinghouse experience, (2) The much larger relative shrinkage of two sides (cold leg and hot leg) of each tube that can occur within the tubesheet drilling tolerances. Differences in axial shrinkage of tube legs can change the shape of the U-bends and reduce in-plane clearances between tubes from what was installed prior to hydraulic expansion, (3) The potential for the ends of the lateral sets of AVBs (designated as side narrow and side wide on the Design Anti-Vibration Bar Assembly Drawing that are attached to the AVB support structure on the sides of the tube bundle to become displaced from their intended positions during lower shell assembly rotation, (4) The potential for the 13 orthogonal bridge structure segments that are welded to the ends of AVB end cap extensions to produce reactions inside the bundle due to weld shrinkage and added weight during bundle rotation, and (5) Control of AVB fabrication tolerances sufficient to avoid undesirable interactions within the bundle. If AVBs are not flat with no twist in the unrestrained state they can tend to spread tube columns and introduce unexpected gaps greater than nominal inside the bundle away from the fixed weld spacing. The weight of the additional support structure after installation could accentuate any of the above potential issues. There is insufficient evidence to conclude that any of the listed potential issues are directly responsible for the unexpected tube wear, but these issues could all lead to unexpected tube/AVB fit-up conditions that would support the amplitude limited fluid-elastic vibration mechanism. None were extensively treated in the SCE root cause evaluation.”

    A.4 -HAHN Baba concludes that SONGS Unit 3 RSG’s were operating outside SONGS Technical Specification Limits for Reactor Thermal Power and Current Licensing Basis for Design Basis Accident Conditions. HAHN Baba further agrees with MHI that high steam flows and cross-flow velocities combined with narrow tube pitch-to-diameter ratio caused elastic deformation of the U-tube bundle from the beginning of the Unit 3 cycle, which initiated the process of tube-to-AVB wear and insufficient contact forces between tubes and AVBs. Tube bundle distortion is considered a major contributing cause to the mechanism of tube-to-tube/AVB/TSP wear seen in the Unit 3 SG’s. After 11 months of wear, contact forces were virtually eliminated between the tube and AVBs in the areas of highest area of Unit 3 wear as confirmed by ECT and visual inspections. Therefore, based on a review of MHI, AREVA and Westinghouse excerpts shown below, the HAHN Baba concludes that FEI and MHI Flowering effect redistributed the tube-to-AVB gaps in Unit 3 RSG’s. It is the HAHN Baba’s opinion that NRC and SCE claims that insufficient contact forces in Unit 3 Tube-to-AVB Gaps ALONE caused tube “to” tube wear are misleading, erroneous and designed to put the blame on MHI for purposes of making SCE look good in the public’s eyes and for collecting insurance money from MHI’s manufacturing so called defects.

    B. Let us now examine of effects of modeling errors, that the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.

    B.1 – NRC AIT Report states, “The ATHOS thermal-hydraulic model predicts bulk fluid behavior based on first principals and empirical correlations and as a result, it is not able to evaluate mechanical, fabrication, or structural material differences or other phenomena that may be unique to each steam generator. Therefore this analysis cannot account for these mechanical factors and differences which could very likely also be contributing to the tube degradation.”

    B.2 – Ivan Cotton states, “Fluid elastic instability is one of the most damaging types of instabilities encountered in heat exchangers and steam generators and can impose a severe economic penalty on the power and chemical industries. At present our understanding of the mechanisms leading to fluid-elastic instability is very limited and more experiments are needed to more fully delineate the conditions for the onset of fluid-elastic instability.” Such experimentation should only be done in a sealed lab, NOT our environment with the lives of eight million local residents at stake in the outcome!

    B.3 – Ishihara, Kunihiko and Kitayama state, “Tube vibrations become large as tube thickness/diameter ratio (T/D) increases and tube length/diameter ratio (L/D) decreases, and the tube vibrations strongly depend on the dynamic characteristics of tubes such as the natural frequency and the damping ability.”

    B.4 – Fairewinde states, “Realistically, the 3-D steam analysis is not accurate enough to apply to such important safety related determinations. To make such mathematical risk 3-D analysis, a very large margin of error must be applied, and that has not been done. For example, if the 3-D steam analysis determines that plugging 100 tubes is a solution, then plugging ten times that number might be the appropriate solution due to the mathematical errors in the 3-D analysis being applied by Edison and Mitsubishi.”

    B.5 – Mitra, V.K. Dhir, I. Catton state, “ Flow induced vibrations in heat exchanger tubes have led to numerous accidents and economic losses in the past. Efforts have been made to systematically study the cause of these vibrations and develop remedial design criteria for their avoidance. Instability was clearly seen in single phase and two-phase flow and the critical flow velocity was found to be proportional to tube mass. It is also found that nucleate boiling on the tube surface is also found to have a stabilizing effect on fluid-elastic instability.

    B.6 – SCE states that SONGS Unit 3 Damage (FEI) was caused due to outdated MHI Thermal-Hydraulic Computer Models. According to NRC AIT Report, SONGS did not specify the value of FEI in its Design and Performance Specifications SO23-617-1. Academic Researchers have discussed and warned about the adverse effects of fluid elastic instability (tube-to-tube wear) in nuclear steam generators since 1970’s. Westinghouse and Combustion Engineering (CE) have designed CE engineering replacement steam generators (RSGs) to prevent the adverse effects of fluid elastic instability since 2000’s (e.g., PVNGS).

    B.7 – The NRC AIT Report dated November 9, 2012 states, “the FIT-III thermal-hydraulic model was still in-progress at the time of the inspection and no final conclusions were reached for the cause of the non-conservative flow velocities, which were used as inputs in the tube vibration analysis and resulted in non-conservative stability ratios. Since the licensee had not completed the cause evaluation for this unresolved item, the inspectors were not able to make a final
    determination of whether a performance deficiency or violation of NRC requirements
    occurred. The inspectors were informed that Mitsubishi was performing an evaluation of the
    potential factors that contributed to the low flow velocities in FIT-III relative to the velocities calculated by the ATHOS model developed after the tube leak event in Unit 3. This evaluation was included in Document SO23-617-1-M1530, Revision 1, which also intended to demonstrate the validity of FIT-III results for the original tube vibration analysis. This evaluation was still being finalized and not yet approved by Edison. The licensee and Mitsubishi continued to evaluate this unresolved item and no final conclusions were reached at the time of the inspection. The NRC is continuing to perform independent reviews of existing information, and will conduct additional reviews as new information becomes available. In another related finding, NRC inspectors stated, “SCE Engineers did not meet Procedure SO123-XXIV-37.8.26 requirements to ensure the design of the retainer bar was adequate with respect to the certified design specification. Specifically, the licensee failed to ensure that there was sufficient analytical effort in the design methodology of the anti-vibration bar assembly to support the conclusion that tube wear would not occur as a result of contact with the retainer bars due to flow-induced vibration. The inspectors determined that the requirements for flow-induced vibration in the certified design specification, along with the expectations in Procedure SO123-XXIV-37.8.26, provided sufficient information to reasonably foresee the inadequate design of the retainer bars during
    the review and approval of design Calculations SO23-617-1-C749 and SO23-617-1-C157, including the associated design drawings provided by Mitsubishi.

    B.8 – Arnie Gundersen states, “Not only is Mitsubishi unfamiliar with the tightly packed CE design, but Edison’s engineers added so many untested variables to the new fabrication that this new design had a significantly increased risk of failure. As a result of the very tight pitch to diameter ratios used in the original CE steam generators, Mitsubishi fabricated a broached plate design that allows almost no water to reach the top of the steam generator.

The maximum quality of the water/steam mixture at the top of the steam generator in the U-Bend region should be approximately 40 to 50 percent, i.e. half water and half steam. With the Mitsubishi design the top of the U-tubes are almost dry in some regions. Without liquid in the mixture, there is no damping against vibration, and therefore a severe fluid-elastic instability developed.

Because of the Edison/Mitsubishi steam generator changes, the top of the new steam generator is starved for water therefore making tube vibration inevitable. Furthermore, the problem appears to be exacerbated by Mitsubishi’s three-dimensional thermal-hydraulic analysis determining how the steam and water mix at the top of the tubes that has been benchmarked against the Westinghouse design but not the original CE design.

The real problem in the replacement steam generators at San Onofre is that too much steam and too little water is causing the tubes to vibrate violently in the U-bend region. The tubes are quickly wearing themselves thin enough to completely fail pressure testing. Even if the new tubes are actively not leaking or have not ruptured, the tubes in the Mitsubishi fabrication are at risk of bursting in a main steam line accident scenario and spewing radiation into the air.”

    B.9 – John Large, Internationally Known Scientist and Chartered Nuclear Engineer from London says about the SONGS Unit 2 Replacement Steam Generators (RSGs) AVB Structure, “It impossible to reliably predict the effectiveness of the many thousands of AVB contact points for when the tube bundle is in a hot, pressurized operational state. The combination of the omission of the in-plane AVB restraints, the unique in-plane activity levels of the SONGS RSGs, together the very demanding interpretation of the remote probe data from the cold and depressurized tube inspection, render forecasting the wear of the tubes and many thousands of restraint components when in hot and pressurized service very challenging indeed.”

    B.10 – HAHN Baba Comment to Limitations of ATHOS thermal-hydraulic Models: SCE and MHI are both negligent because they did a very poor job of Industry and Academic Research benchmarking regarding the applicability of thermal-hydraulic computer models during the redesign of SONGS original CE SGs. SCE is negligent because they did not check the results of MHI’s outdated Thermal-Hydraulic Computer Models to meet their specification requirements. This does not meet the NRC Chairman’s Standards. Therefore, the DAB Safety Team concludes that SCE claims as stated above are not factual. SCE engineers did not check the work of MHI with a critical and questioning attitude and did not meet the 10CFR50, Appendix B, Quality assurance Standards and or NRC Regulations

    C. Let us now examine the other operational factors, which were significant contributors to the “fluid-elastic instability” in SONGS Unit 3 and the tube-to-tube wear resulting in the tube leak.

    C.1 – Low steam generator pressures (1 causes the onset of FEI). At the onset of FEI, U-tube bundle tubes start vibrating with very large amplitudes in the in-plane directions. Extremely hot and vibrating tubes need a little amount of water (aka damping, 1.5% water, steam-water mixture vapor Fraction 99.5%). When this happens, the extremely hot and vibrating tubes cannot dissipate their energy and return to their original in-plane design position. In effect, one unstable tube drives its neighbor to instability through repeated violent and turbulent impact events which causes tube leakage, tube failures at MSLB test conditions and or unprecedented tube-tube wear, Tube-to-AVB/Tube Support Plates wear, as we saw in SONGS Unit 3. So in review, due to narrow tube pitch to tube diameter, low tube wall thickness/diameter ratio, high tube length/diameter ratio, low tube clearences, in certain portions of the RSGs U-tubes bundle, fluid velocities exceed the critical velocities due to extremely high steam flows (100% SONGS power conditions outside the industry NORM). These high fluid velocities cause U-tubes to vibrate with very large amplitudes in the in-plane direction and literally hit other the tubes with repeated and violent impacts. Due to lower steam operating pressures (required to generate more heat, electricity and profits) and excessive pressure drops due to high flows and velocities, steam saturation temperature drops. This lowering of steam temperature combined with high heat flux in the hot leg side of the U-tube bundle causes steam dry-outs to form (Vapor fraction >99%), known as “NO Effective Thin Tube Film Damping.” Thin film damping refers to the tendency of the steam inside the generators to create a thin film of water between the RSG tubes and the support structures. That film is enough to help keep the tubes from vibrating with large amplitudes, hitting other tubes violently, and protect the Anti-Vibration Bar support structures and maintain the tube-to-AVB gaps and contact forces. These adverse conditions in SONGS at 70% power operation (RTP) with the present defective design and degraded of RSGs known as fluid elastic instability (Tube-to-Tube Wear, or TTW) can lead to rapid U-tubes failure from fatigue or tube-to-tube wear in Unit 2 due to a main steam line break as seen in SONGS Unit 3 RSG’s. In summary, FEI is a phenomenon where due to SONGS RSGs design intended for high steam flows causes the tubes to vibrate with increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit (critical velocity) for a given tube and its supporting conditions and a given thermal hydraulic environment. This occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports. Nucleate boiling on the tube surface or a little amount of water (aka damping, 1.5% water, steam-water mixture vapor fraction <98.5%) is found to have a stabilizing effect on fluid-elastic instability.

    C.2 – For more information, please see comments posted by HelpAllHurtNeverBaba, January 18, 2013 at 12:20 am on this blog

  83. Edwin Hackett, Executive Director ACRS ==> Main number is 301-415-7360
    Thanks for your reply, and staying aware of what is happening at SONGS aka SanO.

  84. SanO is now a 1.5 Billion Dollar RED FLAG that illustrates how easy NRC regulations can be gamed (without ANY enforcement penalties) which allow Utilities/Operators to make changes that have enormous implications to safety and the Public Health, with little to N☢ actual oversight, until it is to late!

    The two basic problems at Fukushima, Japan were that:
    (1) TEPCO’s regulator pushed too much paper instead of being “hands on”.
    (2) TEPCO had total control over what data the public had access to, which prevented any real oversight by the public.

    The USA cannot afford a Trillion Dollar Eco-Disaster like Fukushima, that is why the NRC needs to “overhaul” how it enforces its current regulations and develop new regulations ASAP to patch all the regulatory holes that now exist!

    A major first step should be to really open up the entire NRC process to the public, so that true public oversight can take place, instead of the flawed system we now have, as SanO illustrates all to well! As it is now, the public does not have enough access to NRC reports and/or data which prevents all knowledgable people from providing input into the decision making process.

  85. Great comment – I’m looking forward to additional posts from you – Salute!

    Getting far more qualified people involved and especially professionals from outside of the NRC and most importantly from outside of Region IV, is the first step toward answering basic reactor fatigue safety questions that we now know, affect the entire US Nuclear Fleet. If we learned nothing else from the Fukushima tragedy, we now know that when it come to reactor safety, the widest possible public review can only help insure against future nuclear accidents.

    Since you are a QA professional I urge you to read :

    “Press Release 13-01-22 ATHOS Validity Questioned, Qualifying Investigation Required”

    Validity of ATHOS computer model requires NRR Qualifying Investigation. (3 Pages)

    https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?docId=1ltCb57ciXRaOkhK1rhc2BaB0ACXf7MwcSDZZyEAkFDI

    or this one for much more in-depth technical information:

    “Response to NRR RAI #32 – Technical”

    The SCE cannot provide an ACCEPTABLE operational assessment to the NRR, therefore NO RESTART IS POSSIBLE and here ARE THE TECHNICAL REASONS WHY (50 Pages)

    https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?docId=0BweZ3c0aFXcFX05DMWxKNmZXUTA

    and/or the even the longer paper:

    “SCE NRC Presentation analysis + 14 Questions 12-12-17”

    Technical document includes 14 questions affecting US Reactor SAFETY, that the NRC, NRR and RES Regulators need to ask SCE at their 12/18/12 NRR/RES Meeting. (78 Pages)

    https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?docId=0BweZ3c0aFXcFRzBqZUJROWRYNlE

  86. As a professional for many years in manufacturing quality assurance, the first thing that comes to mind is effective root cause analysis. Have all the factors relating to the root cause of the problem been solidly determined? And if so, has this potential for failure been examined at all other plants that might have similar equipment setups?

    Has a failure mode and effects anaysis (FMEA) been conducted to ensure that all potential aspects of failure are considered for retrofitting, including the potential that something at the plant contributed to the failures?

    As another poster opined, over 8 million people live in the area. I think root cause analysis and FMEA study are crucial pieces to help ensure the safety of the plant and the surrounding population.

  87. San Onofre is rated by the Institute of Nuclear Operations (INPO) as an INPO 4 Plant (The Worst Nuclear Plant Rating) and it should also should be rated in NRC Region IV Response Column V (Worst rating) and not in the NRC Response Column I (Best Nuclear Plant Rating).

    San Onofre is the worst nuclear plant in the country with the worst safety record, worst retaliation record, an INPO 4 rating and it is a mockery to place it in NRC Response Column I. NRC Region IV by listing San Onofre in NRC Response Column I, is putting its credibility on line and is displaying clear trends of collusion with SCE. It would be informative to learn who made the decision on San Onofre’s current ranking and why…

    If the NRC San Onofre Special Review Panel wants to be welcomed by Southern Californians at their upcoming February 12 Public Meeting with SCE , the NRC needs to change San Onofre’s rating to NRC Response Column V, which will reflect current reality instead of just wishful thinking.

    Definitions of NRC Response Columns:
    Column I – All performance indicators and NRC inspection findings are GREEN
    Column II – No more than two WHITE inputs in different cornerstones.
    Cornerstone objectives fully met.
    Column III – One degraded cornerstone (two WHITE inputs or one YELLOW input
    or three WHITE inputs in any strategic area).
    Cornerstone objectives met with minimal reduction in safety margin.
    Column IV – Repetitive degraded cornerstone, multiple degraded cornerstones,
    or multiple YELLOW inputs, or one RED input. Cornerstone objectives
    met with long-standing issues or significant reduction in safety margin.
    Response at NRC Agency level
    • Executive Director for Operations to hold public meeting with senior
    utility management
    • Utility develops performance improvement plan with NRC oversight
    • NRC team inspection focused on cause of degraded performance
    • Demand for Information, Confirmatory Action Letter
    Column V. Unacceptable Performance, Unacceptable reduction in safety margin
    Response at NRC Agency level
    •Plant not permitted to operate

  88. ACRS briefings on event-driven issues typically occur after the NRC staff has finished with inspection and oversight activities, which continue with SONGS. The ACRS main number is 301-415-7360

    Edwin Hackett
    Executive Director
    ACRS

  89. @ Mr. Dricks “No, the Advisory Committee on Reactor Safety (ACRS) has not requested a meeting with the NRC technical staff on SONGS related issues.

    Why not?
    Especially since SanO’s RSG tubing now has more damage that ALL the rest of the nuclear fleet combined!

    What are they waiting for, and how would a public person contact the Chief of ACRS?

  90. No, the Advisory Committee on Reactor Safety (ACRS) has not requested a meeting with the NRC technical staff on SONGS related issues.

    Victor Dricks

  91. Reasonable assurance is given when licensees comply with NRC regulations. That said, the NRC is always looking at the adequacy of its regulations to ensure safety.

    Victor Dricks

  92. In reply to Mr. Silberberg: You sir are correct, we need MORE not LESS information made public in order that knowledgeable people can fact check exactly what is happening at SanO. To hide most of the data behind a veil of secrecy, is no longer acceptable especially since that practice is what has resulted in the current 1 to 1.5 billion dollar debacle at SanO.

    This is the first time in the US Nuclear Fleet that what Dr. Joram Hopenfeld, (who also retired from the NRC staff) first described (what we now call the Hopenfeld Effect) as a cascade of SG tube failures, has actually been observed in a Steam Generator (See Response to NRR RAI -32 – Technical ==> Attachment 3 https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?docId=0BweZ3c0aFXcFX05DMWxKNmZXUTA).

    snip
    “The concerns raised by Dr. Hopenfeld are extremely important safety issues. As the ACRS stated:

    • Steam generators constitute more than 50% of the surface area of the primary pressure boundary in a pressurized water reactor.
    • Unlike other parts of the reactor pressure boundary, the barrier to fission product release provided by the steam generator tubes is not reinforced by the reactor containment as an additional barrier.”
    • Leakage of primary coolant through openings in the steam generator tubes could deplete the inventory of water available for the long-term cooling of the core in the event of an accident.

    In the decade since Dr. Hopenfeld first raised his safety concerns, the NRC has allowed many nuclear plants to continue operating nuclear power plants with literally thousands of steam generator tubes that are known to be fatigue cracked! The ACRS concluded that the NRC staff made these regulatory decisions using incomplete and inaccurate information. After receiving the ACRS’s report, the NRC staff considered Hopenfeld’s concerns “resolved” even though it had taken no action to address the numerous recommendations in the ACRS report. The NRC must now formally address Dr. Hopenfeld’s concerns as soon as possible. In the interim, the NRC must stop making decisions affecting the lives of millions of Americans when it lacks “defensible technical basis” because the US cannot afford a Trillion Dollar Eco-Disaster like Fukushima, due to RSG tube failures caused by poor design, fatigue or any other combination of reasons.”

    Because the Hopenfeld Effect has now been proven as factual, the NRC must re-evaluated it’s “dated” thinking and its computer modeling about SG failures which now only allows for a single SG tube failure ASAP… In fact, I predict that time will show that a nuclear accident (not a nuclear incident) was narrowly avoided at SanO on January 31, 2012 only because of shear luck, due to the timing of the discovery of Edison’s poorly in-house designed replacement steam generators (RSG). Had that Unit 3 tube been just a tiny bit stronger and not leaked when it did; then with both Unit 2 & 3 back online when a MSLB occurred, we now know that it would have resulted in the complete venting of the core coolant within minutes…

    This is why what happened at SanO (as the locals like to say) is so important and why the NRC has to “get it right” this time; the safety of the entire US nuclear fleet depends upon it! Just as many basic design problems were discovered after the Fukushima tragedy, Sano has become the model of what NOT to do for all future RSG design engineers globally and demonstrates beyond a shadow of a doubt why having a qualified public review process is so important, especially where the risk of a radioactive “Trillion Dollar Eco-Disaster” is involved.

  93. Mr. Dricks. Would you please make the documents containing the findings of these experts public by posting them on the NRC website, because these are NRC documents and not Licensee documents. Please do it to assure the public of NRC independent conclusions, because public pays all the bills for the government via taxes. Thanks.

  94. A DAB Safety Team Request to the Office of Nuclear Regulatory Research (NRR) Thermal-Hydraulic Experts.

    Please carefully review the SONGS Unit 2 Restart Reports (done by SCE, Westinghouse, AREVA and MHI), SCE Unit 3 Root Cause Evaluation, NRC AIT Report, ATHOS Modeling Results and Unit 2 Operational Data and then arrive: (1) At an unanimous, clear and concise conclusion whether FEI occurred in Unit 2 or not, and (2) Provide a GAP ANALYSIS (The scientific and engineering reasons why all these reports are so different) prior the February 12, 2013 NRC Public Meeting.

    This will be most helpful for everyone on the Special Hearing Panel and the public at large.

    ===> BTW: The DAB Safety Team will show you ours after the NRC shows US theirs…

  95. From Mr. Dricks’s NRC Feb. 12, 2013 Meeting notice “NRC TO MEET PUBLIC TO DISCUSS SAN ONOFRE NUCLEAR GENERATING STATION STEAM GENERATOR ISSUES”
    snip:
    ” A leak in a Unit 3 steam generator tube on Jan. 31, 2012, led to the shutdown of that unit. The other reactor, Unit 2, was shut down for maintenance and refueling at the time. Subsequent inspections of the nearly new steam generators in both units found unexpected wear. Both units remain safely shut down and will not be permitted to restart until NRC has reasonable assurance they can be operated safely.”

    To Mr. Dricks’s: PLEASE define “reasonable assurance,” as the Health and Safety of 8 Million people living in southern California (within 50 miles of SanO), who are depending upon the NRC to “get it right this time after failing to get it right last time,” because the USA cannot afford a Trillion Dollar Eco-Disaster like Fukushima where the Japanese nuclear regulators thought they had everything covered before 3/11/11 and time proved them tragically wrong.

  96. Using “technical experts from headquarters” is not the same thing as having them DIRECT this SPECIAL panel’s “discovery” process! As populated now, this review panel can insure that Region IV stays in charge of its own investigation, which should not be the case, since SanO problems were caused in part by Region IV in the first place due to lax enforcement!

  97. Salute to Mel Silberberg for his great comment!
    As I have also posted, the NRC needs to populate this panel with people from outside Region IV for obvious reasons and also technical reasons as Mr. Silberberg mentions above.
    +
    Perhaps Mel would consider helping the DAB Safety Team’s “Battery of Nuclear Experts”, if so our contact info is listed on any of our documents posted here: https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit

  98. Thank you Victor. Was the SONGS problem discussed with the ACSR? If so please send a reference to the meeting. Why wasn’t their an intensive, public peer review meeting (conference ) involving experts from around the world, including EPRI, comparing their analyses. The SONGS issues were so surprising – we’ve been using steam generators for so long, one has to suspect some new phenomena and or condition never seen before. Given the financial impact and safety significance–the public demands reassurance. Peer reviews are done for this reason. The cost of the shutdown, new generators, and replacement power cost to SCE is over a billion dollars!
    If I were the industry I would be concerned– this is not a nuclear problem- but the general public doesn’t know the difference. You need to answer these questions at the Public Meeting next month in Carlsbad.

    Mel Silberberg

  99. Region IV used technical experts from headquarters, including the Office of Nuclear Regulatory Research, as part of the Augmented Team Inspection following the steam generator tube leak. These technical experts have continued to advise and make recommendations to the Oversight Panel, as the NRC has conducted follow-up inspections and reviews the SONGS CAL response. Before the NRC makes a restart decision, it will ensure all the appropriate discipline experts, including thermal hydraulics and materials, have reviewed their respective areas of technical expertise.

    Victor Dricks

  100. Respected Mr. Mel Silberberg, NRC RES Retired; former Chief of Severe Accident Research Branch … I totally agree with your comments. Just as a reminder…. NRC website states, “As an independent regulatory agency that prides itself on openness, the U.S. Nuclear Regulatory Commission (NRC) is pleased to take an active role in President Barack Obama’s Open Government Initiative, with its focus on open, accountable, and accessible government. The NRC has a long history of, and commitment to, transparency, participation, and collaboration in our regulatory activities.”
    During FOE presentation: Top NRC official fell asleep during presentation — “His eyes were rolling back and his head was bobbling like a little bobble doll” — Process designed to freeze public out. This type of behavior during public presentations on matters of life and death for Southern Californians conflicts with NRC’s commitment to participation in regulatory activities and President Barack Obama’s Open Government Initiative, with its focus on open, accountable, and accessible government. This is one example, so what is new. NRC needs to shift gears and take a very aggressive, prudent and super-conservative approach, commensurate with its authority granted by the Public, President Obama and United States Congress, when it comes to SCE and San Onofre’s profit-motivated and repeatedly dangerous public safety-ventures.

  101. Hi,
    Thanks for your very kind comments. I am just a very average person, lucky to be working with a very dedicated and highly technical team trying to establish facts about San Onofre.

  102. The “Corporate Support” portion of my title refers to oversight of budget and staffing for the NRC’s program for licensing and oversight activities involving operating reactors. Most NRR staff members are actively involved in public support and outreach through, among other things, timely posting of public records to the agency’s document management system, ADAMS, and through planning and participating in public meetings on many diverse topics. In addition, NRR is supported in public outreach by other offices, including the role of the Office of Public Affairs in providing social media such as this blog.

    Dan Dorman

  103. SanO Nuclear Denial*?
    Perhaps this panel will also explain why “severe accident” is not even listed in it’s 130 page NRC Collection of Abbreviations, especially since there are two classes of accident: postulated accidents and severe accidents.

    * http://is.gd/XPjMd0
    The illogical belief that Nature cannot destroy any land based nuclear reactor, any place anytime 24/7/365!

  104. Question: If Dan Dorman, is the deputy director for engineering and CORPORATE support in the Office of Nuclear Reactor Regulation (NRR), who at the NRR is tasked with providing PUBLIC SUPPORT?

  105. Trying to help San Onofre Special Panel,,, Thanks to NRC Moderation for Posting… HAHNBaba
    Credit of DAB Safety Team Press Release 12-12-20
    Prior to Issue of any decision regarding restart for Unit 2, SCE needs to demonstrate the viability of Operator Actions for an earthquake, main steam line break or other unanticipated transients in a Full NRC/FEMA Evaluated Emergency Plan Exercise collocated by NRC Head Quarters and evaluated by IPC/Industry Emergency Preparedness/Reactor Oversight and NRR Experts using the following:
    • Fully Staffed Control Room or Simulator, Technical Support Center, Operations Support Center, Emergency Offsite Facility, Joint Information Center and Fire Department
    • Ability for Accurate & Timely detection of a tube leak using N-16 radiation detection system and initiation of operator actions
    • Ability for Accurate & Timely Identification, Trouble Shooting, Diagnostics and Mitigation of the above events using VLPMS accelerometers for detecting actual tube vibrations for fluid elasticity Mitigation
    • Ability for Accurate & Timely demonstration of actual tube vibration noise from background noise and the required threshold identification criteria, that would be applied to reach the conclusion that tube vibration is occurring and the number of affected damaged and worn tubes
    • Ability for Accurate & Timely use of Emergency, Abnormal & Severe Accident Management Procedures
    • Demonstration of Excellent Communications, Solid Team Work & Alignment, Critical Questioning & Investigative Attitude between all Emergency Operating Facilities, NRC Headquarters, Federal Emergency Management Agency, State of California and Offsite Agencies including Offsite Dose Assessment Committee, California Highway Patrol, Fire Departments, News Media, Emergency Medical Facilities and Public Interest Groups
    • Ability for demonstration of Accurate & Timely Emergency Declarations, Offsite Notifications / Communications, and Protective Actions Recommendations & Decisions
    • Ability for prompt notification, evacuation and/or sheltering of disabled, transient and permanent residents in the Emergency Planning Zone during rush traffic hours
    Acceptance Criteria:
    • 100 % Accuracy in Emergency Declarations, Offsite Notifications / Communications, and Protective Actions Recommendations & Decisions
    • No more than 5 Drill/Exercise Weaknesses

  106. correction–on the fifth line, after ‘peer-reviewers’ please add [should be added to the panel.]

  107. I am disappointed in the composition of the special panel! Where is the representation from NRC-RES? The issues at SONGS involve thermal hydraulics and material science. The NRC-RES and its contractors are experts in these areas. The Office of Research was created by the Congress for such situations. Two RES staff covering these disciplines and one or two consultants, serving as peer-reviewers. Perhaps there needs to be a separate peer review. Public confidence can only be gained using logical, informed measures as I described above.
    Mel Silberberg, NRC-RES, Retired [Chief, Severe Accident Research Branch; Waste Management Branch.

  108. The issues involved in the SONGS steam generator encompass thermal-hydraulics and material science and technology. I am extremely upset and disappointed in the lack of judgement displayed by senior NRC management and the Commission in the glaring omission of the NRC Office of Research[RES] from playing a major role in this special panel. It is at times like this that RES was created by the Congress.to get an independent, confirmatory assessment of abnormal behavior in a nuclear plant.hThe Chairman should insist on the following additions to the panelOne staff expert on thermal-hydraulicsand one staff expert in material science. In addition two consultants from the unverities and or national labs serve on the panel, as peer reviewers. you can not win public confidence in your findingwithout these additions to the to the panel.
    Mel Silberberg, NRC RES Retired; former Chief of Severe Accident Research Branch.

  109. Just trying to help the NRC San Onofre Special Panel with some of the facts:
    1. NRC Augmented Inspection Team Report and SCE Cause evaluations on both San Onofre Unit 3 and 2 FEI are still unresolved and open based:
    A. ATHOS limitations disputed by John Large, Arnie Gundersen, Academic Research Scholars and DAB Safety Team,
    B. Insufficient tube-to-AVB contact forces on Unit 3 disputed by DAB Safety Team, Westinghouse, MHI, AREVA and John Large, and
    C. Operational Factors based on the information from San Onofre Plant Data disputed by SONGS Root Cause Team Member and DAB Safety Team.
    Therefore, the decision of NRC San Onofre Special Panel should take into account the above facts. Thanks to the NRC Moderator for posting this information.

  110. Just trying to help the NRC San Onofre Special Panel with some of the facts:
    1. San Onofre Emergency Preparedness DEP Indicator Value is consistently amongst the lowest in the US Nuclear Power Plants,
    2. The Shift Manager Training Guru was on duty at the time of San Onofre Unit 3 Accident, so the reactor was shutdown in a timely and safe manner. Southern Californians were lucky,
    3. The other best known Shift Manager resigned due to differences with plant management,
    4. The best known Station and Corporate Emergency Directors have retired,
    5. The other Shift Managers, Station and Corporate Emergency Directors record of accomplishment is for NRC San Onofre Special Panel to judge,
    6. The Manager of Plant Operations is very knowledgeable, and
    7. Therefore, the probability of success to avert another potential accident due to Restart of Defectively-Designed and Degraded Unit 2 Replacement Steam Generators at 70% power is 50% based upon who is on duty at the time of the accident (due to a design bases main steam line break or other anticipated operational occurrences).
    Therefore, the decision of NRC San Onofre Special Panel should take into account the above facts. Thanks for posting.

  111. EPRI, NRC, Westinghouse, AREVA and MHI ATHOS thermal-hydraulic computer models cannot accurately account for all the mechanical and structural unknowns, and extremely narrow tube-to-tube clearance differences, which would very likely govern the catastrophic tube-to-tube wear (fluid elastic instability) in San Onofre Unit 2 during a main steam line break or other anticipated operational occurrences at 70% power. Computer Modeling predictions are as good as the input based on the as-built hot pressurized U-Tube Bundle Anti-Vibration Structure behavior, which nobody knows. John Large, Internationally Known Scientist and Chartered Nuclear Engineer from London says about the SONGS Unit 2 Replacement Steam Generators (RSGs) AVB Structure, “It impossible to reliably predict the effectiveness of the many thousands of AVB contact points for when the tube bundle is in a hot, pressurized operational state. The combination of the omission of the in-plane AVB restraints, the unique in-plane activity levels of the SONGS RSGs, together the very demanding interpretation of the remote probe data from the cold and depressurized tube inspection, render forecasting the wear of the tubes and many thousands of restraint components when in hot and pressurized service very challenging indeed.”

  112. The 50.59 is required for ANY change in the plant. The 50.59 was performed. 50.59 is what says you don’t need a license amendment.

    I think you’re talking about the 50.90 process for a license amendment.

  113. It’s sloppy reading like that that makes your opinions so mistaken. I didn’t ask any questions, so you can’t possibly answer them. All your responses were in error.
    1. Unit 2 steam generator has no leaks, exactly as I said.
    2. All tubes have been inspected and tubes with wear blocked. Your nonsensical “visual inspection” is not an option and in any case is not the best technology for assessment.
    3. Undertaking a test run, especially one with no adverse consequences, is the rational way to proceed to improve understanding of the situation and possible future scenarios.
    4. Understanding likelihood is an essential part of risk management.
    5. A leak or break in steam generator tubes has no prospect of causing a loss of coolant accident, no matter what fantasy incident you decide to give an acronym to. Adding “that is factual” to your daydreams does not make it any more likely.
    6. Your “calculations” are immaterial; reality dictates that no such outcome will occur.

  114. Wish nuclear opponents would just come clean and just say you’re not concerned of any fix or even perfect reactors but just want them all the hell out of here as a matter of “conscience”. This way meetings can deal with the more open-minded concerned who don’t belittle and disparage the family-loving engineers and technicians who are investing their time and effort not only fixing reactor issues but making them even safer than they are now. Yes, safe reactors are an contradiction to hard-core antis, but the historic mortality record of nuclear reactors is unassailable and enviable, especially since it’s pretty hypocritical to other energy sources slide with their score of the tens thousands killed and maimed just by accidents alone. It seems some safety concerns are barking up the wrong bogeyman to me.

    James Greenidge
    Queens NY

  115. The more that take interest and demand an open 50.59 process before any restart decision, the safer all of California and the rest of the US will be…

  116. It’s thinking like that that has gotten Edison and too many of the NRC’s regulators into hot water!
    BTW in answer to your questions:
    1. One RSG failed in less than a year of operation (starting from NEW) and now has more damaged, worn and plugged tubes that the entire rest of the US Nuclear fleet.
    2. Both Unit 2 and Unit 3 have a unknown amount of wear because Edison has not visually inspected all the thousands of tubes using the best technology to insure safety, because then they would be forced to exceed their tube plugging limitation!
    3. This is nothing but using San as a TEST site instead of a proven safe reactor/RSG as the law demands.
    4. Even thinking 50-50, either it will or will not have a nuclear incident, it’s N☢T worth the Gamble!
    5. Wrong, worse case is a Fukushima-type event/disaster after a MSLB… That is factual!
    6. We calculated that within 5 minutes, a cascading tube failure would doom the reactor core, because of loos of coolant which would be vented to the atmosphere, all beyond the ability of the Operators to prevent!

  117. Chairman Allison Macfarlane said Unit 2 would not be permitted to restart unless the NRC has reasonable assurance it can be operated safely. Let us examine at the scenario below and determine whether NRC can have that reasonable assurance or not? Thanks to NRC for posting this blog. Just trying to help… HelpAllHurtNeverBaba

    Let us examine, what John Large, Internationally Known Scientist and Chartered Nuclear Engineer from London says about the SONGS Unit 2 Replacement Steam Generators (RSGs) AVB Structure, “It impossible to reliably predict the effectiveness of the many thousands of AVB contact points for when the tube bundle is in a hot, pressurized operational state. The combination of the omission of the in-plane AVB restraints, the unique in-plane activity levels of the SONGS RSGs, together the very demanding interpretation of the remote probe data from the cold and depressurized tube inspection, render forecasting the wear of the tubes and many thousands of restraint components when in hot and pressurized service very challenging indeed.”

    Let us examine, what MHI says about tube-to-AVB Gaps, “The most likely cause of the observed tube-to-tube wear is multiple consecutive AVB supports becoming inactive during operation. This is attributed to redistribution of the tube-to-AVB-gaps under the fluid hydrodynamic pressure exerted on the tubes during operation. This phenomenon is called by MHI, “tube bundle flowering” and is postulated to result in a spreading of the tube U-bends in the out-of-plane direction to varying degrees based on their location in the tube bundle (the hydrodynamic pressure varies within the U bend). This tube U-bend spreading causes an increase of the tube-to-AVB gap sizes and decrease of tube-to-AVB contact forces rendering the AVB supports inactive and potentially significantly contributing to tube FEI.”
    Let us examine, what AREVA says about tube-to-AVB Gaps, “Contact forces significantly increase at normal operating temperature and pressure due to diametric expansion of the tubes and thermal growth of the AVBs. After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location.”

    Let us examine, what Westinghouse says about tube-to-AVB Gaps, “Test data shows that the onset of in-plane (IP) vibration requires much higher velocities than the onset of out-of-plane (OP) fluid-elastic excitation. Hence, a tube that may vibrate in-plane (IP) would definitely be unstable OP. A small AVB gap that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion.”

    Let us examine, what John Large says further, “There is no account of the changes that have been made in the evaluation of the tube structural and leakage integrity, that is from the stage of predicting those tubes at risk of TTW and other forms of wear, the tube thinning wear rates, through to the nature of the tube failure being unique to the type and extent of the wear pattern and tube thinning; and the methods of deducing, mainly by unproven inference, from the probe inspection results particularly to determine the in-plane AVB effectiveness, includes unacceptably large elements of test and experimentation that are inconsistent with the analyses and descriptions of the FSAR. I provide a number of explicit examples where I consider that the circumstances and risks accompanying the proposed restart of Unit 2 will result in unacceptable levels of test and experiment.

    What these World Known Experts are saying is that this degraded tube bundle cannot prevent multiple tube ruptures from fluid elastic instability as we saw by the failure of 8 tubes in Unit 3 RSGs under Main Steam Line Break (MSLB) test conditions.

    Main Steam Line Break (MSLB) Scenario: The most severe design basis accident to meet the SONGS Unit 2 TS 5.5.2.11.b.1 steam generator structural integrity is a MSLB at the first weld outside containment. This assumption minimizes the flow resistance between the break and the affected SG and maximizes the mass & energy (M&E) release. The analyses focus on M&E releases at licensed Rated Thermal Power (RTP or 100% Power). The outside containment case includes the assumption that the main steam isolation valve (MSIV) in the steam line with the least flow resistance fails to close following the main steam isolation signal (MSIS). This assumption maximizes the M&E release during a MSLB outside of the containment. Super-heating within the SG initiates upon U-tube uncovery as specified in the NRC Information Notice 84-90. The turbine stop valves are assumed to close instantaneously at the time of the reactor trip. This assumption is conservative for a MSLB event because the entire steam inventory at the time of reactor trip is assumed to be forced out the break in 300 seconds or 5 minutes. No Operator action outside Control Room can be credited, if it takes less than 30 minutes.

    The depressurization of the non-isolable steam generator would result in 100% void fractions in the degraded Unit 2 U-Tube bundle due to instant flashing of the sub-cooled 440 degrees Fahrenheit feedwater into steam. This condition of ZERO Water in the steam generators would cause fluid elastic instability (FEI), flow-induced random vibrations and excessive hydrodynamic pressures (Mitsubishi Flowering Effect). The force of the flashing steam would create high-energy jets, lifting loose parts and debris present in the steam generator, which would do additional damage by cutting holes into the already degraded tubes and creating additional loading (See Note A below) on the tube support plates (TSPs) due to heavy build-up of deposits on trefoil/quadrifoil-shaped holes from SG blowdown and crack the high cycle fatigued U-bend tubes not supported by Anti-Vibration Bars (AVB). These cumulative adverse conditions in all likelihood would result in a massive cascading of RSG’s tube failures (tubes would excessively rattle or vibrate, hitting other tubes with violent impacts) due to extremely low tube-to-tube clearances and no effective or non-existent in-plane anti-vibration bar support protection system. This jackhammering effect would involve hundreds of degraded active SG tubes along with all the inactive (plugged /unstabilized) tubes causing a catastrophic amount of simultaneous tube leaks/ruptures. Under this adverse scenario, approximately 60 tons of very hot high-pressure radioactive reactor coolant would leak into the secondary system. The release of this amount of radioactive primary coolant, along with an additional approximately 200 tons of steam in the first five to fifteen minutes from a broken steam line would EXCEED the SONGS NRC approved offsite radiological release doses safety margins based on assumption of a single tube rupture in the SONGS FSAR. So, in essence, these RSG’s are like loaded guns, or a Fukushima-type nuclear accident, waiting to happen. Any failure under these conditions would allow significant amounts of radiation to escape to the atmosphere and a major Loss of Coolant Accident (LOCA) could easily result causing much wider radiological consequences and even a potential nuclear meltdown of the reactor.

    SCE states, “A MSLB alone does not generate sufficient differential pressure to cause tube rupture. The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators (See Note B below) prevent RCS re-pressurization in accordance with Emergency Operating Instructions.” SCE’s suggested DID Actions and proven unreliable operator actions to detect a leak and/or to re-pressurize the steam generators as claimed by Edison are not practical to stop a major nuclear accident from occurring in Unit 2 in the first 5-15 minutes of a MSLB during the proposed 5-month trial period.

    NOTES:

    A. Plugging of the at-risk tubes is not a satisfactory solution because it is the retainer bar that vibrates via random fluid flow processes at sub FEI critical velocity levels – these are likely to continue in play or, indeed, exacerbate at the proposed U2 restart at 70% power, leading to through-tube abrasion, the detachment of tube fragments, lodging at other unplugged and in-service tube localities, resulting in the so-called ‘foreign object’ tube wear. This additional loading would exceed: (1) the safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials, and (2) significantly affect burst or collapse pressures determined and assessed in combination with the loads due to a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads. [emphasis added

    B. SCE’s suggested “defense-in-depth” actions are insufficient to stop multiple tube ruptures due to the short duration of a main steam line beak event. Human performance weaknesses, such as mis-diagnoses, substantial delays in isolating the faulted steam generator, communication errors and delayed initiation of the residual heat removal system, have been identified in past events at SONGS and other US Nuclear Power Plants. The events also involved unnecessary radiation releases, lack of RCS subcooled margin, excessive RCS cooldown rates, and overfilling the SG because of human or procedural problems.
    TRILLION DOLLAR ECONOMIC-ENVIRONMENT-HUMAN DISASTER QUIZ: If NRC Region IV Special Technical Panel can determine whether Southern Californians have reasonable assurance that a Fukushima will not occur by granting Unit 2 Restart Permission to Profit-Motivated Gambler SCE? If Fukushima occurs, who is liable, Nuclear Utilities, Insurance Carriers, Federal Government, State of California, CPUC, NRC Commissioners, NRC Region IV, EIX/SCE Shareholders & Employees or Affected Southern Californians?

  118. Lets be scientifically rigorous, SCE’s Richard St. Onge has as much of a chance of explaining how FEI destroyed Unit 2 and 3 at SanO by pointing to their poorly in-hous designed 620 ton replacement steam generator, as NRC Chairman Allison Macfarlane would have of explaining how dip/slip effects earthquakes by pointing to the earth! This photo is an insult to the intelligence of your readers and the Director of the NRC, who happens to be a World Class Geologist…

  119. SCE’S PR MACHINE IS CAPABLE OF OVERCOMING ALL HURDLES, EXCEPT GOOD SCIENCE AND SAFETY

    Here is much more about: NRC Violating Presidential Directive and the Public Trust https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?docId=1QnQRbpsWgxn5xVftt-fmBBaPohF5m5OXyQgyWuz8vJI

    San Onofre Unit 2 Restart Decision by NRC Imminent – SCE’s PR Machine Is Capable Of Overcoming ALL Hurdles, Except Good Science And Safety

    NRC’s enforcement history, drama and pre-rehearsed tough questions, press reports, casual relationship and/or protection of SCE officials and utility biased public meetings are just old and cheap regulatory tricks that are now being used to protect the NRC’s own public image and to fool the public into believing that the NRC is really concerned about public safety regarding SCE’s Restart Plan. The Justice Department & NRR Officials need to set up a legal/technical taskforce to publically question Edison’s design and MHI Engineer’s listed below under oath regarding their:
    Understanding of their legal obligations under the 10 CFR 50.59 Process,
    Understanding of problems with the original steam generators,
    Critical questioning and professional/investigative skills,
    Efforts made in industry and academic benchmarking to identify and resolve problems
    with the original steam generators
    What part did they play in the preparation of design specifications, fabrication, computer
    modeling, mock-up testing, anti-vibration bar structure, and research required to prevent
    the adverse effects of fluid elasticity and flow-induced random vibrations in these unique
    San Onofre Combustion Engineering replacement generators.

    Any NRC decision to grant a restart of Unit 2 without a formal 50.90 licensing review along with public participation will be seen as an invitation to risk a Fukushima-type disaster happening in Southern California.

  120. S0 NRC does not want to shut it down, they hand it off to others that will and wipe their hands of it, cowards. We now know they will allow it, what a scam.

  121. Gee, it looks so shiny and new you can see their reflections! And why does St. Onge have different color booties on? Is that like a leadership color or something?

    I too believe an unbiased review team shouldn’t have any Region IV NRC personnel on it. They’ve already shown gross incompetence, industry bias, and unsound judgement. We don’t need any more of that here. The AIT already utterly failed to find the root cause, and failed to declare Units 2 and 3 inoperable as they should have, thus allowing SCE to proceed to ask for a restart license — maybe even the same lousy engineers and executives who built/approved the first failed redesign. It makes further sense to have non-region IV personnel for the simple reason that this problem — FEI — potentially affects the entire PWR nuclear industry, so they might as well get used to it. Other reactors (and their inspectors) should also be reanalyzing what might happen in a MSLB condition with degraded tubes, or in any event — design basis or other — that ruptures multiple SG tubes. Boeing ain’t got half the problems our nuclear fleet has — and by the way, what kind of batteries do we use at SanO???

  122. Dear Mr. Art Howell
    NRC’s Region IV Deputy Regional Administrator

    FYI and Help – Courtesy of DAB Safety Team

    Please read Media Alert 13-01-17 Allegation – NRC Violating President’s Directive And the Public Trust and other San Onofre Papers by searching DAB Safety Team & Related Information on Google Drive.

    Personal Note: If you want to know more SCE Safety and Retaliation Issues and examine the evidence with my and NRC attorneys present about SCE Management, Retaliation, Safety Issues and San Onofre $1 Billion Radiation Steaming Crucible Watergate Insider Secrets, please visit me in Southern California for a Face-To-Face Technical Meeting. These concerns have already been relayed to Senator Barbara Boxer. Please feel free to send me an email helpallcqiascnp@yahoo.com. Thanks HELPAllHurtNeverBaba

  123. OK, let’s be efficient here. I can do this panel’s remaining work in one comment.

    1. Past performance of this steam generator: no leaks
    2. Past work: block tubes showing any sign of wear
    3. Proposal: operate the reactor for a shorter period than produced the previous wear, reevaluate afterwards
    4. Likely outcome: safe operation with intact steam generator due to points 1,2,3 @ probability > 99%
    5. Worst case outcome: tube leak/break in steam generator
    6. Worst case consequence: immediate detection, reactor shutdown, no safety implications

    Conclusion: Restart SONGS 2 immediately.

  124. The above Panel should N☢T be co-chaired by anyone in Region IV, since their supervision of Edison has been called into question and the panel should include at least one and preferably two outside experts to insure that this HISTORIC NRC/NRR Panel is not just covering up for the NRC (and Edison) to protect its own public image!

    FEI does not care about NRC internal politics, nor does it follow inter-office memo’s or yield to graft.

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