U.S. NRC Blog

Transparent, Participate, and Collaborate

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)

134 responses to “NRC Forms Special San Onofre Review Panel

  1. CaptD January 30, 2013 at 9:35 pm

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

  2. Moderator January 30, 2013 at 5:11 pm

    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.

  3. CaptD January 29, 2013 at 6:54 pm

    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.”

  4. HelpAllHurtNeverBaba January 29, 2013 at 1:04 am

    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

    • 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

      • Mel Silberberg January 29, 2013 at 2:55 pm

        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

      • CaptD January 31, 2013 at 8:03 am

        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

  5. HelpAllHurtNeverBaba January 28, 2013 at 10:05 pm

    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.

  6. HelpAllHurtNeverBaba January 28, 2013 at 8:19 pm

    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.

  7. HelpAllHurtNeverBaba January 25, 2013 at 3:07 pm

    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

  8. Dee January 25, 2013 at 5:22 am

    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.

  9. CaptD January 24, 2013 at 7:55 pm

    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

  10. CaptD January 24, 2013 at 11:25 am

    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.

Leave a Reply

Fill in your details below or click an icon to log in:

WordPress.com Logo

You are commenting using your WordPress.com account. Log Out / Change )

Twitter picture

You are commenting using your Twitter account. Log Out / Change )

Facebook photo

You are commenting using your Facebook account. Log Out / Change )

Google+ photo

You are commenting using your Google+ account. Log Out / Change )

Connecting to %s

Follow

Get every new post delivered to your Inbox.

Join 1,489 other followers

%d bloggers like this: