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NRC Keeps an Eye on Gulf Coast Flooding

Victor Dricks
Senior Public Affairs Officer
Region IV

Torrential rains have been battering the Gulf Coast since Friday, but have not adversely affected any of the nuclear power plants in Louisiana, Mississippi, or Arkansas.

louisiana map_sealThough skies have now cleared over Baton Rouge, the area has been especially hard hit by flooding. But this has had no significant impact on the River Bend nuclear power plant, about 25 miles northwest of the city, or the designated routes that would be used to evacuate the public in the event of a nuclear emergency.

The Waterford 3 nuclear plant, located in Killona (about 25 miles west of New Orleans), has been similarly unaffected. “We’ve had some heavy rain here over the weekend but there has not been any real impact on the plant,” said NRC Resident Inspector Chris Speer.

Flooding is one of the many natural hazards that nuclear power plants must be prepared for. Every nuclear power plant must demonstrate the ability to withstand extreme flooding and shut down safely if necessary. Most nuclear power plants have emergency diesel generators that can supply backup power for key safety systems if off-site power is lost.

All plants have robust designs with redundancy in key components that are protected from natural events, including flooding. These requirements were in place before the Fukushima accident in Japan in 2011, and have been strengthened since.

As of Tuesday, Arkansas Nuclear One, in Russellville, has gotten about five inches of rain since Friday, NRC Resident Inspector Margaret Tobin said. “It’s a little muddy at the site, but that’s about it.”

At Grand Gulf plant in Mississippi, 20 miles southwest of Vicksburg, only light rain has been reported. “We actually had very little rain at the site, compared to what was expected,” said Matt Young, the NRC’s Senior Resident at the plant.

The NRC is closely following events and getting periodic updates from the National Weather Service on conditions that might affect any of the Gulf Coast nuclear plants. Additionally, the resident inspectors are monitoring local weather conditions to remain aware of conditions that could affect continued safe operations of the plants.

REFRESH — SATAN’s Code: The Early Years of Accident Models

Thomas Wellock
Historian

When the first mass-produced computers hit the stage in the 1950s, nuclear engineers saw the opportunity to use them to help run accident scenarios. It was a good idea that took decades to become reality and the computer limitations created early uncertainty about reactor safety.

refresh leafIn 1954, Westinghouse experts put together a homemade digital computer that read punch tape. With a practiced ear, you could tell from the computer sounds which program was being run.

In 1959, Battelle Memorial Institute developed an early Loss-of-Coolant-Accident model for a heavy-water plutonium reactor. The program was run on an IBM-650/653, the first mass production computer ever developed. The 650 weighed more than a 1955 Cadillac Deville, had vacuum tubes, and used a punch-card reader. Even if it had the memory and someone willing to load the 50 million cards, it would take six months to boot up Microsoft Windows 7.

Fortunately, Battelle’s code was a mere 166 cards. It calculated the behavior of just one fuel rod (modern reactors have thousands of rods) and took minutes to produce one data point.

For the sake of speedier results, gross simplifications were made. For example, an ideal accident code would have broken a reactor cooling system into many small volumes and done extensive calculations on each one to accurately simulate the complex conditions that existed throughout the reactor core and piping. But to run it on mid-1960s computers could take days. As a result, Westinghouse’s FLASH code used just three volumes to represent the whole reactor system.

At least they had computers. Neither the Idaho National Labs, a center for accident-code development, nor the Atomic Energy Commission had them. INL relied on weekend visits to the University of Utah. At the AEC, engineer Norm Lauben begged time from the National Bureau of Standards. Norm drove to the Bureau’s headquarters in Gaithersburg, Md., in the morning to submit his job on the 12,000-line RELAP-3 code, and returned after lunch to pick up the results.

Engineers were confident that the codes would prove reactor designs were overly conservative. Early results dashed their optimism.

When Westinghouse proudly unveiled its 70-volume SATAN code in 1970, AEC staffers discovered errors in the code indicating that the company’s Emergency Core Cooling System might fail in an accident. The problems of the SATAN code helped lead to a major rulemaking hearing in 1972 on the adequacy of both emergency cooling system designs and accident codes. Those hearings revealed just how embarrassingly uncertain and rudimentary the early codes were about what happened during an accident.

The AEC and later the NRC had to make a huge investment in creating more robust – and accurate – codes. Additional research that produced the RELAP-5 Code that became an industry standard worldwide.

REFRESH is an occasional series where we revisit previous blog posts. This post first ran in July 2011.

NRC Issues Comprehensive Inspection Report on Arkansas Nuclear One

Victor Dricks
Senior Public Affairs Officer
Region IV

The NRC has just issued a lengthy report documenting the results of a comprehensive inspection conducted earlier this year at Arkansas Nuclear One, or ANO. The plant, operated by Entergy Operations, Inc., is located in Russellville, Ark.

anoOn March 4, 2015, the NRC moved ANO into Column 4 of the agency’s Action Matrix (where operating plants with significant performance issues receive the second highest level of NRC oversight). This followed inspection findings of substantial safety significance stemming from a heavy equipment incident as well as degraded flood protection at the site. As part of its increased oversight of ANO, the NRC conducted a rigorous, independent, diagnostic assessment of the performance, programs and processes at the site earlier this year.

A team of 27 NRC inspectors from all four NRC regional offices and NRC headquarters spent about 3,800 hours of direct inspection. They concluded that despite its problems, ANO has been –- and continues to be — operated safely. This judgement was based on the fact there have not been any safety-significant inspection findings since the plant’s move to Column 4. Inspectors also concluded the robust plant design had not been compromised and the ANO staff’s operational focus has improved.

In addition to inspecting ANO’s program for evaluating and correcting performance issues, the NRC team developed insights into the causes for the performance decline at ANO. The team also evaluated the adequacy of a third-party safety culture assessment ANO commissioned. Additionally, the NRC inspection included an assessment of how well ANO determined the root causes for its performance deficiencies and developed performance improvement programs.

PI_ROPBased on Entergy’s review of the causes of the performance decline, the independent third-party nuclear safety culture assessment findings, and the results of the NRC’s independent diagnostic evaluation, “the team determined that Entergy understands the depth and breadth of performance concerns associated with ANO’s performance decline,” NRC Region IV Administrator Marc Dapas said in a letter to Entergy officials accompanying the report.

Dapas further stated that “effective implementation of the comprehensive recovery plan, supported by the allocation of adequate resources and continued enhanced oversight by Entergy leadership, should lead to substantial sustained performance improvement.”

The NRC team documented 16 findings of very low safety significance. The 243-page report goes into a high level of detail about the results of the inspections, but here are some highlights:

  • Resource reductions and leadership behaviors were the most significant causes for ANO’s declining performance. Entergy reduced resources across its fleet in 2007 and 2013, but it did not adequately consider the unique staffing needs for ANO created by having two units with different designs.
  • ANO management did not reduce workloads through efficiencies or the elimination of unnecessary work, as was intended as part of the resource reduction initiatives. Leaders attempted to prioritize work with the available resources, but they did not address expanding work backlogs. Over time, this contributed to equipment reliability challenges.
  • An unexpected increase in employee attrition between 2012 and 2014 caused a loss in experienced personnel, which led to a reduced capacity to accomplish work, and an increased need for training and supervision.
  • Since 2007, the reduced resources created a number of changes that slowly began to impact equipment reliability. The Entergy fleet reduced preventive maintenance and extended the time between some maintenance activities.

Although ANO is in the early stages of implementing its comprehensive recovery plan, there have been some notable improvements in station performance. ANO implemented prompt action to improve operator performance, for example, and ANO management decision making has increased in rigor and conservatism. The NRC resident inspectors have noted a number of examples that indicate the operations department is taking a leadership role and raising the standards across the station. Also, employees are engaging in discussions of the potential risk of plant activities, and the corrective action program rigor has improved. And ANO staff have begun to question the status quo and emphasize the need to assess and address problems.

The NRC inspection team reported its preliminary findings at an April 6 public meeting held in Russellville. The NRC is preparing a Confirmatory Action Letter to document Entergy’s commitments to address performance issues identified in the inspection report and the key actions needed to ensure sustained improvement in safety performance. The NRC expects to issue this letter and make it publicly available later this month.

NRC will conduct quarterly inspections to verify that ANO successfully undertakes all necessary improvement actions. From these inspections, NRC will independently determine whether ANO’s corrective actions have been effective and whether, after a period of sustained good performance, the NRC can return ANO to normal oversight.

An Outage Twist: Degraded bolts at New York Nuclear Plant Warrant Attention

Neil Sheehan
Public Affairs Officer
Region I

When the Indian Point Unit 2 nuclear power plant entered a refueling and maintenance outage in early March, the to-do list included a task born of industry operating experience, both in the United States and overseas.

BaffleBoltsGraphic1_cleanbigfontSpecialists were geared up to check on the condition of bolts employed in the reactor vessel at the Buchanan, N.Y., facility. These are the kind of bolts you likely wouldn’t find at your local hardware store. Rather, they are made of a stainless-steel alloy capable of withstanding decades’ worth of neutron bombardment, as well as extraordinarily high temperatures and pressure.

Measuring about 2 inches in length and 5/8ths of an inch in diameter, the bolts hold in place a series of vertical metal plates. Known as baffle plates, they help direct water up through the nuclear fuel assemblies, where it is heated and subsequently used for power production.

The baffle plates are attached to eight levels of horizontal plates called baffle-former plates, which are in turn connected to the reactor core barrel.

As far back as the late 1980s, cracking was identified in baffle-former bolts – the bolts securing the baffle plates to the baffle-former plates — in pressurized-water reactors (PWRs) in France. (Both Indian Point Units 2 and 3 are PWRs.) The degradation is caused by what is known as irradiation-assisted stress corrosion cracking.

In response, the U.S nuclear industry began checking on these bolts in a small number of domestic PWRs on a sample basis.

The NRC staff also made use of a communications tool called an Information Notice to advise U.S. plant owners of what had been observed in Europe. A March 1998 notice let U.S. plant owners know that the baffle-former bolt cracking reported in foreign PWRs had occurred at the juncture of the bolt head and the shank, a location not accessible for visual examination.

Several U.S. plants subsequently evaluated their baffle-former bolts and in some cases replaced a sizable number.

Jumping ahead, the Electric Power Research Institute developed a standard industry program for the aging management of PWR reactor vessel internals and submitted it to the NRC in January 2009. The NRC staff approved the approach in an agency safety evaluation issued in December 2011 and then published in January 2012 as MRP-227-A. (MRP is short for Materials Reliability Program.)

Under this new standard, U.S. PWRs were to conduct an initial ultrasonic examination of all of their baffle-former bolts when the plant had between 25 and 35 effective full power years of service.

This is exactly what was being done at Indian Point Unit 2 during the current outage. It was adhering to the standards of MRP-227-A. In the course of this review, it was determined that 227 of 832 baffle-former bolts at the plant were degraded, which means any indication of cracking. What’s more, two bolt heads were missing.

The number of degraded baffle-former bolts was the largest seen to date at a U.S. reactor.

Entergy, Indian Point’s owner, is in the process of analyzing the condition and replacing the degraded bolts. It will also assess any implications for Indian Point Unit 3, though that reactor is believed to be less susceptible to the condition for several reasons, including fewer operational cycles.

As for the NRC, we will independently review the company’s analysis and bolt-replacement plans to ensure safety. The results of those reviews will be documented in an upcoming inspection report for the plant.

We have already had a metallurgical specialist on-site reviewing the company’s evaluations of the bolts and have agency experts reviewing the matter.

More information will be forthcoming on the issue. However, it’s important to note that the NRC staff will ensure the condition is fully understood and addressed prior to the plant returning to service. The NRC staff will also consider all available information in evaluating if changes are needed to the current inspection programs for these bolts across the industry.

 

NRC Talks Research in Tennessee

Salman Haq
Reactor Engineer
Office of Nuclear Regulatory Research

We recently issued the draft report summarizing detailed research and analyses into what might happen during an accident at a nuclear power plant. Tomorrow, we’ll head to the third plant we analyzed, Sequoyah Nuclear Plant, to discuss the results with the surrounding communities. The plant is located in Soddy-Daisy, Tenn.

Cover of SOARCA Communications Brochure (NUREG BR-0359 Rev2)The project, called the State-of-the-Art Reactor Consequence Analyses, or SOARCA, looked at potential situations that could disable a reactor’s normal safety systems. The project used powerful computer programs to predict the plants’ behavior based on decades of real-world experiments into issues such as how reactor fuel responds during the extreme temperatures expected during these accidents.

SOARCA then plugged up-to-date information about the plants into the programs and examined how a potential accident might unfold.

We found that safety equipment the NRC required after the 9/11 attacks, or additional equipment that industry voluntarily added following the Fukushima event, if used according to plan, would help prevent or mitigate a reactor accident. Even for the most severe accidents the research came to three basic conclusions:

  • Accidents occur more slowly than we originally thought;
  • Accidents release less radioactive material than we originally thought; and
  • The emergency plans every U.S. reactor has in place can help keep people safe.

The project came to some more specific conclusions about accident effects around the three plants, Surry (southeast of Richmond, Va.), Peach Bottom (southeast of Lancaster, Pa.), and Sequoyah. For example, the slowly developing nature of the accidents and the existing emergency plans would help keep people safe, even during uncontrolled accidents.

Some of the NRC staff involved in SOARCA discussed the Sequoyah project on April 20, at the TVA Sequoyah Nuclear Training Building.

If you have comments on the draft report, you have until May 12, 2016 to send them in. The best way to comment is through regulations.gov, using Docket ID NRC-2016-0074. You can also mail comments (referencing the Docket ID) to Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

If you submit comments in writing or in electronic form, they will be posted on the NRC website and on regulations.gov. The NRC will not edit or remove any identifying or contact information; please don’t include any information you wish to keep private.

We’ve also developed a public communications brochure to help explain the SOARCA project to a wider audience of stakeholders using plain language and more illustrations.

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