*The present generation of nuclear reactors has had a good safety record, with the major exception of the Chernobyl-type reactors. Outside the former Soviet Union, about 8500 reactor-years of commercial nuclear power-plant operation have been realized until now, with no accident involving a large external release of radioactivity and only one accident with fuel melting: the 1979 accident at Three Mile Island (TMI).
These numbers suggest that the risk of an accident with fuel damage has averaged approximately 10 E-4 per reactor-year, corresponding under common assumptions to a large external release of radioactivity at a rate of 10 E-5 per reactor-year. But this performance would not suffice for a world with ~4000 reactors, because the expectation would then be for a TMI-scale nuclear accident every several years.
However, changes in equipment and operating procedures since TMI suggest considerably improved safety. The likelihood of an accident that proceeds all the way to core damage can be estimated by analyzing data on the occurrence of individual system malfunctions (precursor events). Such analyses of actual U.S. reactor performance show a drop of roughly a factor of 100 in the inferred core damage probability, when comparing the 1994-1998 record with that for the pre-TMI period of 1974-1978 (See T. E. Murley, Nucl. Saf. 31:1 (1990); T. E. Murley, "MIT safety course" (July 1999); and W. D. Travers, SECY-99-289, NRC, 1999).
There are also well-developed designs for a next generation of reactors, which promise still greater safety. Of these, the advanced boiling water reactor (ABWR) is the first to have been ordered, with two now operating in Japan. The probability of core damage is estimated by the ABWR designers to be 2 x 10 E-7 per reactor-year and by the staff of the U.S. Nuclear Regulatory Commission to be "on the order of 10 E-6 or less" if the plant is built and operated as specified (See The ABWR General Plant Description, GE Nuclear Energy, 1999; and NRC, "Final safety evaluation report related to the certification of the advanced boiling water reactor design," main report, NUREG 1503, vol. 1, 1994, p. 19-6).
Research is under way to design a further generation of advanced reactors that differ from previous generations in that they make greater use of passive safety systems, based on simple physical laws. Because they will require no immediate operator intervention in the case of malfunction, they are expected to operate with extremely low levels of risk to the public.
[Source: William C. Sailor et al., "A Nuclear Solution to Climate Change?", Science 288:1177, May 19, 2000]
* Probability is expressed as a number between 0 and 1 (0 to 100 percent likelihood of the occurrence of an event). The notation 3 × 10-6 can be read 0.000003, which means that there are three chances in 1,000,000 that the associated result (for example, a fatal cancer) will occur in the period covered by the analysis.
Nuke Protect - Potassium Iodide / Selenium Blend
90 cap
* The International Nuclear Event Scale (INES) was developed jointly by the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organization for Economic Cooperation and Development.
* The NRC's safety goals are described in the Safety Goal Policy Statement, which was released in August 1986 (See 51 Federal Register 30028). The development of the Policy Statement began not long after the Three Mile Island accident, and was a first attempt by the Commission to come explicitly to grips with the integration of the quantitative assessment of risk into the regulatory system. A few years earlier, the NRC had funded the Reactor Safety Study, known as WASH-1400 and perhaps even better known as the Rasmussen study. That study represented the first use of probabilistic techniques to estimate the frequency of accidents and their ultimate consequences, thereby allowing a quantitative estimate of risk. The primary issue for the NRC in developing safety goals was to use these techniques to help articulate a level of acceptable risk -- in other words, to define "how safe is safe enough."
The Commission established two goals that are stated in terms of public health risk -- one addressing individual risk and the other addressing societal risk. The risk to an individual is based on the potential for death resulting directly from a reactor accident -- that is, a prompt fatality. The societal risk is stated in terms of nuclear power plant operations, as opposed to accidents alone, and addresses the long-term impact on those living near the plant. In both cases, the Commission based its acceptable level of risk on a comparison with other types of risk encountered by individuals and by society from other causes, applying the rule that the consequences of nuclear power plant operation should not result in significant additional risks to life and health. The goals were expressed in qualitative terms, perhaps so the philosophy could be understood by all.
In both cases, however, the Commission also expressed the qualitative goals for the safety of nuclear power plants in terms of individual and societal "quantitative health objectives" or "QHOs." These were established at one one-thousandth of the risk arising from other causes presenting the same type of risk.
It is important to note that the QHOs per se have never been directly reflected in the NRC's regulations, but were promulgated to provide guidance as to the level of "public protection which nuclear plant designers and operators should strive to achieve." They were also meant to provide guidance to the NRC staff to use in the regulatory decision-making process. However, the Commission was clear that the safety goals were not meant "to serve as a sole basis for licensing decisions." In fact, the Commission disclaimed an intent to use the goals in making plant-specific regulatory decisions.
While the safety goals provided a metric to address the question of "how safe is safe enough," practical implementation of the Commission's guidance proved to be difficult. This was the result of the large uncertainties involved in calculation of risk in the mathematical sense of probability times consequences. As a result, the NRC staff began looking for other metrics to use as surrogates for the QHOs in regulatory decision-making.
In 1990, the Commission provided additional guidance to the staff regarding the Safety Goals, endorsing surrogate objectives concerning the frequency of core damage accidents and large releases of radioactivity (see Staff Requirements Memorandum on SECY-89-102, ÒImplementation of the Safety Goals,Ó June 15, 1990). The numerical value of one-in-ten-thousand for core damage frequency (CDF) was cited as a "very useful subsidiary benchmark...." In addition, a conditional containment failure probability of one-tenth was approved for application to evolutionary light water reactor designs. This resulted in a large release frequency of one in one-hundred-thousand, since containment failure is necessary for a large release to occur. These values have evolved into the "benchmark" values of 10-4 for CDF and 10-5 for large early release frequency (LERF), as discussed in Regulatory Guide 1.174 for use in risk-informed regulatory decision-making.
The application of these goals as an underpinning of the regulatory system has evolved over time from the philosophical to the practical. Now they serve as the basis for many regulatory initiatives. An early example of explicit consideration of risk in a regulation is the NRC's Backfit Rule, originally issued in 1988 (10CFR50.109). But we have moved on to a much more comprehensive application of risk in our regulations, as most in this audience are undoubtedly aware. The aim, of course, is to use risk as the tool for dissecting and reforming our regulatory system so that the NRC focuses on risk-significant activities, thereby both enhancing safety and reducing needless regulatory burden. In implementing this approach we still adhere to many of the basic concepts discussed in the original Safety Goal Policy Statement, such as the use of risk as only one factor among many in making regulatory decisions.
In short, the development of a practical application of the safety goals and the ancillary tool of PRAs have taken many years, but they have growing significance as the foundation for the NRC's work. That being said, there are challenges that must be confronted. Let me mention a few.
First, we recognize that risk, at least for the foreseeable future, will be only one factor that can guide regulatory decisions. In this connection, I want to emphasize the relationship of risk insights to defense in depth. If one had complete confidence in the accuracy of PRAs, one might conclude that defense in depth could be ignored if the risk were sufficiently low. But the Commission is not prepared to jettison the deterministic processes and the defense-in-depth philosophy that are integral parts of the regulatory system. Defense in depth is to be applied at a high level -- that is, to require both prevention and mitigation -- and then as well at lower levels to compensate for uncertainty. There has been much discussion within the NRC and with the Advisory Committee on Reactor Safeguards as to how defense in depth should be incorporated into a risk-informed regulatory approach and this discussion will no doubt continue.
Second, we may need to reconsider the subsidiary objectives. Although the CDF and LERF goals have proven to be quite useful and valuable in implementing the Commission's safety philosophy, they do tend to skew the focus of attention to severe reactor accidents. While it is unquestionably true that the societal risk from nuclear power is dominated by accidents that have low frequencies and high consequences, the perception of risk on the part of the public is influenced by events of low consequence in terms of radioactive releases, but which have much higher frequencies. This is illustrated, for example, by the reaction following the steam generator tube failure at the Indian Point 2 station in February 2000. The event was widely reported to have involved a release of radioactivity to the environment, although the release was determined to be so slight that the monitoring equipment around the plant could not detect it. Nonetheless, there was an intense public reaction to the event, which continued for several months and has only recently begun to subside. The safety strategy should address plant operations, not just accidents, and should consider the full spectrum of events on a frequency/consequence continuum rather than just extreme events. That is, even a low-consequence event is of concern if its frequency of occurrence is high.
Finally, while we wrestle with incorporating risk insights into regulatory processes, we face other practical challenges as well. As you know, in the past few months there has strong interest in exploring new construction. We fully expect to see aggressive use of PRAs in connection with new reactor designs as means of satisfying the Commission's goal of assuring that advanced reactor concepts meet or exceed the level of safety provided by the current generation of reactors. Of course, PRAs are now used in the design process itself, to pinpoint and correct vulnerabilities based on risk insights. In this connection, we are grappling with the possibility that we may have to develop a new regulatory system that, unlike the focus of the current rules on light water reactors, will be independent of technology. The foundation of any such system must inevitably include compliance with the safety goals-or their subsidiary objectives-as demonstrated by PRAs.
Despite these many challenges, the NRC is clearly moving in the direction of greater reliance on quantitative tools and goals -- thereby achieving the promise first signaled by the Commission's Safety Goals nearly 15 years ago. I believe the next 15 years will see accelerated progress.
[Source: Dr. Richard A. Meserve (NRC Chairman), "The Evolution Of Safety Goals And Their Connection To Safety Culture", speech at Topical Meeting On Safety Goals And Safety Culture, Milwaukee, Wisconsin, June 18, 2001]
March 7, 2008
February 27, 2008
This is from the front page of today's The Arizona Republic.
See full text of this story, and many public comments, via the web version of this article.
May 28, 2007
"Having reactors close to a city can be worse than storing nuclear bombs in it"
That's how Dr Pervez Hoodbhoy, cited by Pakistani newspaper Dawn as a renowned professor of nuclear physics, is quoted in an article yesterday. "While a reactor cannot explode", he explains, "even one as small as a 200-megawatt reactor, after a year of operation, contains more radioactive cesium, strontium and iodine than the amounts produced in all the nuclear weapons tests ever conducted." He emphasizes that these deadly materials can be released if the containment vessel of the reactor is somehow breached. "A Chernobyl-like accident, leading to a loss of coolant, is the nightmare that dogs nuclear plants everywhere." The article cites a Dr Nayyar as another leading authority on the subject who seconds Dr Hoodbhoy regarding the devastating perils of nuclear plants: "In their normal functioning, they do not produce harmful radiation. But in an accident the most lethal radioactivity is released into the environment which kills in thousands and not only instantaneously but over decades." The article notes that spokesperson for Pakistan Atomic Energy Commission vehemently refutes the views of the two scientists quoted above. The spokesperson, Mohammed Tariq Rasheed, is quoted thusly: "In all these decades, KANUPP has not experienced any event which could be considered a potential threat to either plant workers or the public."
[Ref: Reemi Abbasi, "New nuclear power plant sets alarm bells ringing", article on Dawn website, May 27, 2007, distributed by BBC Monitoring under headline "Pakistan to establish three nuclear power plants near Karachi", May 28, 2007]
March 4, 2007
February 27, 2007
Cook request (to allow reactor trip system and ESFA system instrumentation to be tested in bypass) in danger of failing acceptance review
DC Cook wants to allow reactor trip system (RTS) and engineered safety feature actuation system (ESFAS) instrumentation to be tested in bypass. The plant submitted an application about this to NRC, requesting amendment to tech specs, on September 15, 2006. NRC staff reviewed the application, and determined that insufficient information was provided. The following details are from an email from NRC to Cook, proposing that a conference call be scheduled for Cook to discuss the issues with NRC reviewer Hukam Garg.
The most notable omission, NRC told Cook, was the lack of information on the changes to the RTS and ESFAS systems to include the bypass capability. Page 5 of the submittal, titled Technical Analysis, states that: "Hardware changes necessary to be made to the NIS and Foxboro analog/digital protection system to facilitate testing in bypass will be implemented in accordance with 10 CFR 50.59." Standard Review Plan, Chapter 7, states that the review should include an evaluation of the protection system design against the requirements of ANSI/IEEE Std 279, or Reg. Guide 1.153, which endorses IEEE Std. 603, depending upon the applicant/licensees commitment regarding design criteria. The RTS review should address all topics identified as applicable by Table 7-1. Major design considerations that should be emphasized in the review of the RTS are:
* Design basis
* Single failure criterion
* Quality of components and modules
* Independence
* Defense-in-depth and diversity
* System testing and inoperable surveillance
* Use of digital systems
The same requirements are also applicable to ESFAS instrumentation. NRC noted that the application does not provide information showing how the bypass system capability meets all of the above guidance. Also, the application does not state that Cook has done the failure mode and effects analysis to determine if the failure of bypass system will not have any impact on accident analyses, or would not create new potential accidents. The current DC Cook licensing basis does not include any mass addition transients or accidents. As such, a change in the ESFAS or RTS system that creates the possibility for an inadvertent injection of water into the reactor coolant system (such as the inadvertent start of a charging pump), would require the licensing basis to include mass addition transients in the DC Cook licensing basis. If the licensing basis did not include these transients, neither the licensee nor the NRC could conclude that the proposed change poses no safety concern. There was no discussion on the brand and type of system, or the method used for the qualification of this system.
The email from NRC to DC Cook, dated February 15, 2007, is available as ADAMS ACN ML070460159.
February 24, 2007
Thorp plant personnel lacked 'questioning attitude', which allowed leak to continue for months after contamination detected
The Thorp facility at Sellafield was shut down in April 2005 after 83,000 litres of acid containing uranium and plutonium escaped from a broken pipe. Operator British Nuclear Group was fined £500,000 last year after it pleaded guilty to breaching aspects of the Nuclear Installations Act 1965. The Health and Safety Executive (HSE) has now issued a 28-page report on the matter, with a total of 55 recommendations and actions for company improvements. The report said a number of failures in management meant the leak remained undetected for eight months. It highlighted a lack of a "questioning attitude" or "challenge culture" at the company. The review said: "An underlying cause was the culture within the plant that condoned the ignoring of alarms, the non-compliance with some key operating instructions, and safety-related equipment, which was not kept in effective working order for some time, so this became the norm." The first indication of a leak was on August 24, 2004 when 50 grams of uranium was detected following a sample test. But the full extent of the leak was finally uncovered on April 14 and Thorp was shut down four days later and remains closed.
[Source: BBC News, " Bosses slammed over nuclear leak", Feb 24, 2007]
February 22, 2007
N-plants - Technology Neutral Licensing Framework
On March 8, 2007, the Advisory Committee on Reactor Safeguards (ACRS) is scheduled to hear presentations by and hold discussions with representatives of the NRC staff regarding the Technology Neutral Licensing Framework, and the Commission request in the November 8, 2006 Staff Requirements Memorandum that the ACRS provide its views to the Commission with respect to the staffÕs work on Technology Neutral Licensing Framework with the focus on ensuring the value of such an approach versus the development of a licensing framework for specific designs. This session is slated for 3:45 pm to 5:15 pm. The meeting is to be held at NRC's Rockville Maryland headquarters -- building TWFN, in conference room T-2B3.
[Source: Meeting Notice for Federal Register, 540th Meeting of the Advisory Committee on Reactor Safeguards -- ML070400481]
October 25, 2006
*
Nuclear industry has good record
Toronto Star ( Canada)
While it might be convenient for Amory Lovins to tar the entire nuclear industry with his comments about Areva's troubled project in Finland and elsewhere, a ...
*
Nuclear Power - Saviour or Sinner?
Earthtimes.org
In his speech at the 15th Pacific Basin Nuclear Conference in Sydney, Australia in mid-October, John Ritch, director-general of the World Nuclear Association ...
*
Reactors may face suspension: nat'l assembly
hani.co.kr (South Korea)
Two South Korean nuclear reactors are on the brink of being suspended for 6 months and one year, respectively, after the process to extend their usage period ...
*
Nine of Sweden's 10 nuclear reactors back online
Monsters and Critics.com (UK)
Stockholm - Nine of Sweden's 10 nuclear reactors, including four that were taken offline last July over flaws in their backup systems, were Wednesday back ...
*
Two nuclear reactors are closed
BBC News
* [2006-03-01] Fly a jet into our N-plant, it WON'T hurt it
Mark Hookham, icNorthWales
February 7, 2006
Tornado missile hazard
Duke is reviewing "Draft Regulatory Guide DG-1143, Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants", which has recently been made available for public comment. Duke said that it will reach a decision by March 1, 2006, regarding possible use of DG-1143. Oconee will be using Fiber Reinforced Polymer (FRP) technology to strengthen selected masonry walls against the effects of tornado wind and differential pressure loads. A license amendment request will be submitted by June 1, 2006, for use of FRP technology. NRC staff said that the use of this technology will be a first-time review for the NRC staff. Duke is using TORMIS, a computer model for analyzing tornado-induced missiles, to help in evaluating risk. The meeting summary prepared by NRC is available on ADAMS as ML060580176.
May 31, 2005
* Wolf Creek - pressurizer piping and valves okayed to remain as Class 2
April 16, 2005
*
US panel: Fuel pool attack could trigger zirconium fire
Thecla Fabian, Nuclear Engineering
A terrorist attack on the spent fuel pools at some US nuclear plants could trigger a high-temperature zirconium fire that would lead to a significant release of radioactivity, though not on the scale of the 1986 Chernobyl explosion, concluded a blue-ribbon panel of scientists assembled by the National Research Council of the US National Academies. The unclassified academies' report, Safety and Security of Commercial Spent Nuclear Fuel: Public Report, contains all the findings and recommendations of the classified report, but with all national security and safeguards information removed, said Louis Lanzerotti, who chaired the 15-member expert panel pulled together by the academies' Board on Radioactive Waste Management in response to a mandate from congress. The panel spent six months gathering and analyzing data, and meeting with regulators, nuclear industry experts, and independent scientists. Lanzerotti is a geophysics expert consulting for Bell Laboratories and Lucent Technologies and a distinguished professor for solar-terrestrial research at the New Jersey Institute of Technology. Other panel members included a former NRC division director in nuclear materials management, and experts in the behaviour of nuclear materials at high temperatures, penetration mechanics, ballistics and weapons technology, health physics, actinide chemistry, heat transfer, thermal hydraulics, structural engineering and terrorism. The panel unanimously concluded that an attack that caused either partial or complete draining of a plant's spent fuel pool might be capable of starting a high-temperature fuel cladding fire that could lead to the Òrelease of large quantities of radioactive material into the environment.Ó The risk depends on a number of factors, including the type of attack, the design of the fuel pool, and the configuration of the fuel in the pool. The panel recommended two immediate measures that could reduce the potential for fuel cladding fires: (1) The reconfiguration of the position of fuel assemblies in the pools to more evenly distribute decay heat loads; and (2) Making provisions to cool the fuel with water spray systems that could continue to operate even after a pool or the building housing it is damaged. The panel noted that water spray systems might not be needed at plants where the fuel pools are located below ground or otherwise protected.
... Pools are and will continue to be needed at all nuclear plants for the foreseeable future, the panel stressed, noting that fuel newly removed from the reactor needs about five years cooling time in a water pool before it can be loaded into casks. For older fuel, however, dry storage has two advantages. It is a passive system that relies on air circulation for cooling, and it divides the spent fuel inventory into a number of individual, robust containers that contain only a small amount of the total inventory. Different dry cask systems available on the US market differed only slightly in robustness under different terrorist attack scenarios, the panel found.
April 15, 2005
* ANO gets permission to operate with some containment isolation valves open
October 22, 2004
Materials Performance In Naval Reactors Program
The Naval Reactors (NR) program has published extensive information on reactor plant and secondary plant materials performance in scientific literature where security classification requirements permit. These publications help share program experience with the technical community. Naval plants operate at different temperatures and pressures than most commercial plants and the flow velocities and chemistry conditions may also be different. The applicability of information in the cited NR program publications to commercial plants must be carefully considered in view of these differences. In response to request from Nuclear Energy Institute, NRC forwarded a list of publicly available information (a bibliography of NR program prime contractor publications) that NR is able to share, on the following topics: pressure vessel steel, low strength nickel-based alloys, high strength nickel-based alloys, stainless steel, core materials, other, and secondary plant materials. NRC staff contact is Margie Kotzalas 301-415-2737 (NRC/NRR Project Licensing and Technical Analysis).
[Source: Brian W. Sheron (NRC/NRR Associate Director for Project Licensing and Technical Analysis), "Materials Performance In Naval Reactors Program", letter to Alex Marion (Engineering Director, Nuclear Energy Institute), October 13, 2004 -- ML042780195]
* Browns Ferry - pipes carrying > 200¡F fluids will be 316 SS NG material, not susceptible to IGSCC
Sept 2004
Will next bign-accident be on night shift too? Is it mere coincidence that TMI accident began shortly after 4 am and Chernobyl disaster began at 1:23 am? A study by Kenneth Fortson suggests not. He argues that there are inherent physiological implications of late-night work that make off-hours jobs more hazardous than daytime jobs. Mr. Fortson, a Ph.D. candidate in Economics at Princeton, bases his conclusion in part on Texas data on workers' compensation claims. [Citation: Kenneth N. Fortson, The diurnal pattern of on-the-job injuries, Monthly Labor Review 127(9):18-25, September 2004 Click here for pdf of Mr. Fortson's paper.
September 3, 2004
* Oyster Creek - reactor power exceeded licensed limit for three days
* Brunswick's EDG model susceptible to catastrophic piston defect
* Grand Gulf - evaluation of effect of degraded pipe supports
* St. Lucie - Hurricane Francis warning triggers e-plan
* Turkey Point - Hurricane Francis warning triggers e-plan
August 21, 2004
* Japan - followup on why the Mihama-3 steam pipe wasn't previously inspected
August 16, 2004
* Columbia - reactor manually scrammed during startup
* River Bend - reactor scrammed upon partial loss of offsite power
* Susquehanna-1 - SPDS wasn't working for 11 hours
* Susquehanna-2 - SPDS wasn't working for almost 10 hours
July 29, 2004
Callaway's analysis of applying leak-before-break (LBB) in analysis of proposed steam generator sludge lance platforms
Callaway has requested use of the LBB methodology for the accumulator, pressurizer surge, and residual heat removal (RHR) lines to exclude the dynamic effects associated with large reactor coolant system branch line ruptures. NRC approved use of LBB for accumulator and RHR lines, but requested more information about pressurizer surge line. Calloway has committed to providing the information by November 30, 2004. For more info, see Callaway withdraws request to apply American Society of Civil Engineers (ASCE) 4-86 "100-40-40" load combination methodology for combining components of seismic response loads.
May 19, 2004
* PWR safety experiments planned for German primary coolant loop test facility
May 18, 2004
March 29, 2004
* Vogtle-1 - manual reactor trip due to feedwater pump overspeed which could not be controlled
* Cooper - both EDGs found inoperable due to closed valve between two fuel oil tanks
March 27, 2004
* Irradiated metal - weldability questions from NRC to BWR Vessel Internals Project
March 26, 2004
* BWR - bigger wetwells in Mark III containments minimize pressure during DBA-LOCA
* Sump strainer clogging - ACRS issues of interest re draft Rev. 3 of Reg Guide 1.82
March 24, 2004
* Susquehanna-1 - reactor vessel penetration crack discovered
March 18, 2004
* Peach Bottom-3 - inoperable HPCI - 14-day LCO
* Wolf Creek - floor drains in ESF switchgear rooms may be too small to handle rupture of fire main
March 17, 2004
March 13, 2004
* Restart after August 2003 blackout took 4-9 days for the 9 n-plants; only 1 NOED
* Grid stability increases n-plant safety margin
* Fire Protection - slow progress in resolving n-plant fire protection issues
* Security - consumes 40% of Commissioner Merrifield's time lately, up from 5% before 9-11
* Safety Conscious Work Environment initiatives - innovative approaches to old issues
March 12, 2004
* N-plant safety concerns: slight increase in # of scrams and other reportable events, sez NRR chief
* Davis-Besse faces minimum of 3-5 years of heightened scrutiny by NRC
* Davis-Besse - two NRC inspectors on duty each shift until further notice
* Davis-Besse - independent audits required for at least 5 years as condition for restart
* Davis-Besse - maintenance outage scheduled for years before refueling date
* Davis-Besse's close call 'should not have been possible', sez NRC Chairman Diaz
* US nuclear industry can be proud of 50-year safety record, sez NRC chairman Diaz
* Nuclear industry 'requires constant vigilance', sez Davis-Besse owner
* North Anna - vessel head replacement example of proactive approach
March 9, 2004
March 8, 2004
* Susquehanna - outage work involves deenergizing safety parameter display system
March 2, 2004
* Pilgrim - Valve failure due to obstruction by loose fasteners for cartridge mounting plate
Feb 28, 2004
February 27, 2004
* BWR - Potential Adverse Flow Effects from Power Uprates
February 24, 2004
* Oyster Creek violation - cable failures affect safety functions
* PWR reactor vessel head inspections - New NRC framework based on susceptibility and obstructions
December 31, 2003
* Defense-in-depth as it applies to spent fuel pools
December 12, 2003
* McGuire - reducing frequency of PORV block valve testing would improve safety
December 11, 2003
* Peach Bottom-2 - HPCI system Torus check valve found to allow reverse flow
December 10, 2003
November 29, 2003
* N-plant Probabilistic Risk Assessment - pilot applications planned
October 21, 2003
Reversing 20-year trend, fuel failures up in last two years
In comparison to where we were 20 years ago, the performance of fuel is greatly improved. The number of light-water-reactor fuel failures has steadily declined during this time. I believe a strong contributor to this improved performance is the increased market competition between the current fuel vendors here in the United States: Framatome-ANP, Westinghouse, and Global Nuclear Fuel/General Electric. Increased competition has forced these companies to review their manufacturing processes and focus on process improvements in the area of new technologies to identify issues such as, chipped fuel pellets and flawed tubes before they are put in service. In addition, vendors are focusing on performance improvements in fuel and cladding design, and other areas to support higher fuel burnups, longer operating cycles, and power uprates.
Yet, despite these successes, the number of fuel failures in the past two years has noticeably increased. Fuel issues are back on the radar screen of many plant operators and calls for improved reliability are common. Thus, the fuel vendors are left with balancing their resources to remain competitive, but still perform the needed research to safely advance their designs.
From where I sit, it appears that industry may be overly focused on the economic issues and may be pushing the fuel too hard. I get concerned when I hear industry folks question whether fuel manufacturers have budgeted sufficient research dollars toward meeting the demand of the new, more aggressive operating environment. From my perspective, increased burnup, longer operating cycles and power uprates are key drivers for the fuel performance desired by our licensees. The fuel environment is going to be more challenging but, as a safety regulator, we need to be assured that the plants can continue to operate safely under these new conditions. To continue to insist on rock bottom fuel prices at the expense of debilitating and costly fuel failures is penny-wise and pound foolish. The industry must leverage its overall experience and utilize initiatives such as the Electric Power Research Institute (EPRI) Robust Fuel Program to effectively deal with fuel reliability.
For our part, the NRC developed a research program to confirm the current fuel burnup limit of 62 gigawatt days per metric ton and to develop a strategy for assessing future requests for burnup extensions beyond the current NRC limit to ensure the adequate protection of public health and safety at our operating reactors. Utilizing a variety of U. S. and international facilities, the NRC research effort is appropriately focused on demonstrating that recent increases in energy output for new cladding alloys can meet our regulatory expectations for postulated accidents. Nonetheless, given the recent spike in fuel failures, I think that both the NRC and industry need to consider additional research to determine how we can get a better handle on new designs and materials that can reverse the recent increase in fuel failures.
[Source: Jeffrey S. Merrifield (NRC Commissioner), "Practical Not Perfect", keynote speech at Nuclear Safety Research Conference Washington, D.C., October 21, 2003 (ACN ML032940500)]
[Note: These comments were the basis for front page story in Oct 27 issue of trade newsletter Platts NuclearFuel, titled "Industry may be too focused on fuel economics, Merrifield says".]
September 16, 2003
Risk-Informed Rules for Control of Combustible Gases (10 CFR 50.44)
On September 16, 2003, the Commission issued the final risk-informed rule for control of combustible gases inside reactor containment buildings (68 FR 54123). This final rule is based on information from extensive NRC-sponsored research over the last 20 years to understand the generation and behavior of combustible gases during nuclear power reactor accidents. The changes will allow power reactor licensees to eliminate hydrogen recombiners and to relax containment hydrogen and oxygen monitoring requirements based on the low risk significance of that equipment. On September 25, 2003, the Commission issued a model application for revising nuclear power reactor technical specifications in accordance with the new rule. This model application will make it easier for licensees to submit license amendments to reduce these requirements.
[Source: "Monthly Status Report On The Licensing Activities And Regulatory Duties Of The United States Nuclear Regulatory Commission, September 2003", enclosure to Nils J. Diaz (Chairman, NRC), letter to Sen. George V. Voinovich (Chairman, Subcommittee on Clean Air, Climate Change and Nuclear Safety, Senate Committee on Environment and Public Works), November 25, 2003, ACN ML032900350]
September 9, 2003
Risk-informed 50.69 option no good if full scope PRA required, NMC tells NRC
Nuclear Management Company (NMC) filed comments with USNRC about proposed 10 CFR 50.69 ("Risk-Informed Categorization And Treatment of Structures, Systems and Components for Nuclear Power Reactors"). The company notes that the Federal Register notice (68 FR 26511, May 16, 2003) raises the question of whether an NRC- reviewed full scope, all modes PRA should be a prerequisite for Implementation of this rule. "Such a requirement", the company told NRC, "Is neither technically necessary, nor feasible and would eliminate 10 CFR 50.69 as a viable option for NMC." NMC is the licensed operator of Duane Arnold, Kewaunee, Monticello, Palisades, Point Beach, and Prairie Island plants.
Two other organizations' comments on the proposed rule were released to public by NRC today, and nuclear.com found their difference in characterizing the purpose of the proposed rule interesting. Tennessee Valley Authority emphasized safety -- "substantially enhancing the safety focus, coherence, and efficiency of current regulations governing nuclear power plant operations." The Nuclear Utility Group on Equipment Qualification (NUGEQ) emphasized "reduc[ing] unnecessary regulatory burden" and "relax[ing] or eliminat[ing] certain special treatment requirements". NUGEQ members operate some 90 nuclear plants in US and Canada.
The comment letters are reproduced here on nuclear.com as pdf files: NMC, TVA, and NUGEQ.
September 1, 2003
Braidwood - lack of exercise hardened grease in (most) circuit breakers
Braidwood has a lot of circuit breakers, including about 3,000 of a "molded case" design known as Westinghouse adjustable magnetic HFB style, many of which are used on safety-related systems. The plant established a formalized testing program for these breakers some years ago, and identified that many of the breakers were failing the tests. In fact, of the 90 tested between June 2002 and June 2003, more than half failed. The problem appears to be that the grease inside the circuit breakers hardens over time. Westinghouse recommends cycling the breaker every month to move the grease around. This helps extend the operational life of the breaker. If left undisturbed, the grease will become quite like cement after as few as six or so years. Most of Braidwood's MCCBs were manufactured in 1970s-1980s. A recent NRC fire protection team evaluated the problem, and the plant's follow-up. The inspectors concluded that the plant should have had the grease problem under control long ago. NRC put out an information notice ten years ago describing the importance of exercising these breakers. The violation was categorized by NRC as a Non-Cited Violation. For more on this story, click here.
Source: Z. Falevits (Senior Reactor Inspector, NRC Region III), et al., Braidwood inspection report 50-456-2003-5, August 21, 2003
August 29, 2003
Undersized air-operated valve actuators has been a common problem in nuclear plants, sez Davis-Besse
Here's part of the section titled "Apparent Cause of Occurrence" from the Licensee Event Report (LER 50-346-2003-001-01) describing the finding that eight valves a Davis-Besse were not capable of performing their intended safety functions for all required conditions:
"Lessons learned from the nuclear power industry's motor-operated and air-operated valve programs indicate that AOV performance can be enhanced by improvements in valve and actuator sizing, setting, testing, and maintenance. It was found that during the original procurement cycle, many AOV actuators were undersized. This was a result of vendors being provided with inaccurate system conditions in combination with less than conservative sizing methodology used at the time, and a lack of formal calculations supporting the design basis and appropriate settings for AOV actuators. There was also the practice of sizing AOV actuators with minimum built-in margin. Similar analytical deficiencies resulted in the design of the air accumulators, used to provide a source of motive power in the event of a loss of non-safety related instrument air, not being sufficient to ensure the valves would perform their intended safety function under all design conditions. This apparent cause applies to valve CC1495."
Other causes were identified for the other seven valves. The full LER is available as a pdf.
August 26, 2003
* Calvert Cliffs - containment sump strainer review
August 22, 2003
Perry - breaker problems mistakenly assessed as inapplicable to some types
A non-cited violation at Perry involved condition of 5kv cell switches. Rather than closing out the matter, the plant's corrective actions prompted NRC to find additional violation -- a mistaken assessment of the similar equipment -- auxiliary and 15kv breakers -- affected by the root cause identified. This finding was considered "more than minor because the failure to adequately identify extent of condition and take corrective actions to address degraded conditions could reasonably be viewed as a precursor to a significant event." Although the 15kv breakers are categorized as non-safety-related, the inspector noted "the risk significant loads which could have been adversely impacted included the 13.8kV bus L10 which supplies the Class 1E 4.16kV buses, the motor driven feedpump, and the recirculation pump motor fast speed supply breaker". The inspection report suggests that Perry should have been more curious as to why some 20% of the auxiliary switches inspected (by the same switch expert who inspected the cell switches) were found to require adjustment. Had this opportunity not been missed, the misanalysis of the 15kv breaker problem might have been identified. [full report, in pdf]
August 18, 2003
MSIV LLRT approach - potential generic issue at BWRs
NRC inspectors questioned Columbia Generating Station's practice of using instrument air to close main steam isolation valves (MSIVs) before local leak rate testing (LLRT). The instrument air system provides more pressure than the safety-related air accumulators that serve as design basis for MSIV operation, so the seal tightness conditions being tested weren't the same conditions desired to be tested. Calls to five other BWRs revealed that none of them actually tested the design basis conditions. Columbia's MSIVs did pass proper test when performed. For more info, see Columbia - MSIV LLRT had never been done right, and error may be pervasive at BWRs.
* Columbia - MSIV closure tests since 1989 used inadequate GE SIL instead of required ASME method [note: this is different than the LLRT story described above]
* Columbia - Loss of Shutdown Cooling events
August 15, 2003
* Fermi - potential unmonitored release path identified by NRC inspectors
August 6, 2003
* EU considering raising insurance coverage required of n-plants, from EU21-million to EU700-million
August 2, 2003
* STP-1 Restart Approved By NRC- July 31 Letter From Region IV's Dwight Chamberlain
* STP-1 Cracks Were in the 2,200 psi, 600-degree System
* South Texas Project - BMI Penetration Crack Repair Description
* South Texas Project - BMI Penetration Crack Cause Sugested: Fabrication Problem
* South Texas Project - Missing Metal Sample From BMI Penetration
July 26, 2003
July 22, 2003
V.C. Summer risk of core damage now considered much lower than estimated in original IPE
The baseline core damage frequency (CDF) estimated for V.C. Summer nuclear plant is approximately 5.6E-5 per year, and the baseline large early release frequency (LERF) is approximately 7.0E-7 per year. The CDF and LERF are based on the risk assessment for internally-initiated events only. The CDF represents a sizeable reduction from the original IPE CDF value of 2.0E-4 per year.
[Source: NUREG-1437, Suppl. 15, Appx. G - Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Virgil C. Summer Nuclear Station, Draft for Comment. ACN ML031900024. July 2003]
July 14, 2003
* Russia - Top n-regulator Yury Vishnevsky forced to retire - bad news for safety
July 13, 2003
* Callaway-1 10-yr pipe test ISI - UT approved in lieu of radiography
* RPV penetration weld repair method - evaluation of Palo Verde's request for exception from code
July 7, 2003
* New from NRC ADAMS system today: Nuclear Reactor Risk - Fact Sheet (NRC, June 2003)
July 2, 2003
* Monticello - door between motor control centers not latched for a few minutes
July 1, 2003
* MIT - operator asleep at reactor controls
* Perry - Unusual Event prompted by Ohio earthquake (3.4 on Richter scale)
* Columbia-2 - reactor trip after turbine trip (cause unknown)
* Brunswick-1 - RCS leakage rose from 0.63 gpm to 2.69 gpm, prompting shutdown
* Seabrook-2 - fire in containment during disassembly of containment dome
June 24, 2003
* Boric Acid Corrosion - EPRI embarks on 4-year research program
June 14, 2003
* Safety culture - everybody wants it; but nobody can define it, much less measure it
* Safety culture regs would allow NRC to remove "toxic leadership" at n-plants
* Safety culture regs not needed - it's the manifestations that are the violations
May 22, 2003
April 29, 2003
Debris and emergency sump operation at PWRs
NRC plans to soon issue an NRC Bulletin titled "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors". A draft was presented at a meeting with industry today, and a variety of comments were received. For example, industry urged the NRC staff to make a decision as soon as possible as to whether or not the Nuclear Energy InstituteÕs process for determining breach size in support of local debris generation following a design-basis LOCA is a reasonable approach since the process is critical to NEI in the development of its own guidance on the subject.
[Ref: John G. Lamb (NRC/NRR/DLPM project manager), "Meeting between the Nuclear Regulatory Commission Staff and Stakeholders concerning Generic Safety Issue 191, ÒAssessment of Debris Accumulation on PWR Sump PerformanceÓ (TAC No. MA6454)", April 29 meeting summary memo, May 15, 2003]
* Example of probabilistic analysis difference with defense-in-depth: ECCS could be simplified