ML041060662

G20040241 - Draft Gao Report - Nuclear Regulation: NRC Needs to More Aggressively and Comprehensively Resolve Issues Related to the Davis-Besse Nuclear Power Plant'S Shutdown

ML041060662
Person / Time
Site:	Davis Besse
Issue date:	05/05/2004
From:	Office of Nuclear Reactor Regulation
To:
Rosenberg S, NRR/ADPT, 415-3868
References
BL-01-002, CORR-04-0076, G20040241, GAO-04-415, LTR-04-0199, TAC MC2563
Download: ML041060662 (14)

Text

NRC Comments on Draft Report, GAO-04-415

1. The draft report does not speak to a key issue, the responsibility of licensees to provide complete and accurate information to the NRC. In carrying out its safety responsibilities, NRC must rely heavily on our licensees to provide us with complete and accurate information. Title 10 of the Code of Federal Regulations Section 50.9 requires that information provided to the NRC by a licensee be complete and accurate in all material respects. By not recognizing this explicitly and its role in this matter, the draft report conveys the expectation that the NRC staff should have known about the thick layer of boron on the reactor vessel head. The Davis-Besse Lessons Learned Task Force (LLTF),

which NRC formed to ensure that lessons from the Davis-Besse experience are learned and appropriately captured in the NRCs formal processes, noted that the information that FirstEnergy provided in response to Bulletin 2001-01, Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles was inconsistent with information identified by the task force. Further, the LLTF report stated that had this information been known in the fall of 2001, the NRC may have identified the vessel head penetration (VHP) nozzle leaks and reactor pressure vessel (RPV) head degradation a few months sooner than the March 2002 discovery by the licensee. See also the related information in response #2.

2. Page 7, first sentence of the last paragraph states: NRC should have but did not identify or prevent the vessel head corrosion at Davis-Besse because both its inspections at the plant and its assessments of the operators performance yielded inaccurate and incomplete information on plant safety conditions.

Response: This statement is misleading. We agree that our oversight program should have identified certain evolving plant conditions for regulatory follow-up. This was also Enclosure 1

identified in the report of the Davis-Besse Lessons LLTF. It is the responsibility of licensees to provide the NRC with complete and accurate information. In fact, Title 10 of the Code of Federal Regulations Section 50.9 requires that information provided to the NRC by a licensee be complete and accurate in all material respects. The report should clearly indicate that NRCs licensees are responsible for providing us with accurate and complete information. While the NRCs Davis-Besse LLTF concluded that the NRC, the Davis-Besse licensee (FirstEnergy), and the nuclear industry failed to adequately review, assess, and follow up on relevant operating experience, the LLTF also noted that the information that FirstEnergy provided in response to Bulletin 2001-01 was inconsistent with information identified by the task force. Further, the LLTF report stated that had this information been known in the fall of 2001, the NRC may have identified the vessel head penetration nozzle leaks and the reactor vessel head degradation a few months sooner than the March 2002 discovery by the licensee. As you are aware, there is an ongoing investigation by the Department of Justice regarding the completeness and accuracy of information that FirstEnergy provided to the NRC on the condition of Davis-Besse.

3. Page 8, last sentence states: Further, the risk estimate indicated that the likelihood of an accident occurring at Davis-Besse was greater than the level of risk generally accepted as being reasonable by NRC.

Response: This is incorrect. NRC staff explained to the GAO consultants that NRC guidance produces an estimate for the change in core damage frequency of 5x10-6 per year, not 5x10-5 as indicated in the GAO report. According to Regulatory Guide (RG) 1.174, for Davis-Besse, this estimate is within acceptable bounds. NRC specifically documented the acceptability of the estimate in the December 2002 assessment. Thus, the December 3, 2002, safety evaluation concluded that the delta core damage frequency was consistent with the guidelines of RG 1.174.

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4. Page 15 states that borax (i.e., sodium borate) is dissolved in the water. This is incorrect.

Please replace the word borax with boric acid crystals.

5. Page 18, first full paragraph states: NRC, in deciding on when FirstEnergy had to shutdown Davis-Besse for the inspection,...

Response: In addition, the staff relied upon information provided by the licensee regarding the condition of the vessel head (i.e., previous leakage and action taken to repair leaks and clean the vessel head).

6. Page 26, beginning on line 4, states: According to the NRC regional branch chiefwho supervised the staff responsible for overseeing FirstEnergys vessel head inspection activities during the 2000 refueling outagehe was unaware of the boric acid leakage issues at Davis-Besse, including its effects on the containment air coolers and the radiation monitor filters.

Response: According to the individual to whom this statement is attributed, the statement would be correct if the phrase, he was unaware...filters is changed to he was unaware that boric acid was found on the reactor vessel head during the outage.

7. Page 27, first sentence states: Similarly, NRC officials said that NRC headquarters had no systematic process for communicating information in a timely manner to its regions or on-site inspectors.

Response: If the information in question refers to issues of potential safety significance into which inspectors should look, then this statement is inaccurate. The systematic process for temporarily focusing inspection activity in a coordinated program-wide manner on high-priority issues is the Temporary Instruction (TI) process, which is well established within the NRC Inspection Manual and frequently used. The legitimate point 3 Enclosure 1

to be made is that until the Davis-Besse event, the NRC had not concluded that boric acid corrosion was a sufficient safety concern that reached the threshold for using the TI process.

8. Page 33, middle paragraph states: For example, concern over alloy 600 cracking led France, as a preventive measure, to develop plans for replacing all of its reactor vessel heads and installing removable insulation to better inspect for cracking.

Response: French regulators instituted requirements for an extensive, non-visual nondestructive examination inspection program for vessel head penetration nozzles that resulted in plant operators deciding, on the basis of economic considerations, to replace vessel heads in lieu of conducting such examinations.

9. Page 34, last paragraph states: If such small leakage can result in such extensive corrosion...

Response: Small leakage alone was not the cause of the corrosion. It was a combination of prolonged leakage in conjunction with allowing caked-on boron to remain on the vessel head.

10. Page 36, middle paragraph states: However, NRC decided that it could not order Davis-Besse to shut down on the basis of other plants cracked nozzles and identified leakage or the managers acknowledgment of a probable leak. Instead, it believed it needed more direct, or absolute, proof of a leak to order a shutdown.

Response: As discussed at the NRC-GAO exit conference, plant Technical Specifications, as well as many other NRC requirements and processes, provide a number of circumstances in which a plant shutdown would or could be required, including the existence of reactor coolant pressure boundary leakage while operating at power.

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Please note that there was no legal objections to the draft order and the stated basis for deciding to not issue the order was not an insufficient legal basis.

11. Page 36, last paragraph states: ...NRC does not have specific guidance for shutting down a plant when the plant may pose a risk to public health and safety even though it may be complying with NRC requirements.

Response: We disagree with this finding and with the reports related recommendation on Page 63 identifying the need for NRC to develop specific guidance and a well-defined process for deciding when to shut down a nuclear power plant. We believe our regulations, guidance, and processes that cover whether and when to shut down a plant are robust and do, in fact, provide sufficient guidance in the vast majority of situations.

Plant technical specifications, as well as many other NRC requirements and processes, provide a spectrum of conditions under which plant shutdown would be required. Plants have shut down numerous times in the past in accordance with NRC requirements. From time to time, however, a unique situation may present itself wherein sufficient information may not exist or the information available may not be sufficiently clear to apply existing rules and regulations definitively. In these unique instances, the NRCs most senior managers, after consultation with staff experts and given all of the information available at the time, will decide whether or not to require a plant shutdown. Risk information is used in accordance with RG 1.174. This process considers deterministic factors as well as probabilistic factors (i.e., risk information). We regard the combined use of deterministic and probabilistic factors to be a strength of our decisionmaking process.

12. Page 38, third paragraph states: At some point during this time, NRC staff also concluded that the first safety principle was probably not being met, although the basis for this conclusion is not known.

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Response: The report should clarify GAOs basis for this statement. NRC staff believed that the regulations were met.

13. Page 40, last paragraph states: However, NRC did not provide the assessment until a full year laterin December 2002. In addition, the December 2002 assessment, which includes a 4-page evaluation, does not fully explain how the safety principles were used or metother than by stating that if the likelihood of nozzle failure were judged to be small, then adequate protection would be ensured.

Response: The attachment to the December 3, 2002, letter is an 8-page evaluation, not 4 pages. We note this to make sure GAO is referring to the same document. The assessment addresses four of the five safety principles. In the NRCs December 2002 safety evaluation, the staff stated that the criterion related to compliance with the regulations was being met because the inspections performed by the licensee were in conformance with the ASME Code. In addition, the safety evaluation stated that Davis-Besse met the criterion related to defense-in-depth because all three barriers against release of radiation were intact and reliable; they met the margin criterion because even the largest circumferential cracks found in pressurized-water reactors had considerable margin to structural failure, and they met the low-risk impact criterion based on a comparison of delta core damage frequency estimates with the guidelines of RG 1.174.

The fifth safety principle, requiring a monitoring program, was not relevant to a decision that lasted only 6 weeks.

14. Page 42, first paragraph states: Multiplying these two numbers, NRC estimated that the potential for a nozzle to crack and cause a loss-of-coolant accident would increase the frequency of core damage at Davis-Besse by about 5.4x10-5 per year, or about 1 in 18,500 per year. Converting this frequency to a probability, NRC 6 Enclosure 1

calculated that the increase in probability of core damage was approximately 5.0x10-6, or 1 chance in 200,000. While NRC officials currently disagree that this was the number it used, this is the number that it included in its December 2002 assessment provided to FirstEnergy. Further, we found no evidence in the agencys records to support NRCs current assertion.

Response: These statements mischaracterize the facts. NRC estimated that the probability of nozzle cracking leading to a loss-of-coolant accident during the first 6 weeks in 2002 would increase the annual core damage frequency (CDF) by about 5.4x10-6 per year, or about 1 in 185,000 per year. The estimate of 5x10-5 was an intermediate step in our calculation. The estimate of 5x10-5 represents the change in CDF if Davis-Besse were allowed to operate for one year without shutting down for inspection of the vessel head.

Allowing Davis-Besse to continue to operate for one year was never a consideration.

Thus, multiplying by the fraction of time in one year under consideration (in this case 7 weeks) was the final step in the calculation of delta CDF. The confusion about the estimate NRC used in the decisionmaking process may be due to NRCs method of calculating delta CDF for plant conditions which do not persist for the entire year. If this final step (the fraction of the year the plant is allowed to operate) were not part of the calculation, then the risk estimate of allowing the licensee to continue to operate for 7 weeks, as compared to one year, would be the same. Logically, this does not make sense. Therefore, the estimate of 5x10-5 does not automatically convert to a probability, as GAOs statement implies. Because the period of operation under consideration was approximately 0.13 years, the annual average change in CDF was about 5x10-6 per year, and the increase in the probability of core damage was about 5x10-6 as well. NRC officials agree that 5x10-6 was the estimate used in the decisionmaking process and is the estimate provided in the December 2002 assessment.

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15. Page 42, second paragraph states: For example, the consultants concluded that NRCs estimate of risk was incorrectly too small, primarily because the calculation did not consider corrosion of the vessel head.

Response: An underlying assumption in any risk assessment is that you have complete and accurate information from the licensee. NRC staff was of the understanding that efforts had been made to remove boric acid accumulation from the vessel head during previous outages. For all six B&W plants that found signs of penetration leakage, the leakage manifested itself in the form of small amounts of dry boron crystals on the vessel head, which are not corrosive, and did not produce any corrosion on the vessel heads of these six B&W plants. Boron leaking onto a clean vessel head does not cause corrosion.

Therefore, corrosion this extensive was not anticipated at the time. Also, it is important to note that had Davis-Besse shut down on December 31, 2001, the same corrosion would have been found.

16. Page 43, first full paragraph discusses the experience at French nuclear power plants.

Response: The NRC staff was aware of the issue as illustrated in an internal memorandum dated December 15, 1994, from Brian Grimes to Charles Rossi.

17. Page 44, first full paragraph states: Third, NRCs analysis was inadequate because the risk estimates were higher than generally considered acceptable under NRC guidance. Despite PRAs [probabilistic risk assessments] important role in the decision, our consultants found that NRC did not follow its guidance for ensuring that the estimated risk was within levels acceptable to the agency. Page 45, first paragraph states: ...NRCs PRA estimate for Davis-Besse resulted in an increase in the frequency of core damage of 5.4x10-5 or 1 chance in about 18,500 per year was higher than the acceptable level.

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Response: This conclusion is not supported by the facts and it is misleading. The estimate referenced by GAO is an intermediate calculation in our process, and was not used, and should not be used, in the decisionmaking process. NRC staff explained to the GAO consultants that NRC guidance produces an estimate for the change in core damage frequency of 5x10-6 per year, not 5x10-5 as indicated in the GAO report.

According to RG 1.174, for Davis-Besse, this estimate is within acceptable bounds. NRC specifically documented the acceptability of the estimate in the December 2002 assessment. Thus, the December 3, 2002, safety evaluation concluded that the delta CDF was consistent with the guidelines of RG 1.174.

18. Page 45, first paragraph states: NRCs guidance for evaluating requests to relax NRC technical specifications suggests that a probability increase higher than 5x10-7 or 1 chance in 2 million is considered unacceptable for relaxing the specifications.

Thus, NRCs estimate would not be considered acceptable under this guidance.

Response: This criterion in RG 1.177 is not relevant to the Davis-Besse decision. It is confined to decisions on allowed outage times (AOT) for equipment, and is defined to avoid very high instantaneous risks (CDF > 10-3) for very short periods (5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />).

19. Page 46, first full paragraph states: Lastly, NRCs analysis was inadequate because the agency does not have clear guidance for how PRA estimates are to be used in the decision-making process.

Response: The NRCs process for risk-informed decision-making is considerably more robust than characterized in this section. Regulatory Guide 1.174 comprises 40 pages of guidance on how to use risk in decisions of this type, and it is backed up by equally detailed guidance for specific types of decisions such as technical specifications, in-service inspection programs, in-service testing, and quality assurance. The NRC has 9 Enclosure 1

amassed a great deal of experience in application of the guidance. Risk assessment is a tool to help better inform decisions that are based on engineering judgements.

20. Page 46, last paragraph states: It is not clear how NRC staff used the PRA risk estimate in the Davis-Besse decision-making process.

Response: The December 3, 2002, safety evaluation clearly states how the PRA estimate was used in the decisionmaking process; the estimate was compared with the guidelines of RG 1.174. The safety evaluation also points out that NRC staff who are expert in non-PRA disciplines such as probabilistic fracture mechanics, gave more weight to deterministic factors, such as the structural margin that remains in the nozzles with circumferential cracks. The NRC considers the combined use of deterministic and probabilistic factors to be a strength of our decisionmaking process.

21. Page 48, last paragraph states: ...NRC had made progress in implementing the recommendations, although some completion dates have slipped.

Response: The schedules for implementation of all high priority recommendations have not slipped. The implementation schedule for certain low or medium priority recommendations slip only in accordance with NRCs Planning, Budgeting and Performance Management (PBPM) process, which explicitly considers safety significance when making budget priority decisions.

22. Page 51, top of page, first full bullet states: One recommendation is directed at improving NRCs generic communications program. NRC is...

Response: We recommend re-wording this as follows: One recommendation is directed at improving follow up of licensee actions taken in response to NRC generic communications. A Temporary Instruction (Inspection Procedure) is currently being 10 Enclosure 1

developed to assess the effectiveness of licensee actions taken in response to generic communications. Additionally, improvements in the verification of effectiveness of generic communications are planned as a long-term change in the operating experience program.

23. Page 51, last paragraph states: ...NRCs revised inspection guidance for more thorough examinations of reactor vessel heads and nozzles, as well as new requirements for NRC oversight of licensees corrective action programs, will require at least an additional 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> of inspection per reactor per year.

Response: It is unclear where this number comes from, but the changes to the corrective action program procedure require only about 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> per reactor year for the trend review.

24. Page 53, first paragraph discusses the NRCs Office of the Inspector Generals (OIGs) findings on communications.

Response: The NRCs actions are not limited primarily to improving communication about boric acid corrosion and cracking. There are multiple task force recommendations, and other NRC initiatives, that are aimed at addressing the broader implications stemming from communication lapses noted by the task force and the OIG. For example, actions have been implemented to more effectively disseminate operating experience to end users, reenforce a questioning attitude in the inspection staff, and discuss Davis-Besse lessons learned at various forums.

NRCs initial response to the OIG did not directly address the broader actions we are taking to improve communications. Our response to the OIG only indirectly addressed this by discussing the operating experience program enhancements. Part of the 11 Enclosure 1

enhancements to the operating experience program is the expectations for improved communications. In addition, communication improvement initiatives with internal and external stakeholders are in progress to address shortcomings in this critical area. Our revised response to the OIG on this issue, dated April 19, 2004, is provided as Enclosure 2.

25. Page 53, second paragraph states: NRCs Davis-Besse task force did not make any recommendations to address two systemic problems: evaluating licensees commitment to safety and improving the agencys process for deciding on a shutdown.

Response: The LLTF did not make a recommendation for improving the agencys process for deciding on a shutdown. This area was not reviewed in detail by the task force because of coordination with the OIG. Moreover, the task force review efforts were focused on why the degradation cavity was not prevented. While related, the shutdown issue had little to do with the degradation cavity.

The task force made multiple recommendations aimed at enhancing NRCs capability to evaluate the licensees commitment to safety, by indirect means. Refer to task force recommendations: 3.2.5(1), 3.2.5(2), 3.3.2(2), 3.3.4(5), and Appendix F.

26. Page 54, last paragraph states: This problem identification and resolution inspection procedure is intended to assess the end-results of managements safety commitment rather than the commitment itself.

Response: This statement is inaccurate. Regarding its accuracy, the PI&R inspection procedure (IP 71152) actually has six stated inspection objectives (refer to section 71152-

01) including: (1) provide for early warning of potential performance issues that could 12 Enclosure 1

result in crossing threshold in the action matrix and (2) to provide insights into whether licensees have established a safety-conscious work environment. Using this IP, inspectors seek factual evidence of the licensees assumed commitment to safety (by reviewing their identification and correction of actual problems). Inspection issues routinely are raised with regard to a licensees weakness in correcting recurrent problems or in adequately addressing issues that could become a future significant safety concern.

The statement on Page 55 of the report, Furthermore, because NRC directs its inspections at problems that it recognizes as being more important to safety, NRC may overlook other problems until they develop into significant and immediate safety problems does not accurately reflect the stated objectives and demonstrable implementation of IP 71152.

27. Pages 55-56, discuss safety culture.

Response: To a significant degree, the areas referenced in this draft report are addressed either by NRC requirements or inspection activities. For example, the NRC has requirements limiting work hours for critical plant staff members such as security officers and plant operators. The NRC has requirements governing operator training.

Inspectors routinely monitor various licensee meetings and job briefings to evaluate the licensees emphasis on safety.

Moreover, the NRC has a number of other means to indirectly assess safety culture.

Other NRC tools that provide indirect insights into licensee safety culture include:

• inspection procedure for assessing the licensees employee concerns program,

• NRCs allegation program,

• enforcement of employee protection regulations, 13 Enclosure 1

• Safety-Conscious Work Environment (SCWE) assessments during problem identification and resolution inspections,

• lessons-learned reviews such as the one conducted for the Davis-Besse reactor pressure vessel head degradation; and

• Reactor Oversight Process cross-cutting issues of human performance, problem identification and resolution, and SCWE.

28. Page 58, paragraph under the first header states: It recognized that NRCs written rationale for accepting FirstEnergys justification for continued plant operation was not prepared until 1 year after its decision...

Response: For clarification, the documentation of the decision about one year later was corrective action from a task force finding.

29. Page 58, paragraph under second header states: The NRC task force did not address NRCs failure to learn from previous incidents at power plants and prevent their recurrence.

Response: This sentence is factually inaccurate. The task force performed a limited review of past lessons-learned reports and actually identified many more potentially recurring programmatic issues as a result of that review than the three examples cited by the GAO in this section of the draft report. As discussed during the NRC-GAO exit conference, the task force made a recommendation to perform a more detailed effectiveness review of the actions stemming from other past NRC lessons learned reviews (Appendix F). This review is currently in progress.

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