ML083180986
| ML083180986 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/05/2008 |
| From: | Shack W Advisory Committee on Reactor Safeguards |
| To: | Klein D NRC/Chairman |
| References | |
| LTR-08-0587 | |
| Download: ML083180986 (8) | |
Text
OFFICE OF THE SECRETARY CORRESPONDENCE CONTROL TICKET Date Printed: Nov 13, 2008 08:00 PAPER NUMBER: LTR-08-0587 LOGGING DATE: 11/12/2008 ACTION OFFICE: EDO *-"--
AUTHOR: William Shack DEDMRT DEDR AFFILIATION: ACRS DEDCM Dale Klein AO ADDRESSEE:
SUBJECT:
Summary Report-55th meeting of the ACRS, October 2-3, 2008, and other related activities of the Committee ACTION: Appropriate DISTRIBUTION: RF LETTER DATE: 11/12/2008 ACKNOWLEDGED No SPECIAL HANDLING:
NOTES:
FILE LOCATION: ADAMS DATE DUE: DATE SIGNED:
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UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 November 5, 2008 The Honorable Dale E. Klein Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
SUBJECT:
SUMMARY
REPORT - 556th MEETING OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS, OCTOBER 2-3, 2008, AND OTHER RELATED ACTIVITIES OF THE COMMITTEE
Dear Chairman Klein:
During its 556th meeting, October 2-3, 2008, the Advisory Committee on Reactor Safeguards (ACRS) discussed several matters and completed the following reports, letters, and memoranda:
REPORTS Reports to Dale E. Klein, Chairman, NRC, from William J. Shack, Chairman, ACRS:
- Report on the Safety Aspects of the License Renewal Application for the Shearon Harris Nuclear Power Plant, Unit 1, dated October 16, 2008 Status of Resolution of Generic Safety Issue-191, "Assessment of Debris Accumulation on PWR Sump Performance," dated October 22, 2008 LETTERS Letter to R. W. Borchardt, Executive Director for Operations, NRC, from William J. Shack, Chairman, ACRS:
Interim Letter 5: Chapters 19 and 22 of the NRC Staffs Safety Evaluation Report with Open Items Related to the Certification of the ESBWR Design, dated October 29, 2008 Letter to Dr. Brian Sheron, Director, Office of Nuclear Regulatory Research, NRC, from William J. Shack, Chairman, ACRS:
ACRS Assessment of the Quality of Selected NRC Research Projects - FY 2008, dated October 22, 2008 MEMORANDA Memoranda to R. W. Borchardt, Executive Director for Operations, NRC, from Edwin M. Hackett, Executive Director, ACRS:
Proposed Interim Staff Guidances COL/ESP-ISG-004 and DC/COL-ISG-07, dated October 8, 2008 Draft Final Revision to Regulatory Guides 1.114 and 3.11, dated October 8, 2008 Withdrawal of Regulatory Guide 6.8, "Identification Plaque For Irretrievable Well-Logging Sources," dated October 8, 2008 HIGHLIGHTS OF KEY ISSUES
- 1. License Renewal Application for the Shearon Harris Nuclear Power Plant, Unit 1 The Committee met with representatives of the Carolina Power & Light Company (CP&L) (the applicant) and the NRC staff to discuss the license renewal application (LRA) for the Shearon Harris Nuclear Power Plant (HNP), Unit 1, and the associated NRC staff's final Safety Evaluation Report (SER). The operating license for HNP, Unit 1 expires on October 24, 2026. The applicant has requested approval for continued operation for a period of 20 years beyond the current license expiration date.
In the final SER, the staff documented its review of the license renewal application and other information submitted by CP&L and obtained during the audits and an inspection conducted at the plant site. The staff reviewed the completeness of the applicant's identification of structures, systems, and components (SSCs) that are within the scope of license renewal; the integrated plant assessment process; the applicant's identification of the plausible aging mechanisms associated with passive, long-lived components; the adequacy of the applicant's aging management programs; and the identification and assessment of time-limited aging analyses requiring review.
The applicant addressed the open item, which is related to the scoping classification of feedwater regulating valves and bypass valves. HNP included these valves in the LRA within the scope of license renewal per the criteria of 10 CFR 54.4(a) (2), because it's believed that these valves are nonsafety-related components. However, the staff initially believed that these valves perform main feedwater isolation function under certain postulated transients and fulfill a safety-related function; therefore, these valves should be included within the scope under 10 CFR 54.4(a) (1). The applicant pointed out a past NRC precedence which allowed credit be taken for the backup nonsafety-related components to mitigate the consequences of a main steamline break inside containment. The staff agreed with the applicant's assessment and determined that the feedwater regulating and bypass valves are properly categorized as nonsafety-related components, and that the requirements of 10 CFR 54.4(a) (2) apply to these Valves. The applicant also discussed electrical manholes issue. The applicant described the cable design, manhole inspection program, historical operating experience, and corrective actions.
The staff described its review of the applicant's scoping, screening, aging management programs, time-limited aging analyses, and the resolution of the open item, described above.
The staff concluded that the requirement of 10 CFR 54.29(a) has been met.
Committee Action The Committee issued a report to the NRC Chairman on this matter, dated October 16, 2008.
The Committee concluded that the programs established and committed to by the applicant to manage age-related degradation provide reasonable assurance that the HNP, Unit 1 can be operated in accordance with the current licensing basis for the period of extended operation without undue risk to the health and safety of the public. The Committee recommended that the CP&L application for renewal of the operating license for HNP, Unit 1 be approved. In addition, the Committee recommended that prior to entering the period of extended operation, the staff inspect the applicant's programs for managing water intrusion into underground cable vaults and cable insulation testing.
- 2. Status of Resolution of Generic Safety Issue - 191 The Committee met with representatives of the NRC staff to discuss the status of resolution of Generic Safety Issue (GSI)-191, "Assessment of Debris Accumulation on PWR Sump Performance."
The Committee noted that significant progress has been made towards resolving GSI-191. All PWR licensees have installed significantly larger sump screens and some have undertaken further actions, such as changing fibrous insulation and chemical buffer. The staff described progress made toward resolving GSI 191, including the response of licensees to Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors." The presentation focused on sump strainer head loss testing, chemical effects, and in-vessel sump strainer downstream effects. The staff indicated that testing of the new strainer design has been performed by nearly all licensees and that the initial review of the industry submittals by the staff is nearing completion. The staff provided an overview of the testing that has been performed and the results obtained.
Committee Action The Committee issued a report to the NRC Chairman on this matter, dated October 22, 2008, stating that significant progress has been made towards resolving GSI-1 91. The Committee recommended, to ensure the prototypicality of tests for extrapolation to plant conditions, further guidance be developed for the test cases in which a significant portion of the debris is allowed to settle out upstream of the screens. Also, the staff should develop guidance with regard to the conditions and protocols for the PWR Owners Group tests to facilitate closure of issues related to in-vessel downstream effects.
- 3. Selected Chapters of the SER Associated with the ESBWR Desiqn Certification Application The Committee met with representatives of the NRC staff and General Electric - Hitachi Nuclear Energy (GEH) todiscuss Chapters 19 and 22 of the NRC Staff's SER with Open Items associated with the Economic Simplified Boiling Water Reactor (ESBWR) Design Certification Application.
GEH discussed the objectives of the Design Certification Probabilistic Risk Assessment (PRA),
and stated that 386 of 450 requests for additional information on this topic have been resolved.
GEH intends to issue Revision 4 of the PRA in December 2008. GEH described the nature and approach of the risk analysis calculations, discussed the results and their implications for the ESBWR risk profile, and noted that the PRA was a major influence on the ESBWR design.
They also discussed the criteria used to classify structures, systems, and components (SSCs) as part of the "regulatory treatment of non-safety systems" (RTNSS) program. A high relative risk importance is one of several criteria which can add SSCs to this list.
NRC staff stated that its review of these Chapters covered overall PRA quality as well as the seismic margins analysis, high winds analysis, non-power operational modes, severe accident management, and severe accident mitigation. The staff discussed a number of open items which need to be resolved and also summarized its review of RTNSS. The objectives of this review were to confirm that all non-safety SSCs requiring treatment are identified, to confirm that the reliability and availability missions for active systems are consistent with the risk assessment, and to confirm that the level of treatment is based on the ability to meetreliability/availability missions.
Committee Action The Committee issued a letter to the Executive Director for Operations on this matter, dated October 29, 2008. The Committee recommended that a more detailed explanation of the applicant's analysis of the low failure probabilities of the passive systems and a more systematic evaluation of the relevant uncertainties be provided. In addition, the technical basis for the failure probability of the digital instrumentation and control systems should be provided and specific issues need to be clarified to ensure the functionality of the Basemat-Internal Melt Arrest and Coolability (BiMAC) device. The Committee also stated that it is awaiting the completion of the staffs review of Revision 3 of the ESBWR PRA, and that it would review the resolution of open items in SER Chapters 19 and 22 in future meetings.
- 4. Quality Assessment of Selected Research Proiects The Committee completed its report on the quality assessment of the research projects on:
"Assessment of Predictive Bias and the Influence of Manufacturing, Model, and Power Uncertainties in NRC Fuel Performance Code Predictions," and NUREG/CR 6943, "A Study of Remote Visual Methods to Detect Cracking in Reactor Components."
Committee Action The Committee issued a letter to the Director of the Office of Nuclear Regulatory Research, dated October 22, 2008, transmitting its report on the quality assessment of the research projects noted above.
- 5. Historical Perspectives and Insiqhts on Reactor Consequence Analyses The Committee discussed the draft White Paper prepared by Dr. Nourbakhsh, ACRS Senior Technical Advisor, regarding historical perspectives and insights on reactor consequence analyses.
Committee Action The Committee plans to issue a letter transmitting the White Paper to the Executive Director for Operations during its November 2008 meeting.
- 6. Materials, Metallurgy, and Reactor Fuels Subcommittee Report The Chairman of the Materials, Metallurgy, and Reactor Fuels Subcommittee provided a report to the Committee summarizing the results of the October 1, 2008, meeting with the NRC staff to review the proposed amendment to 10 CFR 50.61: "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events." He stated that the NRC staff provided a detailed discussion on the technical basis for the proposed supplemental rule, including treatment of uncertainties, PRA event sequence analysis, thermal hydraulic analysis, and probabilistic fracture mechanics analysis. The Subcommittee requested additional information regarding the independent review of the FAVOR code, the studies conducted to show the general applicability of the rule to the entire fleet of PWRs, and the justification for not treating thermal hydraulic parameter uncertainties. The staff is expected to provide additional information to the Subcommittee on these topics. The Subcommittee will review the draft final revision to 10 CFR 50.61 after reconciliation of public comments.
- 7. Reliability and PRA Subcommittee Report The Chairman of the Reliability and Probabilistic Risk Assessment Subcommittee provided a report to the Committee summarizing the results of the September 28, 2008, meeting with the NRC staff and EPRI to discuss draft NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," and the latest version of the companion EPRI report, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments."
The draft NUREG-1 855 is intended to provide the needed guidance recommended by the ACRS. EPRI, in parallel with the NRC, has been developing guidance documents on the treatment of uncertainties. This work is meant to complement the guidance included in NUREG-1855. Where possible the NUREG-1 855 refers to the EPRI work for acceptable treatment of uncertainties. The staff plans to complete the NUREG report by the end of the calendar year and then hold a public workshop.
OTHER RELATED ACTIVITIES OF THE COMMITTEE During the period from September 8, 2008, through October 1, 2008, the following Subcommittee meetings were held:
- Thermal Hydraulic Phenomena - September 23, 2008 The Subcommittee discussed the status of resolution of GSI-191 and heard presentations from the staff, the PWR Owners Group, and the Union of Concerned Scientists, in this regard.
- Reliability and Probabilistic Risk Assessment - September 30, 2008 The Subcommittee-discussed NUREG-1 855 and EPRI-1016737 documents, and heard presentations from the staff and the EPRI, in this regard.
- Materials, Metallurqy, and Reactor Fuels - October 1, 2008 The Subcommittee reviewed the technical basis for the amendment to 10 CFR 50.61: "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events."
- Planning and Procedures - October 1, 2008 The Subcommittee discussed prop5osed ACRS activities, practices, and procedures for conducting Committee business and organizational and personnel matters relating to ACRS and its staff.
LIST OF MATTERS FOR THE ATTENTION OF THE EDO
- The Committee will review the draft final revision to 10 CFR 50.61, after reconciliation of public comments. Additional information was requested on the independent review of the FAVOR code, the studies conducted to show the general applicability of the rule to the entire fleet of PWRs, and the justification for not treating thermal hydraulic parameter uncertainties.
The Committee would like to have an opportunity to review the final versions of Interim Staff Guidances COL/ESP-ISG-004 and DC/COL-ISG-07, after reconciliation of public comments.
The Committee plans to review draft NUREG-1 855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," during a future meeting.
The Committee plans to review the resolution of open items in SER Chapters 19 and 22 associated with the ESBWR design certification application.
PROPOSED SCHEDULE FOR THE 5 5 7 th ACRS MEETING The Committee agreed to consider the following topics during the 5 5 7 th ACRS meeting, to be held during November 6-8, 2008:
" Chapter 14 of the SER Associated with the ESBWR Design Certification Application
- Position Paper on Incorporating the International Commission on Radiological Protection (ICRP) Recommendations into 10 CFR Parts 20 and 50 0 Status of License Renewal Activities
- Current Issues Associated with Fire Protection and Related Matters
- Proposed Changes to. the Review Process for Subsequent Combined License Applications (SCOLAs)
In addition, the Committee will meet with the Commission on November 7, 2008, to discuss several items of mutual interest, including BWR Power Uprate Issues, PWR Sump Performance Issues, and TRACE Code Development.
Sincerely, IRA/
William J. Shack Chairman