Environmental Impact Statement - radioactive monticello

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Appendix G Table G-3. Monticello PSA Historical Summary PSA Version Summary of Changes from Prior Model CDF 1992 IPE Submittal 2.6 x 10. 5 1995 * Added non-safety diesel generator to supply battery chargers 1.37 x 10"5 - Added hard pipe containment vent - Improved safety/relief valve (SRV) pneumatics * Added cross-tie for diesel fire pump for low-pressure makeup - Replaced instrument air compressor with one not dependent on service water - Established more realistic success criteria for service water pumps 1999 * Incorporated effects of extended power uprate 1.44 x 10-5 2003 * Updated failure rate data 4.43 x 10"5 • Revised operator error structure to explicitly model dependencies * Credited manual alignment of low pressure safety systems when control power not available * Incorporated new findings on two significant flood scenarios * Modified recovery modeling for offsite power and emergency diesel generators * Credited control rod drive injection under certain conditions SAMA * Small number of event failure probability changes resulting 4.47 x 1O"s from data update tasks * Lowered truncation limit to 1 x 10"11 The staff considered the peer reviews performed for the Monticello PSA, and the potential impact of the review findings on the SAMA evaluation. In the ER, NMC described the previous peer reviews, the most significant of which was the Boiling Water Reactor Owners Group (BWROG) Peer Review of the 1995 PSA model conducted in 1997. The BWROG review concluded that the Monticello PSA can be effectively used to support applications involving NUREG-1437, Supplement 26 G-6 August 2006
Appendix G relative risk significance. NMC stated that all peer review comments, or the evolutions of those peer review comments, are captured by the 2003 model, and that no outstanding model issues exist outside the normal PSA maintenance program and that none of the PSA maintenance tasks are known to have the potential to impact the SAMA conclusions. Given that the Monticello internal events PSA model has been peer-reviewed and the peer review findings were either addressed or judged to have no adverse impact on the SAMA evaluation, and that NMC satisfactorily addressed staff questions regarding the PSA, the staff concludes that the internal events Level 1 PSA model is of sufficient quality to support the SAMA evaluation. Further consideration of the Level 2 PSA model is provided below. As indicated above, the current Monticello PSA does not include external events. In the absence of such an analysis, NMC used the Monticello IPEEE to identify the highest risk accident sequences and the potential means of reducing the risk of posed by those sequences, as discussed below. The Monticello IPEEE was submitted in March 1995 (NSP 1995a), in response to Supplement 4 of Generic Letter 88-20. A revision to the IPEEE was submitted in November 1995 (NSP 1995b). Northern States Power did not identify any fundamental weaknesses or vulnerabilities to severe accident risk in regard to the external events related to seismic, fire, or other external events. In a letter dated April 14, 2000, the staff concluded that the submittals met the intent of Supplement 4 to Generic Letter 88-20, and that the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities (NRC 2000). The Monticello IPEEE uses a focused scope Electric Power Research Institute (EPRI) seismic margins analysis. This method is qualitative and does not provide numerical estimates of the CDF contributions from seismic initiators. The seismic IPEEE identified a number of outliers of items within the scope of the Unresolved Safety Issue (USI) A-46 program. Resolution of these outliers was accomplished in the context of USI A-46. Given the satisfactory resolution of these outliers, Monticello found that none of the plant's high confidence in low probability of failure values were less than the 0.3g review level earthquake used in the IPEEE. The NRC-review and closure of USI A-46 for Monticello are documented in a letter dated November 12, 1998 (NRC 1998). Notwithstanding the conclusions of the IPEEE, as part of the SAMA evaluation NMC reviewed the seismic analysis results and history to determine whether there were any unresolved issues that could impact the seismic risk at Monticello, particularly, any unfinished plant enhancements that were needed to ensure equipment on the safe shutdown list would be capable of withstanding the review level earthquake, or any additional plant enhancements that were identified as means of reducing seismic risk but were discarded due to cost considerations. Based on their review, NMC concluded that there were no outstanding issues that could impact the SAMA results. August 2006 G-7 NUREG-1437, Supplement 26 I
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NUREG-1437 Supplement 26 Generic En
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Generic Environmental Impact Statem
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Abstract The U.S. Nuclear Regulator
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Contents Abstract........... ......
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Contents 4.8.7 Conclusions Regardin
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Figures Figure 2-1. Figure 2-2. Fig
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Contents Table 5-1. Category 1 Issu
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Executive Summary The Commission ha
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Executive Summary NMC and the staff
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0 degree pCi microcurie(s) pCi/mL m
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Abbreviations/Acronyms IPE IPEEE J
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Abbreviations/Acronyms rem REMP RHR
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Introduction and (4) present the st
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Introduction (1) The environmental
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Introduction I assessment of the re
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Introduction This definition of pur
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2.0 Description of Nuclear Power Pl
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Description of Site and Environment
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Descfiption of Site and Environment
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Table 2-10. Land Use in Sherburne C
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Environmental Impacts of Refurbishm
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9.0 Summary and Conclusions By lett
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Summary and Conclusions The GElS co
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Summary and Conclusions not reached
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z a :m G) CA, Table 9-1. Summary of
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Summary and Conclusions U.S. Nuclea
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Appendix A: Comments Received on th
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Appendix A Commenter ID Commenter A
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Appendix A A.1 Comments and Respons
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Monticello Nuclear Generating Plant
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3. General Comments Regarding Licen
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Appendix A nuclear power (I) and Mo
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Appendix A scoping process. If an i
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Appendix A Comment: The NRC does no
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Appendix A Comment: The EIS needs t
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Appendix A Comment: The EIS needs t
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Appendix A cognitive impairment. We
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Appendix A Despite available eviden
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Appendix A Comment: The Hospital Di
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Appendix A Comment: The NRC also fa
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Appendix A The NRC fails to acknowl
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the coal chain What about all the s
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gasification generation balanced ag
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Appendix A 14. Comments Concerning
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Appendix A Part II Comments Receive
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Appendix A Table A-2. (contd) Comme
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Appendix A Table A-2. (contd) Comme
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Appendix A A.2.4 General Comments i
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Appendix A Comment: Page 2-7, Line
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Appendix A A.2.12 Comments Concerni
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Appendix A Response: As described a
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information in the Final SEIS as pa
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Appendix A Response: The impacts to
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aging management aspects of the lic
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Appendix A A-2 PURVES TODD: Just ba
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Appendix A PURVES TODD: Thank you v
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Appendix A So you don't look for wh
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Appendix A Yes CHUCK HOSTOVSKY: I t
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Appendix A information a little bit
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Appendix A Return-path: Received:
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Appendix A tAY-W8-2006 11! 07 'ENCL
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Appendix A ta W.Y United States Dep
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Appendix A MAY-17-206 13:11 P.04 U.
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Appendix A IMAY-17-20O6 13:12 P.06
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Appendix B Contributors to the Supp
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Appendix B Name Karen McWilliams Na
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Appendix C: Chronology of NRC Staff
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Appendix C May 27, 2005 May 27, 200
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Appendix C June 1,2005 June 2, 2005
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Appendix C January 23, 2006 Februar
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Page 413 and 414: NRC FORM 335 U.S. NUCLEAR REGULATOR
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