Environmental Impact Statement - radioactive monticello

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Appendix G G.2 Estimate of Risk for Monticello Nuclear Generating Plant NMC's estimates of offsite risk at the Monticello are summarized in Section G.2.1. The summary is followed by the staff's review of NMC's risk estimates in Section G.2.2. G.2.1 Nuclear Management Company, LLC's Risk Estimates Two distinct analyses are combined to form the basis for the risk estimates used in the SAMA analysis: (1) the Monticello Level 1 and 2 PSA model, which is an updated version of the IPE (NSP 1992), and (2) a supplemental analysis of offsite consequences and economic impacts (essentially a Level 3 PSA model) developed specifically for the SAMA analysis. The SAMA analysis is based on a slight modification of the 2003 Monticello Level 1 and 2 PSA model, referred to as the SAMA model. The scope of the Monticello PSA does not include external events. The baseline CDE for the purpose of the SAMA evaluation is approximately 4.5 x 10-s per year. The CDE is based on the risk assessment for internally-initiated events at extended power uprate conditions. NMC did not include the contribution from external events within the Monticello risk estimates; however, it did account for the potential risk reduction benefits associated with external events by doubling the estimated benefits for internal events. This is discussed further in Section G.6.2. The breakdown of CDP by initiating event is provided in Table G-1. As shown in this table, events initiated by internal floods are the dominant contributors to CDE. Station blackout (SBO) sequences contribute 1.52 x 106 per year (about 3 percent of the total internal events CDE), while anticipated transient without scram sequences are insignificant contributors to CDE (8.24 x 10'8 per year). NMC defined SBO as loss of offsite power and both emergency diesel generators. This definition excludes the SBO-like conditions resulting from flooding-induced loss of electrical buses which are large contributors to the internal flooding CDE. When the contribution from flooding events is also included, events resulting in SBO-like conditions account for the majority of the CDE. NUREG-1437, Supplement 26 G-2 August2006 I
Appendix G Table G-1. Monticello Core Damage Frequency CDF % Contribution Initiating Event (per year) to CDF Fire protection system line break in turbine building (TB) 3.2 x 1 0V 71 931-ft elevation west Service water (SW) line break in TB 931-ft elevation east 5.8 x 10" 13 SW line break in TB 911-ft elevation 1.8 x 10 - 4 Loss of offsite power 1.8 x 10-6 4 SW line break in residual heat removal (RHR) A room 8.9 x 10-7 2 SW line break in residual heat removal (RHR) B room 8.9 x 10.7 2 SW line break reactor building 896-ft elevation 4.5 x 10"7 1 Turbine trip 4.5 x 10"7 1 Loss of feedwa~ter 4.5 x 10 1 Other 4.5 x 10" 7 1 Total CDF (from internal events) 4.5 x 10-5 100 The Level 2 Monticello PSA model that forms the basis for the SAMA evaluation represents an adaptation and updating of the IPE Level 2 model. The IPE Level 2 model involved the development of containment event trees that describe the response of the containment to the severe accident phenomena for each of the Level 1 accident classes. The current Level 2 model retains the IPE containment event tree logic and is directly linked with the Level 1 model via the linked fault tree process. In addition, the SAMA model incorporates several modeling changes to better reflect an improved understanding of Level 2 PSA issues as suggested by an independent peer review, most notably, drywell shell failure due to contact with core debris, several items related to radionuclide release states, and net positive suction head (NPSH) limits following containment venting. The result of the Level 2 PSA is a set of release categories with their respective frequency and release characteristics. The results of this analysis for Monticello are provided in Table F.2-4 of the ER. The frequency of each release category was obtained from the quantification of the linked Level 1 - Level 2 models. The release characteristics were obtained from the results of modular accident analysis program (MAAP) analyses that were determined to bound the release fraction for the sequences in each release category. The offsite consequences and economic impact analyses use the MACCS2 code to determine the offsite risk impacts on the surrounding environment and public. Inputs for these analyses August 2006 G-3 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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9.0 Summary and Conclusions By lett
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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 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 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 A.2.4 General Comments i
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Appendix A A.2.12 Comments Concerni
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information in the Final SEIS as pa
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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 So you don't look for wh
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Appendix A Yes CHUCK HOSTOVSKY: I t
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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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Page 337: Appendix D Organizations Contacted
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Page 413 and 414: NRC FORM 335 U.S. NUCLEAR REGULATOR
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