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| | number = ML20059E555 | | | number = ML20059E555 |
| | issue date = 08/31/1990 | | | issue date = 08/31/1990 |
| | title = Individual Plant Exam for Severe Accident Vulnerabilities. | | | title = Individual Plant Exam for Severe Accident Vulnerabilities |
| | author name = | | | author name = |
| | author affiliation = NORTHEAST UTILITIES | | | author affiliation = NORTHEAST UTILITIES |
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| =Text= | | =Text= |
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| I NUSCG 171 N
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| < 1 2 Millstone Unit 3 individual Plant Examination for Severe Accident Vulnerabilities
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| ; , 4 PREPARED BY Probabilistic Risk Assessment Section Northeast Utilities Service Co.
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| August 1990 oooo1oo m 00001 PDR ADOCL 050004K P PDC
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| o c '
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| ; General Offic:s o Selden Street. Berlin. Connecticut 4
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| '9 Ei
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| ',f7 r,$ecN P.O. BOX 270 ' ..
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| HARTFORD. CONNECTICUT 06141-0270 k C $ $NE.U.C. (203) 665 5000
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| , y .
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| August 31, 1990 Docket No. 50-423
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| . B13596 Re: Generic Letter 88-20
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| +> IPEL
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| " U.S. Nuclear Regulatory Comission Attention:, Document Control Desk-Washington,:DC 20555' Gentlemen:
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| Millstone Nuclear Power Station, Unit No. 3-Response to_ Generic Letter 88-20 m Individual Plant F< amination for Severe Accident Vulnerabilities
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| [ 5umary Report Submittal
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| ~ The purpose of this submittal is to forward the report summarizing. the Indi-
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| :vidual P1 ant Examination for Severe Accident. Vulnerabilities (IPE)' conducted
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| :by members of- the Northeast . Utilities Service Company (NUSCO) for Northeast Nuclear.- Energy . Company- (NNECO) on behalf of Millstone. Unit No. 3. This
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| *, ~" ' examination was conducted ~ in response to " Individual Plant Examination for H4 Severe s Accident Vulnerabjyties--10CFR50.54(f)
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| The enclosed summary (Generic Letter report was - No.
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| prepared88 20),"in J dated November 23, 1988.-
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| accordance- with . " Individual Plant Examination: Submittal Guidance
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| , (NUREG-1335):- Fi ) Report," issued .as Supplement No.1- to GL 88-20, ' dated
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| 'AugustL29, 1989
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| ~
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| v' ' On July F ' 11989,(3b NNECO and Connecticut _ Yankee Atomic Power Company (CYAPC0) sun 4.*ad a' comprehensive letter in response to GL 88-20 for the l
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| ' 't' Millstone unas and 'the. Haddam Neck ~ Plant. - That submittal outlined the
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| ~
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| genera 11 approach,-' key ' points of past involvement in related issues, and key
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| ' milestones whicht would 'be utilized during' the IPE process.- Later, on -
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| , , -(1)- Dennis Crutchfield letter to All Licensees and Construction. Permits for
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| - - Nuclear ~ Power ' Reactor Facilities, " Individual P1 ant Examination. for Severe Accident, Vulnerabilities--10CFR50.54(f) .(Generic Letter
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| -Nov 88-20)," dated November 23, 1988.
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| -(2) ~ James G. Partlow letter to Licensees and Construction-Permits for Nuclear 4~
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| Power Reactor Facilities, " Initiation of the Individual Plant Examination
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| ' for Severe Accident Vulnerabilities--10CFR50.54(f)--Generic Letter 88-20, i Supplement Not 1," dated-August 29, 1989. .
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| ;(3) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit ^ Nos.1, 2, and 3, Response
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| -tor Generic letter 83-20, Individual Plant Examinations for Severe Accident Vulnerabilities," A07693, dated July 27, 1989.
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| 0S3422 AEV 4 88 I-
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| : i. '
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| h
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| U.S. Nuclear Regulatory Commission '* .
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| B13596/Page 2-August 31,.1990 :
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| October 31,1989,N NNECO and CYAPC0 submitted a second letter, which when Aombined with the July 27 letter constituted the formal 60-day response to'
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| %L 88 20. NNECO's response to GL 88-20 committed to sutmitting the Millstone
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| ' Unit No. 3. IPE Sumeni Report by mid 1990. The NRC of the response h . letter dated November 7,1989,pffand acknowledged receipt stated that their i review of the submitted IPE plans would begin. Followi..g the codetion of their review, the NRC Staff issued a letter on January 9, 1990, stating ;
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| their conclusion that the IPE approach, methodology, and schedule were accept- :
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| able for Millstone Unit No. 3. '
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| This enclosed sucimary report was prepared in accordance with the approach, y methodology, and schedule which were reviewed and judged acceptable by NRC Staff. This summary document, formatted extremely closely to NUREG-1335 Table 2.1, provides the " road map" link to the detailed support information submitted prior to the IPE initiation date, as allowed by NUREG-1335, Appen- ,
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| dix C. The results of this report indicate that there are no major severe accident vulnerabilities requiring immediate corrective action beyond those
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| + already implemented or in the process of being implemented. Plant improve-ments, increased emphasis on operator training, increased surveillance,'and/or procedural changes have been or are being accomplished for the most important t
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| 'PRA issues such as station blackout (SBO) and interfacing system loss-of- 1 coolant accident (ISLOCA) prevention and mitigation.
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| The Millstone Unit No. 3 Probabilistic Safety Study '(PSS), discussed exten-sively within this submittal, is the cornerstone on which the IPE was con-ducted. The PSS, based on a Level III Probabilistic Risk Assessment (PRA),
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| ,I including internal flooding, seismic, and other external events has previously l undergone extensive NRC Staff' and contractor review (Brookhaven~ National l Laboratory, Lawrence Livermore National Laboratory, et al . ) . Since the l Level III PRA .was completed, submitted, and extensively reviewed, several i updates have been completed and submitted to maintain the model according to the as-built plant configuration (i.e., "living PRA"). Subsequent formal revisions will likewise continue to be submitted. This "living" PRA program,- i discussed in Section 5.4 of the report, is utilized in many different ways to a support routine ' operation, beyond the IPE requirement (e.g., operator- train- ,
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| .ing, safety evaluations for plant design changes and proposed license
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| [ ,
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| ! '(4) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit Nos.1, 2, and 3, Respose to Generic Letter 88-20, Supplement 1, Individual Plant Examinations for Service Accident Vulnerabilities," A08283, dated October 31, 1989.
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| -(5) John F. Stolz letter to E. J. Mroczka, " Receipt of 60-day Response to
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| 'L_ Generic Letter 88-20, Individual Plant Examination (IPE) (TAC No. 74417, 74432, 74433, and 74434)," dated November 7, 1989.
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| (6) John' F. Stolz letter to E. J. Mroczka, " Review of 60-Day Response to Generic Letter 88-20, Individual Plant Examinations (IPE) (TAC Nos. 74417/74432/74434)," dated January 9, 1990.
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| i
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| Ues. Nuclear Regulatory Commis'sion >
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| :i*
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| * s B13596/Page 3.
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| August 31, 1990 I;
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| amendants,etc.). A number of plant design and procedural changes, discussed in Section 6, have been implemented based on PRA insights.
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| s This summary report addresses external events as- part 'of the IPE program, ,
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| consistent with the option provided in GL 88-20, Supplement 1 (NUREG-1335), '
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| Section P.5. During the. final' stages of internal review and approval of the attached report,, the Staff issued Individual Plant Examination of Exterg Events (IPEEF) for Severe. Accident Vulnerabilities GL 88-20, Supplement 4, in draft ' for industry comment at the upcoming IPEEE Workshops. Although-review of thic draft is ongoing, our initial finding is that the external e events section of -our attached submittal- generally fulfills the intent of ,
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| GL B8-20, Supplement 4, and NUF.EG-1407.- >
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| Although this numn.ary - document does not contain Accident Management (AM)
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| L discussion, the July 27, 1989 initial IPE submittal presented our AM framework l and existing ctpabilities at some length. . Later, these rudiments (e.g., ,
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| organization, instrumentation, PRA and thermal-hydraulic models on personal
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| ' computers, equipment, procedures, and training) were discussed and _ demon- -
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| strated in a joint meeting with the Staff at our headquarters on August 29,
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| -1989. As our AM car,abil &ies continue to evolve, the information contained in ,
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| ~
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| GL 88-20, Supplemcat 2,W will be addressed as appropriate. Our position-continues 'to be consistent with that of the Nuclear Management and Resources l, Council (NUMARC). In addition, we will address NRC's guidance contained in the yet-to-be-issuegenericletteronAMthatwasdiscussedinSECY-90-180, dated May 18, 1990 1 The complementary relationship existing between the IPE and Integrated Safety l Assessment (ISA) was discussed in both GL 88-20 and NUREG-1335. Ncrtheast r Utilities has utilized -the Integrated Safety Assessment Program (ISAP) for L several years (prior to the IPE initiation date) at two other units to evalu-ate proposed plant modifications based on an integrated and plant-specific evaluation methodology. The relationship between IPE, ISAP, -and our (7) James G. Partlow to All Licensees for Nuclear Power Reactor- Facilities,
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| " Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities--10CFR50.54(f) (Generic Letter No. 88-20--Supplement 4), Draft for Comment," not dated, but issued July 1990.
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| '(8) James G.- Partlow letter to All Holders of Operating Licenses and Construction Permits for Nuclear Power Reactor Facilities, " Accident l
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| Management Strategies for Consideration in the Individual Plant.
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| ;i. Examination Process--Generic Le:ter 88-20, Supplement No. 2," dated I
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| ' April ',, 1990.
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| (9) James M. Taylor Policy Issue (Information ) to the Commissioners, " Status of Implementation Plan for Closu e of Severe Accident Issues and Status of The Individual Plant Examinations (IPE), SECY-90-180," dated May 18, 1990.
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| 1
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| . U.Si Nuclear Regulatory Comission 813596/Page 4 -
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| L. August'31, 1990
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| ; intentions ~ to.. expand the application t f ISAP methodology to Millstone Unit No; 3 were discussed extensively in our July 27, 1989 submittal and summarized sagain in this submittal. The initial ISAP report for this unit is currently
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| - 'edergoing internal- review, with submittal planned in the near future. Any
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| . contemplated potential design modifications will undergo integrated, plant-specific ISAP - evaluation and be scheduled according to its integrated ranked value. Further, our intention to pursue the " multiple-unit" option, discussed 5 NUREG-1335 Appendix C, is well documented.
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| Based on the above discussion, NNECO believes that the steps to achieve
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| ' closure on severe accidents for Millstone Unit No. 3 are understood and well on their way to being completed. This document submits the IPE, which has identified that there are no major severe accident vulnerabilities- requiring imediate corrective action beyond those already completed or initiated. An AM framework has been developed and implemented which can accommodate new information as it is developed. The NRC Staff Containment Performance Improvement (CPI) program specific insights on pressurized water reactor (PWR) largegy 1990, containments, were considereddiscussed in GL duri"g the 88-20, Supplement discussion 3, dated of specific plant July 6, containment
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| ; configuration, although no Comission-approved generic requirements have been issued.. IPEEE is generally addressed in the attached report.
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| As at direct result of >NECO's long-standing management recognition, proactive comitment, and aedication of significant in-house- resources to the nuclear safety ethic, we have benefited significantly as a result of PRA and inte-
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| : grated assessment. This speaks to our inherent comitment to the intent of
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| : IPE. The same proactivity has allowed this M111 atone Unit No. 3 IPE to be conducted and sumary report issued, in accordance with Generic Letter 88-20 and NUREG-1335, for your review very early in the three-year response window.
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| This is a clear indication of how risk man;gement has been institutionalized at NU on a L plant-specific basis. Accordingly, we hereby certify that:
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| (1) the updated Millstone Unit No. 3 Level III 'Living" PRA, in conjunction with this- submittal, meets the intent of the generic letter, especially concerning utility staff involvement; (2) it ref1! cts current plant design and o>eration, and (3) results are being submitted as soon as completed, on a .
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| s iorter schedule than 3 years, as comitted in the July 27, 1989 initial response to GL 88-20.
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| Since the IPE submittal represents one of the early industry responses, perhaps a meeting with the Staff would be useful in clarifying any questions i
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| g.
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| (10) James G. Pe-tlow letter to All Licensees Holding Operating Licenses and Construction Permits for Nuclear Power Reactor Facilities Except
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| !- Licensees for Boiling Water Reactors With Mark I Containments, t- " Completion of Containment Performance Improvement Program and Forwarding 3
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| of Insights For Use In the Individual Plant Examination For Severe r Accident Vulnerabilities - Generic Letter No. 88-20, Supplement No. 3,"
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| dated July 6,1990.
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| ev
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| ly 0 "U.S. Nuclear Regulatory Comission
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| ' B13596/Page 5
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| : August 31, 1990' that may arise. during your review. We remain available for'such a meeting,'
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| 'should you . judge it beneficial.
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| Yary truly yours, NORTHEAST NUCLEAR ENERGY COMPANY
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| .2/
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| E . J .'1 Senior)p66zki'.47-Vice President cc/att: E. S. Beckjord, NRC Director, Nuclear Regulatory Research M. L. Boyle, NRC Project Manager, Millstone Unit No. 1 T. H. Cox, NRR Policy Development and Technical Support Branch J. H. Flack, NRC Research, Severe Accident Issues Branch
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| ~
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| - R. W. Houston,. Nuclear Regulatory Research 1 D. H. Jaffe,-NRC Project Manager, Millstone Unit No. 3 K.'S, Kolaczyk, Resident Inspector, Millstone Unit No. 3 T. T.. Martin, Region-1-Administrator 6 R. L. Palla, Jr...NRC, NRR Probabilistic Risk Applications Branch
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| + W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 -
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| y T. P. Speis, NRC Deputy Director, Nuclear Regulatory Research I4 F
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| cc: .J. M. Taylor. Executive Director for Operations
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| . T.:E. Murley, Director, Office of Nuclear Reactor Regulation
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| .E. L.. Jordan,- Director, Office for- Analysis & Evaluation of Operational Data W. T.: Russell,, Associate Director for Inspection and Technical Assessment-J. G. Partlow, ^.ssociate Director for Projects A. C. Thadani Director, Division of Systems Technology M. C. Thadan,, Division of Reactor Projects--l/II J. A. Zwolinski, Assistant Director for Region 111 s K. M. Carr, Chairman, U.S. Nuclear Regulatory Comission y K. C. Rodgers, U.S. Nuclear Regulatory Comission' '
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| J. R. Curtiss, U.S. Nuclear Regulatory Comission A F. C. Remick, U.S. Nuclear Regulatory Comission STATE OF' CONNECTICUT)
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| : ) ss.- Berlin "
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| -COUNTY OF HARTFORD ).
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| t Then personally appeared before me, E. J. Mroczka, who being duly sworn, did state.thatihe is Senior Vice President of Northeast Nuclear Energy Company, a
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| [F . Licensee herein, that he is authorized to execute and file the' foregoing information in the name and on behalf of the Licensee herein, and that the (T : statements contained.in said information tre true and correct to the best of I his knowledge and belief.
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| $/1(}4'l 'L70l' R Kary Fublic '~ ~
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| CommissionEsplesMyth31,1995 1
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| . NUSCO 171 -i I
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| MILLSTONE UNIT 3 Individual Plant Examination ;
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| . for Severe Accident Vulnerabilities !
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| Probabilistic . Risk Assessment Section Northeast Utilities Service Co. :
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| I P o DISCLAIMER 4
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| 1
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| (
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| The information'' contained in this topical report was prepared for- ,
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| the specific requirements of Northeast Utilities Service Company (NUSCO) and its affiliated companies, and may contain materials
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| : subject to privately owned rights. Any use of all oc any' portion
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| ~
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| of the information, analyses, methodology-or data' contained in-Lthis topical report by third parties shall be undertaken at such
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| : party's sole riski NUSCO and its affiliated companies hereby disclaim any liability (including but not" limited to tort, i t
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| contract, statute, or codrse.of dealing) or warranty (whether- '
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| express or implied)-for the accuracy, completeness, suitability for.a particular purpose of merchantabilityLof the information.
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| i, TABLE OF CONTENTS i EAGE .,
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| . . . . . . . . . . . . . .- . . . 1 1 EXECUTIVE
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| | |
| ==SUMMARY==
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| 1.1 . Background and Objectives . . . . .. . . . .- . 1 1-
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| -1.2 Plant Familiarization . . . . . . . . . . . . .
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| 2' 1.3 Overall Methodology . . . . . . . . . . . . . .
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| };
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| L 1.4 Summary of Major Findings . . . . . . . . . . . 3 9
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| 2 EXAMINATION DESCRIPTION . . . .. . . . . . . . . . 9 2 .1' Introduction . . . . . . . . . . . . . . . . .
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| 2.2 Conformance with Generic Letter and Supporting Material . . . . . . . . .. . . . . . - . . . . . 11 General Methodology . . . . . . . . . . . . . . 12 2.3 Information Assembly . . . . . . . . . _ . . . . 13 2.4 3 FRONT-END ANALYSIS . . . . . . . . . . . . . . . . . 16 3.1- Accident Sequence Delineation . . . . . . . . . 16 3 .1.1 - Initiation' Events.. . . . . . -. . . . 16 3.1.2 Front-Line Event Trees . . . . . . . 17' 3.1.3 Special Event Trees . . . . . . . . . 18 3.1.'4 _ Support System Event Trea . . .- .c . . 18 d.1.5- Sequence Grouping and Back-end Inter-face . . - , . . . . . . . . .. . .- , 19 3.2 System Analysis . . . . . . . . . . . . . . . . 20 3.2.1 System Descriptions . . .. . ' . . . . 20 3.2.2 System Analysis-. . . . . .. . . . . 20 3.2.3 System Dependencies . . . . . .. . . . 21 3L.3 Sequence'Quantification . . . . .. . . . . . . 22 3 . 3 .1' List of Generic Data . . . . . .. . .- 22 3.3.2 Plant Specific Data and Analysis . . 23 3.3.3 Human Failure Data-(Generic and Plant Specific) . . ... . . .. . . .- . . 23-3.3.4 Common Cause Failure Data . . . . . . 25 3.3.5 Quantification of Unavailability of Systems and Functions . . . .. . . -. 26 3.3.6 _ Generation of Support System States
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| -and Quantification of Their Probabil-ities . . . . . . . . . . . . . .. . . 26 3.3.7 Quantification of Sequence Frequen-cies . . . . . . . . . . . . - . . . .. 27 3.3.8 Internal Flooding Analysis . . . . .
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| 28 e
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| '3. 4 Results and Screening Process . . . . . . . . . . .; 30-3.4.1 Application of Generic Letter Screen-ing Criteria . . . . . . . <. s . . . . -30 3.4.2 Vulnerability Screening . . . . . . . 31 3.4.2 Decay Heat Removal Evaluation . . . . 33 ,
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| 3.4.4 USI and GSI Screening . . . . . . . . 36 3.5 External Events . .. . . . . . . . . . . . . . 37 3.5.1 Seismic Initiating Events . . . . . . 38 3.5.2 Fires . . . . . . . . . . . . . . . . 41 04LW X.0 2
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| r 8 *
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| : 3. 5. 3 - -External' Flooding . .. . . . . . . . 44 1
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| : 3. 5. 4 ' -Extreme' Winds and Wind-Generated '!
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| Missiles . . . . . . . . . . .. . . 45 l 3.5.5 ' Aircraft Accidents . . . . . . . . .
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| 46- 't 1
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| 3.5.5- Transportation and Storage of- .
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| Hazardoils Materials . . . . . . . . . 47-3.5.7 Turbine Missiles . . . . . . . . . . 47 4 BACK-END ANALYSIS. . . . . .. . . . . . . . . . . . 52 #
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| 4.1 Plant Data and Plant Description. . . . . . . . . 52 4.2 Plant Models and Methods for' Physical Processes 54 4.3 Bins and Pl' ant Damage Staties .. . . . . . . . 54 '
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| 4.4- Containment. Failure Characterization . . . . . 55 4.5 Containment Event Trees . .. . . . . . . . . . 55
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| ^
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| 4.6 Accident Progression.and CET Quantification . . 56-4.7 Radionuclide Release Characterization . . . . . 58 4.8 Follow-On Back-End Analyses . . . . . . . . . . 59
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| . 4.9 Insights from NUREG-1150 . . . . . . . . . . . 62 5 UTILITY PARTICIPATION AND. INTERNAL REVIEW TEAM . . - . 65 5.1 IPE Program Organization . . . . . . . .. . . . ,
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| 65 5.2 Composition of Independent Review Team . . . . 65
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| '5.3- Areas of Review, Major. Comments, and Resolution- ,
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| of Comments . . .- . . . . - . . . . . . . .. . .. - 66
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| -5.3.1 Utility-Sponsored Rev d ew . . . . . . 66' f 5.'3.2- LLNL Review . . . . . . . . . . . . . -69, ,
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| 5.3.3 BNL Review of the Back-End Analysis . 73 {
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| 5.3.4 NRC Review . .. . .. . . . . . . . . 76 I 5.4 'Living PRA Program . .- . . . . . . . . . . . . - 78 6 PLANT IMPROVEMENTS AND UNIQUE SAFETY FEATURES . . . 83 V -
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| L 7
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| ==SUMMARY==
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| AND CONCLUSIONS . . . . . . . . . . . . . . 86 i L .
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| -APPENDIX A: . Front-Line Event Trees , l APPENDIX,B: System and Event Tree Success Criteria-APPENDIX'C: Support State Event Trees for 1
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| Support State System Related Initiators APPENDIX D: Human Error Probabilities J l> 1 i
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| l l l
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| 04LWX.080
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| _p
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| .- ._ .~ ---- .. . - . . . . - - . . _ . . . --
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| 1
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| )
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| 3 I
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| LIST OF TABLES g PAGE
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| ~
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| 1.4-1 Core-Melt Frequency Summary 91 Summary of Release Category Frequencies 92 o
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| .1.4-2
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| -3.'l-1 -Internal Initiating Event Vector 93 3~.1-2 Simplified Support States _
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| 94 3.1-3 Definition and Nomenclature of-Plant
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| . Damage States' 95 98*
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| System Summary
| |
| -3.2-l' 13.2-2 System Dependency Matrix. .
| |
| 101- ,
| |
| 3'.3-1 Summary of-System / Function Unavailabilities 102 1 3.3-2 Support State Probabilities 106 1
| |
| 3.3-3 Significant Contributions to CDF from Internal Flooding 107 .
| |
| 3.4-1 Core Damage Frequency by Initiator '108 i Core Damage Frequency by Plant Damage State 109-
| |
| '3.4-2 3.4-3 Summary of: Significant Sequences 110-3.4-4 Summary of Significant Containment Bypass _'
| |
| p Sequences 121 3.4-5 Release Spectrum due to Core Damage 122 3.4-6 : Dominant Containment Failure Sequences 123 3.4-7 Decay Heat Removal Vulnerability Insights and' Applicability to MP3 124 L
| |
| 3.5-1. Seismic Initiating Event Definition :
| |
| and-Mean Frequencies 125 3.5-2 Seismic Core-Melt State Definitions H and Calculated Mean Frequencies 126-
| |
| ;. 3.5-3 Seismic Release Category Dafird". ions L
| |
| and calculated Mean Frequencies 127 3.5-4 Fire Initiating 1 Event Frequencies 128 3.5-5. Fire Areas-versus Plant: Damage States Mean Frequencies 129 ;
| |
| _ 3.5 Fire: Release _ Category Definitions and Calculated Mean Frequencies 130 ,
| |
| - 4.3-1 Containment Response Class Definitions 131- t 4.6-1 Best Estimate Accident Chronology. 132 g 4.6-2 Summary of Containment Response'for. Core ,
| |
| Melt Accidents 133 i 4 4.7-1 Notation and Definitions for Release-l Categories- _
| |
| 134 Release Category Summary Cesium - Iodine 4.7-2
| |
| !: Model 135
| |
| +
| |
| 0 4.9-1 Comparison of Millstone Unit 3 and Surry Unit 1 136
| |
| . 4'.9 Applicability of NUREG-1150 Insights for Surry to Millstone Unit 3 137 5.3-1 SimplifiedLContainment Matrix 142 5.4-1 Sample Millstone Unit 3 PRA Support Activities 143 5.4-2 PDCRs to be' Incorporated into Future Update 145 1 Sample Plant Improvements with High PRA Basis _145 6-2 Addressment of Significant PRA Findings 147 04LW7X.080
| |
| | |
| 4 LIST OF FIGURES ZlGGLE ,
| |
| 3.1-1 Support System Event Tree for Nonsupport .
| |
| -Systan Related Initiators 150 3.5-1 PGA Results for Millstone 152 4.1-1 . Reactor Cavity Elevation View 153,
| |
| ' 4.8-1 Containment Analysis Benchmark'(TMLB) , 154:
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| 5.1-1 Project Organization 155 A
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| i 04LW7X,08D 9
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| . _ , ~.- . . _ . _ . . _ . _ - ._ _. _ _ _ _ . . _ _ _ _ _ _ . _ _ . .
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| 1 1 1
| |
| MIT.T. ATONE UNIT 3 POWER STATION-INDIVIDUAL PLANT EXAMINATION SUBMITTAL )
| |
| '' I J
| |
| 11~ EXECUTIVE-
| |
| | |
| ==SUMMARY==
| |
| | |
| i 1.1. Background and Objectives p Generic Letter 88-20 (Reference 1-1) issued by the United States Nuclear Regulatory Commission (NRC) requested all licensees holding operating licenses and construction permits
| |
| [ -for nuclear power reactor facilities to perform individual plant examinations (IPE) oof their plant (s) for severe accident vulnerabilities and submit the results. This document h constitutes the Millstone Unit No. 3 (MP3) IPE. submittal. 1N)
| |
| L 'the extent possible, the guidance provided in NUREG-1335 h ,
| |
| (Reference 1-2) was utilized in the preparation of this '
| |
| document.
| |
| l .>
| |
| The MP3' Level III PRA formed the basis to address Generic
| |
| ~
| |
| Letter.88-20. Since the MP3 Probabilistic Safety Study (MP3 PSS) (Reference 1-3) has been previously submitted and-
| |
| ~G reviewed by NRC, Northeast Utilities-(NU) has elected to ,
| |
| _ provide a summary document and " road map" of the existing probabilistic risk assessment (PRA) results as part of this s IPE submittal. In this report, internally initiated events and external events are addressed.
| |
| 1.2 Plant Familiarization When the MP3 PRA was performed, some two dozen Northeast Util-ities.(NU) engineers, scientists, and operators participated, in-the study. NU and contractor personnel became familiar j with'the plant by performing plant walkdowns of virtually all ]
| |
| major systems and plant areas. Since the plant was under con- l
| |
| .struction at the time the PRA was performed, the personnel had access to many systems and areas which are not easily 04LW7X.080
| |
| ------_---i_-_-__ . _ - - - _ . - - - . - - . _ _ - - - . . - - - - - - - - - - - - - . - - - _ . - - , --- -- , - ~ ,-
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| | |
| 2 accessible in:an operating plant, including reactor containment andilower cavity.
| |
| _Since completion of the PRA, NU,has amended and updated the- .
| |
| PRA six times. NU personnel have kept abreast of design and operational. changes at the plant by virtue of established
| |
| - i" procedures in. place which require the PRA engineers to review.
| |
| :and prioritize essentially all-major design _ changes to the plant. This relatively high PRA personnel 1 involvement and the' PRA section participation on the Nuclear Review Board has created an environment where there is frequent exchange-of information between the plant's. operations personnel and the PRA staff. Therefore, the PRA staff continues to keep up-to 4 date with the design and procedural changes at the plant.
| |
| 1.3 .overall Methodology The IPE submittal of MP3 utilizes the already documented MP3
| |
| 'PSS as its basis.- Since its completion, this PRA has ;
| |
| -undergone several in-house reviews, contractor review, an- .
| |
| expert panel review (Reference 1-4), and NRC reviews ,
| |
| (References.1-5, 1-6, and 1-7). -,j During the IPE,_in' order to ensure that'the intents of the .)
| |
| ' generic letter'are met, the following tasks were perforined: j'
| |
| : 1. Review of the plant design change records (PDCRs). 'i
| |
| : 2. Review of the back-end analysis to incorporate insights gained from NUREG-1150 (Reference 1-8).
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| : 3. Update of the initiating event frequencies to .
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| i
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| ; reflect MP3 experience.
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| w .
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| 4
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| : 4. Incorporation of significant comments of NRC reviewers that had not been addressed previously.
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| 04LW7X.060
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| 3 . ;
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| i
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| '' 5. -Completion of an integrated PC-based PRA model for
| |
| .MP3:to address the needs of the IPE and as a step in ,
| |
| the living PRA program at MP3.
| |
| L 1.4 Summary of Major Findings I:
| |
| This section summarizes the major findings of the Millstone 3 ;
| |
| -Probabilistic Safety S,tudy. Additional discussion can be '
| |
| found in.Section V of the Introduction and Summary tc the MP3.PSS, as well as Section 7 of the PS'S. While the order of some'of the dominant accident sequences has changed as the PSS L .
| |
| has been updated with time, the significant insights have not l
| |
| 'been greatly affected.
| |
| For loss of coolant accidents, a major contributor to core melt frequency: is the failure of high or low pressure contain- ;
| |
| ment sump recirculation, particularly for large and medium LOCAs. These, in turn,'are dominated:by human error as well as common cause failure of motor operated valves. A large [
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| numberLof MOVs, especially in the service water system, are ,
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| tested only during. cold shutdown, and hence have a long test '
| |
| interval. This, in turn, increases the failure probability.
| |
| For small break LOCA, the large. refueling water storage tank capacity (1.2 million gallons) substantially delays the. time D
| |
| to sump recirculation. Substantial-conservation of RWST
| |
| ~ inventory for core cooling is also possible because of the-reduced need for conta'inment spray. These, in turn, can delay ,
| |
| the need for sump recirculation in many instances;to nearly 24 hours,.and greatly reduce the human error probability. ,
| |
| Moreover,-alternatives to sump recirculation such as primary i
| |
| .depressurization and entering normal residual heat removal' cooling (cold shutdown) are viable alternatives on this time frame. Based on PRA insights, relatively high priority has been assigned to the sump recirculation procedure for operator L training. 'Also,-a design change for cold leg recirculation i
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| 04LW7X.080
| |
| | |
| p ._ _ _ _ _ _ _ _ . _ _ _. . _ _ _ _ _. _. _ _ _ _ . .
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| I l
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| 1 4
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| imprcvement-(not'yet credited in the PSS) has been l
| |
| impletented.
| |
| Failures of.the safety injection accumulators also appear in : -l h several top accident sequences. A large reason for.this is again the long test interval between testing of th? discharge check valves. Based on PRA insights, partial stroke testing l of the valves,is now performed every. refueling outage.
| |
| Another reason for the somewhat noticeable contribution ofL ,
| |
| : accumulator failure is the relatively strict success criterion
| |
| -used' (3 of 3 accumulators on the unf aulted loops for both large and medium break LOCA). In the absence of additional
| |
| ! best-estimate LOCA analyses, this criterion cannot be relaxed.
| |
| However, it is an area for future investigation. .
| |
| b Other than the accumulator issue, failure of emergency core cooling injection for all break sizes.is found to be a relatively small contributor to core melt frequency. This is ,
| |
| l because of the numerous high pressure injection pumps for t small/ medium break LOCA (two (2) high head Safety Injaction pumps, two (J) normally available large capacity centrifugal ,
| |
| L charging pumps). Also, based on analyses, any two of-these-four pumps have been credited for large LOCA mitigation should Lboth low pressure safety injection pumps fail.
| |
| ! For transient events, the diversity and redundancy of L auxiliary feedwater (two (2)-motor-driven pumps, one (1) steam-driven) is beneficial. Likewise, the four high pressure
| |
| -pumps provide multiple means of primary feed and bleed cooling when used in conjunction with the pressurizer PORVs.. Feed and bleed cooling is found to be a very important means of provid -
| |
| 'ing decay heat removal at MP3 for loss of all feedwater events. Again, high priority for operator training has been assigned to the feed.and bleed procedural steps based on PRA insights.
| |
| 04LW7X.080
| |
| .J . . r .-_ _..______r____._.
| |
| | |
| . ~ . . . - - . - - . - . . - - - . -. . .-
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| ,, +
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| 5 Anticipated: transients without scram (ATWS) were not found to !
| |
| LJ - be major contributors:to core melt frequency (CMF) in the 1 original MP3 PSS. Because of the reduction in transient l initiating event frequency in the most recent PRA update, the 5 I contribution of ATWS has been further reduced. Additional reduction:is anticipated when the ATWS Mitigation System
| |
| ' Actuation Circuitry (AMSAC) system.is credited in the PRA.
| |
| Station blackout sequences were found not to be major contri-butors to core melt frequency in the PSS, but have been shown i to be important contributors to large scale fission product releases and public risk. Since publication of the original-PSS, additional station blackout analyses have been performed
| |
| - to incorporate. insights from on-going activities such as-industry research on reactor coolant pump seal behavior. In 4 response to the Station Blackout rule, Northeast Nuclear Energy Company (NNECO) has committed to the installation of a ,
| |
| b, third diesel generator, which would be air-cooled. This-will 3
| |
| further reduce the station blackout contribution.
| |
| * The interfacing systems L CA (ISLOCA) analysis.has continued to be refined es insights ,com ongoing PRA activities are ,
| |
| ' incorporated. Based on evaluation of the structural capability of the third (normally closed) Mov in the RHR suction lines, some credit has been taken for the capability- .
| |
| of the valves to prevent ISLOCA. Also, based (n) industry experience (no observed catastrophic disk rupture of MOVs (Reference 1-9)), the probabilities'ofLcertain failure modes i have been reduced. On the other hand, insights from'the- ~
| |
| ! Haddam-Neck Plant Event V audit by NRC in July 1989, as well as the MP3 RHR Autoclosure Interlock Removal study (Refer-ence 1-10), have resulted in the addition of other failure modes not previously considered, such as human error. None-theless, the V-sequence has been reduced'by approximately one order of magnitude to about 2 x 10-7/yr from the original PSS. 1 Event V remains a major contributor to large scale offsite releases for internal events.
| |
| 04LW7X.000 M' ,
| |
| | |
| ~
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| y j 6
| |
| ' The MP3 PSS evaluated the full spectrum of initiating events, L both internal-and external. For some external initiators, a l .
| |
| L conservative bounding approach was used. In spite of this, -
| |
| external; events except seismic events have been found not to .
| |
| i be major contributors to core melt frequency or public risk. {
| |
| Seismic events contributes 9 x 10~'/yr to the CMF which is .,
| |
| l approximately 13 percent of the total CMF. Because contain-L ment systems and containment isolation capability can be affected for very severe seismic events f'ar beyond the Safe A Shutdown Earthquake (SSE), seismic events are found to be major contributors to public risk. ,
| |
| Major reasons for the generally low contribution of externally initiated events to core melt frequency are the hardened facilities for major equipment, and physical separation of the two trains of. engineered safeguards: equipment. U
| |
| :- With regard to the back-end analysis in the MP3 PSS, the .
| |
| diversity and redundancy of the two containment spray systems ]
| |
| (quench' spray and recirculation spray) has been.found to be !
| |
| advantageous. For most' core melt scenarios where both systems are available,. containment failure is not expected. 'The containment ultimate capacity (median value) is calculated to be about two and one-half times design pressure based on
| |
| ' detail'ed' plant-specific structural analyses.
| |
| 4 Because the containment is operated at subatmospheric.condi-tions during normal power operation, the probability of a significant, undetected pre-existing failure of conh inment isolation is judged to be low. (A plant design change and technical specification change have been proposed which would ,
| |
| increase the allowed containment pressure to 14.0 psia in i operating modes 1 to 4. The capability to detect gross ,
| |
| failure of containment isolation would not be greatly affected).
| |
| 04LWX.080 a
| |
| | |
| t e 7 i The lower-reactor cavity design'is characterized to be " dry" at the time of reactor vessel failure for MP3. Substantial
| |
| -hydrogen generation during molten-core / concrete interactions could be expected before a coolable debris bed could be'estab- i
| |
| +
| |
| lished and/or further hydrogen generation could be arrested. ;
| |
| 'With containment sprays operating, hydrogen burns are not (
| |
| expected to be major threats to containment integrity.
| |
| i There remains a great deal of uncertainty regarding High ,
| |
| Pres'sure Malt Ejection and Direct Containment Heating ,
| |
| phenomenon, as is the case for other PWRs analyzed in NUREG-1150.. Many of the insights from the NUREG-1150 analysis for Surry Unit 1 are applicable to MP3-as well. This area of research in' severe accident behavior is being followed closely.-
| |
| Based on the numerous updates to the PRA model,'the living PRA program, and consideration of reviewer comments, NU is '
| |
| confident that the risk profile provided in this IPE, submittal report is a reasonable representation of the actual risk L profile at MP3. Tables 1.4-1 and 1.4-2 which summarize the j
| |
| estimates of the core-melt frequency and the containment l
| |
| ' failure frequency are representative of the MP3 risk profile.
| |
| No major severe accident vulnerabilities requiring immediate corrective action have been identified or are outstanding.- :
| |
| l -.
| |
| REFERENCES
| |
| 'l-1 U.S. NRC Generic Letter 88-20, Supplement No. 1, "Initia -
| |
| tion of the Individual Plant Examination for Severe Accident Vulnerabiiities - 10CFR50.54(f).," dated August 29, 1989.
| |
| L t 1-2 U.S. Nuclear Regulatory Commission, "Indivi' dual Plant Examination Submittal Guidance," NUREG-1335, August 1989.
| |
| l; 04LW7X.080
| |
| : l. . . . - . .
| |
| | |
| p 8 .
| |
| 1-3 uNillstone Unit 3'Probabilistic Safety Study," Northeast Utilities,-August 1983 (as amended).-
| |
| 4 i
| |
| 1-4' N.C.-Rasmussen, S. Levine, P. J. Wood, " Final Report of <
| |
| the-Level 3 Review Board on the Millstone Point Unit 3 Probabilistic Safety Study," August 1983.
| |
| 1-5 A, A. Garcia, et al, "A Review-of the Millstone 3 Probabilistic Safety Study," NUREG/CR-4142, Lawrence
| |
| ~
| |
| Livermore National Laboratory, April 1986.
| |
| 1-6 M.-Khatib-Rahbar, et al, " Review and Evaluation of-the-Millstone Unit 3 Probabilistic Safety Study Containment Failure Modes, Radiological Source - Terms and'Offsite Consequences," NUREG/CR-414~*, Brookhaven National-Laboratori'., September 1985.
| |
| 1-7 U.S. Nuclear Regulatory Commission, " Millstone 3 Risk Evaluation Report,"'NUREG-1152, August 1985.
| |
| 1-8 U.S. Nuclear Regulatory Commission, " Severe Accident Risks: Surry Unit 1," NUREG-1150, Volume 1, Second Draft for Peer Review, June 1989.
| |
| 1-9 G. Bozoki, et al, " Interfacing Systems.LOCA: Pressurized Water Reactors," NUREG/CR-5102, BNL-NUREG-52135,
| |
| - Brookhaven National Laboratory, February 1989.
| |
| 1-10 "RER Autoclosure Interlock Removal at Millstone Unit 3,"
| |
| . - NUSCO 170, April 1990.
| |
| 04LW7X.08D . ,
| |
| | |
| 9 !
| |
| s 2 EXAMINATION DESCRIPTION ,
| |
| This section describes how the examination process was
| |
| : 1) the methods used to perform the <
| |
| conducted to ensure that:
| |
| IPE conform with the provisions of.the generic letter, and l 2)Jthe-primary objectives of the IPE are met..
| |
| The examinatiQn' description discusses two distinct elements ,
| |
| e associated with this IPE submittal. These two elements are: l a) the MP3 PRA upon which the I'PE was founded, and b) the k additional tasks performed to ensure that the IPE objectives
| |
| -are met. l 2.1 Introduction The four specific objectives of the IPE are for each utility l D i to:
| |
| L L 1) develop an overall appreciation of severe accident ,
| |
| behavior,
| |
| : 2) understand the most likely severe accident sequences that could occur at its plant, f 1:
| |
| L 3) gain a more quantitative understanding of the L overall-probability of core damage and radioactivity material releases, and a
| |
| s 4) 11f necessary,. reduce the overall probability of core ,
| |
| damage and radioactive material releases,by appro-'
| |
| priate modifications.
| |
| As notedLearlier, a Level 3 PRA has already been performed for MP3. This IPE submittal for MP3 draws extensively from the <
| |
| already performed MP3 PRA. In making plans to respond to the IPE, and in order to ensure that the above-mentioned objec-tives are met, NU identified and performed the following tasks:
| |
| i' 04LnUX.06D
| |
| | |
| - 10
| |
| : i. 1.. Plant design and'operationalichange records (PDCRs)
| |
| 'were reviewed to investigate the impact of those changes on the MP3 PSS.- .The MP3 PSS was' completed in 1983 while the plant was under construction. -
| |
| Since then, the single major design modification-perform $d that may significantly affect the risk g profile of the plant is the-ATWS Mitigation System Actuation Circuitry (AMSAC). The effect of this
| |
| . modification is to reduce risk attributed to the ATWS events. A major procedural change that affected the.MP3 PRA is'the Inservice Test Pump and l
| |
| Valve Program (IST) (Reference 2-1). The IST program has already been incorporated during the 1987 update of the PRA.
| |
| : 2. The back-end analysis for MP3 was reviewed in light of more recent insights regarding severe ~ accident behavior. . In particular, the Surry Plant analysis in NUREG-1150 was carefully reviewed for applica-bility to-Millstone Unit 3. Based on this' review, and the earlier Brookhaven National Laboratory review of the MP3 PSS (NUREG/CR-4143);
| |
| (Reference 2-2), it'is concluded-that the existing back-end analysis for MP3 is' acceptable 1for the purposes of identifying severe accident vulnerabilities. As insights into severe accident phenomena such as High' Pressure Melt' Ejection and Direct Containment Heating evolve, consideration will be given to future updates if justifiable.
| |
| : 3. The initiating event frequencies', specifically those of transients, were. updated to. incorporate' plant-specific experience (Reference 2-3). When the'MP3 a
| |
| PRA was performed, no plant-specific data existed for failure rates or initiating event frequencies.
| |
| However,.after several years of operation, MP3 has 04LW7X.060 ih 4
| |
| s
| |
| | |
| l 4
| |
| , + 11 .;
| |
| sufficient data to use plant-specific frequencies for-some initiating events.
| |
| : 4. The comments made by external reviewers were .
| |
| addressed. The MP3 PRA was subjected to four independent expert reviews. They are a) expert panel review (Reference 2-4), b),NRC review (Refer-ence 2-5), c) LLNL review (Reference 2-6), and d) -
| |
| BNL review (Reference 2-2). Most of the significant concerns of the external reviewers have already been ,
| |
| 1 responded to by NU. The purpose of this task was to ,
| |
| assure that all significant comments have been
| |
| - addressed. Details of this task are included in L
| |
| Section 5.3 of this report.
| |
| : 5. A PC-based integrated model of the PRA was created.
| |
| While the PC-based model was driven by needs of tho' living PRA concept pr,acticed by.NU, this model was l completed on an accelerated schedule so that' e :
| |
| analysis sup' porting the IPE could be performed efficiently (Reference 2-7).
| |
| l-2.2 - Conformance with Generic Letter and Supporting Material K Section 4 of Generic Letter 88-20 identifies three approaches )
| |
| ] that satisfy'theLIPE.- According to Generic Letter 88-20, the l:
| |
| staff will consider those PRAs that follow the PRA procedures
| |
| [ guider (Reference 2-8, NUREG/CR-2300) to be adequate for
| |
| [z.; performing-.the IPE, provided the amendment considers-the most-current severe accident phenomenological issues and the PRA'is- ,
| |
| -based'on the procedures delineated in NCREG/CR-2300, "PRA
| |
| \ . Procedures Guide." The MP3 PRA followed guidance established ;
| |
| by the PRA Procedures Guide. Section 2.1 discusseo six tasks, that were completed by NU to assure conformance with Generic Letter 88-20. It is noted here that in addition to conforming p with Generic Letter 88-20, the submittal, whenever practical,
| |
| ! used the guidance provided in NUREG-1335, " Individual Plant o4unx.oso l4
| |
| | |
| 12 L Examination ' Submittal Guidance." Therefore, the MP3 IPE
| |
| . submittal conforms with the Generic Letter 88-20 and the-supporting material.
| |
| 3.3 General Methodology The_IPE submittal of MP3 is based upon thw MP3 PSS and several tasks performed in order to meet the intents of Generic Letter 88-20. This section discusses briefly the general methodology employed in performing tho MP3 PRA.
| |
| The initial task in the plant analysis was the determination of a complete set of initiating events. Section 1.1,
| |
| " Internal Accident Initiators," of the PSS describes the methods used to select and group initiating events 'and quantify their frequencies. Event trees were used to madel the plant behavior in response to these initiators.
| |
| Section 2.2 of the MP3 PSS, "Evant Trce Analysis," provides detailed information on various aspects, such as success criteria, damage states, system interactions, and recovery actions of event _ tree modeling. Section 2.3 of the MP3 PSS,
| |
| " Plant System Fault Tree Analysis," discusses the overall 1
| |
| methodology used to model and quantify the unavailability of systems.
| |
| The MP3 PRA used the support state methodology'to integrate
| |
| -initiating events, event trees, and fault tree information to generate sequences and highlight plant vulnerabilities.
| |
| Section 3 of this report discusses the methods used for initiating events, event trees, and systems analysis in detail.
| |
| The dependency analysis was considered under several t' asks of the PRA. First, in identifying initiating events, the loss of key support systems were screened as potential initiators.
| |
| The MP3 PRA modeled partial loss of service water, partial and total loss of vital DC, loss of off-site power, and partial osunx.oso !
| |
| ..' ~-
| |
| | |
| i 13 -
| |
| -loss of. vital 1Ac as common cause initiators. System dependencies were examined carefully when the support state r event tree.was created.- Functional dependencies. ware handled via the frontline system event trees. Finally, the common cause f ailure - (ccF) between components was accommodated during system modeling.
| |
| i
| |
| )
| |
| 2.4_ Information Assembly B
| |
| The MP3 PSS documents much of the information used in support of the Individual Plant Examination. Most of the. plant, layout information can be found in the Final Safety Analysis Report. .
| |
| .m L Additional information.on the containment design is found in- [
| |
| Section 4.1 of the PSS.
| |
| At the time the PRA was performed in 1983, insights from the Zion, Indian Point, and Sizewell B PRA studies were used extensively. Insights from PRAs performed by NU such as the Haddam Neck PRA (Reference 2-9) and the ongoing Millstone
| |
| . Unit No. 2 PSS have also been considered when updates were made to the MP3 PRA'models. These-insights include the use of l-l plant-specific failure rate data, methods, the addition of failure modes for various equipment, determination of success criteria, and so on. In addition, the NUREG-1150 analysis of Surry; Unit 1 has been carefully reviewed, particularly from L the'back-end perspective. In order to. respond to Generic L Letter 88-20, NU personnel have also addressed the NRC and contractor reviews of the MP3 PSS.
| |
| Because of the Living PRA program, documented in Section'5.4,
| |
| 'dU'has confidence that the IPE represents the as-built, Briefly, this conclusion is based on: I as-operated plant.
| |
| e full technology transfer of the PRA models from the'PRA contractor to utility personnel, e periodic update and frequent exercise of the PRA models, 04LW7X.06D
| |
| .___r_- :--__-__._____-_________ , ,
| |
| | |
| 'T 14- -
| |
| e full doc antation of model changes, e- systematic screening of past' design changes for incorpor-ation=in model updates, e active participation of PRA personnel on the Nuclear Review Board, e formalized procedure for the review of all new projects, e an Integrated Safety. Assessment Program (ISAP) whereby PRA-insights are used in part for project prioritization,.
| |
| . ready accessLto controlled plant procedures including Emergency.and Abnormal Operating Procedures, and
| |
| # o ready access to plant LERs and Plant Incident Rsports.
| |
| As noted previously in this report, when the MP3.PRA was performed, some two dozen NU engineers, scientists,'and operators participated in the study.. NU and contractor per-sonnel performed plant walkdowns of virtually all systems and.
| |
| plant' areas, including containment'. Plant walkthroughs'and interaction'with plant operations personnel are routine activities whenever' situations at the plant require inputs from the PRA section of NU. NU PRA personnel have also made use of the plant-specific control room simulator on an es-needed basis.-
| |
| REFERENCES .
| |
| 2-1 " Millstone Unit 3 Inservice Test Pump and' Valve Program,"
| |
| Reo.;2,' Northeast' Utilities, June 1988.
| |
| 2-2 M. Khatib-Rahbar, et al, " Review and Evaluation of the Millstone' Unit 3 Probabilistic Safety Study: Containment 04LW7X.00D
| |
| | |
| i
| |
| > 15 Failure Modes, Radiological Source - Terms and offsite Consequences," NUREG/CR-4143, September 1985.
| |
| 2 S. D. Weerakkody,'"MP3 Initiating Event Frequency a Update," Northeast Utilities Calc. File #W3-517-950-RE (Rev. 0),'May 2, 1990.
| |
| 2-4 N.C. Rasmussen, S. Levine, P. J. Wood, " Final Report of-the Level 3-Review Board on the Millstone Point Unit-3 Probabilistic Safety Study," August 1983.
| |
| 5 U.S. Nuclear Regulatory Commission, " Millstone 3 Risk Evaluation Report," NUREG-1152, August 1985.
| |
| l L
| |
| 2-6 A . 1A . Garcia, et al, "A Review of the Millstone 3 Probetilistic Safety Study," NUREG/CR-4142, April-1986.- ,
| |
| 2-7 .S. D. Weerakkody,"IntegratedCoreDamageModeland L Level II PRA Model on PC," Model Version MP3PRA4B, June. ]
| |
| 1990.
| |
| l ,
| |
| 2-8 'U.S. Nuclear Regulatory Commission, "PRA Procedures ,
| |
| . Guide,".NUREG/CR-2300, January 1983.
| |
| i L
| |
| 2-9 '" Connecticut Yankee (Haddam Neck Plant) Probabilistic Safety Study," NUSCO 149,~ February 1986.
| |
| ;I j
| |
| \
| |
| 1 i
| |
| 04LW7X.080
| |
| \
| |
| | |
| 1 16
| |
| - 3. FRONT-END ANALYSIS
| |
| ' 3 .'1 - Accident Sequence Delineation 3.1.1 Initiation Events Sections IV.2 and 1.1 of the MP3 PSS address initiation events analyzed during the PRA. The succeeding sections of this-report summarize the methods used to identify and grcup ,
| |
| . initiating events and discuss how che initiating event frequencies were quantified.
| |
| The determination of a complete set of initiating events was accomplished by a systematic identification of events capable of-presenting significant challenges to plant safety. A Master Logic Diagram was used to screen internal initiating events from full power operation. -An extensive variety of sources-was used to' establish a comprehensive list of tran-sient initiators. The transient initiators were grouped into categories based on plant response, signal actuation, anda
| |
| . systems required for mitigation and subsequent plant related effects.'
| |
| After compiling a complete list of MP3 specific initiators,
| |
| -the initiator.. frequencies were estimated. The initiators, together with their estimated frequencies, are shown in Table 3.1-1. The statistical method of estimating the initiating event frequency is dependent on the amount of available data. .For relatively frequent transients for which
| |
| -adequate plant speciffe data are available (e.g., tu;Line trip, loss of main feedwater), the classical statistical techniques are used to obtain frequency estimates from plant specific data. For relatively frequent transients for which sufficient plant specific data are unavailable (e.g., loss of RCS flow), classical statistical techniques, together with U.S. PWR population data, yield mec66naful results. For those events that are rare or have not ocu n rad (e.g., large LOCA, 04LW71.0ED
| |
| .. , u,
| |
| | |
| 1
| |
| , 17'- 'l l
| |
| medium LOCA)', Bayesian approach was used to estimate the l
| |
| ~
| |
| frequencies.:
| |
| 13.1.2 Front-Line Event Trees .
| |
| i Section 2.2'of the MP3 PSS provides detailed information on !
| |
| the various' aspects of event tree analysis. The paragraphs j that' follow briefly discuss some aspects of-the front-line e'Jent trees. Specifically, the methodology used to generate the front-line event-trees, success criteria, and event tree i branching dependencies are included.
| |
| i L -As a first step, a plant functional event tree model was ,
| |
| l implemented by identifying various safety functions, such as !
| |
| u reactor subcriticality, core cooling, maintaining core coolant ;
| |
| L inventory, and containment cooling. Next, plant systems-which' ,
| |
| h could: accomplish the safety functions were' identified. As:the '
| |
| L ' result, a-plant response matrix which-identifies the specific.
| |
| q
| |
| . plant response for each initiating event was constructed. {
| |
| These matrices. ware developed using the MP3 Final Safety g ,
| |
| ' Analysis Report (FSAR) and functional logic-diagrams. '0he event tree for each. initiator was constructe2 based on.the l functional' event tree, the. plant response matrix, the MP3-h FSAR; and the Westinghouse optimal Recovery Guidelines and-Emergency Contingencies. In addition.to the event tree models, two other models were developed to aid in the analysis of internal accident. sequences. These models, the supporting systems model and the: consequential failure model, were used ,
| |
| i 4
| |
| }. in conjunction with the event tree models. Appendix A includes the front-line event trees for all initiators. For .:
| |
| 9 each initiator, there are eight or less number support states.
| |
| ,c
| |
| ]" The event trees included in Appgndix A contain the first 100 h dominant contributors to core damage. At least one support 3 state for each event tree is also included.
| |
| m I
| |
| The intersystem functional dependencies were explicitly incorporated into the plant system event tree models. If a 04LW71.080
| |
| | |
| 18 system was not required for a particular accident sequence, boccuse of previous system failures or successes, the system was siwply not addressed. For example, in the event trees where high pressure safety injecti.n initially fails, followed -
| |
| by failure of quench spray, contait ..c recirculation spray is not addressed due to the conservative assumption of insuffi-
| |
| ** cient water in the sump. Similarly, if the failure or the success of a system is certaine because of the status of preceding systems, the branching event tree sequence was eliminated. This was often done for signal actuation type failures. ' Anytime a functional intersystem dependence was treated in this manner, a description of what was done is given in the event tree structure discussion sections.
| |
| The' success criteria were developed after an exhaustive search of available plant analysis for various accident sequences.
| |
| Even so, some specific analyses were performed to define system success criteria. In many cases, bounding transient ana3yses were used to develop the required success criteria.
| |
| Tables 2.2.2.2-1 and 2.2.2.2-2 of the MP3 PSS lists the system and event tree success criteria. These tables are reproduced ~
| |
| and attached as Appendix B to this report.
| |
| 3.1.3 Special Event Trees Several special event trees were modeled to accommodate
| |
| , consequential events (ATWS, Consequential LOCA, Consequential Steamline Break Inside Containment, and consequent steamline
| |
| - Break Outside' Containment), and shared support system failures (loss of one train of service water, vital DC power, and vital AC power). These event trees are also 1ncluded in Appendix A. ,
| |
| 3.1.4 Support System Event Tree The MP3 PSS utilized the support state methodology. Hence, a support state event tree which contains dominant support systems as its headings was constructed. Figure 3.1-1 depicts 04DaX,060 ,
| |
| | |
| I 19 I i
| |
| 3 the support state event tree. The support stGua event tree l was modified for the support system related initiators such as.
| |
| . loss of Vital AC or DC. These trees are included in Appendix c. ..
| |
| l Section 2.2.1.3 of the MP3 PSF provides details on the basis of the MP3 support state event tree. The dominant support systems were identified as electric power; service water; the !
| |
| engineered safety features system which monitors plant l parameters, processes the input, and actuates the ESF equip-ment as necessary; and the emergency generator load sequencer system (CGLS) which would load ESF equipment on ESF buses.
| |
| While DC power was not included ir. the support state event )
| |
| trees, DC dependency was reflected through the DC power depen- I dance of ESF. ]
| |
| 4 The support system sequences were classified and grouped into eight support states based upon their effect on representative ]
| |
| systems. Table 2.2.1.3.5-2, which is reproduced in this report as Table 3.1-2, summarizes the simplified support states.
| |
| ~
| |
| )
| |
| 3.1.5 Sequence Grouping and Back-end Interface Groups of possible accident sb7uenc's were classified into ,
| |
| plant, damage states using a disirate set of accident conditions. At Mp3, the accidelt conditions utilized are the type of accident (Large LOCA, St all Loc.\, Transient) , timing l of co'ra-melt (Early, Late), and the operational status of the engineered safety features. Pt.r MP3, in all, there are approximately .11 plant dam 89e states which represent the '
| |
| hundreds of possible &ccident sequences modeled by the system event trees. 1hese damage stater present the plant analysis I
| |
| results in a munageable form, passing on sufficient informa-tion about each accident seguence to permit a naaningful analysis of the containment response. Table 3 2.4-1 of the i 04Df7X 0 2 ?
| |
| | |
| 20 MP3 PSS, included in this report as Table 3.1-3 summarizes the plant damage states used in the MP3 PRA.
| |
| 3.2 System Analysis In summarizing the method used to generate front-line event trees, it was noted that the systems needed to perform safety functions were identified before constructing the plant response matrix. Section 2.3 of the MP3 PSS entitled, " Plant Systams Fault Tree Analysis," includes the followingt a) the overall methodology used to model and quantify the unavail-ability of systems represented in the event trees, b) guide-lines used for fault tree construction, c) system description together with details on how each of the system fault trees was modeled and analyzed.
| |
| 3.2.1 System Descriptions The detailed system descriptions have been documented in the MP3 PSS. Table 3.2-1 summarizes key features of ms3or systems modeled during tse PRA analysis.
| |
| 3.2.2 System Analysis This section summarizes the methodology used to model and quantify the unavailability of systems. In order to assure consistency in the fault tree modeling, a set of basic guide-lines was developed. These guidelines follow the principal methods used in past risk assessment studies, and generally conforming to the principles described in the "PRA Procedures Guide" and the " National Reliability Evaluation Program (NREP)
| |
| Procedures Guide" (Reference 3-1). In certain system specific cases, exceptions were taken to these guidelines based on engineering judgment.
| |
| 04LW7X 06D
| |
| | |
| t 21 3.2.3 System Dependencies system dependencies were investigated during different tasks ,
| |
| of the PRA. First, in establishing a support state model, there was a need to identify the dominant support systems. At this stage, a system dependency matrix was drawn up. This ,
| |
| dependeacy matrix appears as Table 2.2.1.3.1-1 in the MP3 PSS.
| |
| It is reproduced and included in this report as Table 3.2-2.
| |
| The dependency matrix shown in Table 3.2-2 is not ,
| |
| divisionalized in that dependsncies are not shown for separate {
| |
| trains. MP3 is a relatively new plant which began operation in 1996 and the concept of physical and electrical train i separati'n has been strongly adhered to. Therefore, train A support systems support train A front-line systems and train B j
| |
| ; support systems support train B front-line systems.
| |
| 'The PRA Procedures Guide (NUREG/CR-2300) identifies.four intersystem dependencies: a) intersystem functional dependencies, b) intersystem shared-equipment dependencies, c) intersystem physical interactions, and d) intersystem human interactions. The intersystem functional dependencies were explicitly incorporated into the plant system event tree models. The majority of intersystem shared-equipment dependencies were accounted for by using a support state model.- other shared equipment, such as RWST, shared pipe segment in the injection path to the cold legs, and dependency between containment spray and the recirculation system were
| |
| . accommodated using the front-line event trees. For internal 4 events, including common-cause initiators, the physical interaction dependencies are embodied in the success criteria and damage limits for system components. Some of the important physical interaction dependencies were modeled in conjunction with the intersystem functional dependencies. l Ir',arsystem human interactions due to acts of commission and .
| |
| laterdependencies associated with design errors were not modeled in the plant system event trees. However, acts of omission which could affect one or more systems were modeled ,
| |
| 04WTX. '8D
| |
| _ . _ _ _ . . ___.___c._.a....._._..___..____ __ _,
| |
| | |
| 22 by incorporating operator actions as top events in the con-struction of the plant system event trees. In addition, separate operator actions, modeled within one sequence, were evaluated to ensure that no double counting existed. #
| |
| 1 3.3 Sequence Quantification 3.3.1 List of Generic Data ,
| |
| Appendices 2A and 2B of the MP3 PSS provide detailed discussions on the unavailability medel and the failure rate data base used for randon component failures and for test and
| |
| ; maintenance outages. The MP3 PRA was performed prior to the initiation of the commer:ial operation in April 1986 knd plant specific data was unavailable. Therefore, the Westinghouse data base (proprietary) was extensively used for the random failure rates. The Westinghouse data base csvers nuclear ,
| |
| plant operating experience over the time span of 1972 to 1981 and contains over 200 reactor-years of plant operation. In addition, for those cases where little or no nuclear plant data exist, other data sources were reviewed. Some of the scye frequently used data sources were WASH-1400 (keference 3-2), Military Standardization Handbooks (Reference 3-3) , Nuclear Reliability Data Manual (IEEE-500)
| |
| (Reference 3-4), and National Reliability Evaluation Program (NREP).
| |
| A pre-established set of guidelines were utilized to determine which components were modeled for test and maintenance unavailability. These guidelines are summarized in Section 2.3.2.2 of the MP3 PSS, and these guidelines follow the principal methods used in present risk assessment studies and generally conform to the principles described in the "PRA Procedures Guide" (NUREG/CR-2300) and in the National Reliability Evaluation Program (NREP).
| |
| . osunx.oso
| |
| | |
| i L
| |
| 23 3.3.2 Plan % Specific Data and Analysis Plant specific data were used whenever it was reasonable and practical to do so in the fault tree analysis and event tres .-
| |
| quantification when the original MP3 PRA was performed.
| |
| Subsequent updates were made to the PRA in order to accom-modate operaticus experience. This section summarizes the key areas in which plant specific dr.a have been incorporated into the PRA.
| |
| Initiating Event Frequencies The PSS was updated recently to accommodate the frequency of transient initiators which have moderate frequencies.
| |
| Speci.'ically, the frequencies of reactor trip, turbine trip, loss of main feedwater, and primary / secondary power mismatch transients were updated.
| |
| Demand Failure Probabilities The basic event probabilities of demand failures were updated to account for the in-service test program (IST) for pumps and valves as implemented at MP3. The IST pump and valve program for MP3 established a test plan for in-service testing to verify the operational readiness of Class 1, 2, and 3 pumps :
| |
| and valves. The test intervals specified in the IST together with the test frequencies required in the Millstone Station Technical Specifications and the reparted test durations af components were utilized to calculate basic event probabili-ties.
| |
| 3.3.3 Human Failure Data (Generic and Plant Specific) i Section 2.2.3.2 and Appendix 2D of the MP3 PSS discuss numerous operator actions used in the MP3 PSS. This section summarizes the overall methodology utilized to identify and quantify major operator errors. Since probabilities used for osum.no I
| |
| | |
| I 84 :
| |
| operator actions can dominate the results of front-end {
| |
| analysis, a list of major operator actions together with their {
| |
| probabilities is provided as an Appendix 0 to this report.
| |
| The approach used in the MP3 PRA for human performance modeling is contained in NUREG/CR-2815, " National Reliability Evaluation Program Procedures Guide." Following the binary ,
| |
| approach suggested by the NREP guide, the study accommodates t the human errors in the even' trees and the fault tress as followet
| |
| : 1) Event Trees For operator actions appearing in the event trees, !
| |
| the human error probability is calculated con-sidering the time available for the operator to j
| |
| identify the sequence, decide on a course of action ,
| |
| !> and initiate corrective action. Whether the
| |
| ! potential failure modes are dominated by cognitive ;
| |
| cr procedural-based behavior, is also factored in ,
| |
| when' probabilities are selected for human errors.
| |
| Since the probabilities chosen for human errors can dominate the. core damage frequency resulting from the internal event front-end analysis, the proba- .
| |
| bilities of major human errora, directly affecting I the event treo branch probabilities are considered !
| |
| as a significant input to IPE and therefore, are included in Appendix D.
| |
| L
| |
| : 2) Fault Trees ,
| |
| operator errors in restoring equipment for those systems determined to be risk significant are
| |
| ! modeled in the fault trees. In those cases where observation or screening values showed a significant operator contribution to system failure probability, detailed THERP (Technique for Human Error Rate ,
| |
| 04LW7X.080
| |
| | |
| ,f 25' Prediction, NUREG/CR-1278) (Reference 3-5) was performed. Since MP3 was not in operation when the PRA was performed, these analyses were not as detailed as such models would be for an operating <
| |
| plant, but do reflect the conditions existing at the other Millstone plants combined with the canditions applicable to MP3. Table D.2 of Appendix D provides a summary of these human error rates..
| |
| 3.3.4 Common Cause Failure Data The types of common cause failures (CCF) handled during the MP3 PRA are a) support system failures, b) command failures,
| |
| ~
| |
| c) human errors, d) environment conditions, e) intersystem dependency, and f) external events. During the MP3 PRA, the CCF due to external events (fires, seismic) were analyzed using deterministic methods. CCF due to support state (Emergency AC, Service Water) dependency was explicitly modeled via use of support states logic. CCF due to con-ditional dependencies (HPSI and High Pressure Recirculation (HPR)) and shared components (RWST) are explicitly modeled in the event tree structure and via use of conditional probabili-ties in:the quantification process for the event tree. Common cause failure due to command type failures are modeled via incorporating ESF actuation logic commands directly into the support state logic. The residual CCF due to multiple human errors and environmental factors were handled using the Binomial Failure Rate (BFR). The failure rates used for the computations drew from Atwood, et al (References 3-6, 3-7, 3-8). . The CCF of diesel generators, running and standby
| |
| " pumps, air operated valves, and motor operated valves, were incorporated into the system fault trees. Tables 2-C-2 through 2-C,23 of MP3 PSS' summarize the lower (0.05), median
| |
| (.50) and upper bounds (0.95) of the failure rates used during the MP3 PRA for different components, n i 04LW7X 080
| |
| | |
| f, 26 .,
| |
| 3.3.5 Quantification of Unavailability of Systems and {
| |
| runctions A fully integrated PRA model was installed on the personal i i
| |
| computer (PC) as'a significant enhancement to the living MP3 PRA program at NU. This integrated model is based on a software package that quantifies fault trees and event trees t
| |
| and analyzes results. This section addresses the quantifica-tion of the unavailability of systems and functions. Sec- ,
| |
| tions 3.3.6 and 3.3.7 discuss the quantification of support i
| |
| system states and a9quence frequencies, respectively.
| |
| i The fault trees of systems were quantified using the CAFTA software package (Ref.erence 3-9) to generate system unavail-abilities. As needved, system fault trees were combined with fault trees of other systems and/or operator action to generate unavailabilities of functions such as feed and bleed and 16econdary depressurization. Table 3.3-1 is a list of event tree heading probabilities that summarizes the system l
| |
| and function unavailabilities.
| |
| For a given function or system, the unavailabilities vary depending upon the support state and sometimes depending upon the initiators. This variation is attributed to the system or functional probabilities being impacted by the available
| |
| [
| |
| support states and the initiator. ,
| |
| I 3.3.6 Generation of Support System States and
| |
| * Quantification of Their Probabilities [
| |
| Section 3.1.4 of this report briefly describes the support state event tree and the basis for the generation of the ,
| |
| support states defined in Table 3.1-2. This section discusses how the support system states were quantified.
| |
| Several support state event trees exist for the MP3 PRA:
| |
| a) the basic support system event tree for the nonsupport 04LW7X.00D
| |
| | |
| I 27 system related initiators, and b) support system event tree !
| |
| for the support system related initiators. Under category (b) J five support state event trees exist for the following e
| |
| initiators J
| |
| : 1. Loss of Offsite Power. !
| |
| f 1
| |
| : 2. Loss of a Single Service Water train. {
| |
| : 3. Loss of one vital Dc bus. .
| |
| : 4. Total loss of DC power.
| |
| : 5. Loss of vital AC bus 1 or 2.
| |
| : 6. Loss of vital AC bus 3 or 4. .
| |
| Table 3.3-2 summar'izes the probability of each support state l for the above.six special initiators and the nonsupport system -
| |
| related initiators. These support system probabilities were l
| |
| calculated by summing the probabilities' assigned to sequences in the event trees. The PC based PRA software package at NU was used to generate and sum the support state probabilities. ;
| |
| The Support System event trees are included in Appendix C of this report.
| |
| 3.3.7 Quantification of Sequence Frequencies ;
| |
| A PC based PRA software packege (Refe rence 3-9) was used to ,
| |
| -quantify the sequence frequencies. Since MP3 PSS used a support state method sequence quantification, the process b involved simple multiplication of the event tree branch point l< split fractions. However, there were some sequences in which dependencies existed between event tree functions. The soft-ware package, with the appropriate inputs, accommodated these dependencies during quantifications. The sequence frequencies were quantified for all event trees for each of the support 04W7X.02
| |
| * l 28 states. The results of sequence quantification are reported in this report in Section 3.4.2 under " Vulnerability Screening."
| |
| e 3.3.8 Internal Flooding Analysis Section 1.2.4 and Appendix 1-C of the MP3 PSS evaluate the potential of internal flooding as a contributor to core melt accident sequences. This section summarizes the rathodology and the results of the internal flood evaluation.
| |
| 3.3.8.1 Information Assembly The following information was assembled and tabulated in Appendix 1-C of MP3 PSS:
| |
| : 1. Flood Zones -
| |
| The flood zones were chosen to correspond to the existing fire zones developed for analysis of~com-pliance with 10CFR50 Appendix R requirements.
| |
| : 2. Internal Sources for Flooding -
| |
| Internal sources for flooding and the location of the piping associated with these systems were identified and summarized.
| |
| : 3. Location of Components Critical to Operation or Safe Shutdown -
| |
| Components that are considered critical to operation or safe shutdown were summarized against'their location.
| |
| : 4. Matrix of Flood Events -
| |
| For each flood zone, break sources and the additional zones that are affected by progressive flooding were identified.
| |
| 04DUX. 06D
| |
| | |
| I 29 1
| |
| 3.3.8.2 Major Assumptions
| |
| .c 1
| |
| \ r' \
| |
| The following major assumptions were made during ths analysis 8
| |
| : 1. All equipment within a zone that is subject to a flooding event is disabled.
| |
| : 2. Progressive flooding occurs if the accumulation of (
| |
| volume discharged is sufficient to degrade the l integrity of the fire boundary. ;
| |
| )
| |
| i 3.3.8.3 Methodology The methodology used for flooding analysis can be summarized I as follows: ,
| |
| : 1. A qualitative screening was performed to identify i i
| |
| the significant. flood events. The capability of a flooding event leading to transient event and the [
| |
| poscivility of disabling scfe shutdown components as
| |
| : r. result.of the flood were the two factors con-l sidered as significant during the screening i analysis. If the total loss of components within a '
| |
| flood zone would not initiate a transient or a LOCA, or if no safe shutdown equipment would be destroyed, then the flood zone is removed from further analy-sis.
| |
| : 2. Quantitative analyses were performed for those areas ~
| |
| identified significant. For quantitative analysis the WASH-1400 estimate of 2 x 10'3/ year was used as the frequency of pipe breaks greater than 6 inches.
| |
| ~
| |
| 04LW7X.000
| |
| | |
| 30 3.3.8.4 Results Table 3.3-3 lists the significant contributors to Core Damage Frequency (CDF) from internal flooding and a brief description of each significant sequence.
| |
| 3.4 Results and Screening Process 3.4.1 Application of Generic Letter screening criteria Following the guidance provided in NUREG-1335, the following screening criteria were utilized to determine potentially important systemic sequences and system failures that might "
| |
| ler.d to core damage er unusually poor containment performance:
| |
| : 1. Any systemic sequence that contributes 1.00E-07 or more per reac'.or year-to-core damage. -
| |
| : 2. All systemic sequences within the upper 95% of the total core damage frequency.
| |
| : 3. All systemic sequences within the upper 95% of the total containment failure frequency.
| |
| : 4. Systemic sequences that contribute to a containment bypass frequency in excess of 1.00E-08 per reactor year.
| |
| Tables 3.4-1 and 3.4-2 list the core damage frequency by the initiator and the plant damage state, respectively.
| |
| Table 3.4-3 contains the first 100 systemic sequences that lead ta core damage. Each sequence cc.ntains the initiating event, the support state, failed-func11ons or systems, sequence frequency, percent contribution to core damage, and a brief description of the accident scenario. For each sequence, the plant damage state (PD3) is also listed. The plant damage state descriptions presided in Table 3.1-3
| |
| ~
| |
| 04LW71.03D
| |
| | |
| 31 i
| |
| provide the timing of core dadage, and the systems available r for containment heat and radioactivity removal, j The first 100 sequences listed in Table 3.4-3 also contain all 5 l
| |
| systemic sequences that contribute 1.005-07 or more per ;
| |
| reactor year to core damage. Further, this table contains the {
| |
| upper 95 percent of the total core damage frequancy, ;
| |
| Table 3.4-3 also lists the contribution by each core damage ;
| |
| sequence to each of the release classes M1A through M12. !
| |
| 3.4.2 Vulnerability Screening 1
| |
| Information summarized in Tables 3.4-1 through 3.4-6 forms the ,
| |
| bases for vulnerability screening. Table 3.4-1 identifies the Large LOCA, Medium LOCA, and S/L Break Outside Containment (SLBOC) as the three dominant contributors to the CDF contri-buting 14.54%, 18.55%, and 14.69% to CDF, respectively.
| |
| Sequences #1, 2, 6, 11, 15, and 19 listed in Table 3.4-3 are ;
| |
| helpful in interpreting the significant contributions from the '
| |
| Large and Medium LOCAs. The relatively high unavailability of the recirculation function (in the order of 10-2) drives .
| |
| sequence frequencies of sequences #1 and #2. This recircula-tion function unavailability consists primarily of the CCFs of a large number of MOVs that should change state and the opera-tor error failure rates. Sequence #15 and #16 are dominated by the CCFs of MOVs in the service water to recirculation heat exchangers. Sequence #6 and #11 are driven by accumulator failure probability which uses a conservative success criteria- ,
| |
| (3/3 accumulators). ,
| |
| Sequences #3 and #9 in Table 3.4-3 show that the primary depressurization function contributes to the CDF domination by SLBOC. Tne failure probability of primary depressurization accomplir.hed by opening relief valves is 8.16E-02. This ,
| |
| failure probability is high due to relatively high operator error contributions (1.50E-02) and the high random failures j 04DCX.02 I
| |
| | |
| 7._ .
| |
| 4 32 driven by 18 months test interval of solenoid valves (4.5E-02) ,
| |
| and the associated circuitry. ,
| |
| As illustrated in Table 3.4-2 the plant damage states that dominate the CDF are TEC and ALC with contributions of 2.171F-05 (~39% of CDF) and 1.317E-05 (~24% of CDF),
| |
| , respectively. The dominance of ALC sequences is primarily attributed to relatively high recirculation function unavail-ability. This recirculation function unavailability, as I
| |
| pointed out in the previous paragraphs is dominated by a large number of CCFs and operator actions. TEC sequence domination is attributed to the relatively high operator error assigned to bleed operation (1.5E-02) and random failure probabilities {
| |
| of solenoid valves and associated circuitry based on an 18-month test interval.
| |
| The bypass sequences at MP3, summarized in Table 3.4-4 con- ,
| |
| tribute less than 1% to CDF. However, these sequences are ,
| |
| significant since they dominate the risk to public. The high ;
| |
| consequence release e,ategory M1A is contributed to solely by the event V sequencen. The high consequence release category l
| |
| M1B is contributed to solely by the V2 sequences attributed to l
| |
| steam generator tube rupture.
| |
| l Table 3.4-5 and 3.4-6 sunmarize the containment failure frequencies at MP3. As illustrated in Table 3.4-5 release !
| |
| ca.egories (or containment failure modes) M7, M9, and Mi.
| |
| dominate the containment failure frequency. ,
| |
| In order to l interpret Table 3.4-5 to draw insights on risk to public, this table must be examined together with the information on condi-tional consequences associated with each release category.
| |
| This information is available for MP3 since a level III PRA has already been performed. The following conclusions are ;
| |
| drawn based on the hindsights gained from having performed the l level III PRA and the information in Table 3.4-5:
| |
| 04LW7X.06D e . - - - - - --_ - _ e -
| |
| | |
| ~!
| |
| 33 j 1.' Latent cancer fatality risk is dominated by the-release category M7. This is attributed to high I frequency of this release category (55% of the total [
| |
| containment failure probability) compared to the <
| |
| )
| |
| release categories M1A through M5. a l
| |
| 1
| |
| : 2. Early fatality risk is dominated by the release category M1A. Although M1A contributes to approxi- j nately 2% of the containment failure probability, the conditional early fatality risk associated with !
| |
| this release category is very high. Further, the l other dominant categories (e.g., M7, M9, Mll) have long warning times associated with them and there- !
| |
| fore, do'not pose high early fatality risks. l Table 3.4-6 lists the systemic sequences that dominate the containment, failure frequency. As shown in the first column -
| |
| of this table, it is associated with Table 3.4-3 of this ,
| |
| report. At.the top of the list in this table is. sequence #7, a station blackout sequence that contributes 14.3% of tho' total' containment failure probability. This is followed by a f medium LOCA sequence which contributes 8.59% and a consequen- l tial small LOCA sequence which contributes 8.36%. Since M7 ;
| |
| has higher conditional consequences associated with it compared to the conditional consequences of M9 knd Mll, the j true relative significance of these sequences differs from the order in which they are listed. Finally, note that the dominant contributions to core-malt are spread over a large numP'r of initiators.
| |
| 3.4.3 Decay Heat Removal Evaluation ,
| |
| This1 discussion provides a brief evaluation of the decay heat removal function at MP3. The purpose of the evaluation is to examine whether or not risks attributed to the loss of decay heat removal can be lowered in a cost-effective manner. !
| |
| 04LW7X,08D O _ . _ . , _ ____.__ _ _ __ _ _ _ _ _ _ _.
| |
| | |
| 34 ;
| |
| The decay heat removal during power operation (Mode 1) to j Mode 3 at MP3 is accomplished by the following functions:
| |
| : 1. During a medium or large LOCA, decay heat is trans- !
| |
| ferred from the core to the containment using high -
| |
| or low pressure safety injection systems. The ;
| |
| recirculation function which utilizes the Contain-ment Recirculation, Charging and Safety Injection
| |
| ]
| |
| systems, removes heat from the containment. !
| |
| : 2. During transients and small break LOCAs, the !
| |
| Auxiliary Feedwater (ATW) or Main Feedwater (MFW) system removes the secondary heat, and thereby cools l
| |
| the primary. In the event that AFW and MFW fail, j feed and bleed operation, together with the recircu- l lation function are needed for successful decay heat l
| |
| l removal.
| |
| l l )
| |
| Given (1) and (2) above, with regards to successful shutdown ;
| |
| L decay heat removal, the following special features of the ATW and MFW systams and feed and bleed operation, are noteworthy .
| |
| : 1. The ATW system at MP3 has three redundant trains j
| |
| feeding the four steam generators (SG). Two'of the l motor-driven feedwater pumps feed two_SGs each. The diverse turbine-driven auxiliary feedwater pump can t feed all four SGs.
| |
| : 2. The main feedwater (MFW) systen consists of a moto't-driven feed pump in addition to the two steam-driven- 4 MFW pumps.
| |
| : 3. The feed and bleed procedure is in place and is a priority for operator training.
| |
| 04LW7X.C6D
| |
| | |
| 35 The following significant features of the Recirculation related systems are relevant to a shutdown heat removal capability evaluation. -
| |
| : 1. There is redundancy and diversity in all systems, both front-line and support, utilized in the recir-culation function. The success of 1 of 2 contain-ment Recirculation pumps, and= success of 1 of 4 safety Injection or charging pumps is adequate for the successful recirculation function.
| |
| : 2. For the operator actions associated with the recir-culation function, the probability of cognitive
| |
| - error is relatively low due to the relatively long time window available to the operator to decide to initiate the recirculation function. This relatively long time window is attributed to the following At MP3, the RWST has a capacity of 1.2 million gallons. This far exceeds the RWST capacity at similar units. Therefore, the operator has a relatively long time available to decide the need to initiate the recirculation function.
| |
| : 3. The probability of error of commission associated' vith t,he recirculation function is relatively low
| |
| , due to the special control room design feature, specifically the cold leg recirculation array.
| |
| Due to the above special features related to the AFW, MFW and
| |
| - Recirculation system, and the feed and bleed operation for MP3, shutdown heat removal is not perceived as a concern. The following information based on quantitative analysis is provided to back up this NU position on USI A-45.
| |
| 06W7X ,0lD
| |
| | |
| _. . _ _ . ~ . _ . _ _ . . _ . . . . _ _ __ __ _ ._ _ _ . .
| |
| 36
| |
| : 1. Failure probability of AFW is relatively low J (7.12E-05) when all support state systems are avail- I able. This increases to 9.63E-04 when all train B (or A) support systems are unavailable. +
| |
| ]
| |
| l
| |
| ~
| |
| : 2. Probability of loss of AFW and failure to recover MFW or AFW is estimated at 1.00E-06. I Appendix 5 to the Generic Letter 88-20 was reviewed in order to utilize the insights ga.ined from six PRAs sponsored by HRC to address USI A-45. Table 3.4-7 summarizes the Appendix 5 insights and the applicability to MP3.
| |
| 3.4.4 USI and GSI Screening In response to Generic Letter 89-21, NNECO provided informa-tion on behalf of Millstone Unit 3 regarding the status of .
| |
| L unresolved safety issues (Reference 3-10). The NRC staff, in turn, provided its assessment which identified only A-44 and i
| |
| A-47 as USIs yet to be implemented at MP3 (Reference 3-11).
| |
| t l
| |
| In response to Generic Letter 90-04, NNECO provided informa-L tion on behalf of Millstone Unit 3 regarding the status of Generic Safety Issues (GSIs) to be resolved with imposition of requirements or corrective actions (Reference 3-24).
| |
| For USI A-47, Safety Implications of Control Systems, NNECO noted that the steam generator overfill protection design'for i
| |
| MP.? fully meets the intent of the requirements to mitigate mhan feedwater overfeed events. Also, the MP3 Technical Specifications currently require periodic testing of the overfill protection system (see Reference 3-12).
| |
| For USI A-44, Station Blackout, NNECO has noted its intention 1to provide a dedicated air-cooled diesel generator, of suffi-cient capacity, to provide power for the MP3 station loads ;
| |
| (Reference 3-13). This Alternate AC power source will meet onmx,eso ,
| |
| i
| |
| | |
| 37 '
| |
| l the criteria specified in Appendix B to NUMARC 87-00 and will t
| |
| be available within one (1) hour of the onset of the station ;
| |
| t blackout event. :
| |
| i For both these issues, PRA input to the evaluation process was {
| |
| provided. The implementation /close-out of A-44 and A-47 will J resolve all USIs.
| |
| For GSI 99, RCS/RHR Suction Line Valve Intarlocks on PWRs, NNECO plans to remove the interlock during the third refueling <
| |
| outage of MP3.
| |
| \
| |
| NU intends to expand participation in the Integrated Safety Assessment Program (ISAP) to MP3. Experience to date with ISAP for Millstone Unit No. 1 and the Haddam Neck Plant has demonstrated that the program is an efficient and ;
| |
| cost-effective complementary process for enhancing nuclear power plant safety and operation.
| |
| Under ISAP, many generic safety issues have been addressed for Millstone Unit 1 and the Haddam Neck Plant. When possible, ,
| |
| probabilistic safety assessments were performed. Very similarly, those generic safety issues which are open items at -
| |
| MP3 and are suitable for PRA or deterministic approaches will
| |
| ' be' addressed under ISAP.
| |
| i 3.5 External Events The MP3-PSS performed by NU and its contractors addressed the i p
| |
| effects of external events on the plant. Events treated include earthquakes, fires, external flooding, internal flooding, extreme winds, aircraft accidents, hazardous I materials, and turbine missiles. Of these events, the fire and seismic events were found to be important to risk.
| |
| Sections 1.2, 2.5, and Appendix 1-B of the MP3 PSS provide i details of the external event analysis for MP3.
| |
| l 04LW7X 06D
| |
| | |
| f j 1
| |
| 38 i This section summarizes the external event analysis performed !
| |
| as part of the MP3 PRA. For those external events that were ;
| |
| i considered as insignificant contributors to risk, this summary provides the justifications. For fires and seismic events, -
| |
| the methodology, results, and the insights are summarized in i this report.
| |
| 3.5.1 Seismic Initiating Events !
| |
| Several major sections of the MP3 PSS were devoted to the detailed analysis of the risk attributed to seismic events.
| |
| This section provides an overview of the methodology, results, j and insight gained related to the seismic ' event PRA analysis. j 3.5.1.1 Methodology A summary of the different elements of the 4eismic PRA addressed in several sections of the MP3 PJS is representative of the methodology used. This summary ir, provided below:
| |
| : 1. Sections 1.2.1 and Appendix 1-B to the MP3 PSS considers the seismic hazard assessment performed by l
| |
| Dames and Moore (Reference 3-14). The purpose of the seismic hazard study was to provide estimates of the frequency of exceeding various seismically induced ground acceleration levels at the site. , L
| |
| : 2. Sections 2.5.1.1 and the Appendices 2-I and 2-J.of the MP3 PSS address the seismic fragility analysis.
| |
| - The fragility analysis provides the seismicteapacity of structures and equipment that are necessary to l
| |
| mitigate the consequences of accidents caused by an l earthquake. Information provided by Structural !
| |
| L Mechanics Associates (Reference 3-15) was used in this part of the analysis.
| |
| l 04LW7X,000
| |
| | |
| i 39 l
| |
| : 3. Section 2.5.1.2 of the MP3 PSS addresses the seismic plant systems / accident sequence analysis. During i
| |
| this task the seismic induced initiating events and i the resulting plant damage states (PDS) were /
| |
| identified. Fault tree models that mimic event tree '
| |
| paths leading to plant damage states were identified. l I
| |
| : 4. Section 2.5.1.3 of the MP3 PSS describes how the PDS fault trees are quantified and coupled with the seismically induced initiating event frequencies to j
| |
| [ provide core-melt and PDS frequencies. l l
| |
| : 5. Section 7.5.1 of MP3 PSS addressed the impact of ;
| |
| seismic events on the containment and consequence analysis. ;
| |
| L . 1 3.5.1.2 Results j i
| |
| Table 3.5-1 provides the results of the Dames and Moore (Refer,ance 3-14) seismic hazard analysis for MP3. Table 3.5-2 summarizes the contributions to core-melt of different plant damage states from seismic events. Finally, Table 3.5-3 .j provides the contributions to different release categories l 1 from seismic induced events.-
| |
| J Figure 3.5-1 compares the hazard analysis results of Dames and' l Moore (Reference 3-14) with the two most recent seismic hazard studies, LIRL (NUREG/CR-4885) (Reference 3-16) and EPRI (EPRI L
| |
| NP-6395-D) (Reference 3-17). Dames and Moore results are in:
| |
| reasonable agreemant with the EPRI and LLNL results at low l ground acceleration levels. Although there are deviations at L high acceleration levels, the MP3 seismic PRA is considered as ,
| |
| adequate to meet the; intents of the IPEEE (Reference 3-25) due to the following:
| |
| r 1
| |
| 04LW7X.08D
| |
| | |
| w .. .
| |
| 40 l 1. As reflected by the External Event Steering Group (EESG) positions on seismic events, a hazard analysis is not essential to meet the intents of the IPEEE.
| |
| : 2. A comprehensive margins analysis has been performed in support of the MP3 PSS (Reference 3-15). This analysis fulfills the objectives of the EESG recom-mended margins methodology.
| |
| 3.5.1.3 Insights The total contribution to core-melt frequency from seismic events is 9.08E-05 per reactor year. This is approximately a 13% contribution compared to the total CDF attributed both the internal and the external svents.
| |
| Insights on plant seismic vulnerabilities are not drawn based on comparisons between internal events since the uncertainties and the method of analysis associated with the internal events are different from those associated with seismic events.
| |
| Rather, insights on vulnerabilities were derived as the dif-forent phases of the study (hazard analysis, fragility analysis, etc.) progressed. The only vulnerability discovered that required modifications was the replacement of anchor bolts on the diesel generator ol'1 coolers with bolts of stronger material.
| |
| The risk due to seismic events is dominated by the release
| |
| -classes M4 and M6. The release class M4 is attributed to a scenario in which 7. LOCA occurs together with a loss of con-tainment isolation dha to the collapse of the containment crane wall for seismic events far beyond the safe shutdown earthquake (SSE). The dominancy of the release class M6 is attributed to the plant damage state AE which results from a large LOCA followed by early core-melt, ownx.oen l
| |
| | |
| o 41 3.5.2 Fires Sections 1.2.2 and 2.5.2 of the MP3 PSS provide the detailed l fire analysis for MP3. This section will summarize the methodology and the results of fire analysis and derive -
| |
| insights on plant vulnerability and risk to the plant fron l fire events.
| |
| 3.5.2.1 Methodology l
| |
| The major steps performed during the fire analysis were ,
| |
| a) selection of critical fire areas, b) estimation of frequency of fires in critical areas, and c) quantification of fire related core-melt frequencies. Sixty-eight fire areas ,
| |
| were screened to identify the critical fire areas and zones. ,
| |
| The following information related to each fire area / zone-were -i
| |
| * utilized for screening purposes:
| |
| : 1. , Type and amount of combustible materials, estimate l of fuel loading, and expected duration of a fire.. i
| |
| : 2. Potential for the initiation of a transient or a LOCA.
| |
| 3.. .
| |
| Potential for the loss of safe shutdown components, ,
| |
| and 4 i
| |
| : 4. Potential for the propa'gation to adjacent areas.
| |
| This screening analysis was based on information contained in Refekences 3-18 and 3-19.
| |
| The frequency of fires for each critical area was determined ,
| |
| using summary information of the American Nuclear Insurers i (ANI) data bases, as provided in References 3-20 and 3-21, and the number of compartment years as reported in Reference 3-22. ;
| |
| For the areas for which explicit generic data did not exist, r
| |
| 04LW7X.06D i
| |
| , ., ,- __ _ m - .~
| |
| | |
| f i
| |
| 43 l e
| |
| frequencies were estimated using MP3 specific combustible l loadings and related generic data. ,
| |
| l The quantification model used for fire related core-melt ,
| |
| analysis of the critical fire areas can be summarized by its i following features
| |
| : 1. Analysis'of fire detection and suppression in each l area. l
| |
| : 2. Event tres model with stages of ignition, detection, suppression, and propagation to determine the condi- ,
| |
| tional probabilities of partial, total, or no safety losses, given a fire, r
| |
| Some of the significant assumptions used in the analysis are: '
| |
| Fire Detection: ,
| |
| i
| |
| : 1. Unless an area such as the control room is occupied l at all times, no detection is initially credited to humans.
| |
| : 2. Smoke detectors are assumed ,to detect a fire before heat detectors and a're credited with detect ~ng a t fire in the earliest, or discovery stage.' .
| |
| g
| |
| : 3. Human detection could occup because of 2 spurious ,
| |
| signal.if the smoke detectors have not responded.
| |
| I Suncreazion aivan Detection:
| |
| : 1. If the-smoke detectors detect the fire and the fire is suppressed with portable extinguishers, it is assumed that this is an early stage and no safety loss has occurred.
| |
| l.
| |
| / 04LW7X.08D l
| |
| l
| |
| | |
| 43
| |
| : 2. If the portable extinguishers do not suppress the fire and the smoke detectors-have not detected the fire, but either the heat detectors or humans have responded, then suppression may depend on either ..
| |
| automatic systems or hose stations. If the fire is suppressed at this stage, it is assumed that a partial safety loss has occurred, based on the pro-portion of safe shutdown components in the area.
| |
| Fraauanev of Fire:
| |
| : 1. Although some of the fires in the generic data base may have been started, detected, and extinguished by individuals without causing a safety loss, this type of fire may also be under-reported. Therefore, this ..
| |
| scenario is not included.
| |
| : 2. All fires are assumed to increase in size, none are L assumed to self-extinguish.
| |
| : 3. Specific fire scenarios were not modeled, instead, it is assumed that any fire in an area has the potential of causing.a loss of either electrical equipment or cables, regardless of the source of ignition. An oil fire is assumed to cause loss of a pump, and if it propagates, it is assumed to propc-gate to cables or electrical components.
| |
| '3.5.2.2 Results Tables 3.5-4 and 3.5-5 summarize the results of the Fire PRA f
| |
| .for MP3. While the MP3 PSS estimates include 5th, 50th, and 95th percentiles, the table in this report provide only the mean values. Table 3.5-6 lists the contribuhions to different ;
| |
| release categories due to fires.
| |
| 1 3.5.2.3 Insights 04Df7X.0SD
| |
| | |
| 44 p
| |
| The total contribution to core-melt frequency from fire events is estimated at 4.80 x 10-6 per year which is approximately 7%
| |
| of the total core melt frequency. Fires in the Charging and the Component Cooling Pump (CCP) zone, cable spreading room, and the control room contribute more than half of the total core damage frequency contribution from fires.
| |
| Table 3.5-6 provides the contributions to some of the contain ,
| |
| ment failure classes from fires. Based on Table 3.5-6 it may be concluded that the impact on the latent fatality risk due to fires is somewhat significant. Note that frequency of the M7 category is approximately 25% of the contribution to this category from internal events (6.12E-06 per reactor year) and that M7 dominates the latent fatality risk. On the other hand, the contribution from fire events to early fatality risk, which is tied to frequencies of M1A and M1B, is minimal.
| |
| Some of the significant qualitative insights derived from the fire analysis are that (a) at MP3 there are no fire zones in which a fire can lead directly to core-melt, and (b) the fire risk is dominated-by potential fires in the cable spreading room (CB-8), control-room (CB-9), and charging and component cooling pump zone (AB-1).
| |
| 3.5.3 External Flooding Discussion of external flooding is provided in section 1.2.3 of the MP3 PSS.-
| |
| External flooding is considered as an insignificant contributor to risk due to the following:
| |
| : 1. There are no major rivers or streams in the vicinity of MP3, nor are there any water courses on*the site.
| |
| : 2. The North Atlantic coastline has an extremely low probability of tsunamis, osumt.cao
| |
| | |
| t h' 45
| |
| : 3. In view of the maximum storm surge flood levels, the very low frequency associated with these events, and l
| |
| * the plant features designed for protection against I such extreme flooding condition, tidal flooding.is e f an insignificant contributor. l t
| |
| " 4. .In view of the maximum precipitation intensities, '
| |
| * the very low associated frequencies, and the plant l
| |
| ' features Josigned to withstand such conditions, intense precipitation is an insignificant contributor.
| |
| 3.5.4 Extreme Winds and Wind-Generated Missiles This issue is discussed in Section 1.2.5 of the MP3 PSS. l It has been judged that extreme winds do not pose a signifi-cant risk owing to the following:
| |
| : 1. The frequency of exceeding the design tornado wind ;
| |
| I speed of 360 mph at MP3.is calculated to be approxi-mately 5.35 x 10-6 per year. >
| |
| p
| |
| : 2. All MP3 safety-related structures are of reinforced ,
| |
| concrete with wall thicknesses of at least two feet.
| |
| : 3. Compressive stresses induced in the walla due to the weight of the roof and above floor loads reduce the tendency toward the development of tensile stresses
| |
| - in the concrete. 1 It is concluded that tornado missiles do not constitute a significant risk to the safe operation based on the following: ,
| |
| : 1. The estimated frequency of a tornado missile striking a safety-related structure and the frequency of missile-induced backface scabbing of
| |
| +
| |
| 04LW7X.000 l ._. . , . - . . - . - - . . _ - _ . _ _ . _ _ _ _ ________:______________
| |
| | |
| ' - 46-12"-18" reinforced concreta. wall for MP3 are
| |
| .3.5 x 10-5 per year and 1.7 x 10~' per year, respecti'sely. ,
| |
| : 2. Since all MP3 safety-related walls have thicknesses exceeding 24" and since scabbing and penetrations do not=necessarily lead to loss ~of vital components leading to core-melt, the frequency of a missile induced core-melt is estimated to be one to several orders _ of magnitude below 1.7 x 10'' /ysar.
| |
| 3.5.5 ' Aircraft ~ Accidents 1
| |
| This issue is discussed in Section.1.2.6 of the MP3 PSS.
| |
| The risk from airc.'sft accidents due to near-airport' accidents and inflight'accidaits was estimated using information on j e nearby air facilitiss and airways. The MP3 PSS estimated.the j
| |
| - aircraft crash frequency at the MP3 site-from general- 'i aviation,' commercial aviation, and military aviation to be ,
| |
| : 11. 5 x 10-6, 1. 2 x 10-7, ' and 3. 4 x 10'' per year, respectively. ,
| |
| These frequencies were estimated in accordance with-procedures described in the NRC Standard Review Plan (Reference 3-23).
| |
| Since completion of the MP3 PSS, one of the two airports near ll the Millstone sits (Waterford-New London) has closed. There-
| |
| * fore, these risks are even lower.
| |
| It has been-determined that a general aviation crash.cculd ;
| |
| initiate a' loss of offsite power, but a core melt accident would.further require the independent random failure.of onsite ,
| |
| power. In view of the extremely low frequencies associated 1 with this~ scenario, and with commercial and military aircraft
| |
| : crashes, it is concluded that aircraft hazards do not consti-tute a significant contribution to core melt risk.
| |
| i 04LW7X,060
| |
| | |
| 47 1
| |
| .3.5.6 ' Transportation and Storage of Hazardous Materials l In Section-l.2.7.of the MP3 PSS, the' potential for initistion j of a core-melt accident sequence as a result of offsite and e
| |
| .onsite incidents involving transportation facilities and hazardous materials was evaluated. After detailed consider-' I ation of road, rail, and waterway traffic routes, and the potential-for damage to safety-related structure attributed to
| |
| , transportation of hazardous materials, potential missile generations, vapor cloud explosions, and control room r uninhabitability, it was determined that transportation accident pose an insignificant risk. .
| |
| Storage of large quantities.of hydrogen and liquid chlorine j were viewed as' potential contributors to risk. (Since publi-cation.of the MP3'PSS, the chlorine system has been replaced '
| |
| by a less hazardous sodium hypochlorite system.') After evalu-ations, the risk due to storage of these hazardous materials were found to be insignificant.
| |
| m 3.5.7 Turbine Missiles This issue is' addressed in Section 1.2.8 of the MP3 PSS.
| |
| The frequency of significant damage to safety-related struc- i tures and equipment from turbine missiles was estimated to be '
| |
| approximately 7.5 x'10-9 for the 30-year liist-me of the plant.- This above number was generated after estimating the , ,
| |
| frequency ~of: turbine missile generation due to the'two.plau-p sible aechanisms,< ductile fracture and stem corrosion.- This L
| |
| frequency was coupled with estimates of conditional probabili-ties of turbine missile striking a critical structure and the W conditional proSability of the striking missile to cause significant damage.
| |
| l l
| |
| l 11- 0'LW7X,04D 1
| |
| | |
| um v
| |
| 48 g
| |
| REEEEENCES 3-l' ' R A. Barij et al, " National Reliability Evaluation Program," NUREG/CR-2815,-1984.
| |
| 3-2-' U.S. Nuclear Regulatory Commission, " Reactor Safety Study: An Assc-sment of Accident Risks in U.S. Commer-cial~ Nuclear Power Plants," WASH-1400, October 1975.
| |
| 3-3 Military Standardization Handbook--Reliability Prediction of Electronic Equipment (MIL-HDBK-217C).
| |
| 3-4 IEEE Guide to the Collection and Presentati0n of Electrical, Electronic, Sensing Component, and Mechanical Equipment. Reliability Data for Nuclear-Power Generating Stations, IEEE Std. 500-1984.
| |
| 3-5' A. D. Swain, and H. E. Guttmann, " Handbook of Human
| |
| - Reliability Analysis with Emphasis-on Nuclear Power Plant
| |
| ~ Applications," NUREG/CR-1278, August 1983. - .
| |
| 3-6' C.: L. Atwood, and J. A.Stevenson, "Commen Cause Fault Rates for Diesel Generators: Estimatea Based on. Licensee Event Reports at U.S. Commercial Nuclear Power Plants 1976-1978," U.S.N.R.C.,-NUREG/CR-2099,'Rev. 01, issued June 1982.
| |
| 7 C. L. Atwood, " Common Cause Fault Rates for Pu=ps:
| |
| -Estimates Based on Licensee Event Reports'atLU.S.
| |
| Commercial Nuclear Power Plants, January'1, 1972-through Septerber 30, 1980", E.G. and G. Idaho' EGG-EA-5239, Rev. 01, issued August'1982.
| |
| 3-8 C. L. Atwood, and J. A. Stevenson, " Common Cause Fault Rates for Valves: ' Estimates Based on Licensee Event Reports at U.S. Commercial Nuclear Power Plants, i
| |
| 04LWTA.00D
| |
| | |
| 3 49 1976-1980^," E.G. and G. Idaho, EGG-EA-5485,- Rev. 01, issued ~ September 1982. <
| |
| 3-9 "A Computer Program Package'for PC-Based PRA Work- .-
| |
| stations," Science Application Inte.rnational Corporation, ;
| |
| 1987. !
| |
| 3-10 E. J. Mroczka letter.to U.S. Nuclear Regulatory Commis-sion, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, Response to Generic '
| |
| Letter 89-21, Status of Unresolved Safety Issues,"' dated
| |
| -November 27, 1989.
| |
| 3-111D. H. Jaffe letter to E. J. Mroczka, "Unimplemented Unresolved Safety Issues at Millstone Unit 3," dated '
| |
| March 13, 1990.
| |
| ,; t 3-12 E. J.'Mroczka letter to U.S. Nuclear Regulatory Commis-sion, "Haddam Neck Plant, Millstone Nuclear Power f Station, Unit Nos. 1, 2, and 3, Response'to Generic Letter 89-19, Request for Action Related-to Resolution ;
| |
| Unresolved Safety Issue A-47," dated March 27, 1990.
| |
| 3-13 E. J.: Mroczka letter to T. E..Murley, "Haddam Neck Plant, r Millstone Nuclear = Power Station, Unit Nos. 1, - 2 , and:3, Response to Station Blackout, Additional Information,"
| |
| dated March 30, 1990, i
| |
| 3-14 " Seismic' Hazard and Design-Spectra at Millstone Nuclear
| |
| ' Power Plant Unit 3," Dames and Moore,_ October 26, 1983.
| |
| )'
| |
| 3-15 " Seismic Fragilities of Structures-and Components'at the Millstone 3 Nuc?. ear Power Station," SMA 20601.01-R1-0,
| |
| > Structural Mechanics Associates, March 1984."
| |
| 3-16 U.S. Nuclear Regulatory Commission, " Seismic Hazard ,
| |
| l Characterization of the Eastern United States: Compara- i 04LW7X,080 m
| |
| , p - , - + .- , ,
| |
| | |
| 4 m
| |
| 50-
| |
| [4 tiveLEvaluation of the LLNL and'EPRI Studies," NUREG/CR-4885, 1987..
| |
| 3-17 Electric Power Research Institute, "Probabilistic Seismic $
| |
| Hazard-Evaluation at Nuclear Plant Sites in the Central and Eastern' United States: Resolution of the charleston Earthquake Issue," EPRI NP-6395-D, April 1989.
| |
| 3-18 Millstone Nuclear Power Station Unit 3 Fire Protection Evaluation Docket No. 50-423, June 1977..
| |
| 3-19 Appendix R Compliance Evaluation for Fulfillment of
| |
| , Requirement of Fire Protection Rule, 10CFR50.48 and Appendix R to 10CFR50, provided by Stone and Webster, August 1982.
| |
| 3-20 Hockenbury, R., Yeater, A., " Development and Testing of a Model for Fire Potential in Nuclear Power Plants,"
| |
| (NUREG/CR-1819) Rensselaer Polytechnic Institute, Troy, N.Y., November, 1980.
| |
| 3-21 Garlington, D., " Fire Incident Data Analysis System" Masters Project, Rensselaer Polytechnic Institute, Troy, N.Y., May 1978.
| |
| 3-22 Kazarians, M., Apostolakis, G., " Fire Risk Analysis for Nuclear Power Plants," (NUREG/CR-2258) UCLA-ENG-8102, University of California at Los Angeles, Los Angeles, ,
| |
| California, May 1981.
| |
| 3-23 Nuclear Regulatory Commission, " Standard. Review Plan,"-
| |
| Section 3.5.1.6, Rev. 2, July 1981.
| |
| 3-24 E. J. Mroczka Letter to U.S. Nuclear Regulatory Commis-
| |
| -sion, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, Response to Generic Letter 90-04, Request for Information on the Status of O O 7X D**
| |
| l
| |
| | |
| 51 GSIs Resolved with Imposition of Requirements or Correc-tive Actions," dated June 27, 1990.
| |
| 3-25 Nuclear Regulatory Commission, " Procedure and Submittal i Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities,"
| |
| NUREG-1407 (Draft for Comment), August 1990.
| |
| 04LW7X.080
| |
| | |
| i 52 4 'BACK-END ANALYSIS 1
| |
| The MP3 Probabilistic Safety Study-consists of a full Level 2 (and Level 3) PRA including externally initiated events. A >
| |
| comprehensive analysis of the response of the MP3 containment to severe accidents was ptrformed in 1982-1983. In this section, a brief summary of the MP3 back-end analysis is providad. A summary can also be found in Reference 4-1.
| |
| In addition, Section 4.8 summarizes the results of additional back-end analyses performed s!.nce the MP3 PSS submittal.
| |
| ~
| |
| ' Finally, in Section 4.9, a comparison is made between the Millstone Unit 3.containmer.c features and those of Surry Unit 1 in light of the NrJREG-1150 severe accident risk evalua-i tion report.
| |
| l N
| |
| l -4.1 Plant Data and Plant ~ Description l
| |
| Section 4.1 of the MP3 PSS provides a complete description of the relevant plant parametern and containment features as they l relate to the back-end analysis. Included are emergency core ..
| |
| cooling system and containment spray' flow rates e.nd para-meters, containment dimensions,' material properties, contain- -
| |
| ment structure, heat sinks, basemat concrete composition, ]
| |
| E drawings o* the containment layout, and information on the 1
| |
| ' lower-reactor navity design. A simplified drawing of the lower reactor cavity configuration is provided in Figure 4.1-1 of this report.
| |
| l The significant features of MP3-from a severe accident .
| |
| response perspective include the- following:
| |
| l
| |
| * Reactor power of 3,411 MWt. l l
| |
| *. DC powered pressurizer PORVs (total of two). j 1
| |
| i 1
| |
| 04LW7X.08",
| |
| _ . - _ _ - - ._ - = - - -
| |
| | |
| W 93 y ,
| |
| e ' Emergency core cooling system accumulators (four total) discharge setpoint at approximately 650 psig.
| |
| i
| |
| * Refueling water storage tank capacity of 1.2 million gallons.-
| |
| o Loop isolation valves on the four loops.
| |
| e -
| |
| Two containment quench spray pumps of 4,000-gym each.
| |
| e Four containment recirculation spray, pumps of 3,950 gpm
| |
| ~
| |
| each and heat exchangers (all outside containmer.t).-
| |
| 6 e Containment net free volume of-approximately 2.3 x 10 ft 3. ,
| |
| r e Containment operating pressure of approximately 10-11 psia; i.e., subatmospheric (a_ change has been proposed to-increase the allowed pressure to 14.0~ psia).
| |
| a e- Cor.hainment design pressure of -45 psig with~ ultimate failure pressure estimated'at 117 psig.
| |
| y<
| |
| * A lower reactor cavity-which does not.readily' communicate with the containment sump and would be " dry".at vessel failure;.however, because of the large volume of the RWST,.there is the possibility of late spillover of water
| |
| 'from the' containment sump into the lower reactor cavity F for many severe accident sequences.
| |
| * A reactor cavity geometry which may inhibit the dispersal of the molten core debris during high pressure melt ejection scenarios (see Sections 4.8 and 4.9 for
| |
| ;j g
| |
| ( additional discussion).
| |
| fit *- Basalt-based concrete; i.e., silicacus.
| |
| 04LW7X.000 1
| |
| | |
| a 54 e :Jur containment internal structural design which precludes accumulation of hydrogen qas=in a local area.
| |
| 4.2 Plant.Models and Methods for Physical Processes <
| |
| x
| |
| -The back-end methodology which has been employed is similar to that used in the~ Zion and Indian Point PSSs (References 4-2 and 4-3). A full description of the physical processes, severe accident phenomena, and models used in the analysis can be found.in Section;4.3 of the MP3 PSS. Additional analyses related to debris quenching phenomena, molten core concrete interaction, hydrogen deflagration, heat transport from the-reactor cavity, steam production phenomena, reactor vessel failure analysis, other ex-vessel phenomena, and metal water reaction.phonomena can be found in the appendices to section 4 of the MP3 PSS.
| |
| A description of the severe accident computer codes used'in the~ analysis-is p'covided in Appendix 4-E to the MP3 PSS. In summary, the MAT.CH computer code (Reference 4-4).was used for modeling,in-vessel severe accident phenomena. Reactor pres-sure vessel blowdown at'high: pressure was modeled using the
| |
| = Westinghouse based-MODMESH computer program. CORCON-MOD 1 (Reference 4-5) , with modifications as- described in Appendix 4-C to the MP3 PSS, was used to model molten core-concrete interaction. Westinghouse COCOCLASS9.(Reference 4-6) was used to model containment thermal response.
| |
| 4.3 Bins and Plant Damage States
| |
| : j. Section 3.1.5 of this report describes the plant damage states (PDS) that are used in-the MP3 PSS. For the purposes of performing severe accident progression analyses, these plant damage states were-bjnned into ten containment response-classes. Table 4.3-1 of this report summarizes'these ten containment classes. Since no accident progression analyses were performed for the containmwnt bypass sequences (inter-o m ax.oso
| |
| | |
| . - - .- - - - . - . - - . - . . . - . . - . . ~ . . . -
| |
| 55
| |
| } < l E facing system:LOCAs, corn melt initiated by steam' generator ;
| |
| l
| |
| . tube rupture with direct release to'the environment, etc.),
| |
| they are=not assigned to any containment response class. ]
| |
| c
| |
| *i 4.4 Containment Failure Characterization A full' structural analysis of the' Millstone Unit 3 containment ~ I was performed to identify containment failure modes and j failure pressures. The analysis included the consideration of.
| |
| the containment-shell, basemat, major penet.:ations, contain-
| |
| . ment equipment-hatch, and personnel air lock. Variability-in material properties, analysis methods, and construction were also accounted for.- A complete description can be found in Appendix-4-F to the MP3 PSS.
| |
| .The various containment failure modes and corresponding failure pressure and variances were then combined.to give a containment failure distribution curve. This analysis is described in detail'in Appendix 4-G to the MP3 PSS. A concise summarv of t2un analysis can also be found in Reference 4-7.
| |
| l The .-adian failure pressure.isfestimated to be 117'psig. The ;
| |
| fifth'and ninety-fifth percentile values for containment.
| |
| failure pressure are 97 and 132 psig, respectively.
| |
| ( 4.5 Containment Event Trees b'
| |
| L A complete description of the containment event trees can Tua
| |
| -found in Section 4 of the MP3 PSS.- Specifically, Section 4.2 of the*PSS along with Appendix 4-A of that report describes
| |
| -tho' containment. event tree (CET) structure. The CET is
| |
| . divided into six time frames consisting of a total of 17 nodes.- The time frames represented in the CET are:
| |
| 1 - From accident initiation to the beginning of H
| |
| l core degradation.
| |
| l l
| |
| 04DOX.080 IL L
| |
| | |
| 56 ,
| |
| t II '-- -From the beginning of degradation to the. start
| |
| -of significant accumulation of core debris in
| |
| -the reactor vessel lower plenum. ,
| |
| i III - From_the start of.significant accumulation of core debris in the vessel lower plenum to just before reactor vessel failure.
| |
| 4 IV - Reactor vessel failure and reactor vessel depressurization.
| |
| V - From the end of reactor vessel depressurization to approximately four hours following vessel failure.
| |
| VI .
| |
| From four hours after vessel failure to approx-imately one day.
| |
| Beyond this point, thi containment is considered to be at the-same state; i.e., failed or intact) as during Time Frame VI.
| |
| 'The containment event tree is oriented towards core and con-stainment'phenomenological events =which affect either contain-mant' failure probability or fission product release character-Jy istics', The CET nodes address the issues of containment failure probability during the various time frames, the potential for hydrogen burn, debris quench ex-vessel, and long-term debris lcoolability. The assignment of the various CET endpoints.to radionuclide release categories can be found in''Secticn 4.6 of the PSS.
| |
| \
| |
| .6-4 Accident Progression and CET Quantification The degraded core and containment response analysis is
| |
| -described in detail in Section 4.4 of the MP3 PSS. The PSS provides a description of the cases analyzed, major phenomenological issues affecting the results, sensitivity i
| |
| 04LWX.0 2
| |
| | |
| .1 o :
| |
| 1 studies', and so on. Reference 4-1 also provides a good-summary of'the plant response for various classes of severe I accidents. *
| |
| ' Table-4.6-1 of this report, reproduced from the MP3 PSS, ,
| |
| provides"the best-estimate accident chronology for the various plant damage states. This-information, along with the con-tainment pressure calculations for,the various accidentl [
| |
| response classes, is used extensively in the CET quantifica-tion.
| |
| is Quantification'of the CETs is described in Section 4.7 of the PSS.- A sample of the CET uncertainty analysis is provided in l Appendix 4-N of the PSS.
| |
| Table 4.6-2 of this report, reproduced from Reference 4-1,
| |
| . summarizes the results of the CET quantification using best- l estimate probabilities. As can be seen, containment integrity
| |
| . can generally be assured only if both the containment re-
| |
| ~
| |
| circulation spray and quench spray-systems are available.
| |
| While both these systems are effective in providing contain-ment-heat removal,' quench spray system. operation ensures that
| |
| ~
| |
| most of the RWST contents of 1.2 million gallons will be ,
| |
| available in the containment-sump. This water, along with the l
| |
| RCS inventory,'can then be potentially available to spillover into the lower reactor cavity to provide debris cooling after
| |
| ~
| |
| 1.24 million gallons collected in the sump. The recirculation spray' system is necessary to-ensure long-term containment heat removal capability. In most accident scenarios, hydrogen
| |
| _ generation-by molten core-concrete interaction and the likeli-hood of hydrogen burns are found to be high.
| |
| An;important insight is the low likelihood of containment isolation failure because of operation at subatmospheric ,
| |
| conditions. A pre-existing failure of containment isolation is very unlikely since containment subatmospheric conditions could not be maintained with a major breach in containment.
| |
| L 04LW7X.080-i
| |
| | |
| bfD '
| |
| 58=
| |
| The possibility of containment ' solation failure at-the time-Eb of the; accident is also low because of the low probability of operation of the containment vacuum pump coincident with an accident and failure of double containment isolation valves.-
| |
| ; 7 While. exact quantification of this failure mode is-arguable, containment isolation failure is judged to be an unlikely-failure mode in comparison to other' pressurized water
| |
| ~ "
| |
| reactors.
| |
| L With the exception of severe seismic events, the containment analyses including containment event tree structure and quan-
| |
| -tification are identical for internal and external events.
| |
| ' it That is, the conditional probabilities of the various release b ' categories are identical for the various plant damage states; i.e., similar-containment matrix.
| |
| The: sole _ exception is the V3 plant damage state for seismic events. As described.in-Section 4.7.5 of the'MP3 PSS, V3
| |
| ~
| |
| ~ representscthe plant damage state (PDS) for very severe seismic events whereby the polar crane support wall inside j
| |
| Leontainment is postulated to collapse. -Such'a failure is w assumed,to result in a large LOCA, a loss of emerge'ncy core cooling injection capability due either to the size of the breakLor direct loss of safety injectian' components. This.PDS is' assigned to the seismic M4: release category. Table 4.7.5-1 ,
| |
| of the MP3 PSS provides the containment matrix for external events.
| |
| 4'. 7 Radionuclide Release Characterization The radionuclide release characterization, Hor source term,'is
| |
| * described in full in Section 5 of the MP3 PSS. In summary, the CORRAL-II code (Reference 4-8) had been the principal tool utilized in arriving at the source term estimates. A9 described in' Appendix 5-A to the PSS, some modifications were made to the CORRAL-II code. The code calculates gap release, melt release,' and vaporization release. Two sets of release on mx.oso l h
| |
| Ee
| |
| | |
| _,_ q..,. + --..an - , -a _ - -+e - - . . . .
| |
| e.,
| |
| ,' (:" )
| |
| i 59 0
| |
| - fractionsifor, iodine, as predominantly CsI and assuming elemental. iodine (I ) 2, were developed for the . PSS.
| |
| l
| |
| ' A total of 13 release categories are used in the PSS for : )
| |
| internally initiated events. They are described in~
| |
| Section 5.1.2.3 of the-FSS,-and a brief definition'is given in !
| |
| Table 4.7-1 of this report. The. release category summary information including-release start and warning times, release-
| |
| ~ duration and energy, as well as release' fractions, are i
| |
| , - provided in Table 4.7-2 of this report for the CsI form of
| |
| : i. dine.
| |
| The methodology'used in CORRAL-II is widely. acknowledged to not. appropriately account-for primary system retention.of'
| |
| . fission products as'well as particle agglomeration and set-tling within; containment. Because of this, the method of discrete probability distributions (DPDs) was used'in the PSS to quantify the uncertainty in source ter:m (see Appendix 5-B to theLPSS). These DPDs had the net effect of greatly widening the uncertainties in offsite. consequences, and explain the large variation in the final risk curves for the mean, median, and 90th percentile levels.
| |
| * Based on.the offsite consequences analysis (see Section.6_of j the.PSS), only-the-M1A and M4_ release categories result in high: conditional probability of tens to hundreds of early fatalities. The release categories M1A through M7, on the other hand, result in significant latent health effects.
| |
| t 4.8 Follow-On Back-End Analyses Through participation in industry activities and confe'rences, 1 NU-personnel have kept current in the understanding of severe accident phenomena and issues. NU senior management was active on the IDCOR Steering Group, and virtually all IDCOR technical reports were reviewed by cognizant NU personnel. NU personnel were instrumental in the formation of the MAAP Code 04LW7X.080
| |
| | |
| -_- - .. .. . .. _ . - ~ . - _. - -
| |
| I , ;
| |
| 60 '!
| |
| ? >
| |
| 1 4
| |
| - U30r'o Group, ond~ currently havo m3mbarchip on tho Sto3 ring
| |
| . Committee.
| |
| Since completion of the MP3 PSS, a number of-follow-on back- I cnd analyses have been performed. A scoping analysis was performea to= compare the risk impact with and without a deliberate hydrogen-igniter system, and a dry versus a flooded
| |
| . reactor cavity configuration in the MP3' Containment
| |
| -(Reference 4-9). A hydrogen igniter system was not found to
| |
| - be advantageous. Although the igniters would reduce the risk associated with certain low probability sequences, other ,.
| |
| sequences were identified in which the igniters could
| |
| ~
| |
| conceivably increase'the risk to the public. The dry cavity configuration was also determined to be acceptable. .
| |
| Vaporization of water in a flooded cavity could lead to earlier containment overpressurization in certain accident L sequences. H l
| |
| L An analysis of containment spray recovery and its effect on- j containment integrity and fission product source term was also performed (see Reference 4-10 and additional discussion in Section 5.3.3'below). The results indicated that while deinerting containment owing to spray recovery could result in hydrogen burn, the fission product removal characteristics of
| |
| ( even short term spray operation more than compensate for the l l impacts of a potentially earlier containment failure time. <
| |
| In response'to a. recommendation from the NRC, NU evaluated the
| |
| .fsasibility of adding an AC-independent containment spray system at MP3 (References 4-11 and 4-12). Several design optior.s were identified. However, the potential risk reduc-
| |
| ' tion from the AC-independent containment spray design was 1
| |
| found to be very low, and no change has been implemented. l l
| |
| l Some limited scope benchmark and analyses using the MAAP code i have also been performed. Figure 4.8-1 shows a comparison of the MAAP code prediction versus the computer code series used 04Df7X. 08D i
| |
| | |
| k
| |
| ( 61 n
| |
| in the.or2ginal-PSS. Despite the fact that no Ofine tuning"-
| |
| of the coda' para meters was performed, the agreement for the 1
| |
| TMLB'-sequence is shown to be quite good. i Based on insights from industry-sponsored and NRC-sponsored i research, the two most risk significant issues at MP3 from a back-end analysis viewpoint are a) the frequency and pathway for release from an interfacing systems LOCA in the RHR suc-tion line and b) the potential for early containment failure owing,to direct containment heating (DCH).
| |
| As part of a previous update to the PSS,.the frequency of an' '
| |
| intersystem LOCA was reevaluated. Some credit for the third l p series MOV (rated at lower design pressure than the upstream L MOVs) was taken. Additional failure modes, such as those due-to human error, were considered in the RHR system autoclosure interlock removal report (Reference 4-13). Furthermore, a valkdown of the RHR system outside containment indicated that the RHR pump cubicle would likely flood during the postulated i ISL. Based on engineering judgment, there is'a reasonable .
| |
| likelihood of the break remaining under water. However, no detailed analysis has been performed.
| |
| f-NU has also closely followed the evolving. concerns regarding the'DCH issue. The MP3 PSS, as well as IDCOR Technical-i Report 85.2 (Reference 4-14), note that, at.least qualita-tively, the cavity configuration at MP3 is expected to retain ,
| |
| essentially all of core debris during a high pressure melt ejection severe accident. This would minimize the DCH con-cern. However, small scale experiments at Brookhaven National Laboratory appear to contradict this conclusion for the Surry plant (Reference 4-15). For this reason, NU is supportive of additional industry sponsored research to address the HPME/DCH issue and has identified the necessary experiments and analyses'which might shed further insight at MP3.
| |
| l r 04W7X . 08D
| |
| | |
| 62 4 . 9' Insights from NUREG-1150 I
| |
| The Surry Unit 1 nuclear power plant was one of the five-reference plants studied in NUREG-1150 (Reference 4-16).
| |
| Despite the large difference in commercial 1 operation date for MP3 and Surry Unit 1, these two plants have similar features with regard to containment and back-end characteristics.
| |
| Table 4.9-1 highlights the similarities.- In particular, the
| |
| . gg subatmospheric~ containment design, reactor ca'vity configura-tion, and concrete type are generally the same. The greatest difference is in-the refueling water storage tank capacity, where MP3's is over three times the volume of Surry Unit l's.
| |
| Because of these similarities, many of the insights from the NUREG-1150 effort are applicable to MP3, as well. For this reason, NUREG-1150 and supporting documents were reviewed (References 4-15 to'4-19). Major insights based on these studies, principally References 4-16 and 4-19, have been
| |
| -summarized in Table 4.9-2. The applicability of these insights to MP3 on a qualitative basis is also described.
| |
| With the major exceptions of t.4e incore instrument room / seal table room layou't and the potential for reflood of the reactor cavity from the containment sump, almost all the NUREG-1150 insights for Surry are applicable to MP3. There remains a great deal'of uncertainty with regard to HPME and DCH, as is the case at Surry.
| |
| REFERENCES 4-1 D. A. Dube, R. J. Lutz, Jr., " Containment Response During Severe Accidents at Millstone Unit 3," Proc. Intl.'Mtg.
| |
| on LWR Severe Accident Evaluation, Cambridge, MA, August 28, 1983.
| |
| 4-2' " Zion Probabilistic Safety Study," Cormonwealth Edison Company, September 1981.
| |
| 04LW7X,06D
| |
| | |
| i 63 4-3 " Indian Point:Probabilistic Safety Study," Consolidated.
| |
| Edison Company, Power Authority of the State of New York,.
| |
| 1982.
| |
| : 4. R. O. Wooten, H. I. Avci, " MARCH Code Description and User's Manual," NUREG/CR-1711 (BMI-2064),-Battelle Columbus-Laboratories, October.1980.
| |
| 4-5 J. F. Muir, et. al., "CORCON-MODit An Improved Model for Molten-Core / Concrete Interactions," NUREG/CR-2142 (SAND 80-2415), Sandia National Laboratories, July: 1981.
| |
| 4-6 F. M. Bordelon, E. T. Murphy, WCAP-8327 Containment
| |
| . Pressure Analysis' Code-(COCO), July 1974.
| |
| 4-7 J._H. Bickel, "Probabilistic Analysis of Millstone Unit 3 Ultimate Containment Failure Probability.Given High=
| |
| -Pressure,". Proc. Intl. Mtg. on LWR Severe Accident Evalu-
| |
| ~
| |
| 'ation, Cambridge, MA, August 28, 1983.
| |
| 4-8 R. J. Burian, P. Cybulskis, " CORRAL II Users Manual,"
| |
| BattelleLColumbus Laboratories, January 1977.
| |
| 4-9 J.'L. Maneke, D. A. Dube, " Effects of Hydrogen Burns and' l Flooded Reactor Cavity on.Public Risk,"~ presented at the Second Workshop on Containment Integrity, Washington, DC, June 13, 1984.
| |
| ~4-10 B. G. Holmes, L. A. Wooten, D..A. Dube, "An Analysis of Containment Spray Recovery and Its Effect on containment Integrity and Fission Product Source Term," Third Ihtl.
| |
| Top. Mtg. on Reactor TH, Newport, RI, October 15-18, 1985'.
| |
| 4-11 " Millstone Unit 3: Evaluation of AC - Independent Con-tainment Spray System," NUSCO 154, Northeast Utilities Service Company, February 1987.
| |
| 04LW7X.06D
| |
| | |
| 64 i
| |
| '4-12 D._A. Dube, Y. F.'Khalil, " Assessment of containment Spray Designs for Station Blackout Mitigation,"-Fourth '
| |
| -Workshop on Integrity of Containments for Nuclear' Power
| |
| . Plants,, Washington, DC, June 15-17, 1988. ,
| |
| 2 4-13'"RHR Autoclosure Interlock Removal at Millstone Unit 3,"
| |
| NUSCO 170,. April U,90. ,
| |
| i 4-14.IDCOR Technical heport 85.2, " Technical-Support.for Issue !
| |
| Resolution," July 1985. ~5 4-15 N. K.-Tutu, et. al. " Debris Dispersal from Reactor Cavities During High-Pressure Melt Ejection Accident >
| |
| ' Scenarios," Brookhaven National Laboratory, NUREG/ 1 CR-5146, BNL-NUREG-52147, July 1988.
| |
| 4-16 " Severe Accident Risks: An Assessment for Five U.S.
| |
| Nuclear Power Plants," NUREG-1150, Volume 1, Second Draft-
| |
| .for Peer Review, June 1989.
| |
| 4-17 A. S.. Benjamin, et. al., " Containment Event Analysis for Postulated Severe Accidents: Surry Power Station, Unit 1," NUREG/CR-4700, SAND 86-1135,. Volume 1, Draft; February 1987.
| |
| 4-18 A. S. Benjamin, et. al., " Evaluation of Severe Accident '
| |
| Risks and the Potential for Risk Reduction: Surry Power Station, Unit 1," NUREG/CR-4551, SAND 86-1309, Volume 1, draft, February.1987.
| |
| 4-19 R. J. Breeding, et. al., " Evaluation of Severe Accident Risks: Surry bnit 1," Volume 3, Part 1, Revision 'l (draf t) , NUREG/CR-4 551, SAND 86-1309, June 1989.
| |
| r 04LW7X.080
| |
| | |
| 65 )
| |
| 5 UTILITY PARTICIPATION'AND INTERNAL REVIEW TEAM I
| |
| This section~briefly describes the organization of-the '
| |
| Millstone-Unit 3 PSS project' team in 1982-83, the current -
| |
| organization, and the utility-sponsored ..sd NRC-sponsored reviews of'the PSS.
| |
| 5.1 IPE Program Organization .
| |
| 1
| |
| - Section III.3 to the Introduction and' Summary of the MP3,PSS describes tihe organization of the study team. Figure.5.1-1 l shows the overall organization. As noted'in the PSS, Northens't Utilities-provided the overall technical management of-the study and'also performed many of the analyses. There )
| |
| was active' participation by both the engineering and opera- ]
| |
| tions organizations'at NU.
| |
| L
| |
| 'Following-completion of the study, there was full technology l
| |
| ' transfer of the-PRA computer software and PRA models to NU.
| |
| The-PRA models are currently. maintained within the Probabilistic Risk Assessment Section of Northeast Utilities Service Company (NUSCO). The IPE report has been prepared in
| |
| -full'by members of the PRA Section, who are cognizant of the=
| |
| -PRA models for Millstone Unit'3.
| |
| l 5.2 Composition of Independent Review Team l l
| |
| 1 Section III.4 to the Introduction and Summary of the MP3 PSS l describes the review process used in the study. In summary,-
| |
| three tiers of review have been performed. The first tier, or
| |
| )
| |
| level, consisted of the normal engineering Quality Assurance ,
| |
| carried out by the organization performing the analysis.
| |
| Here, assumptions, calculations, and results were indepen-i dently checked.
| |
| The'second' level.of review involved detailed review by NUSCO personnel in the course of performing the PRA. Approximately o m x.oso l
| |
| L
| |
| | |
| ~. . . . _ - _ . . _ _ _ _ _ _. _ _ _ _ . _ . ~
| |
| u 90 .
| |
| 4? q
| |
| :W .66 1
| |
| +
| |
| 77 one dozen: operators, engineers, and-scientists of various disciplines were involved in this review. LAt least'five persons from the NUSCO PRA Section were involved in'the review essentia11y' full' time-from the beginning of the study.- :
| |
| Virtually every portion of the study;(fault trees, event J trees, core and containment analyses, consequence analyses, l external events, etc.) received technical review by NU ,
| |
| personnel.- .
| |
| The third level of review involved the broad technical review l- of the study by the MP3 PSS Review Board. This board-was ,
| |
| chaired by' Prof. Norman Rasmussen of the MIT Nuclear m
| |
| ' Engineering Department and also included Mr. Saul Levine'of NUS Corporation and Dr. Paul J. Wood, formerly with Wood-Leaver Associates. The latter two organizations also supported the third level of review of the MP3 PSS.
| |
| In addition to'the utilit*;-sponsored reviews,cthe NRC had sponsored its own reviews of the front-end and back-end por-L tions of the MP3 PSS. These are briefly' described in 1
| |
| Sections'5.3.2 through 5.3.4 of-this report. .
| |
| l-L 3 ,
| |
| 5.3 Areas of Review, Major Comments, and~ Resolution of Comments
| |
| ?
| |
| Sections 5.3 and 5.4 of the Standard Table of Contents of-NUREG-1335 have been combined for the purpose of better continuity. ,
| |
| 5.3.1 Utility-Sponsored Review Sections III.5 and III.6 to the Introduction and Summary of the MP3 PSS describe the major areas of review performed. At g
| |
| 'the first and second levels of review, comments and recom-l- mandations were incorporated on a continual basis. Much of f the.information regarding the NU technical review has been L
| |
| 04LW7X.06D
| |
| | |
| . _ . _ . _ - . - _ _ _ _ _ _ _. _ _ _ . _ _ _ ___ _ _ _ . _ m_ __ . _
| |
| l f
| |
| 67 documented in the MP3 construction project correspondence P files for the PSS. :
| |
| In many cases, Northeast Utilities personnel performed ' -
| |
| 3 analyses or recommended additional analyses to resolve major [
| |
| comments. For example, to resolve concerns regarding the i
| |
| ad'equacy of'the common cause analysis, NU personnel developed the common cause failure analysis method (see Appendix 2-C to the PSS). Likewise, NU personnel actively participated in the human reliability analysis (see Appendix 2-D to the PSS).
| |
| 1; For the back-end analysis, NU recommended additional l L
| |
| sensitivity studies to address radiation heat transfer from the reactor cavity debris (see Section 4.4.3.5 of the PSS), as L well as the potential for containment failure via over- ;
| |
| l temperature from electrical penetration elastomer failure, l- Changes to the evacdation~model in the consequence analysis !
| |
| 1 L for internal and seismic events were incorporated to reflect ,
| |
| NU comments.
| |
| i Section III.6 to the Introduction and Summary of the MP3-PSS, ;
| |
| as well as Reference 5-1, describe the subject areas reviewed
| |
| .by the MP3-PSS Review. Board. Major areas of review included the methodology of the PSS, plant systems analysis, core melt L . frequency quantification, degraded core and containment analysis, radiological source term analysis, consequence analysis, and overall PRA results. Detailed reviews were made of the common cause quantification, treatment of human factors, external events, and containment analysis..
| |
| A-few of the major comments suumarized from Reference 5-1 include the following:
| |
| + That with a modest number of understandable exceptions, the NUSCO team was quite successful in
| |
| . demonstrating technical and management competence.
| |
| ownx.oso ,
| |
| | |
| '!C 68'
| |
| . h e : That the MP3. PSS utilized- the proven methodology; in the EP.A Procedures Guide (NUREG/CR-2300).
| |
| *- That an error in the CRAC-I1' code analysin had led
| |
| .tofsignificant underprediction of health consequences.
| |
| L i
| |
| e That the common cause failure analysis using the. .'
| |
| . binomial failure rate model was conceptually correct' and~ represented a different view of an important problem.
| |
| e That the method of discrete probability distribution (DPD) to characterize uncertainties in the conse-quence analysis by varying the source term input'to )
| |
| the CRAC II code does not' provide good understanding ]
| |
| o'f all the sources of uncertainty. l
| |
| . m ;
| |
| t
| |
| * That the seismic fragility analysis was' excessively '1 conservative, while the failure to include con- a sideration of ths Decollement Zone developed b'y the !
| |
| U.S._ Geological Survey in-the MP3 site seismic- l hazard. curve definition had significantly biased J both the uncertainties and the median hazard curve j in the-optimistic direction.
| |
| e- To.the best_of the ability of the MP3 PRA Review Board to judge, there were no major omissions in the 1 scope of the MP3 PSS. !
| |
| I o e That the risk assessment process is an interactive y one in'which knowledge. gained during the conduct.of ,
| |
| the study is folded back into model refinements. ;
| |
| Because of tihe constraints, only a modest amount of feedback was possible.
| |
| , o n w x.oso i l
| |
| l
| |
| | |
| ? 69 .
| |
| -Many of the comments expressed by the PSS Review Board were incorporated as the study progressed. Meetings of the Review.
| |
| Board- were held periodically over a one-year period to allo'r ample time to resolve major issues. To correct errors .in tl:-
| |
| consequence analysis, highlighted by the Review Board final reportL(Reference 5-1), the consequence analysis was redone, and Amendment 1 to the PSS was issued in September 1983. To remove the overconservatisms in the seismic fragility analysis, the seismic fragilities of equipment and structures were' reanalyzed by a second contractor, and the results reflected in Amendment 2 to the PSS in April 1984. To take account of potential nonconservatisms in the seismic hazard analysis,.a seismic margins analysis was performed as follow-up to the PSS (see Reference 5-2). Sensitivity studies using the Lawrence Livermore National Laboratory Seismic Hazard Characterization curve were performed.
| |
| Finally, NU_ recognizes the iterative nature of the PIUL process and as part of the "Living PRA" concept, has updated the PSS-to reflect the actual in-service test program (Amendment 4 to ,
| |
| tho'PSS, August 1987), as well as actual plant operating experience. (See Section 5.4 of this report for additional discussion of-the Living PRA program.)
| |
| 5.3.2 .LLNL Review In 1986, NRC published the findings of the Lawrence Livermore National Laboratory (LLNL) review of the MP3 PSS (Refer-ence 5-3, NUREG/CR-4142). The comments of the reviewers were also summarized in Reference 5-4. These comments-were.
| |
| incorporated during the IPE process in the following manner:
| |
| : 1. In order to ensure that the primary objectives of the IPE are met, those comments pertaining to potential significant vulnerabilities were identi-fled during the IPE. These are: .
| |
| 04LW7X.000
| |
| | |
| i 70 {
| |
| q
| |
| ~
| |
| : a. RWST failures
| |
| : b. Common cause failure of check valves in '
| |
| Y the injection-path s
| |
| c '. Common cause failure of ESFAS-l1 ,
| |
| : d. Loss of service water system. initiator j p
| |
| \
| |
| : e. DC power system modeling and dependencies l on emergency AC power i
| |
| j 2. Those suggestions which were perceived as ,
| |
| improvements to the MP3 PSS, but were not 'copridered ,
| |
| as significant since they do not point to plant-l f
| |
| vulnerabilities or understanding of the severe n accident behavior, were identified as action itemst to be addresoed in the next update of the MP3 PSS. )
| |
| : 3. Some comments were discounted due to one or more-of.
| |
| . -the following reasons: ,
| |
| k l- a. The comment:is a general observation on a method (nr some .other item rather than a critique highlighting a deficiency of the MP3 PSS. j
| |
| : b. NU= analysts have wel3-founded disagreement with reviewer's suggestion.
| |
| 's- .
| |
| L , c. .The state-of-the-art knowledge, in a given
| |
| ' area, does not justify a change to the PSS ,
| |
| based upon reviewer's comment, sinc,e the ^
| |
| suggested change does not guarantee cn improve-ment to the model. ,
| |
| Reference 5-5' identifies 112 summary comments based upon NUREG/CR-4142 and NU's planned response to those comments.
| |
| 04LW7X 080
| |
| .u ___--.__.-__________:_ -__-_---__-_________-___________-_-.__.-_-_2.2
| |
| | |
| 71 Resolution of Comments for Internal Events The reviewers questioned the validity of the value used for the connon cause failure (CCF) of ESFAS', especially sinos i'c dominates failure of both ESFAS trains. It is acknowledged that CCF of ESFAS is a significant failure mode. Sensitivity analysis performed (Reference 5-6) showed that a 100 percent (factor of 2) increase in this CCF resulted in a 3 percent increase in core damage frequency (CDF). Even a 1,000 percent increase in this CCF resulted in a 25 percent increase in the CDF. Based on the above results and state of knowledge on CCF o'f ESFAS, NU does not find compelling reasons to maks changes to the model.
| |
| freezina of RWST Lines The reviewers noted that the freezing of the RWST lines was excluded without justification. The present MP3 PSS model uses a failure probability of 1.92 x 10-8 for RWST based on the tank failure rate of 8 x 10-10 per hour. No other failure nodes such as freezing of lines has been considered. In order to examine the impact of RWST failure rate on the CDF, a sensitivity was performed. Even when the probability of RWST failure was increased by a factor of 100, the CDF remains unchanged. In spite of this result, NU recognizes that RWST is a key component essential for a large number of safety functions and plans to investigate failure modes of RWST, other than due to tank rupture, in the future.
| |
| Common Cause Failure of Check Valves MP3 PSS did not consider the common cause failure of redundant pairs of check valves in high pressure systems. NU agrees that introduction of these CCFs will improve the completeness of the MP3 PSS model. However, the random check valve failure probability used by MP3 PSS is approximately a factor of five higher compared to NREP/IREP and approximately a factor of ten 04LW7X.06D
| |
| | |
| 72 higher' compared to the-Zion PRA. ~ Introduction ofLthe suggested CV CCFs, prior to resolving some issues ~related-to
| |
| -relatively high probabilities of check valve random failures.
| |
| .used in the MP3 model, can result in an inflated relative ,
| |
| contribution-of these check valve failures to CDF. -NU is in the process of updating the check valve' failure rates and plans:to include common cause failures in a future PRA update.
| |
| Vital DC Power'Modelina LLNL had several concerns over the modeling of the DC power-systems. Since DC-power;is a significant support system, NU decided to address these concerns during the IPE. The MP3 PSS considered two special initiators, " Loss of one DC Bus" and
| |
| " Loss of.All Vital DC" in its analysis. Therefore, NU believes that MP3 PSS adequately depicts the risk attributed
| |
| ~
| |
| to the loss of DC power. NU accepts the notions that perhaps an improved model would have-resulted if the DC power was included in~the support state event tree.. In fact, any future-update of the model would seriously. consider inclusion of DC power utilizing fully linked fr . t trees. - on the concerns of subtle interdependencies between vitalfDC and the emergency AC, NU believes that while-the model can and will be improved, utilizing the suggestion provided;in NUREG/CR-4142,.these improvements are not-expected to uncover vulnerabilities.not identified to date.
| |
| Loss of SW Initiator The MP3 PSS ruled out modeling of the complete loss of-service water (SW) based upon the low frequency of this event. SW is a significant commonality since it supports a large number of
| |
| ' functions. Therefore, in spite of the low probability of the initiating event, its incitsion as an initiator in a future update of the model will be considered.
| |
| ,o 04LW71.080
| |
| | |
| D.-l 73
| |
| ; '5.3.3 BNL' Review of the Back-End Analysis Brookhaven National Laboratory (BNL) performed a technical review and evaluation of the Millstone Unit.3 PSS containment, :
| |
| source _ tern,.and consequence analyses (Reference.5-7). It was O determined in this review that long-term' health risks (latent fatalities, person-rem, etc.)- are dominated by late failure of ;
| |
| w* .the containment. While acute f atalities arid injuries are dominated by containment bypass sequences for internally initiated events. It was also concluded'that the risks for MP3 are comparable to risks from other nuclear power plants at p high population sites. These conclusions are consistent witn ,
| |
| l- 'those of the MP3 PSS. Potential mitigative features were shown not to be cost-effective for internal events (although ,
| |
| " some features might'be for seismic events).
| |
| Some of the major insights and comments from the BNL are summarized below:
| |
| . The lower reactor cavity geometry is expect 6d to -
| |
| i-suppress the dispersion of core debris from the ,
| |
| L l reactor' cavity to the general containment following L failure of the reactor vessel during core melt sequences (see Section 4.9 for additional discus-sion). The cavity area geometry.also would reduce L the potential for establishing eftective convective air currents between the cavity and general contain-ment area for heat removal from core debris in the cavity area.
| |
| l 1
| |
| . The likelihood of transporting a significant amount of core debris to the containment sump area is con-sidered to be small since the bulk of the corium l
| |
| would remain in the cavity, only the smalle*
| |
| l particles would exit the cavity, and the region of E containment into which the exiting deDris would enter is not swept by the containment sprays. Thus, l,
| |
| i 'c4Lwn.oso i.
| |
| | |
| i
| |
| (
| |
| hY ,
| |
| 74 l 1
| |
| >l%
| |
| the potential for contrinment sump blockage fron l i
| |
| core debris is reduced.
| |
| j e The predicted rates and magnitudes of hydrogen :
| |
| generation and release from severe accidents far :
| |
| 1 exceed the capacity of the hydrogen recombiner system (which is designed for mitigation only design basis accidents, not " severe" accidents). ,BNL (and
| |
| , a I
| |
| utility; calculations indicate that early contain- l ment failure by H2 burning is not likely. However, containment failure in the intermediate time frame (four to sixteen hours after vessel failure) could l fail containment fo.r a narrow band of conditions, !
| |
| such as the AE plant damage state (large LOCA, early l core malt, no containment safeguards) and station blackout induced core melt with recovery of sprays, !
| |
| or from condensation deinerting. l l >
| |
| e In-vessel steam explosions which result in direct containment, failure are believed to be highly un-likely.(alpha failure mode).
| |
| * While BNL is not convi'nced that the probability of containment ir 4ation failure is as low as the 2 x 10'4 statet the PSS, they are confident that .
| |
| it is low enough to be a minimal contributor to risk. ,
| |
| i e For sequences in which early and intermediate .
| |
| failure is not expected to occur, and in which the
| |
| * cottainment spray recirculation is inoperable, the containment would fail late by stcam overpressuriza-tion (Decause of the basaltic-based concrete, molten core-concrete interaction would not release a significant amount of carbon dioxide).
| |
| r 04LW7X,089
| |
| ^ *
| |
| | |
| . t 75
| |
| * In the review, ENL did not assume that early con-tainment failure from direct containment heating l (DCH) would occur. However, it was noted that if DCH induced containment failure vers to occur for i j
| |
| all high pressure sequences, and the release frac- l tions were representative of an in-vessel staam explosion with direct containment failure (alpha !
| |
| . failure mode), the estimates of early and latent '
| |
| fatalities together with the public dose, would increase significantly for internal events. ,
| |
| The BNL review recommended adjustments to the containment ;
| |
| matrix; i.e., conditional probability of containment failure by various failure mouas given plant damage state. A com-parison of the BNL containment matrix and a simplified matrix from the MP3 PSS is shown in Table 5.3-1. As can be seen, the j
| |
| agreement is nearly complete with the exception of containment L response class 2 (SE plant damage state), whern BNL assigned a ;
| |
| ! new release category for intermediate containment failure due to steam deinerting and H2 burn (i.e., release cat?;ory M6S).
| |
| Except-for release category M1A, BNL did not adjust the MP3 PSS release fractions. However, some changes to the release energy, as well as release and warning times used in the L ' consequence analysis, were recommended. The BNL report did note (Section 6.2 of Reference 5-7) that the uncertainties in the radiological source terms are very high and are all-in the direction of much lower releases.
| |
| To addrsos the one major issue from the BNL review regarding intermediate containment failure from containment deinerting and hydrogen burn, additional utility sponsored analyses were performed (Reference 5-8). Specifically, the effect of con-tainment spray recovery folliwing a station AC blackout with ;
| |
| pump seal LOCA (the major cor.tributor to release category M6S) was investigated. The results of the analysis indicate that recovery of containment sprays any time prior to about five C W OX.06D ,
| |
| | |
| 76 hours after predicted reactor vessel failure would result in containment pressure, owing to hydrogen burn, less than the j median ultimate failure pressure. The analyses also indicated that, owing to effective spray operation, the fission product :
| |
| release fraction in M6s for such an accident sequence (i.e., -
| |
| containment failure attrir recovery) should be grouped into a l release category having si:jnificantly lower release fractions !
| |
| Consequence calcula- l than the'non-recovery case; i.e., M7. ;
| |
| tions for this earlier containment failure source ters (M6S) J l
| |
| have shown that offsite consequences are significantly less !
| |
| (two orders of magnitude in the case of latent health effects) j than those calculated for the later/no spray release category. l Thus, the fission product removal characteristics of even j short term spray operation more than compensate for the impacts of a potentially earlier containment failure time. ]
| |
| 1 i
| |
| 5.3.4 NRC Review
| |
| }
| |
| NUREG-1152, " Millstone 3 Risk Evaluation Report," (Refer-ence 5-9) documents the results of the staff *. re. view of the .,
| |
| PSS. In a letter from the NRC to NU (Reference '- at the '
| |
| staff identified four (4) improvements to be maae ac :
| |
| Millstone 3 which could result in substantial additional protection of-public health and safety. These NRC recommenda-tions were:
| |
| : 1. Engineering analysis for upgrading the diesel generator lube oil cooler anchorage system,
| |
| : 2. Feasibility of manually operated AC-independent containment spray system, ,
| |
| I
| |
| . 3. Operator training and procedures to recover from earthquake induced relay chatter, and, !
| |
| : 4. Emergency procedures to recover cooling to selected rooms by. alternative methods.
| |
| l owns on
| |
| _ - - ~ . _ _ . _ . _ _ . . _ . _ _ _ . _ . _ _ _ . _ _ _ _ _ . _ __ _ _ _
| |
| | |
| 1)fIj 77 The NU response actions to these reconnandations were as
| |
| 'follows:
| |
| : 1. As a response to the engineering analysis for upgrading the diesel generator lube oil cooler anchorage system, a design m.>dification was per-
| |
| , formed. During this design modification, existing emergency diesel generatcr lube oil cooler anchor bolts vsre replaced with bolts cade of stronger material (Reference 5-11).
| |
| : 2. As a response to the issue on feasibility of manually operated, AC-independent, containment spray system, an evaluation was performed by NU (Refor-ence 5-12). As a result of this evaluation it was concluded that while a containment spray system independent from the existing onsite emergency AC power system could be constructed, these systems will not ba cost-effective by a substantial margin.
| |
| Therefore, no design change has been sDplemented.
| |
| : 3. As a response to the issue on earthquake induced relay chatter, the emergency procedures were upgraded as necessbry to cope with such events.
| |
| : 4. As a response to emergency procedures to recover cooling, an evaluetion was performed to assess the risk significance of loss of HVAC system scenarios in the control room, switchgear room, and auxiliary feedwater pump cubicles (Reference 5-13). Based on the evaluation, NU concluded that the loss of room cooling at MP3 is not a significant core dr. mage frequency issue. Further, since the.N2C : eviewed the MP3 PSS, steps have been added to the MP3 operating procedures to deal with high temperatures in heat-sensitive vital areas. As a response to the Station Blackout Ru39, the room cooling reliability has been improved further.
| |
| o n m x.oso
| |
| . . . . . . . . = . ,m
| |
| | |
| 78 In addition to the above, NRC analysis had estimated that station blackout dominates internal and external event core damage frequency. In a letter from NRC to NU (Refer-ence 5-14), NU was requested to respor.d to the apparent large contribution to risk due to station blackout. An evaluation performed by NU (Reference 5-15) on the NUh?G-1152 analysis of the station blackout provided a basis to justify not taking any plant specific action pending full generic resolution of the Station AC Blackout Unresolved Safety Issue (USI A-44).
| |
| Since then, thJ Station Blackout Rule was published and NU plans to add a third diesel generator to MP3 to further reduce the rish from loss of offsite power events.
| |
| 5.4 Living PRA Program Northeast Utilities has developed and implemented a comprehen-g sive Probabilistic Risk Assessment (PRA) program in support of nuclear power plant engineering and operations. The key elements of the program include a corporate policy on nuclear >
| |
| safety goals; Level 1 PRAs for four units; plans to develop ;
| |
| Level 3 PRAs; formalized procedures for the review of plant design changes at the conceptual and final approval stagas; formalized procedures for determining the cost effectiveness and integrated scheduling of major projects under the Integrated Safety Assessment Program (ISAP); and formalized >
| |
| procedures for notifying senior management on a regular basis concerning the risk assessment for each unit. FRA is also p
| |
| used extensively in support of plant operations and licensing *
| |
| !' activities (see Reference 5-16).
| |
| Since completion of the MP3 PSS in 1983, six amendments /
| |
| i updates to the PRA have been made.
| |
| The MP3 PSS is used to varying degrees for nuclear plant support in many different areas such as:
| |
| Operator training in risk dominant sequencts 041W7X ,08D
| |
| | |
| s 79 Safety avaluations.
| |
| o Establishment of equipment test interve.s.
| |
| Prioritization of important equipment and systems.
| |
| Establishment of allowable cutage times (ACT in Technical Specifications for safety related equipment).
| |
| Table 5.4-1 provides a sample list of major MP3 support activities plant in which PRA has been used since the beginning of operation.
| |
| As noted earlieri all new plant projects are required by procedure to be reviewed from a PRA perspective for impact on corporate nuclear safety goals.
| |
| effect since early 1988. This procedure has been in In 1989, the NUSCO Prob.abiliscic Safety Studies (PSS) Key Assumptions uccument was developed to make corporate engi -
| |
| neering and plant personnel better informed of the general-assumptions made within the Northeast Utilities PSSs, and to ensure that subtle plant changes get reflected in the PSS models.
| |
| Rather than list all the key assumptions in the PSS ,
| |
| the approach taken in th4,s document is to identify the syst modeled in the PSS, ems identify major assumptions and operator actions for each system and list in a general way types of changes which could-impact the PSS assumptions and results .
| |
| Training on use of the Key Assumptions Document was given to plant personnel to sensitize them to changes to procedures ,
| |
| equipment or practices which could have a potential impact in the safe operation of the plant and, hence on the PSS itself .
| |
| Since 1988, MP3 has been involved in an internal Integrated Safety Assessment Program (ISAP) based on the Haddam Neck Plant and Millstone Unit No. l's formal program.
| |
| The ISAP 04LW7X,06D .
| |
| | |
| 80 d on process involves prioritizing proposed backfits base public safety, economic performance, Thepersonnel public safetysafety, '
| |
| personnel productivity, ALARA and Cost. ii the input to the ranking process is obtained by determ R n ng potential core melt frequency reduction and not i Man-the ea reduction associated with each proposed backfit dur ngBased on the pri .
| |
| remaining operating life of the plant. incor-tization results and available resources, projects areTo date, porated into an integrated implementation schedule. The numerous projects have been evaluated'in MP3's is program.
| |
| expected MP3 initial ISAP submittal is near completion andThe subseq to be submitted mid-1990.
| |
| is scheduled for submittal in early 1991.
| |
| vide The maintenance of the MP3 PRA model is 4ssential Plant to pro an accurate description of the plant risk profile.a collec design changas The NU procedural 'tramework is used to col-update process.
| |
| lact plant design change documentation in the form of plL d design change records (PDCRs). is per-minor plant changes, a preliminary screening process As formed before the remaining PDCRs are reviewed identify in detail.
| |
| a result of the PDCR review process, NU is able to Such changen changes that would impact the current PRA model. l include installation of additional valves, changes finequip- va ve operator type, change of power supplies, relocation i ot ment, changes to control room design and changes to equto tions pmen which will replace operator actions with automatic ac The design changes which are identified toi havethe name a few.
| |
| an impact are then incorpolated into the PRA model dur ng next update.
| |
| h of To date, over 175 PDCRs have been reviewed andthe only PRA t ree these PDCRs have a significant but positive impact on model. Table 5.4-2 into future up'ates.
| |
| 04tW74.080 II
| |
| | |
| - . - _ _ - - - - . . -. .- - - . . .~ - -.
| |
| I 79 \
| |
| e safety evaluations. )
| |
| i l
| |
| e Establishment of equipment test intervals.
| |
| $ l e Prioritization of important equipment and systems. J 9
| |
| * Establishment of allowable outage times (AOT in i Technical Specifications for safety related equipment). ;
| |
| Table 5.4-1 provides a sample list of major MP3 support activities in which PRA has been used since the beginning of plant operation.
| |
| As noted earlier,' all new plant projects are required by j procedure to be reviewed from a PRA perspective for impact on corporate nuclear safety goals. This procedure has been in effect since early 1986.
| |
| In 1989, the NUSCO.Probabilistic Safety Studies (PSS) Key l Assumptions Document was developed to make corporate engi-
| |
| ! nearing and plant personnel better informed of tne general [
| |
| \
| |
| assumptions made within the Northeast Utilities PSSs, and to ensure that subtle plant changes get ruflected in the PSS models. Rather than list all the key assumptions in the PSS, l the approach taken in this document is to identify the systems modeled in the PSS, identify major assumptions and operator actions for each system and list in a general way types of changes which could impact the PSS assumptions and results.
| |
| Training on use of the Key Assumptions Document was given to plant personnel to sensitize them to changen to procedures, l equipment or practices which could have a potential impact in the safe operation of the plant and, hence on the PSS ttself.
| |
| 1 l Since 1988, MP3 has been involved in an internal Integrated Safety Assessment Program (ISAP) based on the Haddam Neck Plant and Millstone Unit No. l's formal program. The ISAP 04LWX.08D I
| |
| | |
| )
| |
| l 80 process involves prioritizing proposed backfits based on public safety, economic performance, personnel safety, l personnel productivity, ALARA and Cost. The public safety 1 input to the ranking process is obtained by determining the '
| |
| l potential core melt frequency reduction and not Man-Rem reduction associated with each proposed backfit during the (
| |
| remaining operating life of the plant. Based on the priori- l tization results and.available resources, projects are incor-porated into an integrated implementation schedule. To date, '
| |
| numerous projects have been evaluated'in MP3's program. The MP3 iniwial ISAP submittal is near completion and is expected to be submitted mid-1990. The subsequent ISAP Summary Report is scheduled for submittal in early 1991. .
| |
| The maintenance of the MP3 PRA model is essential to provide an accurate description of the plant risk profile. Plant design changes are collected and reviewed as part of the PPA '
| |
| update process. The NU procedural framework is used to col-icct plant design change documentation in the form of plant design change records (PDCRs). Because of the large volume of j minor plant changes, a preliminary screening process is per-formed before the remaining PDCRs are reviewed in detail. As ;
| |
| a result of the PDCR review process, NU is able to identify i l changes that would impact the current PRA model. Such changes ;
| |
| include installation of additional valves, changes in valve operator type, change of power supplies, relocation ofiequip-ment, changes to control room design and changes to equipment >
| |
| which will replace operator actions with automatic actions to name a few. The design changes which are identified to have an impact are then incorporated into the PRA model during the next update.
| |
| To date, over 175 PDCRs have been reviewed and only three of ,
| |
| these PDCRs have a significant but positive impact on the PRA model. Table 5.4-2 lists the PDCRs which will be incorporated into future updates.
| |
| cawn.oso 1
| |
| -- , . . _ ~ _ . - - . , - - - . . , - - . . - - -. . - -
| |
| | |
| 81 REFERENCES 5-1 N. C. Rasmussen, S. Levine, P. J. Wood, " Final Report of the Level 3 Review Board on the Millstone Point Unit 3 Probabilistic Safety Study," August 1983.
| |
| 5-2 "A Program to Determine the Capability of the Millstone 3 Nuclear Power Plant to Withstand Seismic Excitation Above the Design SSE," Structural Mechanics Associates Report NiC/SMA 20601.01-R3, November 1984.
| |
| w 5-3 U.S. Nuclear Regulatory Commission, "A Review of the Millstone 3 Probabilistic Safety Study," NUREG/CR-4142, dated April 19CC.
| |
| 5-4 M. Wasim Akhtar and W. J. Parkinson, "NRC Review of Nuclear Plant Probabilistic Safety Studies," Draft Interim Report, DRAF8/ MAR 890, dated March 1990.
| |
| 5-5 Northeast Utilities PRA Calc. File MP3-PRA-90-028 (Sec-tion I), S. D. Weerakkody, dated April 17, 1990.
| |
| 5-6 Northeast Utilities PRA Calc. File MP3-PRA-90-028 (Sec-tion II), S. D. Weerakkody, dated April 18, 1990.
| |
| 5-7 M. Khatibsaahbar, et. al., " Review and Evaluation of the Millstone Unit 3 Probabilistic Safety Study: Containment Failure Modes, Radiological Source - Terms and offsite Consequences," Brookhaven National Laboratory, NUREG/CR-4143, BNL-NUREG-51907, September'1985.
| |
| 5-8 B. G. Holmes, L. A. Wooten, D. A. Dube, "An Analysis of Containment Spray Recovery and Its Effect on Containment Integrity and Fission Product Source Term," Third Intl.
| |
| Top. Mtg. on Reactor TH, Newport, RI, October 15-18, 1985.
| |
| 04Df7X,06D )
| |
| | |
| 88 5-9 U.S. Nuclear Regulatory Commission, " Millstone 3 Risk Evaluation Report," NUREG-1152, August 1985.
| |
| 5-10 Letter from Thomas M. Novak (USNRC) to J. F. Opeka *
| |
| (NNECO), "Probabilistic Safety Study for Millstone Nuclear Power Station, Unit 3," October 17, 1985.
| |
| 5-11 " Emergency Diesel Generator Lube 011 Cooler Anchor Bolts ,
| |
| 1 Replacement," PDCR #MP3-86-126.
| |
| 5-12 " Evaluation of AC-Independent Containment Spray System: l Millstone Unit 3," NUSCO 154, Northeast Utilities Service ,
| |
| Compar.y, February 1987.
| |
| 5-13 Letter from E. J. Mroczka (NNECO) to USNRC, " Millstone Nuclear Power Station, Unit No. 3, Report on Loss of Room Cooling," January 7, 1988.
| |
| 5-14 L'tter a from Harold R. Denton (USNRC) to J. F. Opeka (NU), December 18, 1985. 3 5-15 Letter from J. F. Opeka to Harold R. Denton (USNRC),
| |
| " Millstone Nuclear Power Station, Unit No. 3: Response i to information Requested Regarding Station Blackout," ;
| |
| dated March 18, 1986.
| |
| ; i 5-16 D. A. Dube, "PSA Support of Nuclear Power Plant '
| |
| Engineering and operations," International Topical Meet-ing on Probability, Reliability, and Safety Assessment, Pittsburgh, PA, April 2-7, 1989.
| |
| 9
| |
| . 04LW7X,000 1
| |
| 6
| |
| ._ _ _______ _ . _ b._ _ __._ _ __ _ _, . , , m . , , _ . - _ , , , , - . _ - .
| |
| | |
| f 03 :
| |
| 6 PLANT IMPROVEMENTS AND UNIQUE SAFETY FEATURES A number of design changes and procedural changes have been :
| |
| implemented based on PRA insights. Although some of the changen have been initiated for regulatory reasons, PRA based l 1
| |
| evaluations performed under the auspices of an internal l Integrated Safety Assessment Program often provided further motivation for implementation. Table 6-1 gives a sample listing of changes .nich have provided a measurable improve-ment in core melt f.requency and/or public safety since the beginning of commercial operation. Because of the Living PRA program, future improvement based on PRA insights can be anticipated. .,
| |
| i Section I.2 to the Introduction and Summary of the MP3 PSS describes many of the key plant safety features. While not ,
| |
| all of these features are necessarily unique, they do ;
| |
| contribute significantly to the overall safety of MP3. These -
| |
| include:
| |
| e Two motor-driven and one steam-driven auxiliary feedwater pumps which, on a best estimate basis, provide three-fold ;
| |
| redundancy.
| |
| . Three high pressure, high capacity charging pumps and two relatively high head safety injection pumps. The pumps, together with the pr; aurizer PORVs provide primary feed i
| |
| and bleed capability. The physical separation between
| |
| ( trains, and different physical location of the charging and Safety Injection pumps (auxiliary building and ESF l building, respectively) are of great benefit from a l spatial interactions perspective (fire, flood, seismic).
| |
| * The 1.2 million gallon refueling water storage tank substantially delays the time to switch over during loss of coolant accident. This improves the ability of the operator to properly diagnose the accident, allows addi-04LW7X.0$D 4
| |
| ~c - - , . , - ~n -, -
| |
| | |
| l 84 tional time for equipment recovery, substantially shifts j the timing of containment integrity challenge for certain !
| |
| accident sequences, and reduces and delays fission product release far a number of severe accident .
| |
| scenarios.
| |
| i L
| |
| e DC powered pressurizar PORVs. l
| |
| * The lower reacter cavity design is expected to be mostly !
| |
| dry at the time of vessel failure during core melt !
| |
| accidents. Substantial hydrogen would be generated from j molten core / concrete interactions (McCI). Hydrogen !
| |
| combustion within containment would not threaten conta.in--
| |
| ment integrity, however, as long as containment sprays were available. In addition, the basemat composition is basaltic, so that carbon monoxide and dioxide generation from MCCI is relativuly low. For many severe accident ;
| |
| l l
| |
| sequencee, late spillover of water into the cavity to provide debris cooling is predicted.
| |
| J Two spray systems (the quench spray system and the recir-culation spray system) provide diverse and independent i means of containment depressurization and fission product
| |
| ! scrubbing.
| |
| I e The containment ultimate failure pressure (median value) is estimated to be two and one-half times design.
| |
| * The operation of_the containment at subatmospheric condi- l tions reduces the probability of having a pre-existing failure of containment isolation.
| |
| The 1PE Submittal Guidance, NUREG-1335, requested that the IPE submittal provide enough documentation so that the NRC reviewer can be confident that a reasonable effort to address eaci identified vulnerability has been performed, whether or not a fix has been implemented. Table 6-2 provides this 04 Di7X.0 @
| |
| _____m___ _ _ _ -_-._ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ . . _ _ _ _ - a ,. . - - .,
| |
| | |
| 85 l
| |
| .c information. Many of these issues and resolution of issues q have been highlighted throughout this IPE report. It should !
| |
| ! be noted, moreover, that these issues in almost all instances have'been addressed in the regular course of engineering, :
| |
| j, design, and operations, independent of the IPE process. This
| |
| !- is a clear indication of how risk management has been institu-tionalized at NU on a plant specific basis.
| |
| l' I
| |
| J l
| |
| J I
| |
| J l
| |
| I I
| |
| l \
| |
| l 1
| |
| 'i i
| |
| I 1
| |
| I l
| |
| I 1 !
| |
| l l
| |
| * l s
| |
| 04LW71,080 e a.
| |
| | |
| 1 86.
| |
| 7
| |
| | |
| ==SUMMARY==
| |
| AND CONCLUSIONS A full Level 3 PRA including externally initiated events has beenperfodedforMP3. The RP3 POS documents the PRA models .
| |
| and findings. Northeast Util' ties personnel participated in the analysis and/or review of /irtually every portion of the PRA. Following completion of the PRA study in 1983, full technology transfer of the PRA models, computer software and ]
| |
| hardware to NU took place. Six amendments and updates to the l PSS and/or PRA models'have occurred. The PRA models have i since been transferred to personal computer based systems, consistent with the needs of a Living PRA program. {
| |
| I NU has developed and implemented a comprehensive Probabilistic Risk Ascessment program in support of nuclear power plant engineering and operations. Included in this progre' aret ]
| |
| e A high level Corporate Policy on Nuclear Safety Goals.
| |
| .
| |
| * Formalized procedure for communication of PRA findings to !
| |
| senior management on an annual basis, upon completion of l major PRA studies, and immediately for risk significant conditions.
| |
| * Formalized procedure for the PRA review of all new plant !
| |
| l project assignments.
| |
| . An internal Integrated Safety Assessment Program (ISAP) j for prioritizing proposed backfits based on public safety I impact and other attributes. ,
| |
| I l
| |
| e A controll<3d means of updating the PRA models on an as-needed basis. )
| |
| l PRA support of important design and operational issues is extensive. The percentage of this support in relationship to i all PRA-based activities including development of models has ]
| |
| l l i osunx.no
| |
| (-
| |
| l
| |
| | |
| i 87 l increased significantly in recent years, indicating wide-scale use of PRA.
| |
| For MP3, the " front-end" (Level 1 PRA) and "back-end" (Level -
| |
| 2/3 PRA) have received exte~.sive internal and external reviews. The results of'these reviews have been considered, and have been or will be incorporated into the PRA as deemed ;
| |
| appropriate. While the bottom-line PRA values are believed to f be an accurate representation of the plant risk profile, the results are a'lways arguable because of differences in assump- 1 tions, methods, and data bases. Because NU P'A R section per-sonnel are familiar with the assumptions as well as results in
| |
| < the PRA models, these issues are taken into account when major recommendations are made regarding plant design and opera- ,
| |
| tions. If an issue appears to be a major contributor to core melt frequency, for example, while there are large uncertain-ties regarding dat&, decisions are typically made in the conservative direction. For PRA issues which appear to be ,
| |
| minor, but with large uncertainties usually indicating a fair degree of overconservatisms, the decision-making process is .
| |
| again usually clear.
| |
| Because of the L portance of assumptions in PRA modeling, a Key Assumptions Document for the Probabilistic Safety Studies of MP3 and other NU operated nuclear units has been issued to l
| |
| key engineering and plant operations personnel. Formal instruction has also taken place. The purpose of the document is to alert personnel of the potential impacts of design and L operational changes on the PRA results.
| |
| For MP3, the results of the PRA indicate that there are nr major severe accident vulnerabilities requiring immediate corrective action. As noted in Section 6 of this report, 3 plant improvements, increased emphasis on operator training, i increased surveillance, and/or procedural changes have been effected for the most important PRA issues such as station blackout and interfacing systems LOCA prevention and mitiga-04tW7X 06D
| |
| - , _ . _ - - . . , , _ , ~ - ._. ..
| |
| | |
| 88 tion. As' highlighted in Section 1.4, the major insights from tho PRA analysis indicate the following:
| |
| . Operator training for containment sump recirculation (i.e., Emergency Operating Procedures for " Transfer to Cold Leg Recirculation" as well as " Loss of Emergency Coolant Recirculation") is important.
| |
| . Surveillance, alarm, and position indication on the motor operated valves in the RHR suction lines are important with regards to the potential for interfacing systems LOCA. Emphasis on operator training for the EOP on "LOCA Outside Containment" is also important.
| |
| * That the large volura of the refueling water storage tank is advantageous becaL.se it increases the available time for decision making by operators and emergency organiza-tions.
| |
| * That the redundancy and diversity of the high head safety Injection pumps and charging pumps, as well as their physical separation (dif f erent buli.aings altogether) is advantageous.
| |
| * Feed and bleed' capability is advantageous for loss of all feedwater mitigation.
| |
| * Station blackout sequences are not major contributors to core melt frequency, but have been founi to be important contributors to offsite release frequency. The proposed addition of a third, air-cooled diesel generator will significantly reduce these contributions to CMF and public risk.
| |
| i
| |
| . Externally initiated events are not major contributors to core melt frequency or public risk, with the exception of coax.cn ,
| |
| | |
| 89 I severe seismic events significantly beyond the SSE in l magnitude. i o That a major contributor to seismic core melt frequency s 1
| |
| and public. risk has been alleviated by the replacement of l anchor bolts on the diesel generator oil coolers with bolts of stronger material.
| |
| e That the ultimate capacity of the containment is about two and one-half times design pressure.
| |
| i i
| |
| e For most core melt scenarios leading to reactor vessel failure, containment integrity is expected'so long as the e
| |
| quench and recirculation spray systems remain available.
| |
| e operation of the containment at subatmo.p.aric k conditions 2 minimizes the potential for pre-existing failure of con-i tainment isolation.
| |
| l e The " dry" lower react.or cavity design is expected to result in substantial molten core-concrete interaction and hydrogen generation following vessel failure ,
| |
| scenarios, e The MP3 PSS evaluation of the potential for Direct Containment Heating following High Pressure Melt Ejection >
| |
| is inconsistent with small scale experiments performed at Brookhaven National Laboratory in support of the NUNEG-1150 evaluation for Surry Unit 1. Additional industry-sponsored research on the issue is encouraged.
| |
| Until further clarification is provided, complete
| |
| ! re-analysis of the back-end portion of the MP3 3 PSS would ;
| |
| be premature and an inefficient use of limited resources.
| |
| Because of NU's Living PRA program, plant operating experience as well as increased understanding of severe eswn.cso ,
| |
| | |
| ?
| |
| ,\
| |
| 90 accident behavior can be incorporated into the MP3 PRA model as needed. ,
| |
| i N
| |
| I i
| |
| e E
| |
| l l
| |
| i l
| |
| l l
| |
| l S. ]
| |
| l l
| |
| I i
| |
| i 1
| |
| i l
| |
| 04LW7X.080 ]
| |
| E
| |
| | |
| I 91 TABLE 1.4-1 ,
| |
| 1 CORE-MELT FREQUENCY
| |
| | |
| ==SUMMARY==
| |
| 1 I
| |
| l l
| |
| i CMF Percent i (per reactor year) Contribution Internal Events 5.52E-05 78.88 ,
| |
| Internal Floods 8.50E-07 1.21 Fires 4.85E-06 6.93 l Seismic 9.08E-06 12.98 L
| |
| TOTAL 7.00E-05 100 t
| |
| (
| |
| {
| |
| +
| |
| i 04LW7X.000
| |
| - 1 - , - . . ,- . - . _ _ _ _ _ _ _
| |
| | |
| TABLE 1.4-2
| |
| | |
| ==SUMMARY==
| |
| OF RELEASE CATEGORY FREQUENCIES (Internal and External Events)
| |
| .t
| |
| ! Peroomt contributions !
| |
| Release Description Frequency Internal Fire Seismic Category M1A Containnent Bypass, V-Sequence 2.21E-07 100 - -
| |
| M1B Containment Bypass, SGTR 1.74E-07 99.4 -
| |
| 0.6 M2 Early Failure /Early Melt, No 9.64E-09 73.5 0.7 25.7 Sprays ,
| |
| I ,
| |
| j M3 Early Failure / Lake Melt, No Sprays 6.21E-09 78.9 10.5 10.6 j M4 Containment Isolation Failure 1.17 E-07 9.4 0.8 89.8 MS Intermediate Failure / Lake Melt, No 4.89E-08 35.6 12.7 51.7 4 Sprays M6 Intermediate Failure /Early Melt, 6.07E-07 13.8 -
| |
| 86.2 No Sprays M7 Late Failure, No Sprays 1.49E-05 41.1 10.2 48.7 j M8 Intermediate Failure with Sprays 3.31E-08 92.9 6.2 0.9
| |
| ; i
| |
| ! M9 Late Failure with Sprays 2. 04 E- 06 83.3 7.1 9.6 M10 Basemat Failure, No Sprays 1.28E-06 33.5 10.1 55.8 Mll Basemat Failure with Sprays 2.60E-06 91.8 7.7 0.5 M12 No Containment Failure 4.71E-05 93.4 6.0 0.6 l
| |
| 04LW7X.08D i ..
| |
| | |
| I H ull 93 TABLE 3.1-1 '
| |
| INTERNAL INITIATING EVENT VECTOR WI (FREQUENCIES IN EVENTS PER REACTOR YEAR)
| |
| EVENTh FREQUENCIES (MEAN VALUE)
| |
| : 1. -Large LOCA
| |
| : 2. Medium LOCA 3.88 x 10'4
| |
| : 3. Small LOCA 6.11 x 10-4
| |
| : 4. Steam Generator Tube Rupture 9.07 x 10~3
| |
| : 5. 3.92 x 10-2 Steam Line Break Inside Containment 3.88 x 10*4 6.
| |
| Steam Line Break Outside Containment 3.78 x 10-2
| |
| : 7. Loss of RCS Flow
| |
| : 8. Loss of Main Feedwater Flow 4.91 x 10'1
| |
| : 9. Primary to Secondary Power Mismatch 8.60 x 10~1(*)
| |
| 4.30 x 10~1(*)
| |
| : 10. Turbine Trip 2.25(*)
| |
| : 11. Reactor Trip 7.50 x 10~1("I
| |
| : 12. Core Power Excursion 7.18 x 10-2 13.. Spurious Safety Injection
| |
| : 4. 99 x 10-2
| |
| : 14. Losis of offsite Power 1.12 x 10~1
| |
| : 15. Incore Instrument Tube Rupture 9.20 x 10~4
| |
| : 16. Special Large LOCA Initiators
| |
| .a. Interfacing Systems LOCA
| |
| : b. -Catastrophic Reactor Vessel Rupture 2.21 x 10~7
| |
| '17 . 3 x 10~7 Loss of a Single Service Water Train 1. 81 x 10-2
| |
| : 18. Loss of a Single Vital DC Bus 3.92 x 10'3
| |
| : 19. Total Loss of Vital DC Power 1.40 x 10-8 20.
| |
| Loss of Vital AC Bus 120-VAC-1 or 120-VAC-2 6.15 x 10-2 21.
| |
| Loss of Vital AC Bus 120-VAC-3 or 120-VAC-4 6.11 x 10-2 (a) These initiating frequencies are based upon plant specific data 04W7X,0tD
| |
| | |
| W
| |
| ' TABLE 3.1-2 SIMPLIFIED SUPPORT STATES 2
| |
| 3 SERVICE WATER EGLS. TRAINS TnATMS AVAlf N R ESF BUSES I ESF CABINETS .
| |
| _ ACTUATED _
| |
| SUPPORT OFFSITE ELECTRICAL ENERGIZED .._ ACTUATED _
| |
| A+B STATE _ POWER AVAIIABLE _ A+B A+B A or B 34C+34D A or B Yes 1
| |
| 34C+34D A or B None Yes A or B 2
| |
| 34C+34D A or B None Yes None 3 None A+B Yes 34C+34D A-rB 4 A+B A or B 34C+34D A or B 5 No A or B No 34C or 34D None None 6
| |
| None A or B 7 No None None None 8
| |
| No $
| |
| lied by both offsite power and ucrgency 1.
| |
| ESF buses are the 4.16 kV emergency AC buses supp diesels. i h monitor selected planti parameters, such d actuate ESF equipment, as necessary.
| |
| 2.
| |
| ESF cabinets consist of two identical trains wh cas pressure and i
| |
| d sequentially load ESF equipment on
| |
| : 3. EGLS trains comprise of two identical EGLS tra ns wloads, temporarily ESF buses 34C and 34D during emergency conditions. -
| |
| 04LWTE.08D
| |
| | |
| ~
| |
| Table 3.1-3 KFINITION AIS NGENELATUE F PLAlli BMOGE STATES T YPE APPLICAM E TisIING PHYSICAL K SCRIPTION CONTAleENT $4FEGt3885 PLAIIT
| |
| & E VENT W COE F PLAlsi OPENITIONAL STATUS - SalmGE ACCIDENT TRE E S ELT ACCIDENT SEQUENCE COISI110m UNEKN SP98V IEC-SPImV STATE A-large LOCA 1) Large LOCA E-E arly Rapid loss of RCS inventory safety injection Fallere Ves Yes AEC
| |
| : 2) IIedium LOCA with fallere of safety Ves IIe M C'
| |
| ; fMection IIe Ves AEC*
| |
| IIe IIe AE A-Large LOCA Il Large LOCA L-Late Rapid loss of RC5 inventory Safety IRjection Success Yes Yes ALC
| |
| : 2) 90edf ue LOCA with failure of recircu- AND Ves 10 0 ALC' lation cooling Recirtw14tten Cooling No Yes ALC*
| |
| Fallwre No Iso AL e uv 5-Saall LOCA 1) Srell LOCA E-E arly Slow loss of K5 inventory Safety injection Failure Ves Ves SEC
| |
| : 2) Incore Instru- with failure of safety OR ,
| |
| Ves SIG ) * $r C' mient Tube injection or core heat Core Heat Removal failure No Ves - 5EC*
| |
| Rapture removal functions which OR IIe IIe SE f
| |
| : 3) Ste.m Genere- causes the reactor coolant ATM5 Pressere Spfbes tor Tube Rup- system to reesta pressurized tore
| |
| : 4) Loss of Off-
| |
| - site Power
| |
| : 5) Anticipated Transients Withewt Scram .
| |
| 5-Sea 11 LOCA 1) Small 10CA L-Late Slow loss of RCS inventory Safety Injection success Yes Yes SLC with failure of rectrcu- Als Ves so SLC*
| |
| : 2) Steam Genera- lation cooling accfrtulation Cooling IIe Ves SLC*
| |
| tor Tube Fallere IIe IIe SL Rupture e
| |
| ~ m , , s m w
| |
| -r- r- - ,w.e . . - - c-w 1 _ e v-.- i- - - - . ..w.--- .-,_-, , - - ...g,.
| |
| | |
| +
| |
| Table 3.1-3 (Continued)
| |
| MFimtilon Ace nomart ATUK OF PLANT tuwmCE STATES l
| |
| TYPE APPLICAR E !!alur, PHYSICAL K SCRIPTION CenTAlteEnf 54fEGRAA05 PLAuf l
| |
| 0F EM NT OF COE OF PLAmi GPERRTlemL STATE 5 - SeenGE ACCIDE NT TEES MLT ACCIDENT WOMERCE ConDITien gutKm SPSV MC.$PMT STATE f-Transfest 1) All Transient E-Early Transient behavior with Core seest memoval Yes Yes TEC Event Trees failure of core heat removal Failure Tes me TE C*
| |
| : 2) Steam Genera- which causes the reactor OR mo Ves TE C*
| |
| l tor Tube coolant system to remain Faf1ere te esistain RCS IAe to M i
| |
| Rapture highly presserf aed er fewentory control during
| |
| : 3) Anticipated reactor coolant pump seal a reactor pasy seal LOCA Transfent LOCAs induced ly a Isss of i Without Scram offstte power event
| |
| * T-Transient 1) All Transient L-t ate Slow loss of ACS leventory Core seest Semoval Success Yes Yes SLC Event Trees due to failure of recirce- Amp Tes me 5L C* * -
| |
| : 2) Anticipated latten cooling during feed Recletulation Coeling me Tes 5LC* '
| |
| Transient and bleed operation Failure me no SL Withist scram S*-Incore 1) Incore instra- L-tate Slow loss of K5 inventory Safety injection success me no 5*L Instrument ment Tube with recirculatten cooling ses Tube Rupture pupture fatlure Recirculation Cooling Failure :
| |
| 5*-luore Il Incore E -E arly Slow loss of RCS inventory Safety talection Failure no no 5'E Instrument Instrument with failure of safety OR !
| |
| Tube Rupture Tube fr.jection or core heat Core meat Removal Failure '
| |
| . Ampture removal functions Although this event is actually a sm.11 LOCA f t is classified as a transient event because the loss of ACS feventory is very slem coepered to a nomel small LOCA event.
| |
| | |
| ~A-4 Table 3.1-3 (Continued) e MFIulTInes Ass NOEWC1ATUK & PtAmt DApuGE STATES ComiAllmEspt 5Af tGUAmes Plent T YPE APPLICAM E TIMING PHYSICAL K50RIPTI0le SAspGE PLAmi OPERATleIIAL STAIUS -
| |
| EDE WT OF CosE OF OF SiaTE QUEK H SPARY M C-SPmAF Rti ACCIDFai 20utKE CossBITIon ACCIM mT TREES Tes Yes YNC V2-Steam 1) Steam Genera- E-E arly Stem generator tube rupture Safety injection Failus_
| |
| Yes no VNC*
| |
| with a steam leak and Dit Generator tor Tube Yes WN C*
| |
| failure of safety injection Core Heat Scooval Failure no Tube Rupture no V2E to or core heat removal functions which causes the reactor coolant system to, remain pressortzed e- i Ves Tes V2LC V2-Steam 1) Steam Genera- t -L ate Steam generator tube rupture Safety inf V2tC*
| |
| Assb Ves no ,
| |
| Generator tor Tube with a steam leak and f ail. "
| |
| no Yes V2LC*
| |
| Tube Rupture are of recirculation cooling Recirculattr a n no ago V2L Failure not mot V V-Interf acing 1) Interfacing E-Early Rapid loss of RCS f aventory not Applicable with bypass of contaivaient Applicable Appilcable Systems (OCA Systems LOCA o
| |
| c .. _ _ _ ll 5
| |
| | |
| TABLE 3.2-1 SYSTEM
| |
| | |
| ==SUMMARY==
| |
| | |
| ~
| |
| SYSTEM
| |
| | |
| ==SUMMARY==
| |
| OF KEY FEATURES FUNCTION j LPSI Two-RHR pumps, take suction from Injects into the RWST. RCS.
| |
| Accumulators - Four accumulators. Injects into the l RCS.
| |
| Main Steam Fcur MSIVs, one per loop, close To mitigate the Isolation immediately in the event of impact of postu-rupture of line. lated steamline breaks.
| |
| Quench Spray Two pumps, takes suction from Provides rapid RWST, initiate on contain- short-tera quench-ment depressurization acc.u- ing of steam within ation (CDA) signal. containment.
| |
| Containment Recirculation Four pumps (two modeled for sprayProvides long-term Spray function) take suction from contain- removal of heat ment sump. Actuated automatically from the contain-l, on High-3 containment pressure after 7:= t atmosphere.
| |
| a time delay. .
| |
| l 1 1
| |
| Service Water Consists of 60 trains. Each train Cools a number of i has an in-service and a standby important emergency pump. and normal system heat loads. i Main Electrical Two emergency buses 34C and 34D, Major-support one diesel generator dedicated to' system providing each bus. Auto start on loss of emergency AC power.
| |
| I offsite power and accident con- l ditions.
| |
| 04LW7X.08D !
| |
| ( t - N n e r *. +
| |
| m^ er _ __s __ c _.-__.____m.__-___ _.- . __m.._ ._._--u_.-_---_._
| |
| | |
| TABLE 3.2-1 SYSTDI SINelARY (CONT'D.)
| |
| SYSTEM
| |
| | |
| ==SUMMARY==
| |
| OF KEY FEATURES FUNCTION 120 VAC Four trains, supported.by 125 VDC, Support system
| |
| . and 480 VAC buses. providing control power.
| |
| 125 VDC Four trains, each train powered Support sys?.en from a battery and a battery providing control charger. power.
| |
| ESFAS Two discrete portion, an analog Support system gen-portion consisting of three or erating actuation four redundant channels per signal for ESF
| |
| . system parameter, and a digital equipment.
| |
| portion consisting of two redundant logic trains. .
| |
| 1 EGLS Two identical EGLS cabinets. Creates a predeter-Receives input signals from BUV I , . mined sequence of g 4
| |
| ' CDA 2 , RECIRC, ARBKR 3, DGBKR . actuation signals.
| |
| I .
| |
| AFW Two motor driven pumps and one Provides a supply turbine driven pump, take suction of high pressure from DWST or CST. MD, AFW pumps feedwater to the can inject to two SGs each. TDAFW secondary side of i can inject to all four SGs. steam generators j
| |
| following loss of normal feedwater flow.
| |
| 1 04LW7X.08D
| |
| - s
| |
| - _ . . . . , . . , _ . . . ~. . _ . . .._ __ . . . _ . . . . , _ _ . . . .~
| |
| | |
| ^ ^ ~
| |
| ; +i&:... ~ :.-
| |
| =-- '.
| |
| ._: .r'- - ' '. ..;-
| |
| , . . _ , ~ r_ -
| |
| _ C". _
| |
| s-, ._'' ._~- .,,
| |
| .; _. : = '
| |
| ^ ~
| |
| ~-
| |
| a _ . .
| |
| : TABLE 3.2-1 SYSTEM
| |
| | |
| ==SUMMARY==
| |
| .;(L W 'D.')
| |
| SYSTEM SUMMAD.Y OF KEY FEATURES -
| |
| FUNCTION HPSI Two charging pumps and two safety. Provides reactor- _'
| |
| injection-pumps, takes suction core .ooling and from RWST, receives' auto actua- shutdown capability tion signal fro's ESF. . by injecting borated water into.the vessel together'with. containment
| |
| . spray. pumps: recirculates borated water from-the.
| |
| ~
| |
| . containuent sump to the RCS cold' legs.
| |
| 1 - BUV: ' Bus Under Voltage :
| |
| 2 - CDA: Containment Depressurization Actuation . ;
| |
| 3 - ARBKR: Reserve Breaker ;
| |
| i 4 - DGBKR: Diesel Generator Breaker - g '- .
| |
| O ,
| |
| 1 I
| |
| l.
| |
| i i
| |
| j 4
| |
| 4 f 04LW7X.0eD
| |
| , , ,.w e A. ,.. e,- - .*- - - ,,[
| |
| -. . ...m.,#..w.-. --
| |
| -..i,... w,. . . , .__.. . .c. , ..r." .,,,,y-4 , y- --,
| |
| | |
| ( ' 'I } '
| |
| s .
| |
| tog i
| |
| , 74. '
| |
| 2.2 , [ Table - 3.2-2 c System Dependency Matrix '
| |
| -?
| |
| Plant Systems Dominent Support Systess I
| |
| Cngineered Turbine Plant Reactor Plant Electric Service .
| |
| Safety Component Component Power .W ater
| |
| _ Features cooling Coo 11nt ' ..
| |
| l
| |
| [ngt'neered Safety Fea. 3 tures Actuation M= X 4
| |
| Turbine Plant Component '
| |
| #v Lcoling NA X X Reactor Plant Cnonent
| |
| ,. Cooling NA X X
| |
| . ;c.
| |
| Electric Power I NA X Service Water X X- NA' .
| |
| Reactor Protection X X High Pressure.
| |
| Safety Injection . X X .
| |
| Lw Pressura Safety Injection .X- X X
| |
| .;i CMrging Pump i 1 Cooling X X -X :X.
| |
| - u t. t
| |
| '- 3afety injection Pump Cooling X X X -X-
| |
| .i Accumulators-
| |
| !h ,
| |
| ' Recirculation .
| |
| i Coolig X X X-
| |
| , ., a :
| |
| 71 - Main Feedwater X X n, .
| |
| W- ' Auxiliary Feedwater - X X i Main Steam X X l'
| |
| ., . Condensate X X
| |
| : -t
| |
| \ 4t.:
| |
| I
| |
| : d. ,,,
| |
| k1 - k.
| |
| I
| |
| '4 , [i -' ,s
| |
| . 4 }-
| |
| ; - 4
| |
| , , .o .
| |
| ?, '
| |
| e }/ ,
| |
| | |
| l 102' .
| |
| i :
| |
| TABLE 3.31:
| |
| | |
| ==SUMMARY==
| |
| OF SYSTEM / FUNCTION UNAVAILABILITES :
| |
| System / Function 1 Unavailability : Applicable Support States Remarks -
| |
| LAccumulators . 3.38E 04. ALL
| |
| ~
| |
| "LPSI 1.24E 02 1,5 1.75E 02 2,3,6 HPSI. 8.60E 04 1 9.20E 04 2 For LLOCA Only 1.08E-03 5 9.90E 02 6 2.51 E 04 1 1.76E 03 2 For initiators E 2.64 E 04 - 5 except ROCA .
| |
| p 2.20E 03- 6 AFW- 7.12 E 05 - 1,5
| |
| - 9.63E 04 2,3,6 4.65E 02 7 5.35E 04 1,5 1 Faulted 2.40E 02 2,3,6 Steam Generator 4.66E 02 .7 Station Blackout' 2.58E 03 7 i
| |
| Hlgh Pressure Recirc. 1 16E 02 - 1,5 S.96E 02 2 a.75E 02 6 a
| |
| 1.97E 02
| |
| ~
| |
| 1,5- Given primary 7.89E 02 2- depressurization .
| |
| 9.6 6 E-02 6 failure t.ow Pre %ure Recire. 1.44E 02 1,5 8.36E 02 2 1.05E 01 6
| |
| ^
| |
| 1.53E 02 1 1.33E 01 2 Given LPSI 1.54E 02 5 f allure -
| |
| g 1.51 E-01 6 .
| |
| l.cs3 of AFW & .1.00E 06 1 SLOCA,SSINJ,tiTR, 7 Fallto recover,MFW 2.01 E-05 1 SGTR
| |
| : or. AFW . 7.63E 07 1 Transients except loss of
| |
| .i- MFW 2.49E 06 1 Loss of MFW Failure to Recover FW ' 3.41 E 02 2 1.oss of 1 VAC
| |
| -L 3
| |
| . v.
| |
| ~ ij yg', . S,\
| |
| U b'-} > ,
| |
| | |
| 103 JTABLE 3.3-1:
| |
| | |
| ==SUMMARY==
| |
| OF SYSTEM / FUNCTION UNAVAILABILITES (Cont'd)
| |
| System / Function. Unavailability Applicable Support States Remarks
| |
| \ Feed 8. Bleed = 1.65E 01 1 8.33E 02- 2 8,17E 02 5 -
| |
| 8.29E 02 6 7.82E 02 7 Station Biackout 1.02E 02 7 Sie:'on Blackout (Fesd Only) 8.19 E 02 1 ATWS 1.61 E 01 2 Loss of 1 VAC
| |
| ' 5.00E 01 1,2,5,6 SGTR Primary - Depressurization 1.00E 02 1,2,5,6 SLOCA, llTR, o Action Only 1.65E 01 1 8.17E 02 2.5,6
| |
| ' 5.10E 02 1,5 AFW Avaitat.le 5.58d-02 2,3,6 AFW Available
| |
| [ _
| |
| Secondary Depressurization 2.31 E 01 1,5 MLOCA
| |
| _ 2.35 E.01 2,6 htOCA f 1.00E 02 1,2,5,6 SGTR Quench Spray 1.22E 03 1,5 !
| |
| 1.34E 02 2,3,6 1 2.39E 01 7 Station Blackout 6.04E 01 7 Station Blackout <
| |
| Containment Spray 4.87E 03 1,5 ;
| |
| 4.82E 02
| |
| - ~
| |
| E-4.22E 03 '
| |
| 1,5 Without HPR/CS A. '16 02 2,6 C'F !
| |
| h Stuck Open Turbine 3.o2E 04 1,2,3,4 Relle! Valves 2.39E 02 5 3.52E-02 6
| |
| - 4.63E 02 7 i 2.36 E-02 8
| |
| 7 2.38E 02 2,6 Loss of 1 VOC
| |
| [ 3.50E 02 7 Loss,of1 VDC Stuck Open Turbine 4.17E 04 1,2,3,4 Bypass Valves
| |
| . RWST 1.92E 08 ALL -
| |
| h m
| |
| M
| |
| - - 5Mi / e gg w
| |
| | |
| ',' - 104 [
| |
| TIBLE 3.3-1:
| |
| | |
| ==SUMMARY==
| |
| 0F SYSTEM / FUNCTION UNAVAILABILITES (Cont'd) _
| |
| 4 . System / Function. Unavailabilityj Appfcable Support States . Remarks Reactor Trip 6.54E 06i ALL AWO & Manual 3.00 E-05 ALL E l ATWS (Manual) 2.16 E 01 - ALL .
| |
| RT & Emergency 4.56E 02 1 Boration 4.58 Ee 02 2 ATWS..
| |
| 4.56 E 02. 5-4.59E 02 6 o Turbine Trip 1.44 E 01 ALL ATWS Power Level > 25%' 4.30E 02 ALL ATWS
| |
| . Pressure' Rellef 3.00 E 01' ALL ATWS
| |
| : Steam Leak 5.88E 03 ALL SGTR 1.00E 01 ALL SGTR;Given AFW and ,
| |
| Feed and Bleed Failure, Operator Action
| |
| . Control SI 1.00E 02 1,5 SLBIC 1.00E 03 2,6 SLBIC 1,5 SLBOC i
| |
| : 1.00 E 01 1.00E 02 2,6 St.BOC Terminal SI- 1.00 E O'1 ' 1,5 SSINJ, Operator ,
| |
| 1.'00 E 02 2,6 Action j.
| |
| Station Blackout t
| |
| . RCP LOCA ' 8.58 E 02 L 7 L'
| |
| 1.64 E 01 7 Station Blackout l
| |
| !' Consequential LOCA 7.78E 06 1,2,3,4,5 L 1.54E 05 6-
| |
| [ MSIV Closure - 8.20E 04 ALL ALL SLEIC BLBOC i 1.50E 03 ,
| |
| l! Fall to Recover OSP 6.47E 01 7 1~ Hour 5.91 E 01 7 2 Hours /1 Hour 3.54E 01 7 6 Ho'urs/2 Hour's L SI Signal ~ Actuation . 1.50E 06 1,2,3,5,6,7 .
| |
| l ~ 1.00 E-01 4,8 0
| |
| l; D9tay Cont Recire. 1.00 E-01 1,2,5,6 IITR -
| |
| j .COF Between 1.59E 03 1,5
| |
| -CS & Recire. 1.24F 02 2,6 o
| |
| j h;. n u ,
| |
| L: . - - -_ __ -
| |
| | |
| g.
| |
| 105~
| |
| LTABLE 3.3-1: SOMMARY OF SYSTEM / FUNCTION UNAVAILABILITES (Cont'd)
| |
| Support State Summary Description -
| |
| . SS 1 ' All support states available.'
| |
| 'SS 5 Identical to SS1.except OSP unavailable.
| |
| SS 2 Both EAC available,1 train of ESF, EGLS, &
| |
| SW trains unavailable.
| |
| Identical to SS2 except OSP unavailable and I SS 6 aq only 1 EAC train is available.
| |
| y.
| |
| SS 3 L No service water,1 train of ESF & EGLS,- l
| |
| ~'
| |
| EAC Available since OSP is unavailable.
| |
| SS 7 No EAC, No SW, No EGLS,1 train of ESF il cabinets available. OSP unavailable.
| |
| SS 4 No ESF, NO EGLS, No SW, OSP available.
| |
| SS 8 Same as SS-4 except offsite power unavailable.
| |
| l m 1 7 EAC Emergency AC Power. ;
| |
| EGLSl, Emergency Generator Load Sequencer ESF: Emergency Safety Features Actuation System j
| |
| -FW Feedwater
| |
| [ . IITR Incore Instrument Tube Rupture
| |
| , LLOCA Large LOCA
| |
| - . MFW Main Fe twater q 05P Offsite power RCP Reactor Coolant Pump RWST - Refueling Water Storage Tank SGR Steam Generator Tube Rupture -
| |
| .{
| |
| _ SLBIC . Steamline Break inside Containment SLBCC Steamline Break Outside Containment SI ' Safety injection
| |
| , SLCCA Smal LOCA
| |
| , - SSINJ . Spurious Safety injection SW- ~ Service Water VAC ' Vital AC power t.-
| |
| +!'.
| |
| - s k?.
| |
| E
| |
| _m
| |
| | |
| ll;
| |
| {o r 5 7 4 6 1 fo o 9 0
| |
| 0 0
| |
| 0 1
| |
| 3 9 E E E E E s 0 9 0 5 0 4 8 2 sC.
| |
| oAV 0 6 0
| |
| 3 6 L 2 1 3 2 9 .
| |
| 2 6 9 5 5 4 fr o 0 0 0
| |
| 0 0
| |
| 0 o.. 0 E E E E E s1 0 1 5 0 5 2 4 s 1 3 1 0 0 0 oC - . .
| |
| LA 3 1 3 2 7 V
| |
| C 4 f 0 S
| |
| E o Dse 0 0
| |
| E I s la s 0 0 0 0 0 7 0 T s u 1 0 I o to B - .
| |
| L L T 3 I
| |
| B A
| |
| B j
| |
| O R f s 5 5 4 6 u 0 0 0 0 P oB 0 0 -
| |
| E E 0 E E
| |
| - 'p .
| |
| E T sC 0 1
| |
| 3 7 5 2 6 A sD o -
| |
| 3 1
| |
| 0 0 5
| |
| T L 1 3 1 3 2 3 S
| |
| T R n O f i 5 7 4 5 9 P 0 0 o Tar 0 0 0 0 P 0 - - - - -
| |
| U E E 0 E E E S s 0 1 4 4 2 4 1
| |
| sW oS -
| |
| 0 6 0 7 0
| |
| : L 2 1 3 2 5 2 1 3
| |
| 3 2 4 7 fe 0 0 0 e otr 4 - - -
| |
| l ie 8 E E E b ssw 0 0 0 0 9 4 3 7 a sfo 5 1 0 T ofP . .
| |
| LO 1 2 7 t s 6 8 0 r r 1 3 7 7 4 omdo pe e t 0 0 0 0 0 0 0 1
| |
| pt t a E E E E E E E E us a i 6 5 6 4 2 2 3 3 sy t l 9 0 5 6 0 7 5 0 nS Rn ei . . . . .
| |
| 4 6
| |
| 1 o 9 4 2 1 3 N I t
| |
| re 1 2 3 4 5 6 7 8 ot pa - - - - - - - -
| |
| D pt S S S S S S S S 8 uS S S S S S S S S 0 1
| |
| S 7 W
| |
| L 4
| |
| _ 0 m
| |
| - ~
| |
| 2 i.
| |
| | |
| ?, y - - _ - _ _ _ _ .
| |
| 1
| |
| 's 107 s
| |
| Table 3.3-3 Significant Contributions to CDF from Internal Flooding Sequence Frequency-(YR'l)
| |
| Flood caused by a break in 8 x 10-7 the heating and air condition system cf switchgear or cable spreading room Flood in-diese2. generator 8.6 x 10~9 enclosure with loss of off-site power and the unavail-ability of the redundant diesel-Break in'a SW. pipe in'the 4.94 x 10-8 Circulating and Service Water Building pipe.and the unavailability of the other '
| |
| SW train Source: MP3 PSS'(August 1987 Update) b' I
| |
| +
| |
| (
| |
| k l
| |
| l 04LW7X.040-o4
| |
| | |
| ay 4
| |
| l 108 i
| |
| .- I Table 3.4-1 Core Damage Frequency by Initiator Eveat Initiator Centributies Peroaat svent. to .
| |
| Contri-Frequency (Yr*l) .Corea bution molt (Tr*ll Large LOCA 3.88 x 10** 8.03 x 10"'' 14.54-Medium IACA 6.11 x 10** 1.03 x 10*5' 18.55 Small IDCA 9.07 x 10~3 2.42 x 10*' 4.37 SGTR 3.92 x 10-2 1.18 x 10'' 2.14' St. Break Insida Containment 3.88 x 10**' 5.85 x 10*' O.11 fS/LBreakoutsideContainment 3.78 x 10-2 8.12 x 10~' 14.69' fLossofRCSFlow 4.91 x 10'l 4.49 x 10"7 0.81 Loss,caf MFW P.60 x 10'l 1.05 x 10-6 1.90 3 y Power Misaatch 4.30 x 10*1 3.93 x 10*7 0.71
| |
| 'I u bine Trip _ 2.25 2.06 x 10'' 3.73 R9 actor Trip 7.50 x 10*1 6.86 x 10'7 l'.24 Core Power Excursion 7.18 x 10-2 6.57 x 10'0- 0.12 Spurious SI 4. 99 x 10-2 4.78 x 10'8 0.09 Loss of Offsite Power 1.12 x 1071 4.99 x 10-6 9,o3 Incore Inst. Tube Rupture 9.20 x 10~4 2.46'x 10"7~ 0.45 ISLOCA
| |
| -- 2.21Ix 10~7 0.40 Vessel Rupture -- 3.00.x'10*7 0.54 Loss of 1 SW-' Train 1. 8'd 10-2 2.75 x 10*' 4.98 Loss of 1 VDC Bus 3.92 x 10'3 3.72 x 10-6 6.73 Loss of Total VDC 1.40 x 10-8 6.35'x 10"10 ~0 Loss of VAC -1 or 6.15 x 10-2 1.58 x 1C"' 2.13 Loss of VAC -3 or -4 6.15 x 10-2 4.72 x 10 0.85 ATWS'
| |
| -- 3. 3 8 x 10-6 ' 6.12 consequential Small LOCA -- 1.21 x 10'' -
| |
| '2.19-Consequential S/L B.O.C. -- 6.73 x 10-7 1.22 __
| |
| Consequential S/L B.I.C. 2 1.29 x 10"' '2.33 TOTAL 5.524 x 10~3 100 64U87X.000
| |
| ,g.
| |
| t
| |
| .,j f - : 1 i!
| |
| .]:
| |
| a
| |
| | |
| . . . . - .- . _ _ -- ~. - . . . . . . _ . -.
| |
| m ,
| |
| ,k .
| |
| c 109 Table 23.4-2 !
| |
| , ' Core Damage Frequency by Plant Damage htate DAMAGE CLASS CORE MELT o P,
| |
| AE 6.131E-09' AEC -3.699E-06 l:0 '
| |
| AEC'~ 1.741E-08 AL 2.550E-09 _
| |
| ALC 1.317E-05 1 i- ALC" 2.048E L ALC' 1.671E-06 L S'E 1.623E-09 .
| |
| E S'L 1.239E-03 l SE 1.275E-06 SEC. 1.099E-06 SEC' 6.701E-09 SL 1.136E-08 SLC 6.739E-06 l-SLC"- 5.049E-08 SLC'''. 1.142E-06 TE 3.728E-06 L
| |
| TEC- 2.172E-05 TEC' 4.858E-07 V 2.213E-07 V2E' 3.774E-09 V2EC 1.511E-07 V2EC' .1.191E-08 f I
| |
| : V2L 4.359E-12 H l-p V2LC 1.369E-09
| |
| !- V2LC" 1.804E-11 V2LC'' 3.267E Total: Frequency (Yr-1) 5.524E-05
| |
| ,~ f
| |
| ) [
| |
| I,
| |
| * l l-I t .'jf_n' i
| |
| .,3.04LW7X 080 JE' ,
| |
| I iM
| |
| -.4.
| |
| | |
| % ,c . . . _ - . ~ .
| |
| ~
| |
| [ ' e _ [
| |
| TABLE.3.4
| |
| | |
| ==SUMMARY==
| |
| OF SIGNIFICANT SEQUENCES' -~
| |
| CONTRIBUTION TO MURCE' TERM CIASSES
| |
| -PER- M1A MIB M2 M3 M4 PLANT .
| |
| DAMAGE FREQ. ' CENT MS M6 M7' M8 - M9 ACCIDENT SEQUENCE EVENTS M10 Mll M12
| |
| . CLASS ' I1EASURE (%) 'l 12.6 ET2 M'DIUM'LOCA 0 0 2.02E-09 0 1.39E-09 6.95E-06
| |
| : 1. ALC. E 0 0 0 4.43E-09 8.20E-10 SSI- SUPPORT STATE 1 GH500. HIGU PRESSURE RECIRC. FAILURE O 1.47E-07 6.5 9E-06
| |
| : 2. ALC 5.78E-06 10.5 ET1 LARGE LOCA 0 0 1. 68 E-09 0 1.16E-09 SSI- SUPPORT STATE 1 0 0 0 3.6;E-09 6.82E-10 LPR4 LOW PRESSURE'RECIRC. FAILURE O 2.88E-07 5.4 0E-06
| |
| : 3. TEC 4.58E-06 8.3 ET6 STEAM LINE BREAK OUTSIDE CONTAINMENT .0 8.70E*10 0 4.58E-10 9.16E-10.
| |
| SS1: SUPPORT STATE 1 0 0: ; O 2.92E-09 5.40E-10 GM7001 MSIV CLOSURE FAILURE O .2.29E-07 4.34E-06 GP2006 -PRIMARY DEPRESSURIZATION FAILURE
| |
| : 4. TEC 3.45E-06 6.2 ET18 LOSS OF 1 VITAL OC BUS O 6.55E-10 0 3.45E-10 6.90E-10 sS2DCI SUPPORT' STATE 2 0 0 0 2.20E-09 4.07E-10 CF2002 AUXILIARY FEEDWATER FAILURE O 1.72E-07 3.27E-06 0 4.01E-10 0 2.11E-10 4.22E-10 ~ --
| |
| : 5. TEC 2.llE-06 3.0 ET22SSI ATWS: SUPPORT STATE 1 SSI SUPPORT STATE 1 0 0 0 1.34E-09 2.49E-10 - ((-
| |
| RT3 MANUAL REACTOR TRIP O 1.05E-Oi 2.00E-06 GP2023 IEED AND BLLED FAILURE 6 AEC 2.04E-06 3.7 ET2 MEDIUM LOCA 0 0 5.92E-10 0 4.08E-10 SSI SUPPORT STATE-1 0 0 0 1.30E-09. 2.41E-10 CVA004 ACCUMULATOR FAILURE O 1.02E-07 1.93E-06
| |
| : 7. TE 1.79E-06 3.2 ET14 LOSS OF OFFSITE POWER 0 3.40E 10' O 1.81E-10 3.58E-10 SS7LOSP SUPPORT STATE,7 7.80E-09 0 1.60E-06 0 O E60 FAILURE TO RECOVER OSP.IN 1 HR 1.76E-07 0' 3.54E-11 E120 FAIL TO REC.-OSP IN 2 HRS / FAIL. IN 1 HR.
| |
| E6H FAIL. TO RECOVER OSP IN 6 HRS / FAILURE TI GS3002 QUENCH SPRAY FAILURE
| |
| : 8. SLC 1.76E-06 3.2 ET3 SMALL LOCA. 0 0 0 5.10E-10 3.52E-10 SSI SUPPORT STATE'1 .
| |
| 0 0 0 - 1.12E-09 2.0BE-10 OAX3 PRIMARY DEPRESSURIZATION FAILURE O 8.78E-C8 1. 67 E-0 6 GH8001' HIGH PRESSURE RECIRC. FAILURE
| |
| : 9. TEC 1.63E-06 3.0 ET6 STEAM LINE dREAK OU731DE CONTAINMENT 0 3.10E-10 0 1.63E-10 3.26E-10 SSI SUPPORT STATE 1 0 0 0 .
| |
| 1.04E-09 1.92E-10 GF4001 AUXILIARY FEEDWATER FAILURE O 8.13E-08 ' 1.55 E-06 .
| |
| GP2006 PRIMARY DEPRESSURIZATION FAILURE
| |
| : 10. TEC 1.36E-06 2.5 ET17 . 00SS OF 1 SERVICE WATER '. RAIN -0 2.58E-10' O 1.36E-10 2.72E-10 SS2SW1 SUPPORT STATE 2 0 0 0 8.66E-10 1.60E-10 CF2032 AUXILIARY FEEDWATER FAILURE O 6.79E l'.29E-06 GP2007 FFED AND 11LEED FAILURE W
| |
| ps w .-=.q a- a -
| |
| , .m- ,puf . ,.,.-f .M g--- - i- ,, 3-, -, - - ,,,p.. a ,,m.,.M
| |
| | |
| " 3: .
| |
| ff , ' - ' ,
| |
| L'I ~ ' 7 ~*
| |
| k
| |
| * n' y_ a g
| |
| ,. _ , y , , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
| |
| '2e 7 m--
| |
| *4, m ~ ~
| |
| ~
| |
| 4
| |
| ~
| |
| TABLE 3.4-3:
| |
| | |
| ==SUMMARY==
| |
| OF SIGNIFICANT SEQUENCES (Continued) .
| |
| CONTRIBUTION . TO SOURCE TERM CLASSES M1B M2 - M3 M4 M1A PER- M6 M7 M8 M9 PLANT MS DAMAGE FREQ. CENT .
| |
| M10 Mll M12 CLAS5 tTASURE (4) ACCIDENT SEQUENCE EVENTS 0 0 3.77E-10 0 - 2.60E-10 I I .' AEC 1.30E 2.4 ET1 LARGE LOCA . 0 0 0 8.28E-10 1.53E-10 SSI SUPPORI-STATE 1 O 6.49E-08 1.23E-06 CVA004 ACCUMULATOR FAILURE 0 2.22E-10 0- 1.17E-10 2.34E-10
| |
| : 12. TEC 1.17E-06 2.1 ET14 LOSS OF OFFSITE POWER 0 0 0 7.45E 1.30E-10 SS7LOSP SUPPORT STATE 1 .
| |
| -0 5. 4 8E 1.11E-06 E60 FAILURE TO RECOVER OSP IN 1 HR E120 FAIL TO REC. OSF IN 2 HRS / FAIL. IN 1 HR.
| |
| E6H FAIL.' TO RECOVER OSP. IN 6 HF.SffAILURE TI O O 1.21 E-0 9 0 2.10E-10 1.05E-06 1.9 ET23SS7 CONSEQUENTI AL SMALL LOCA: SUPPOR7 E1 ATE 6.48E-08 9.36E-07 0 0
| |
| : 13. SE 0 SS7 SUPPORT STATE 7 1.47E-11 4.86E-08 0 O O O 2.99E-10 2.007 10 1.03E-06 1.9 ET17 LOSS OF 1 SERVICE WATER TRAIN 0 0 6.56E-10 1.12.:-10
| |
| : 14. SLC 0 SS2SW1 SUPPORT STATE 2 O 5.2.2-08 9.76E-07 ,,
| |
| GF2002 AUXILIARY FEEDWATER FAILURE CH6001 HIGH JRESSURE RECIRC. FAILURE 0 0 2.31E-10 0 1.93E-10
| |
| : 15. A LC' 9.63E-07 1.7 ET2 MEDIUM LOCA 0 0 0 6.52E-10 9.61E-07 SSI SUPPORT STATE 1 9.41E-10 2.08E-11 0
| |
| RSS3 CCF BETWEEN RECIRC. AND CONT. SPRAY 0 1.37E-10 0 7.26E-11 -1.44E-10 7.19E-07 1.3 ET20 LOSS OF VAC BUS 1 OR 3 3.13E-09 0 6. 4 5 E-07 0 0-
| |
| : 16. TE
| |
| -SS4VAC12 SUPPORT STATE 4 7.07E-00 0 9.42E-11 O 1.'30E-10 0 6.84E-11 1.37E-10
| |
| : 17. TEC 6.RSE-07 1.2 ET10 TURBINE TRIP 'O O O 4.36E-10 0.00E-11 SS2 SUPPORT STATE 2 3.42E-08 6.49E ,t, O
| |
| CF2002 AUXILIARY FEEDWATER FAILURE CP2007 FEED AND BLEED FAILURE O 1.16E-10 .0 6.10E-11 1.22E-10
| |
| : 18. TEC 6.11E-07 1.1 ET14 00SS OF OFFSITE POWER 0 0 0 3.89E-10 7.21E-11 SSSLOSP SUPPORT 3 TATE 5 0 3.05E-08 5.79E-07 CF1001 < module >-
| |
| GP2011 FEED AND BLEED FAILURE 0 0 1.45E-10 0 1.21E-10 ALC' 6.04E-07 1.1 ETI LARGE LOCA 0 0 4.09E-10 6.03E 19. 0 SS1 ' . SUPPORT STATE 1 5.90E-10 1.30E-11 0
| |
| RSS3 -CCF BETWEEN RECIRC. AND CONT. SPRAY 0 0 0 1.72E-10 1.190-10
| |
| : 20. S LC 5.94E-07 1.1 ET6 STEAM LINE BREAK OUTSIDE CONTAINMENT 0 0 0 3.78E-10 ?.01E-11 SSI SUPPORT STAJE 1 0 2.96E-08 5.63E-07 CM2001 MSIV CLOSURE FAILURE -
| |
| GH5001 HIC:! PRESSURE RECIRC. F AI LURE
| |
| ~
| |
| d
| |
| _ . . _ _ _ _ h
| |
| | |
| 7 ,- - a s :g n TABLE 3.4-3:
| |
| | |
| ==SUMMARY==
| |
| OF SIGNIFICANT SEQUENCES (Continued)
| |
| CONTRIBUTION TO SOURCE TERM CLASSES -
| |
| 1 P LANT PER- 'M1A M1B M2 . M3 M4 DAMACE FREQ. CENT M5 M6 M7 M8 M9 CLASS MEASURE (%) -ACCIDENT SEQUENCE EVENTS M10' Mil M12 i
| |
| : 21. SLC 5.19E-07 9 ET10 ' TURB1NE TRIP O O O . 1.51E-10 1.04E-10 SS2 SUPPORT STATE 2- 0- 0 0 3.31E-10 6.12E-11.
| |
| CF2002 AUXILIARY FEEDWATER FAILURE O 2.59E-08 4.92E-07 GH6001 HICH PRESSURE RECIRC. FAILURE
| |
| : 22. TE 3.69E-07 1 ET10 . TURBINE TRIP O- 7.01E-11. O' 3.73E-11 s.38E-11~
| |
| SS4 SUPPORT STATE 4 1.61E-09 0 3.21E-07 0 0 3.63E-08 0 7.itE *2
| |
| : 23. TEC 3.55E-07 .6 ET22SSI ATWS: SUPPORT STATE 1 0 6.75E-11 0 3.55E-11 7.10E-11 SS1 SUPPORT STATE 1 0 0 0 2.26E-10 4.19E-11 1.17E-08 RT3 MANUAL REACTOR TRIP O 3. 3 7 E-0 7 TT TURBINE FAILS TO TRIP CP2023: FEED AND BLEED FAILURE
| |
| : 24. TEC 3.50E-07 .6 ET8 LOSS OF MAIN FEEDWATER. 0 4.65E-11 0 3.50E-11 1.00E-11 SSI SUPPORT STATE 1 'O O O 2.23E-10 4.13E-11 E$
| |
| GFN004 LCSS OF AFW & FAIL TO RECOVER FW 0 1.75E-08 3.32E-07 GP2014 FEED AND BLEED FAILURE
| |
| : 25. SEC 3.47E-07 .6 ET22SSI SUPPORT STATE I ATWS 0 0 1.01E-10 0 6.94E-11 SSI SUPPORT STATE 1 0 0 0 2.21E-10 4.09E-11 RT3 MANUAL REACTOR TRIP O 1.73E-08 3.29E-07 PL POWER LEVEL > 254 PR RELIEF VALVES FAIL TO OPEN
| |
| : 26. TEC 3.15E-07 .6 ET14 UOSS OF OFFSITE POWER. 0 5.90E-11 0 3.15E-11 6.30E-11 SS7LOSP SUPPORT STATE 7 0 0 0 2.01E-10 3.72E-11 CF3002 AUXILIARY FEE 0 WATER FAILURE O 1.57E-08 2.99E-07 E60 FAILURE TO RECOVER OSP IN 1 HR E120 FAIL TO REC. OSP IN 2 HRS / FAIL. IN 1 HR.
| |
| : 27. TEC 3.02E-Ol .5 ET20 LOSS OF VAC BUS 1 OR 3 0 5.74E-11 0 3,02E-11 6.04F-11 SS2VAC12' SUPPORT STATE 2-- O. 0 0 1.92E-10 3.56E-11 CF2002 AIIXILIARY FEEDWATER FAILURE O 1.51E-08 2.86E-07 GN3001 MAIN FEEDWATER RECOVERY FAILURE .
| |
| GP2022 FEED AND BLEED FAILURE
| |
| : 28. TEC 3.02E-07 .5 ET21 LOSS OF VAC BUS -3 OR -4 0 5.74E-Il 'O - 3.02E-11 6.04E-11 SSivAC34 SUPPORT STATE 2 0 L _0 GF2002 AUXILIARY FEEDWATER FAILURE 'O 1.51E-08 2.86E-07 ' 1.92E-10 3.56E-11 GN3001 MAIN FEEDWATER RECOVERY FAILURE CP2022 ' FEED AND BLEED FAILURE'
| |
| : 29. AEC 3.00E-07 .5 551- SUPPORT STATE IUNITI :. 0 ' O 8.70E-11' 0- 6.00e-11 VES. VESSEL RUPTURE O O O 1.91E-10 3.54E-11 0 1.50E-08 .2.84E-07'
| |
| - , g n .: m - ,- , ,.,.i - - - .n.. ,. - ._ ,, -A
| |
| | |
| k 5 J@ ~ ,27 j K ], :$g; "
| |
| y H-Q ' - _
| |
| - +--. 7% :q p .-c. _ - y _g < , ,( qq
| |
| . . g,g g
| |
| - , mye w 7. - n. - - 9 ,
| |
| .y.. ~
| |
| y ' -
| |
| w.
| |
| ., g.
| |
| ~ '
| |
| E
| |
| - A J h-
| |
| , . g.9 4
| |
| - = , .m +.
| |
| .m . ,, - . ,,
| |
| NTAB'LEs3.4-3: -1
| |
| | |
| ==SUMMARY==
| |
| O' F SIGNIFICANT SEQUENCES - (Continuedl -
| |
| y gy,
| |
| ~
| |
| CONTRIBUTION TO SOURCE TERM CLASSES e _
| |
| M2. ' M3
| |
| -PLANT- PER _ - 1 M1 A . ~ M1B '
| |
| MS- ;M6- M7- L M8 y DAMAGE FREQ. :- CENT .. . .
| |
| ~
| |
| CLASS. .. MEASURE' (4) ACCIDENT. SEQUENCE EVENTS : M10 - Mil: M12 s
| |
| , r
| |
| : 30. TEC 2.86E-07 .5 ET14 :
| |
| LOSS OF OFFSITE ~ POWER - 0 5.4 3E 0 : -
| |
| ~ 2.86E-11 m- 5.72E-II- -
| |
| SS7LOSP. SUPPORT STATE,7. ,
| |
| 101 .
| |
| 0 0' 1.A2E-10H3.37E-11
| |
| ~ '
| |
| CF3002 LAUXILIARY'FEEDWATER FAILURE O ' 1.43E-08 -2.71E-07.
| |
| E60 FAILURE TO RECOVER OSP IN 1 RR. .
| |
| ~
| |
| ~
| |
| 2.81E-07 . STEAM LINE BREAK OUTSIDE' CONTAINMENT 0 5.34E-11 0 ;2.81E-11 5.62E-11 -
| |
| '31. 'TEC .5'ET6 SS2' ' SUPPORT STATE.2L 0' O. .0 1.79E-10 ^3.32E CF5601- AUXILIARY FEEDWATER FAILURE- 0' 1.40E-08,J2.66E-07 4
| |
| . CP2006- PRIMARY DEPRESSURIZATION FAILURE
| |
| : 32. TEC 2.80E-07 .5 ET10 - TURBINE TRIP O; 5.32E-11'~0 2.80E-11 5.s0E-11" -
| |
| SSI - SUPPORT STATE ~1 0- 0- ..0 1.18E-10 3.30E-11' _
| |
| GFN001 -LOSS OF AFW FAIL TO RECOVER FWL 0 - 1.40E-08. 2.65E-07 CP2014 FEED AND BLEED FAILURE , ,
| |
| : 33. TEC 2.71E-07 .5 ET25SS2 CONSEQUENTIAL LOCA INSIDE CONT:SS2 0' 5.15E-11 0 2.71E-11 '5.42E SS2 SUPPORT ~ STATE 2' .
| |
| 0 0' O-. ..
| |
| 1.73E-10 3.20E-11 CF5601 . AUXILIARY FEEDWATER FAILURE' O - 1.35E-08 2.57E-0 7 - ,, y-CP2006 PRIMARY DEPRESSURIZATION FAILURE: . LJ
| |
| : 34. TEC 2.62E-07 .5 ET4 STEAM GENERATOR" TUBE RUPTURE O 4.98E-11 0 2.62E-11 5.24E-11 ~
| |
| SS2 SUPPORT STATE 2~ 0 0 .
| |
| 0 1.61E-IO 3.09E-11
| |
| - GF5601 AUXILIARY'FEEDWATER FAILURE- 0 1.31E-00 2.48E-07 CP2006 PRIMARY DEPRESSURIZATION FAILURE
| |
| : 35. TEC 2.62E-07 .5 ET8, - LOSS OF MAIN FEEDWATER. 0- 4.98E-11 0~ ,2I62E-11~S.24E-11 SS2 SUPPORT STATE 2. 0 0 0' 1.67E-le 3.09E-11' GF2002 AUXILIARY FEEDWATER FAILUnE 0 1.31E-08 2.48E-07 CP2007 FEED AND BLEED FAILURE
| |
| : 36. SEC 2.54E-07 .5 ET4 . STEAM GENERATOR TUBE RUPTURE. 0 0 7.37E-11 0 . ~5.08E-11' SSI SUPPORT STATE:1 0 0 0 1.62E-10 3.00E-11 ET4 MANUAL OR. AUTO. REACTOR TRIP' 'O 1.27E-08 :2.41E-07
| |
| : 37. TF.C 2.tlE-07 .4 ET24SSI CONSEQUENTIAL S/L BREAK OUTSIDE CONT.: 4.69E-11' O 2.47E-11 4.94E-lb SS1 . SUPPORT STATE-1 .
| |
| 0 0 .
| |
| 0 1.57E-10~2.91E-11' 1 GM2001' MSIV. CLOSURE FAILURE 0, 1.23E-08 2.34E-07 :
| |
| GP2006 ' PRIMARY ' DEPHESSURIZATION FAILURE
| |
| : 18. TEC 2.45E-07 .4 ET6 STEAM LINE BREAK OUTS 1DE CONTAINMENT 0 4.65E 0 2.45E-11 4.90E-11
| |
| ' ES1' SUPPORT STATE'1 .
| |
| 'O 0- 0 1.56E-10' 2.89E-11 t
| |
| .. RT4- . MANUAL OR AUTO. REACTOR TRIP 0L , 1.22E-08 J2.32E-07 - a- -1
| |
| : 39. TEC 2.28E-07 .4 ET11 - REACTOR TRIP" 0- - 4.33E-11. 0; 4.28E-111 4.56E - SS2 - SUPPORT STATE 2- 20' OE rOf 1.45Ei 10 -2.69E-11
| |
| - GF2002 ' AUXILIARY FEEDWATER FAILUkEs 'O# ' 1.14E-08 ~ 2 216E-0 7.. -
| |
| GP2007 FEED AND GLEED FAILURE' , ,
| |
| . , ,.&m.. ,m .. N . y n , ,-y e - &-nw 4 L 'c + < -i< a ,:r, e%- - m - W~. "On , m l ~ l ; , .r. . .',.~,?,,-v+'.,...A,,.,,N,.-, ,,0.,p..m., ,. . , , . , . . + . , - , . ,
| |
| | |
| ~ - - -m_ -r - - --: . me ~- -
| |
| ,.b 1
| |
| ~
| |
| 4 I
| |
| ~
| |
| ~
| |
| ' TABLE 3.4-3
| |
| | |
| ==SUMMARY==
| |
| OF SIGNIFICANT SEQUENCES'(Continued) ~
| |
| e
| |
| ' CON.RIBUTICN TO SOURCE TERM Cl. ASSES PLANT PER--
| |
| M1A MIB M2 M3 ~ M4 DAMAGE FREQ. CENT . .
| |
| . MS M6 M7 una - lM9 CLASS MEASURE (%) ACCIDENT SEQUENCE EVENTS' M10' Mll . M12 -
| |
| : 40. SLC- 2.24E-01 .4 ET4 STEAM GENERATOR TUBE RUPTURE- 0 0 0- 6.50E-11 4.48E-11 SS2 SUPPORT STATc 2 0 0 0- 1.43E-10 '2.64E-11 CF5601 AUXILIARY FEEDWATER FAILURE O 1.12E-08 2.21E-07 GH6001 HIGH PRESSURE RECIRC.' FAILURE
| |
| : 41. SLC 2.17E-07 .4 ET6 STEAM LINE 3REAK CUTSIDE CONTAINMENT 0 0' -0 6.29E-11 4.34E-11 SS2 SUPPORT STATE 2 0 0 0 1.3eE-10 '2.56E GF5601 AUXILIARY FEEDWATER FAILURE O 1.00E-08 2.06E-07 G26001 91GH PRESSURE REC 1RC . FAILURE
| |
| : 42. SLC 2.14E-07 4 ET3 SMALL LOCA 0 0 0 6.21E-11 4.28E-11 SSI SUPPORT STATE 1 0 0 0 1.36E-;; 2.53E-11 GRE001 'HPSI AND HPR FAILURE 0 1.07E-08 2.03e-07
| |
| : 43. SLC 2.12E-07 .4 ET6 STEAM lit'E BREAK OUTSIDE CONTAINMENT 0 0 0 6.15E-11. 4.24E-11 SSI SUPPORT STATE 1 0 0 0 - 1.35E-10 2.50E-11
| |
| "'.5001 AUXILIARY FEEDWATER FAILURE O. 1.06E-08 2. 01 E-01 GH5001 ~HIGH PRESSURE RECIRC. FAILURE (( '
| |
| c~ .
| |
| : 44. SLC 2.10E-07 .4 Er25SS2 SUPPORT STATE 2CONSEQ.ENTIAL LOCA 0 0 0 6.09E-11 4.20E-11 SS2 SUPPORT STAfr. 2 0 0 0 _
| |
| 1.34E-10 2.48E-11 CF5601 AUXILIARY FEEDWATER FAILURE O 1.05E-08 1.99E-07 GH6001 HIGH PRESSURE RECIRCf FAILURE
| |
| : 45. SLC 1.90E-07 .4 ET8 LOSS CF MAIN TEEDWATER O O O 5.74E-11 3.96E-11 SS2 SUPPORT STATE 2 0 0 0 1.26E-10 2.34E-11 CF2002 AUXILIARY FEEDWATER FAILURE O 9.88E-09 1. 8 8 E-0 7 GH6001 HIGH PRESSURE RECIRC. FAILURE
| |
| : 46. SLC' 1.95E-07 .4 ET17 LOSS OF 1 SERVICE WATER TRAIN O O O 5.65E-11 3.90E-11 SS2SW1 SUPPORT STATE'2 1.24E-10 0 1.95E-07 0 0 CF2002 AUXILIARY FEEDWATER FAILURE- 0 1.88E-10 6.84E-12 RSS4 CCF BETWEEN RECIRC. AND CS.
| |
| : 47. SEC 1.95E-07 .4 ET22SSI ATWS: SUPPO~* STATE 1 0 0 5.65E-11 0 3.90E-11 SSI SUPPORT STATF . 0 0 0 1.24E-10 2.30E-11 RT3 MANUA! MEACTOR 1.IP O 9.73E-09 1.85E-07 TT 1 CRC ~NE FAILS TO TRIP PL POWER LEVEL > 25%
| |
| : 48. V 1.93E-07 .3 ET16A INTEkFACING SYSTEMS-LOCA 1.93E-07 0 0- 0 0 SS12 SUPPORT. STATE IUhlTY . 0 0 'O 0. O VRHRSUCT ISLOCA TitROUGH RHR SUCTION PATH 'O- 0 0 .0 0
| |
| : 49. ALC 1.90E-07 .3 ET1 LARGE LOCA C 0 15.51E-11 0 3.80E-11 SS2 SUPPORT STATE 2 'O O O 1.21E-10 ~2.24E-11 GV5601 LOW PRESSURE HECIRC. FAILURE : 0, 9.48E-09 1.8UE-01
| |
| , ..s c , _ , _ . ~ - , , ,- - . . - _. r i , - . . , , - , _ , , .,r.,,..
| |
| | |
| , , .m _
| |
| --z -
| |
| .w .-
| |
| g .-
| |
| .#... 7.~
| |
| " h:
| |
| y:
| |
| TABLE [3.4-3: . SUMidRY OF SIGNIFICANT SEQUENCES (Cor.t in ued) - .
| |
| CONTRIBUTION TO SOURCE TERM CLASSES PLANT PER- MIA MIB M2 . M3 M4 DAMAGE FREQ. CENT MS M6 M7 MB M9 CLASS' MEASURE (t) ACCIDENT SEQUENCE EVENTS M10 M11' M12
| |
| : 50. SLC 1.79E-07 .3 ET15 INCORE INSTRUMENT TUBE RUPTURE O O O 5.19E-11 3.58E-11 SSI. SUPPORT STATE 1 . 0 0 0 1.14E-10 2.11E-11 OAX3 PRIMARY DEPRESSURIZATION FAILURE O 8.93E-09 1.70E-07 GH8001 HIGH PRESSURE RECIRC. FAILURE
| |
| : 51. TEC' 1.74E-07 .3 ET18 LOSS OF 1 VITAL DC BUS 0 3.31E-11 0 1.72E-11 .3.48E-11
| |
| -SS2DC1 SUPPORT STATE 2 1.11E-10 0 1.74E-07 0 0 CF2002 AUXILIARY FEEDWATER FAILURE O 1.68E-10 6.11E-12 GE2004 CONTAINMENT SUMP RECIFC. FAILURE
| |
| $2. SLC 1.73E-07 .3 ET11 REACTOR TRIP O O O 5.02E-11 3.46E-11 SS2 SUPPORT STATE 2 0 0 0 1.10E-10 2.04E-11 GF2002 AUXILIARY FEEDWATER FAILURE O 8.63E-09 1. 64 E-07 GH6001 HIGH PRESSURE RECIRC. FAILURE
| |
| : 53. TEC 1.71E-07 .3 ET22SSI ATWS: SUPPORT STATE 1 0 3.25E-11 0 1.11E-11 3.42E-11 SSI SUPPORT STATE 1 0 0 0 1.09E-10 2.02E-11' GP2024 RT AND E.BORATION FAILURE O 8.53E-09 1. 62 E-07 ,
| |
| PL POWER LEVEL > 25% 1.n
| |
| : 54. ALC 1.61E-07 .3 ET2 MEDIUM LOCA 0 0 4.67E-11 0 3.22E-11 SS2 SUPPORT STATE 2 0 0 0 1.03E-10 1.90E-11 GH6001 HIGH PRESSURE RECIRC. FAILURE O 8.03E-09 1.53E-07
| |
| : 55. TEC 1.55E- 07 .3 ETZ5SSS CONSEQUENTIAL S/L BREAK INSIDE CONT.:SUP 0 2.95E-11 0 m.55E-11 3.10E-11 i
| |
| SS5 SUPPORT STATE 5 0 0 0 9.87E-11 1.83E-11 1.73E-09
| |
| ~
| |
| CM1001 MSIV CLOSURE FAILURE 0 1. 4 7 E-0 7 GP2006 PRIMARY DEFMESSURIZATION FAILURE
| |
| : 56. TEC 1.50E-07 .3 ET7 '1 DSS OF RCS FLOW 0 2 . 8 5E-1 *. 0 ~ 1.50E-11 3.00E-11 SS2 SUPPORT STATE 2 0 0 0 9.56E-11 1.77E-11 GF2002 AUXILIARY FEEDWATER FAILURE O 7.49E-09 1.42E-07 GP2007 FEED AND BLEED FAILURE 57 S LC" 1.43E-07 .3 ET3- SMALL'LOCA .
| |
| 0 0 0 . 4.15E-11 2.86E-11 SSI SUPPORT STATE l' 9.12 E-11 0 1.43E-07 0 0 l
| |
| l' OAX3 PRIMARY-DEPRESSURIZATION FAILORE O- 1.38E-10-' 5.02E-12 j RSS3 CCF BETWEEN RECIRC. AND CONT. SERAY
| |
| : 58. TE 1.41E-07 .3 ET8 LOSS OF MAIN FEEDWATER 0 2.68E-11 0 1.42E-11 2.82E-11
| |
| .SS4 SUPPORT STATE 4 6.15E-10 0 1.26E-07 0. 0 1.39E-08 0 2.79E-12
| |
| : 59. TEC 1.31E-07 .2 ET9 PRIMARY TO SECONDARY 10WER t?ISMATCH 0 2.49E-11 0 1.31E-11 2.62E-11 SS2 SUPPORT STATE 2- 0- 0 .0- 8.34E-11 1.55E-11 i
| |
| GF2002 AUXILIARY FEEDWATER FAILURE O -6.54E-09 1.2 4 E-0 7 i
| |
| CP2007 FEED AND BLEED FAILURE ,
| |
| o a
| |
| ' ' * ' '' ~
| |
| -e ._ t= m _ z.o , m rd
| |
| | |
| t TAdLE 3.4-3:
| |
| | |
| ==SUMMARY==
| |
| OF SIGNIFICANT SEQUENCES (Continued!- .
| |
| CONTRIBUTION TO SOURCE TERM CLASSES M2 P') M4 M1A MIB PLANT ~ PER- M6 M7 Ms M9
| |
| : MS
| |
| ' DAMACE faEQ. CENT M10 Mll . M12 CLASS MEASURE- '(t) ACCIDENT SEQUENCE EVENTS .
| |
| 0 2.45E-11 0 1.29E-11 2.58E-11
| |
| : 60. -TEC 1.29Z-07 .2 ET14 00SS OF OFFSITE POWER 0 0 0 9.22E-11 1.52E-11 2.
| |
| ~ SS7LDSP SUPPORT STATE 7 . O 6.44E-09 1.2"E-07 E60 FAILURE TO RECOVER OSP IN 1 HR E120 FAIL TO REC. OSP IN 2 HRS / FAIL. IN 1 HR.
| |
| UNC CORE UNCOVERY 0 2.36E-11 0 1.24E-11 2.48E-'1
| |
| : 61. TEC 1.24E-07 .2 ET14 00SS OF OFFSITE POWER 0 0 7.9CE-11 1.46E-il 0
| |
| SS6LOSP SUPPORY STATE 6. O 6 19E-09 1.18E-07.
| |
| GF2002 AUXILIARY FEEDWATER FAILURE CP2012- FEED AMD BLEED FAIIURE O 2.34E-11 0 1.24E-11 2.46E-11
| |
| : 62. TE 1.23E-07 .2 ET11 REACTOR TRIP 1.10E-07 0 0 SUPPORT STATE 4 5.36E-10 0 SS4 2.44E-12 1.21E-08 0 0 2.34E-11 0 1.23E-11 2.46E-11
| |
| : 63. TEC 1.23E-07 .2 ET24SS2 CONSEO. S/L BREAK OUTSIDE CONT. 0 0 0 7.84E-11 1.45E-11 SS2 SUPPORT STATE'2 6.14E-09 1.17E-07 O 2 CF5601 AUXILIARY FEEDWATER FAllDRE GP2006 PRIMARY DEPRESSURIZATION FAILURE e
| |
| 0 0 0 3.42E-11 2.36E-11
| |
| : 64. S LC 1.18E-07 .2 ET14 00SS OF OFFSITE POWER ' O O O 7.52E-11 1.39E-11 SS6LOSP SUPPORT STATE 6 O 5.89E-09 1.12E-07 CF2002 AUXILIARY IEEDWATER FAILURE GH7001 HICH PRESSURE RECIRC. FAILURE O 2.20E-11 0 1.16E-11 2.32E-11
| |
| : 65. TEC 1.16E-07 .2 ET4 STEAM CENERATOR TUBE RUPTURE 0 0 0 7.39E-11 1.37E-11 SSI SUPPORT STATE 1 1.10E-07 0 5.79E-09 CFN009 LOSS OF AFW & FAIL TO RECOVER FW GP2013 PRIMARY DEPRESSURIZATION FAILURE 0 0 3.33E-11 0 2.30E-11
| |
| : 66. SEC 1.152-07 .2 ET3 SMALL LOCA 0 0 0 7.33E-11 1.36E-11 SSI SUPPORT. STATE 1 1. 09E-07 O 5.74E-09 GHE116 HICH PRESSURE SAFETY INJECTION FAILURE CVCX01 PRIMARY DEPRESS. OR LPSI FAILURE 0 2.17E-11 0 1.14E-11 2.28E-11
| |
| : 67. TEC 1.14E-07 .2 ET25SSI CONSEQUENTI AL S/L'INSIDE CONT. O O O 7.26E-11 1.35E-11.
| |
| GM1001 MSIV CLOSURE FAILURE O S.69E-09 1.0SE-07 CP2006 PRIMARY DEPRESSURIZATION FAILURE
| |
| ~O O 3.28E-11 2.26E-11
| |
| : 68. SLC 1.13E-Oi ;2 ET7 00SS- OF RCS FLOW 0 0 0 7.20E-11 1.33E-11" SS2 SUPPORT STATE 2 5.64E-09 1. 0 7 E-07 GF2002 AUXILI ARY FEEDWATER FAILURE o-CH6001 HIGH PRESSURE RECIkC. F AI LURE 0 0 0 3.13E-11 2.16E 69. SLC 1.00E-07 .2.ET20 LOSS OF VAC BUS 1 OR 3 0 0 6.88E-11 1.27E-11 0
| |
| ~SS2VAC12 SUPPORT STATE 2 O 5.39E-09 ').02E-07 GF2002. AUXILIARY FEEDWATER FAILURE
| |
| 'Cn3001 MAIN FEEDWATER RECOVERY FAILURE FAILURE ,
| |
| GH6001 .HICH PRESSURE RECIRC.
| |
| x.-_
| |
| | |
| ,, .m . .- _. _
| |
| 7 cc . _
| |
| t '
| |
| ~
| |
| TABLE 3.4-3h .
| |
| | |
| ==SUMMARY==
| |
| OF SIGNIFICANT SEQUENCES (Continued! .
| |
| CONTRIBUTION TO SOURCE TERM CLASSES e4 M3 M4 ~
| |
| M1A~ M1B PER- M6 M7 M8 M9 P! NT ' MS LAMAGE FREQ. CENT M10 M?1 M12 CLASS MEASURE- It) ACCIDENI SEQUENCE EVENTS 0 0 0 3.13E-11 2.16E-11 1.00E-07 .2 ET21 LOSS OF VAC BUS -3 OR -4 0 0 6.88E-11 1.27E-11
| |
| : 70. SLC -0 ,
| |
| SS2VAC34' SUPPORT STATE 2 O 5.39E-09 1.02E CF2002 AUXILIARY FEEDWATER FAILURE
| |
| * CN3001- MAIN FEEDWATER RECOVERY. FAILURE
| |
| * CH6001 ' HICH PRESSURE RECIRC. FAILURE O 1.92E-11: 0 1.01E-11 2.02E-11 1.01E-07 .2 ET25SSS CONSEQUENTIAL S/L BREAK INSIDE CONT.:SUP0 0 0 6.43E-11 1.19E-11
| |
| .71. TEC SSS SUPPORT STATE 5 O 5.04E-09 9.57E-08 CF4001 AUXILI ARY FEEDWATER FAILURE CP2006 FAIMARY DEPRESSURIZATION FAILURE 0 1.89E-11 0 1.00E-11 1.99E-11
| |
| : 72. TE 9.93E-06' .2 ET14 LOSS OF OFFSITE POWER 4.33E-10 0- 8.90E-08 0 0' SS7LOSP SUPPORT STATE 7 9. 76 E-09 0 1.97E-12 GF3002 AUXILI ARY FEEDWATER FAILURE E60 FAILURE TO RECOVER OSP IN 1 HR E120 FAIL TO REC. OSP - IN 2 HRS / FAIL. IN 1 HR.
| |
| CS3001 QUENCH SFRAY FAILURE 0 0 0 2.87E-11 1.98E-11
| |
| .2 ET9 PRIMARY TO SECONDARY POWER MISMATCH 0 6.31E-11 1.17E-11 [$
| |
| : 73. SLC 9.9)E-08 0 0 'J SS2 SUPPORT STATE 2 O 4.05E-09 9.39E-08 CF2002 -AUXILIARY FEEDWATER FAILURE CH6001 HIGH PRESSURE RECIRC. FAILURE O O G 2. 85E'-11 1.97E-11 9.83E-08 .2 ET10 TURBINE TRIP 0 9.81E 08 0 0
| |
| : 74. SLC' 6.27E-11 SS2 SUF20RT STATE 2 O 9.48E-11 3.45E-12 CF2002 AdyILIARY FEEDWATER FAILURE RSS4 CCF BETWEEN RECIRC. AND CS 0 0 0 2.75E-11 1.89E-11 9.47E-08 .2 ET24SS2. CONSEQUENTIAL S/L BREAM OUTSIDE CONT. O O O 6.03E-11 1.12E-11
| |
| : 75. S LC CF5601 AUX 1LIARY.FEEDWATER FAILURE O 4.73E-09 8.98E-08 CH6001 HICH PRESSURE RECIRC. FAILURE 0' 1.78E-11 0 9.34E-12 1 87E-11
| |
| : 76. TEC 9.35E-08 .2 ET11 REACTOR TRIP 0 0 0 5.96E-11 -1.10E-11 S51 SUPPCRT STATE 1 .
| |
| 0 4.67E-09 8. 86 E-0 8 GFN001 LOSS OF AFW & FAIL TO RECOVER FW - '
| |
| CP2014 FEED AND BLEED FAILURE 0 O' 1.05E-10 0 1.83E-11 9.14E-08 .2 ET23SS8 CONSEQUENTIAL SMALL.LOCA: 5.64E-09 8.14E-08 0 0 77 SE 0 SS8 SUPPORT STATE 8 4.23E-09 0 1.2sE-12 0 1.67E-11 .0 8.79E-12 1.76E-11 TEC 8,00E-08 ,2 ET24SSI CONSEQUENTIAL S/L BREAK OUTSIDE CONT. 0 0 0 5.61E-11 1.04E-11 78.
| |
| 551' SUPPORT STATE 1 .
| |
| 4.39E-09 8.34E-08 O.
| |
| CF4001 AUXILIARY FEEDWATER FAILURE. .
| |
| GP2006 PRIMARY DEPRESSURIZATION FAILURE 0 . 0- 0 2.39E-11 1.65E STEAM LINE BREAF OUTSIDE CONTAINMENT 0
| |
| : 79. S LC
| |
| * 8.23E .1 ET6 5.25E-11 0 . 8.21E-08 0 SSI SUPPORT STATE 1 O 7.93E-11 2.89E-12 s GM2001 ' MSIV CLOSURE FAILURE wd mu
| |
| ~ - . - . ,
| |
| | |
| - . m. ;-. .
| |
| 3 . . . -
| |
| w.
| |
| ?", T el .
| |
| :- J L -
| |
| :n.; .
| |
| ~ 1 TABLE 3.4-3:
| |
| | |
| ==SUMMARY==
| |
| OF SIGNIFICANT SEQUENCES (Contirtue@- ,.
| |
| . CONTRIBUTION TO SOURCE TERM CIASSES PER- M1A M1B M2 M3 M4
| |
| - P LANT DAMAGE FREQ. CENT - N5 M6 M7 MS M9-CLASS MEASURE (41 ACCIDENT. SEQUENCE EVENTS M10 Mll M12
| |
| : 80. TE 8.05E .1 ET7 LOSS OF RCS FLOW 0 1.53E-11 0 8.13E-12 1.61E-11 -
| |
| SS4 SUPPORT STATE 4 3.51E-10 0 7.22 E-08 0 0 7.91E-09 0 1.59E-12
| |
| : 81. SLC 7.93E-08 .1 ET14 LOSS OF OFFSITE POWER 'O O O 2.30E-11 1.59E SSSLOSP SUPPORT STATE 5 0 0 0 5.0$E 9.36E-12 CF1001 < module > 0 3.96E-09 7. 52 E-0 8 GH5001 HIGH PRESSURE RECIRC. FAILURE
| |
| : 82. TEC 7.43E-08 .1 ET25SSI CONSEQUENTI AL S/L INSIDE cot?T. O. 1.41E-11 0 7.42E-12 1.49E-11 SSI SUPPORT STATE 1 -0 0 0 4.73E-11 C.77E-12 CF4001 AUXILIARY FEEDWATER FAILURE O 3.71E-09 7.04E-08
| |
| -CP2006 PRIMARY DEPRESSURIZATION FAILURE
| |
| : 83. TE 7.05E-08 .1 ET9 PRIMARY TO SECONDARY POWER MISMATCH 0 1.34E-11 0 -7.3 2E-12 1.41E-Il SS4 SUPPORT STATE 4 3.07E-10 0 6.32E-08 0 0 6.93E-09 0 1.40E-12 ..
| |
| : 84. TEC' 6.88E-08 .1 ET17 LOSS OF 1 SERVICE WATER TRAIN O 1.31E-11 0 6.81E-12 1.30E-11 ,,
| |
| SS25W1 SUPPORT STATE 2 4.39E-11 0 6.67E-08 U O -.
| |
| CD CF2002 AUXILIARY FEEDWATER FAILURE O 6.63E-11 2.41E-12 GP2007 FFED AND BLEED FAILURE GE2004 CONTAINMENT SUMP RECIRC. FAILURE
| |
| : 85. TEC 6.83E-08 .1 ET22SS2 ATWS: SUPPORT STATE 2 0 1.30E-11 0 6.82E-12 1.37E-11 SS2 SUPPORT STATE 2 0 0 0 4.35E-11 0.06E-12 RT3 MANUAL REACTOR TRIP O 3.41E-09 6.47E-08 GP2007 FFED AND BLEED FAILURE
| |
| : 86. TEC 6.82E-08 .1 ET25SS6 CONSEQUENTIAL S/L BREAK INSIDE CONT. O 1.30E-11 0 6.81E-12 1.36E-11 SS6 SUPPORT' STATE 6 0 0 'O 4.34E-11 8.05E-12 .
| |
| CF5601 AUXILIARY FEEDWATER FAILURE O 3.40E-09 6.47E-08 CP2006 PRIMARY DEPRESSURIZATION FAILURE
| |
| : 87. ALC 6.79E-08 .1 ETI LARGE LOCA 0 0 1.97E-11 0 1.36E-11 SSI . SUPPORT STATE 1 0 0 0 4.33E-11 8.01E-12 CVD002- LOW PRESSURE SAFEY INJECTION FAILURE O 3.39E-09 6.44E-08 LPRI LOW PRESSURE RECIRC. FAI LURE -
| |
| : 88. S LC 6.63E-08 .1 ET25SS6 CONSEQUENTIAL S/L BREAK INSIDE CONT. -0 0- 0 1.92E-11 1.33E-11
| |
| . 556 SUPPORT STATE 6 0 0 'O 4.22E-11 7.82E-12 GF5601 -AUXILIARY FEEDWATER FAILURE O 3.31E-09. 6.29E-08 CH7001 HIGH PRESSURE RECIRC. - FAI LURE
| |
| : 89. SLC 6.14E-08 .1 ET14 ' LOSS OF OFFSITE POWER "O O 'O 1.78E-11 1.23E-11 SS7LOSP SUPPORT STATE 7 0 J 0 3.91E-11' 7.2SE E60 FAILURE TO RECOVER OSP IN 1 HR " ~
| |
| O 3.06E-09 5.82E-08 E120- FAIL TO REC. OSP IN 2 HRS / FAIL..IN 1 Hk.
| |
| CH5001 HIGH PRESSURE RECIRC. FAILURE *
| |
| -3_-- m u.o' W --
| |
| ---g_-4 ,-q- .
| |
| ',y .e w, ,.p ,, y
| |
| : m. .
| |
| t .
| |
| TABLE 3.4-3:
| |
| | |
| ==SUMMARY==
| |
| OF SIGNIFICANT SEQUENCES' (Continued)
| |
| COhTRIBUTION TO SOURCE TEPM CLASSES
| |
| , v' M3- M4 M1A M1. 8 M2 PER- M6 M7 M8 M9
| |
| . PLANT MS DAMAGE FREQ. ' CENT M10 +
| |
| Mil M12 C LASS MEASURE (%) ACCIDENT SEQUENCE EVENTS 0 1.16E-11 0 6.11C-12 1.22E-1A
| |
| : 90. TEC 6.12E-08 .1 ET7 LOSS OF RCS FLOW O O -
| |
| 0 3.90E-11 7.22E-12 SSI SUPPORT ~ STATE
| |
| * 0 3.05E-09 5.80E-08 CFN001 : LOSS OF Afd & FAIL TO RECOVER FW .
| |
| GP2014 FEED AND DLEED FAILURE-0 0 1 7C;-11 0 1.17E-11
| |
| : 91. SEC -3.87E-08 .1 ET3 SMALL LOCA L 4 0 0 0 3.74E-11 6.93E-12 SSI SUPPORT STATE 1 O 2.93E-09 5.5(:-08 RT4 MANUAL OR AUTO. REACTOR TRIP O O 6.74. 11 0 1.17E-11 5.86E-08 .1 ET4 STEAM GENERATOR TUBE RUPTURE 3.62E-09 5.22E .8 0 0 _
| |
| : 92. SE SUPPORT STATE ; . 0 SSI 2.71E-09 0 e.20E-13 SA1 SAFETY INJECTION SIGNAL FAILURE 5.65E-09 0 0 0 O
| |
| : 93. V2EC 5.65E-08 .1 Ei4 STEAM GENERATOR TUBE RUPTURE -0 0 0 0 0' SSI SUPPOR1 STATE 1 O O O O O CHE116 HIGH PRESSURE SAFETY INJECTION FAILURE GSG007 STEAM LEAK ASSOCIATED WITH FAULTED STEAM 0 1.07E-11 0 5.71E-12 1.13E-11 5.65E-08 .1 ET6 STEAM LINE BREAK OUTSIDE CONTAINMENT 5.07E-08 0 0 94 TE SUPPORT STATE I 2.46E-10 0 SSI 5.55E-09 0 1.12E-12 SA1 SAFETY INJLCTION SIGNAL FAILURE 0 0 1.60E-11 0 1.10E-11
| |
| : 95. SEC 5.50E-08 .1 ET14 LOSS OF OFFSITE POWER 0 0 0 3.50E-11 .6.49E-12 SS7LOSP SUPPORT STATE 7 0 2.74E-09 5.21E-08 E60 FAILURE TO RECOVER OSP IN 1 HR E120 FAIL TO REC. OSP IN 2 HRS / FAIL. IN 1 WR.
| |
| e?2020 FEED FAILURE O 1.02E-11 0 5.35E-12 1.07E-11
| |
| : 96. TEC 5.36E-08 .1 ET9 PRIMARY TO SECONDARY POWER MISMATCH O O O 3.41E-11 6.32E-12 SSI SUPPORT STATE-l' 0 2.67E-09 5. 0 0 E-08 GFN001 UOSS OF AFW & FAIL TO RECOVER FW. ,
| |
| GP2014 FEED AND BLEED FAILURE 0 0 0 1.54E-11 1.04E-11
| |
| : 07. SLC* 5.22E-08 .1 ET17 LOSS OF 1 SERVICE WATER TRAIN. 3.33E-11 0 5.21 E-08 0 0 SSISW1 SUPPORT STATE 2 _
| |
| O 5.03E 111 1.83E-12 GF2002 AUXILI ARY FEEDWATER FAILURE GH6001 HIGH PRESSURE' RECIRC. FAILURE GE4004 CONTAINMENT SUMP RECIRC. FAILURE O O 5.87E-11 0 1.02E-11 5.10E-08 .1 ET23SS7 CONSEQUENTI AL SMALL LOCA: SUPPORT. STATE 3.15E-09 4.54E-08 0 0
| |
| : 98. SE .0 SS7 SUPPORT STATE 7 7.14E-13
| |
| :2.36E-09 0 GF3002 AUXILIARY FEEDWATER FAILURE 0- 9.35E-12 0 4.97E-12 9.84E-12
| |
| : 99. TE 4.92E-08 .1 ET18 DOSS OF 1 VITAL DC BUS - 2.15E O. 4.41E-08 0 0 SS2DCl SUPPORT STATE 2 4.84E-09 0 9.74E-13 GF2002~ AUXILIARY FEEDWATER FAILURE.
| |
| CS2002.- QUENCl1 SPRAY FAILURE 'I a
| |
| | |
| ~ - ,-g ...=-7-. .__ . . . , . . _ . m_7 , _ , .
| |
| < ~ . _
| |
| 3; ,
| |
| ~
| |
| ~
| |
| .,n...-.~
| |
| - - r.
| |
| y - .-
| |
| f
| |
| . .g
| |
| _.w., . . ~ .
| |
| .z: .
| |
| -TABLE 3.4-3: SUMMARr OF SIGNIFICANT SEQUENCES tContinued) ,
| |
| CONTRIBUTION TO SOURCE TERM CLASSES, _
| |
| HIB M3 ~ M4
| |
| - M1A M2 PLANT PER- ~~
| |
| M9" CENT- MS ' M6 M7 _ M9 :
| |
| ~ DAMACE- FREQ.
| |
| CLASS HEASURE (%)- ACCIDENT SEQU7.NCE' EVENTS M10 M1l' . M12
| |
| .1 ET[ STEAM GENERATOR TUBE RUPTURE . 0 4.86E-OB 0: 0 0 -
| |
| 100. V2EC - 4. 8 6E-08 -- SSI SUPPORT STATE l~ 0 0 0' O CHEll6 ' HICH PRESSURE SAFLTY INJECTION FAILURE O O 0~ 0 0
| |
| 'OAX4 SECONDARY DEPRESSURIZATION FAILURE' OAX5 FEED AND BLEED FAILURE' ........
| |
| .......... .. .. ... ~ ... .a. ............. ....
| |
| TOTA!: 5.30E-05 ' 95.9% of CM Total Frequency 5.52E-05 MIA - Containment bypass directly to the environment. - (V sequence) .
| |
| MlB - Contalement bypass directly to'the environment through steam generator tube rupture.
| |
| M2 - Containment failure due to overpressure shortly after vessel failure. Early core-selt sequences.
| |
| M3 - Containment failure due to overpressura shortly after. vessel failure.- Late core-melt sequences.
| |
| Ms containment isolation failure.
| |
| M5 - Cont ainment failure due to overpressure approximately 4 hours af ter vessel f ailure. No containment spray operation.
| |
| Lati core-melt sequences. '. ..
| |
| M6 - Containment f ailura due to overpressere .approximately 4 hours after vessel failure. No containment spray operation.
| |
| Early core-eelt sequences. . . .
| |
| M7 - Containment f ailure due to ove. pressure approximately 20_ hours af ter ' vessel failure. No containment spray operation.
| |
| HB - Cont inment f ailure due to overpres;ure approximately 4 hours af ter vessel f ailure. Cont. spray operation is successful. -
| |
| M9 - Containment f ailure dwe to overpressure approximately 20 hours af ter venal failure. Cont. spray operation is successful. $.
| |
| M10 - h e-Mat melt-through. No spray operation.
| |
| -Hil - Base-Pat me k i. -t h rough. Spray operation.is successful.
| |
| M1' No cr.ntainment fallure. Design leakage (0.9% per dasy) occurs.
| |
| s e
| |
| t e*
| |
| : w. + *-7 . -'-m.',-we..=--r. , - - e- # an * -..-'r++.s
| |
| * w--y,'bes *- e- % , .- y y- gwe 4- ,y-g y -ese, w ---v-. ,
| |
| | |
| f' 121 TABLE 3.4-4 C '
| |
| | |
| ==SUMMARY==
| |
| OF-SIGNIFICANT CONTAINMENT BYPASS SEQUENCES o
| |
| PLANT ~ PER-DAMAGE- FREQ. CENT CLASS MEASURE. (%) ACCIDENT SEQUENCE EVENTS
| |
| -V 1.93E-07 .3 INTERFACING SYSTEMS LOCA ISLOCA THROUGH RHR SUCTION PATH V2EC 5.65E-08 .1 STEAM GENERATOR TUBE RUPTURE
| |
| " SUPPORT STATE 1 HIGH PRESSURE SAFETY INJECTION FAILURE -
| |
| STEAM LEAK ASSOCIATED WITH FAULTED STEAA V2EC 4.86E-08 .1 STEAM GENERATOR TUBE RUPTURE SUPPORT STATE 1 HIGH PRESSURE SAFETY INJECTION FAILURE-SECONDARY DEPRESSURIZATION FAILURE TEED AND BLEED FAILURE V2EC 2.92E-08 .1 STEAM GENERATOR TUBE RUPTURE SUPPORT STATE 2 AUXILIARY FEEDWATER FAILURE PRIMARY DEPRESSURIZATION FAILURE
| |
| *- STEAM LEAK V .2.81E-08 .1 INTERFACING SYSTEMS LOCA ISLOCA THROUGH LPSI
| |
| <V2EC 1.29E-08 .0 STEAM GENERATOR TUBE RUPTURE SUPPORT STATE 1 AFW FAILURE & FAIL TO RECdVER'AFW OR MFW PRIMARY DEPRESSURIZATION FAILURE STEAM-LEAK-TOTAL: 3.72E-07 .67% of CM Total Frequency 5.52E-05
| |
| , ls I
| |
| f
| |
| | |
| 122 TABLE 3.4-5 RELEASE SPECTRUM DUE TO CORE DAMAGE Percent 3 Contribution to Release Containment-Category Description Frequency Failure M1A' Bypass, Lirect , ,,
| |
| 2.21E-07 1.97-M1B Bypass through S/G -1.73E-07 1.54-M2- Early CM, Early
| |
| , Containment Failure 7.09E-09 0.06 M3 Late CM, Early Containment Failure 4.90E-09 0.04 Md Isolation Failure 1.10E-08 0.10 M5 Late CM, No .Contairiment Sptay,-Containment
| |
| , fails 4 hours after Vessel 1.74E-08' O.16 M6 Early CM, No Contain-ment Spray, Containment -
| |
| fails 4' hours after Vessel 8.38E-08 0. ~ 5
| |
| 'M7 No Containment Spray, Containment fails 20' hours after' Vessel 6.12E-06 ,
| |
| 54.68 M8 Spray available,_ Con-tainment fails 4 hours 0.28 \
| |
| after Vessel 3.08E-08 ;
| |
| M9 Spray available,, Con-tainment fails 20 hours after Vessel 1.703-06 15.19 M10 No' Containment Spray, Base-Mat Melt-through 4.29E-07 3.83 ,
| |
| M11 Spray available, Base-Mat Melt-through 2.39E-06 21.36 (Containment Failure)
| |
| TOTAL 1.12E-05 100 M12 Design Leakage'(No Con-l tainment Failure) 4.40E-05 04Df7X,080
| |
| ..r.
| |
| | |
| - .- . . - . . - - . _ . - . . - - . . .. - _ _ - - ~ _ . . . . . - .- . - _ . .
| |
| 123
| |
| !b a, ,
| |
| % I
| |
| . ,I TABLE 3.4-6 -
| |
| DOMINANT CONTAINMENT FAILURE SEQU3CES j
| |
| , 1 I
| |
| Percent contribu- .
| |
| Sequence # tion to 1
| |
| ~
| |
| (see-. Plant contain -
| |
| Table Support Damage Release. ment 3.4-4) State State Class Frequency Failure 7 7 TE M7 1.60E-06 14.30 15 1 ALC' M9 9.61E-07 8.59 13- 7 SE M7 9.36E-07 8.36 16' 4 TE M7 6.45E-07 5.76 19- 1 ALC' M9- 6.03E-07 5.39 1 1 ALC Mil 3.47E-07 3.10 22 4 TE M7 3.31E-07 2.96 2 1 ALC M11 2.88E 2.57 3 1' TEC Mil '2.29E-07 '2.05 46- 2* SLC' M7 1.95E-07 1.74
| |
| ;: 48 '1 V M1A 1.93E-07' 1.73 l 7 TE M10 1.76E-07 -1.57 l 7
| |
| 51 2 TEC' M7 1.74E-07 1.55 4 2 TEC M11 1.72E-07 1.54 57 >1 SLC' M7 1.43E-07 1.28
| |
| '58 4- TE M7 -1.26E-07' 1.13 62 4 TE 'M7 1.10E-07 0.98 5 1 TEC -M11 1.05E-07 0.93 6 1 AEC M11 1.02E-07 0.91
| |
| . 66.4 Design Leakage (M12).is-excluded from the containment failure. Therefore, total containment failure probability is 1.12E-05.
| |
| i 04 W .001 1
| |
| | |
| i ,
| |
| v i
| |
| 124 1 Table 3.4-7 *
| |
| "i '
| |
| Decay Heat' Removal. Vulnerability Insights and Applicability to MP3
| |
| -1 Insight from_ Ano11embility to MPs .; ;
| |
| 4 Den studies
| |
| : 1. Less redundancy-at MP3 is_a relatively new plant support system level.- with principles'of redundancy built-into the design.
| |
| Therefore, adequate redun-dancy exists at both-the1 sup-port state and front line systems. ,
| |
| : 2. Human Errors of special MP3 study modeled cognitive o significance. errors and the errors of com-mission for the recirculation ],
| |
| function and for feed and i bleed. See Section 3.3.3 for l further details. ,
| |
| : 3. Loss of_offsite power A third diesel generator-to contributes signifi- be installed at MP3 will re-cantly to risk. duce this risk.
| |
| : 4. The " Feed and Bleed" At.MP3, Feed and Bleed proca- ;
| |
| 4
| |
| -operation could have a dure is assigned high significant effect upon priority in operator !
| |
| .the DHR-related core training. ,
| |
| j n damage risk.
| |
| 1 i
| |
| l gj i
| |
| i 1
| |
| s
| |
| * i
| |
| * 'l 4
| |
| (s !
| |
| s i
| |
| 4 1 )
| |
| 04LW7X,08D '
| |
| g, . 'J!
| |
| | |
| m l '' .
| |
| 125 _
| |
| TABLE 3.5-1 SEISMIC INITIATING EVENT DEFINITIONS AND MEAN FREQUENCIES Frequency /
| |
| Event Sy @ l Event Description Reactor Year 0.15G Level Earthquake
| |
| * 3.57E-04 Q15 0.25G Level Earthquake 5.77E-05 $
| |
| Q25 Q35 0.35G Level Earthquake 1.63E-05 0.45G Level Earthquake 5.89E-06 Q45 0.55G Level Earthquake 2.63E-06 Q55 0.65G Level Earthquake 1.18E-06 Q65 m 0.75G Level Earthquake 6.39E-07 Q75 0.80G Level Earthquake ** 8.12F-07 Q80 orspresents earthquakes with a peak ground acceleration between'O.1G and-0.2G.
| |
| corepresents earthquakes with a peak ground acceleration exceeding 0.8G.
| |
| -m a
| |
| 04LW7X.060
| |
| | |
| 126 .
| |
| TABLE 3.5-2 SEISMIC CORE-MELT STATE DEFINITIONS AND CALCULATED MEAN FREQUENCIES Frequency / Percent symbel Description Reactor Year Contribution AEC Large LOCA, Early Melt' 1.96E-09 < 0.1 AEC' Large LOCA, Early Melt,
| |
| . Failure of Recirculation Spray 3.40E-10 < 0.1 ,
| |
| AE Large LOCA, Early Melt, No Containment Cooling 6.54E-07 7.21 ALC Large LOCA, Late Melt 9.82E-09 0.11 ALC' Large LOCA, Late Melt, Failure of Recirculation E Spray 1.39E-07 1.53 AL Large LOCA, Late Melt, No Containment Cooling 6.76E-09 < 0.1 SEC Small LOCA~, Early Melt 2.15E-07 2.37 SEC' Small' LOCA, Early Melt, i Failure of Recirculation Spray 5.65E-08 0.62 L
| |
| SE Srall LOCA, Early Melt, No Containment Cooling- 1.90E-06 20.91 SLC Small LOCA, Late Melt 2.87E-08 -0.32 L SLC' Small LOCA, Late Melt, Failure of Recirculation Spray. 2'.24E-07 2.48 SL- Small LOCA, Late Melt, No Containment Cooling 1.91E-08 0.21 TEC Transient, Early Melt 3.93E-09 < 0.1 TEC' Transient, Early Melt, -
| |
| Failure of Recirculation-Spray- 1.82E-09 < 0.1 TE ' Transient, Early Melt, No Containment Cooling 5.72E-06 62.94 V3 LOCA, Containment Bypass 1.03E-07 1.14 i TOTAL 9.08E-06 100.0 04LW7X.080
| |
| | |
| l l 127 TABLE 3.5-3 I SEISMIC RELEASE CATECORY DEFINITIONS AND CALCULATED MEAN FREQUENCIES l
| |
| Release Frequency / Percent Category Description Reactor Year Contribution M1A Containment Bypass, V-Sequence 0.00 < 0.1 M1B Containment Bypass, SGTR 1.09E-09 < 0.1 M2 Early Failure /Early Melt, No Sprays 2.48E-09 < 0.1 M3 Early Failure / Late Melt, No Sprays 6.56E-10 < 0.1 l M4 Containment Isolation Failure 1.05E-07 1.16 ,
| |
| 1 M5 Intermediate Failure / Late l Melt, No Sprays 2.53E-08 0.28 L M6 Intarmediate Failure /Early Melt, No Sorays '5.23E-07 5.77 l l l M7 Late Failure, No Sprays 7.25E-06 79.85 l M8 Intermediate Failure with Sprays 2.98E-10 < 0.1 M9 Late Failure with Sprays 1.96E-07 2.16 M10 Basemat Failure, No Sprays 7,16E-07 7.88 M11 Basemat Failure with Sprays 1.34E-08 0.15 - 1
| |
| .M12 No Containment Failure 2.46E-07 2.71 TOTAL 9.08E-06 100.0 1
| |
| I 4
| |
| l l'
| |
| i l
| |
| 04Dmt,080 l
| |
| l
| |
| | |
| r
| |
| -128 j TABLE 3.5-4 FIRE INITIATING-EVENT-FREQUENCIES Frequency of Fraction of- Frequency of' '
| |
| Fire-(per Fires Causing Fire Causing Fire' Area reactor year) safety Loss safety Loss C3ntrol Room (CB-9) 3.5 x 10-3 0.128 4.48 x 10~4 Instrument Rack Room ,
| |
| l (CB-11A and CB-11B) 3.5 x 10~3 0.0429 1.50 x 10~4 Cnble Spreading Room i (CB-8) 6.60 x 10~3 0.0925 6.11 x 10~4 l l
| |
| J Switchgear Rooms (CB-1 and CB-2) 1.04 x 10-2 0.132 1.38 x 10~3 1 Electrical Tunnels (SB-2 and SB-3) 6.60 x 10-3 0.169 1.11 x 10-3 I a
| |
| i MCC Rod Control (AB-5 and AB-6) '7.00 x 10-3 v.132 9.06 x 10~4 ,
| |
| Air Conditioning !
| |
| (AB-7 and AB-8) 7.00 x 10~3 0.0251 1.73 x 10-4 !
| |
| i
| |
| ' AB-1 (Charging pumps and ccp pumps) 4.80 x 10-3 0.'0314 1.51 x 10-4 S0rvice Water (CSW-3 and CSW-4) 3.00 x 10-3 0.551 1.62 x-10~3 l.
| |
| l- '
| |
| l - Diesel Generators (EG-3 and EG-4) 3. 4 0 x 10-2 1.00 3.40 x 10-2 <
| |
| 1 i
| |
| . 04LW7X,060
| |
| | |
| ..g ~ , .
| |
| :,g y sL> +... L..
| |
| ; .9 . c;q
| |
| ~
| |
| fX''' '
| |
| . TABLE 3.5-5 ,
| |
| MEAN FRE()UENCY ~
| |
| FIRE AREAS VERSUS PLANT LAMAGE STATES -
| |
| t . .
| |
| -TEC' TE SE C' SEC SLC' 'SLC' Total Fire Zorte TE~
| |
| -# ~
| |
| Control Room 2.53 x 'I'f 7 9.56 x 10~I 4.54 x 10-I' 1.13 x 10-8 7.28 x 10
| |
| ' (15.21)
| |
| (CB-9)
| |
| -I
| |
| -9 1.52 x 10'I 3.80 x 10-9 2.44'x 10 Instrument Rack Room 8.50 x 10-8 3.21 x 10 CBil-A and 118 - (5.11)
| |
| -8 9.89 x 10 -I Cable Sp.eading Room 3.44 x 10~ 't.30 x 10-8 6.16 x 10'I 1.54 x 10 (CB-8)
| |
| _(20.61)'
| |
| l +
| |
| Switchgear Rooms 7.30 x 10~I 2.75 x 10-8 1.26 x 10-8 3.27 x 10-8 8.03 x 10~
| |
| CB-I and CB-2 (16.75) y 6.31 x 10~I - 2.38 x 10-0 1.02 x 10-8 2.82 x 10-8 6.93 x 10'I l Electrical Tunnels 58-1 and 58-2 (14.41) l MCC and Rod Control Areas 7.39 x-10-8 1.49 x 10-10 2.37 x 10-II 1.01 x 10-8 8.42 m'10 (1.85)
| |
| SB-1 and SB-2 Service Water Building '4.27 x 10-8 4.27 x 10-0 C5W'-3 and C5W-4 (0.91)
| |
| Diesel Generator Enclosures 1.45 x 10-7 ' 1.45 x 10 -I' EG-3 and EG-4 (3.05)'
| |
| Charging and CCP Zone 1.63 x 10-9 ' -1.11 x 10'I ~7.64 x 10~I 'l.94 x 10'I' -1.07 x 10-0 AB-1~ (22.3%)
| |
| -0 4.00 x 10 4-Total 2.12 x 10-6 7.72 x 10-8 1.39 x.10 1.45 x 10-I 1.11 x 10'I 7.64 x 10~I' .1.94 x 10-I (44.21) (1.6%) (29.01) (3.01') (2.311 L(5.91) -(4.01) l
| |
| ; - - .~ .. - - - :.- .- . - - . - . _. -,=a,. .- _
| |
| | |
| 130
| |
| ' TABLE 3.5-6 FIRE RELEASE CATEGORY DEFINITIONS AND CALCULATED MEAN FREQUENCIES Frequency /
| |
| Release _
| |
| Percent Category Description Reactor Year contribution M1A Containment Bypass, V-Sequence ,
| |
| M1B Containment Bypass, SGTR - -
| |
| M2 Early Failure /Early Melt, No .'
| |
| Sprays 6.70E-11 < 0.1 M3 Early Failure / Late Melt, No Sprays 6.52E-10 < 0.1 M4 Containment Isolation Failure 9.70E-10 < 0.1 M5 Intermediate Failure / Late 0.12 Melt, No Sprays 6.23E-09 M6 Intermediate Failure /Early Melt, No Sprays M7 Late Failure, No Sprays 1.52E-06 31.34 M8 Intermediate Failure with Sprays 2.04E-09 < 0.1 M9 Late Failure with Sprayo 1.45E-07 2.99 M10 Basemat Failure, No Sprays 1.37E-07 2.82 M11 Basemat Failure with Sprays 2.00E-07 4.12 M12 No Containment Failure 2.84E-06 58.56 TOTAL 4.85E-06 100.0 e
| |
| e 04LW7X,080
| |
| | |
| p"c '
| |
| 131
| |
| - Table 4.'3-1 Containment Response Class Definitions:
| |
| W
| |
| : 1. . -
| |
| l' Containment Class' Plant Damage States l' AE 2 SE 3 AL 4 TE !
| |
| l=
| |
| 5 SL
| |
| != 6- ,AEC, ALC, SEC, SLC, TEC, S'EC 7 TEC', SLC' ;
| |
| 8 AEC', ALC', SEC' ;
| |
| 9 ALC", SLC" f s
| |
| 10 S'E, S'L ,
| |
| l No Assignment V, V2EC, V2EC', V2E, V2LC,.V2LC', '
| |
| V2LC", V2L, V3
| |
| -t l
| |
| e l'
| |
| {
| |
| i.
| |
| l .
| |
| 2 1
| |
| 2 04LW7X.080
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| -__m_-._m_m--_ w 4
| |
| * j
| |
| ,, h 0
| |
| 0 0 0 0 0 -
| |
| - - -. 06 - - - -- -- -- -- -- -- 08 --
| |
| - 0 - - - 4 d - - 1
| |
| - - - 6 - - - 3 -
| |
| o - - 2
| |
| - - 2 - - - 1 - - - 2 ' - - - - - - - - 3 C
| |
| y
| |
| . t i
| |
| l i
| |
| b a
| |
| m 0 0 0 m 0 0 0 0 0 0 0 4 0 0 9 a t 0 07 7 0 00 0 50 00 08 9 - 08 000 00 0 -- 75 7 6 1
| |
| r - 4 4 7 3 6 l
| |
| F a 7 7 7 7 1 - 4 4 3 t 9 31 31 3 3 02 5 7 7 5 0 -1 - 5 6 7 6 - 1 1 5 - 2 2 2 s 4 0 0 y t 5 0 0 t u 0 7 0 3 4 4 i o 2
| |
| - - - 4 - - - 0 - - - - 6 - - - 6 - -
| |
| v y a r
| |
| - - 4
| |
| - - 4 - - - 5 - - - 8 -- -- -- - -B >- -- - > -
| |
| C D - - 6 - - - 5 - - - 5 f
| |
| f h
| |
| o 0 0 0 07 c y 0 0 0 0 0 0 2 2 7 n a 2 2 0 0 0 0 - - 8 8 -
| |
| - - 8 18 1 - - 3 3 - - - 7 7 e
| |
| u pr
| |
| - 4 4
| |
| - 71 7 - - - - 8 8 1 - - - 7 71 - - 2 2
| |
| - - 3 3
| |
| - 1 - - 1 1 - - 1 - - - 1 Q S s h 0 0 1
| |
| 0 0 0 0 03 i c 0 0 7 9 6 9 2 1 r n - 6 5 6 0
| |
| - - 2 1 - - 3 - - - 7 6 - - 74 6 4- -
| |
| b e - - 1 - -- --2 20 0 -- -
| |
| e n D O
| |
| - 0 0
| |
| - 2 2
| |
| - - 6 1 1 - - 2 ' - 3 3 -
| |
| Y G
| |
| O L
| |
| O -
| |
| N r y O e t ~
| |
| R v i H s o va 0 0 0 0 0 0 0 03 1
| |
| C l 0 0 5 7 4 6 6 8 1 d l C 6 0 - - - 4 3 - - 6 5 -
| |
| - T n - 1 0 - - 5 4 - - 0 9 - - - 7 71 - - 2 2 -
| |
| 6 l l o i p o - 7 7 - ~- 1 1 - - 817 1
| |
| - + - - - 1 - - 3 3 -
| |
| E c S t - 1 1 - - 1 1 D e 4 I s C n e C A i 0 0 0 l
| |
| e y 0 0 0 0 0 b E a 0 - -0 80 82 - - 0 0 9 - 5 5 0 m r - 0 - - - 5 5 - 5 5 5 a I R i p n - 0 0 9 - 0 0 9
| |
| - 8 6 1
| |
| 1 - - 9 9 1
| |
| 1 - 1 1 1 T M I
| |
| T S o - 5 5 2 - 5 55 T
| |
| S E
| |
| T S 0 0 0 04 0 8 08 0 8 80 8 0 y t E t u 0 0 B i o v y 0
| |
| 9 0 0 0 1 1 9 286 0 6 20 7 84 4 0 0 0 0 87 664 44 7 7 6 6 6 61 a r 6 7 1 7 61 1 9 9 11 2 5 5 2 3 0 6 6 13 e '
| |
| 1 1 2 2 2 1
| |
| C O 1 1 1 7 7 4 4 4 4 5 9 1 e
| |
| l r 0 0 0 0 0 0 70 0 0 07 e
| |
| s u
| |
| l 0 0 6 0 8 0 09 0 0 0 0 0 9 8 6 e05 05 06 76 64 73 73 6 70 0 0 70 7 s i e a 8 6 3 3 3 3 0 5 5 0 7 7 7 7 3 1 6 6 3 05 02 02 2 1 0 2 V F 1 1 1 1 1 7 7 1 2 2 2 2 5 6 1 0 02 00E 07 0 0 00 000 00 e tl 0 0 0 0 0 0 909 0 1 7
| |
| 3 0 4 4 3 8 8 8 e r 0 0 0 0 27 6 2 25 75 1 6 3 2 2 3 2 0 5 5 25 8 8 8 8 o e 2 1 1 8 8 8 8 9 6 6 9 2 2 2 2 5 6 1 1 1 1 1 1 CM y
| |
| r e 0 0 0 0 0 0 0 04 04 v 4
| |
| _ e r c o
| |
| o n 0 0 0 0 20 1 1 1 1 3 62 6 0 02 06 05 5 2 3 4 4 4 4 6 0 0 0 81 5 24 40 0 R 46 9 7 3 3 9 7 1
| |
| 1 6 6 6 7 7 7 1 1 1 4 5 1 1 4 1 C U 2 2 2 2 8 5 5 8 1 1 1 0 01 0 505 0B 8
| |
| S 0 0 0 0 7 8 8 7 4 5 7 7 4 C f 5 8 1 1 5 C f 6 8 8 6 4 O 0 O 0 E O O 0 O 0 6 3 3 6 O 0 0 0 4 4 1 1 e
| |
| g e t " '
| |
| n a t *
| |
| "t E E CE E'C1 "C 'C E C C "C a ma at "C 'C L C l
| |
| C C E E E E t t A A A A A A A A S t
| |
| t t S S S 5 S S S S T
| |
| * L t t E E E E T T T _
| |
| P D S
| |
| | |
| 7-t-
| |
| L 1
| |
| 133 ,
| |
| TABLE 4.6-2 '
| |
| | |
| ==SUMMARY==
| |
| OF CONTAINMENT RESPONSE FC R CORE MELT ACCIDENTS .,
| |
| (NON-SEISMIC EVENTS)
| |
| CONTAINMENT REEDONRE fA) '
| |
| BYPASS INTERMEDIATE .
| |
| OR EARLY FAILURE LATE BASEMAT 4
| |
| . PLANT . ISOLATION FAILURE BY (4-7 H 2BURN FAILURE PENETRA- NO L-DAMAGE STATES FAILURE f<a HR) HR) (di DAY 1 TION FAILURE AE -(B) - 62% 28% 9% -
| |
| SE - 0.1 6 89 5 -
| |
| AL - - 54 35 11 -
| |
| TE - - 0.4 90 10 -
| |
| SL - - 1 79 20 -
| |
| l ALL C CASES - - - - 5 95 l
| |
| l TEC'/SLC'' - - -
| |
| 100(C)- - -
| |
| 1 1
| |
| [ AEC'/ALC'/SEC' - - - 100(C,D) - --
| |
| ALL C" CASES - - - - 99 1-S'L - - 0.1 ,99 1 -
| |
| V (E)- 100 - - - - -
| |
| -NOTES (A) values are rounded and may not add to 100%.
| |
| .(B)' Means probability is less than 0.1%.
| |
| 1 (C) 20 to 30% are caused by hydrogen burns.
| |
| l :--
| |
| (D). Containment failure is predicted to occur at 15 hr or so, still considered
| |
| -late. 3 (E). Interfacing systems LOCA.
| |
| 'l r
| |
| :j 1-l p:
| |
| 1 I
| |
| ; i -.
| |
| 04pnx.oso ll !- ,
| |
| t -e .-
| |
| = - . - . ____- . . _ _ _ _ _ _ _ _ .
| |
| | |
| -c 134 4 TABLE-4'.7-1 NOTATION AND DEFINITIONS FOR RELEASE CATEGORIES RELEASE CATEGORY DESCRIPTION 1
| |
| -M1A ' Containment Bypass, V-Sequence Containment Bypass, SGTR-l M1B S
| |
| M2 - Early Failure /Early Melt,,No Sprays -
| |
| M3' Early Failure / Late Malt, No Sprays t.
| |
| M4- Containment Isolation Failure M5 Intermediate Failure / Late Melt, No Spraya M6 Intermediate Failure /Early Melt, No' Sprays -
| |
| 1 q
| |
| M7 Late' Failure, No Sprays l
| |
| < 1
| |
| , MS - Intermediate Failure With Sprays ]
| |
| M9 Late Failure With Sprays n l
| |
| l M10 .Basemat Failure', No Sprays- l M11 Basemat Failure With Sprays i i
| |
| M12 No Containment Failure . ..
| |
| .i f .'T i
| |
| l .
| |
| u .
| |
| b' i
| |
| i t
| |
| l' 1
| |
| , 04Di7%.000
| |
| . i ,, . 1- j -
| |
| s s . . . i . . . , ,
| |
| | |
| Ts.bic 4.7-2 L
| |
| RELEASE CATEGORY StBOMRY
| |
| - CESIUM-10DilE MODEL
| |
| ~
| |
| - Release Release Release Point Estimate Source Tern Values Release Start Time Warning Time - - Duration Energy Cs-Rb Te-Sb Ba-Sr. Ru ' La
| |
| - (hrs) (hrs) (Btu /hr) Xe-Kr .01 I-Br Ea tegory thrs) 0.5 0.5 0.3 6E-2 2E-2 4E-3
| |
| - 2. 5 - 1.0 1.0 20 E6 0.9 K-3 M-1A SE-2 K -2 3E-2 K-3 2E-3 4E 4 1.0 1.0 20 E6 0.9 K-3 M-18 2.5 .
| |
| 0.6 0.2 K-2 2E-2 3E-3 2.0 150 E6 0.7 SE-3 0.6 M-2 0.75 0.2 0.6 0.6 0.2 8E-2 3E-2 X-3 6.0 0.5 2.0 190 E6 0.8 SE -3 M-3 0.6 0.6 0.5 K-2 E-2 K-3 l
| |
| 0.2 0.0 , 2.0 70 E6 0.9 6E-3 M-4
| |
| * C 0.5 0.5 0.5 K-2 4E-2 6E-3 0.5 450 E6 0.9 - 6E-3
| |
| ^
| |
| M-5 8.3 4.1 6E-3 0.5 0.5 0.5 SE-2 4E-2 K-3 ~
| |
| 4.3 4.1 ' O.5 440 E6 0.9 M-6 0.3 0.3 0.3 3E-2 2E-2 4E-3 0.5 540 E6 0.9 6E-3 M-7 20.1 16.0 i-IE-5 ' IE-5 IE-5 IE-6 IE-6 2E-7 M-8 4.5 4.0 0.5 22 E6 0.9 7E-3 2E-6 2E-6 IE-6 2E-7 9E-8 IE-8 21.0 20.0 0.5 22 E6 0.9 CE-3 M-9 GE-4 8E 4 IE-3 9E-5 K-5 IE-5 95.0 80.0 10.0 NA 0.3 2E-3 M-10 IE-S lE-5 2E-5 IE-6 .1E-6 2E-7 95.0 80.0 10.0 NA '6E-3 2E-5 M-Il .
| |
| IE-6 IE-6 9E-7 2E-7 8E-8 IE-8 0.5 0.0 15.0 NA lE-3 9E-6 M-12 4
| |
| - e w== w
| |
| % -wn,,,r. g e .,_ . _ . . - ,m_2- .
| |
| % yaw.w.gg - a_c emw-e .h --
| |
| | |
| 136
| |
| - j TABLE 4.9-1 .
| |
| COMPARISON OF MILLSTONE UNIT 3 AND SURRY UNIT 1 FEATURE MILLSTONE UNIT 3 SURRY UNIT 1 Reactor Typa PWR PWR 4 NSSS Vendor . Westinghouse Westinghouse y Architect / Engineer Stone an6 Webster Stone and Webster-Commercial Operation 1986 1972 Thermal Power, NWt 3411 2441 -j Number of Steam Generators 4 3- i l
| |
| Containment Free Volume, ft.3 2.3 X 10 6 1.8'X 10 6 Containment Construction Reinforced Concrete Reinforced Concrete I Containmer.t Operating Pressure, psia 10-11 10 {-
| |
| 1.2 X 10 6 3.5 X 10 5 Refueling Water Storage Tank, gal.
| |
| Containment Design Pressure, psig 45 45 Recirculation Spray Pumps l Number 4 4 Design Flow, GPM'- 3950 3500 1
| |
| . Location All outside 2 Inside, Containment 2 Outside ,
| |
| 4 4 Accumulators Number 4 3 Water Capacity (each), ft.3 900 950 l; Pressure, psig 650 660
| |
| , , 1 Lower Seactor Cavity .
| |
| Concrete Type ~ Basaltic Basaltic !
| |
| Concrete Thickness Above Liner J ft., 9 in. 2 ft. !
| |
| ' Concrete Basemat Thickness 10 ft., 3 in. 10 ft. }
| |
| Total Floor Area, f t.2 - 947- 620 -
| |
| l Water Volume to Spillovar from l Cavity to Containment Sump, ft.3 6,700 12,400 Water t' '''me to Spillover from con + _. nt Sump to Reactor Cavity,, gal.- 1.266 t 0.073.X 10 6 N/A Source of Information for Surry: Reference 4-19 04Df73.040 t
| |
| 61 p. . . g .
| |
| | |
| TABLE 4.9-2 APPLICABILITY OF NUREG-1150 INSIGHTS FOR SURRY TO MILLSTONE UNIT 3 REFERENCE SURRY ISSUE APPLICABILITY TO MILLSTONE 3' 4-19, Abstract Most of the risk is from V- True for ISL; not'necessarily so sequences (interfacing systems for SGTR. The potential use of' LOCA and SGTR). loop isolation valves for beyond design basis accidents has not been fully credited.
| |
| 4-19, Section 1.2 Reactor cavity will remain dry Generally true.
| |
| unless containment sprays operate.
| |
| 4-19, Sections 1.2 and 2.1.5' There is no connection between Because of the large volume of the the sump and the reactor cavity RWST, water spillover into the l- at a low elevation in the. con- cavity is possible for many
| |
| !. tainment. There is no overflow sequences. -
| |
| from the. sump to the~ cavity. g 4-19, Section 2.1.1 Containment failure pressure True (median failure pressure of is.between two and three times 117 psig) ,
| |
| the design (mean failure at 126 psig).
| |
| 4-19, Section 2.1.2 Operation at.sub-atmospheric True containment pressure makes ,
| |
| existence of pre-existing leaks negligible.
| |
| 4-19, Section 2.1.3 Fan coolers are not a viable . True means of containment heat removal during accidents.
| |
| 4-19, Section 2.1.5 The seal table is inside the Not directly applicable. There is crane wall, so molten core an opening in the crane wall allow- .
| |
| debris ejected at high pressure iry; direct line of signt from the cannot attack the containment seal table area to the containment
| |
| . pressure boundary. wall (estimated distance of 22 ft.).
| |
| l.
| |
| osunx.oeD
| |
| | |
| ~
| |
| 7; . _
| |
| ~
| |
| : TABLE 4.9-2 '(CONT'D. )
| |
| APPLICABILITY OF;NUREG-1150 INSIGHTS FOR SURRY TO MILLS 1DNE; UNIT'3 REFERENCE SURRY ISSUE APPLICABILITY TO MILLSTONE 3 4-19, Section 2.1.5 There is no-pathway'from the True reactor cavity to:the contain-
| |
| . ment. wall. -3ence, a direct-impulse from ex-vess31 steam -
| |
| explosions are not possible
| |
| . and early containment failure via-this mechanism is not likely.
| |
| 4-19, Section 2.3.1 With no containment spray Detailed containment analyses- g available, the continual indicate potential failure beyond a' deposition of decay heat into day.for large LOCA, but no conta.in-containment by' operation of ment failure for small break LOCA.-
| |
| of the ECCS.in the recircula-tion mode for LOCA will-lead to eventual-containment failure ;
| |
| in many hours or a few days. :
| |
| Containmentifailure, in turn, C' may leadLto ECCS failure. l 4-19, Section 2.4.1 If the-reactor cavity was dry True (approximately 3.8 ft. would- .
| |
| - at reactor vessel breach and cover the cavity floor) the accumulators: discharge at breach, then the cavity will be about'one-quarter full-(i.e.,
| |
| 4.5 ft. of water). This could delay core concrete interactions. ;
| |
| 4-19, Section 2.'5.1.4 For the V-sequence via failure Because of an additional check-of check valves between'the.RCS . valve,.this ISL pathLis~less likely.
| |
| - and the LPSI, water from~the However, an ISL via the'RHR suction RCSland-the.RWST escaping line would'likely result'in a pool
| |
| , through the break would form a covering the. break within the RHR- '
| |
| pool covering the break by the pump cubicle, based on' engineering-
| |
| ~
| |
| time core. degradation com judgment.
| |
| a menced. 'This would' provide:
| |
| - some fission' product scrubbing.
| |
| 04LW7X.08D a.
| |
| .1--*,w Aqw_%-e-,p.me-.,, y 9.r- - . - - .g .y.- p-* e-.-- r s< -
| |
| -s r . - - - - , f.y_, c y.,, , q ,- .fap. = ,
| |
| | |
| ky . .. -
| |
| ~
| |
| TABLE 4.9-2-(CONT'D.)
| |
| ' APPLICABILITY OF NUREG-1150' INSIGHTS FOR SURRY TO MILLSTONE. UNIT 3-SURRY ISSUE' APPLICABILITY TO MITTAT N1r 3 REFERENCE
| |
| :The probability of a tempera-Not analyzed, but beileved to be 4-19, Section 2.5.1.5 ture-induced failure of.the true. -
| |
| RCS pressure boundary is quite high for' core melts at high-RCS pressure.
| |
| Temperature-induced SGTRs due. Not analyzed, but believed to'be 4-19, Section 2.5.1.5 to core melt accidents at.high true. There remains-large uncertainty,-however.
| |
| pressure are very unlikely. ~
| |
| It.was estimated to be very This is consistent'with the MP3 4-19, Section 2.5.1.9 PSS.
| |
| unlikely to-fall containment early (before vessel breach) due to hydrogen burn.
| |
| True 4-19, Section 2.5.3 Because the basemat is con-
| |
| . structed of siliceous' ~
| |
| g-(basaltic) concrete, long term containment failure is not likely'to be caused by noncon-densible gases from core con-
| |
| - crete interaction..
| |
| With elect.rical, power available This is consistent'with'the MP3 PSS.
| |
| 4-19, Section 2.5.5.1 continuously, hydrogen combus-tion would occur shortly after
| |
| : a. flammable concentration of-
| |
| -hydrogen was attained in the containment.
| |
| +
| |
| 04tW71.08D
| |
| | |
| Y M TABLE 4.9-2 (CONT'D.)
| |
| ' APPLICABILITY OF NUREG-1150 INSIGHTS FOR SURRY TO MILLSTONE UNIT 3 ,
| |
| REFERENCE SURRY ISSUE APPLICABILITY TO MILLSTONE 3 4-19, Appendix A Catastrophic rupture failure Not analyzed, but believed to be' (Question 44) of containment.is unlikely to true.
| |
| severely damage all four.
| |
| independent recirculation spray trains. Hence, spray' .
| |
| operation may still remain available following contain-ment failure.
| |
| 4-19, Appendix A The only hydrogen burns that The likelihood of containment (Question 50) appear capable of challenging failure from hydrogen burn is containment are those which estimated to vary greatly from occur when power is recovered sequence to sequence in the MP3 after the onset of CCI. PSS. The judgment of low likelihood Steam condensation from spray of hydrogen detonation.is. consistent.
| |
| recovery makes large hydrogen witt the MP3 PSS.
| |
| deflagrations possible. -
| |
| Detonations are not likely. $
| |
| 4-15, Section 5.3 Based on 1/42nd scale simula- No detailed review and analysis has tions of the reactor cavity, been made for Millstone 3. The at least 84% of the molten core effect of sub-compartments such as
| |
| -debris will be ejected out of the incore instrument ~ tunnel on l
| |
| the cavity during a high pres- debris dispersal is not clear. The ~
| |
| sure melt ejection accident. degree of " overhang" at the opposite This increases the potential end from the reactor vessel in the for direct containment heating. MP3 cavity is somewhat greater and may tend to further reduce the frac- !
| |
| r tion ejected. In general, 2he DCH issue is of greater concern at MP3 '
| |
| than previously thought. Further L
| |
| research appears to be warranted.
| |
| l Consistent with MP3 PSS..
| |
| 4-16, Section 3.3.1 The mean conditional prob- ;
| |
| ability of early containment !
| |
| l- failure from internal events is-low.
| |
| -i 04LW7X.08D i - , ..
| |
| . . _ , _ _ . . _ u.
| |
| | |
| I -
| |
| TABLE 4.9-2 -r(CONT'D.)
| |
| APPLICABILITY.OF NUREG-1150 INSIGHTS FOR SURRY TO MILLSTONE UNIT-3 ;.
| |
| REFERENCE SURRY ISSUE APPLICABILITY TO MILLSTONE 3 ,
| |
| . 4-16, Section13.3.1 'The principal containment Consistent with NP3'PSS.
| |
| release mechanism.is bypass due-to' interfacing systems ~ -t IDCA.
| |
| 4-16, Section 3.3.1' There is a high probability .Not fully analyzed,1but-believed:
| |
| that the RCS will be at rela- to be generally' applicable to MP3.
| |
| 'tively low pressures (less than 200 psi) at the time of . - ,
| |
| . reactor vessel' failure,-thereby ;
| |
| reducing'the potential'for a high pressure melt ejection and direct' containment heating.
| |
| 4-16, Section 9.2.3 Large,Ldry, and subatmospheric True ~
| |
| containment designs appear to
| |
| 'be quite robust in their ability.to contain severe accident loads. There is a :
| |
| high-likelihood of maintaining containment integrity.through-out the early phases of severe a accidents in which the poten- y tial.for large releases of ~!
| |
| radionuclides is greatest.
| |
| 4-16, Section 9.2.3 Containment' bypass. sequences True ,
| |
| l, represent a substantial frac- .;
| |
| [' tion of high' consequence accidents. .
| |
| 4-16, Section 9.4.2 The subatmospheric containment True is particularly reliable.in
| |
| ; . regard to containment isola-tion because leakage would be identified during operation.-
| |
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| + 143 ) !
| |
| TABLE 5.4-1 SAMPLE. MILLSTONE; UNIT 3 PRA SUPPORT ACTIVITIES l TIME PERIOD F SUPPORT ACTIVITY RHR Autoclosure Interlock Removal Evaluation 1990: f Impact.of Rosemount Transmitter Failure on ,
| |
| RPS-Unavailability 1989-90 Justification for Tech. Spec. Change to ,
| |
| . Section 3/4.5.1, " Accumulators" to increase the allowable time to restore an inoperable accumulator 1989-90' Prioritization of MOVs in order of their i importance to plant safety to support
| |
| : 1) the MOV Operability Issue and 2) the '
| |
| MOV Starter Binding Issue .1989-w Fast Transfer Scheme Evaluation 1989-Tech. Spec. Change to' increase the maximum <
| |
| containment operating pressure 1989 ,
| |
| Service Water-System Temperature Evaluation 1988-89 s D Impact of' Extended Puel Cycles on Core Melt u
| |
| ' Frequency 1988 L Prioritization of Important Systems and Procedures 1) for use in Operator Training-and 2)Lfor use in Quality Assurance Review 1987
| |
| 'l
| |
| ' Review of The Technical Bases of the MP3 Inservice Test-Program -1987 L
| |
| * 1 f Evaluation of AC-Independent Containment Spray System 1987-l Station AC Blackout Assessment 1986 Boron Dilution Operator Response Times Quantification 1985-87 p; Operator Training on PSS Insights 1984 Prioritization of Air-operated Valves 1990 I
| |
| l; 04LW7X.080 h
| |
| ', ~ , =
| |
| | |
| n .. '' - ''
| |
| ~-,
| |
| tw TABLE 5.4 -PDCRs TO1BE INCORPORATED INTO FUTURE UPDATE DESCRIPTION MODEL AREA AFFECTED PDCR NO. TITLE
| |
| .~ Seismic Risk Analysis:-
| |
| - MP3-86-126 Emergency Diesel' Replacement of.the existing Generator emergency DG lube. oil cooler Lube Oil Cooler SAE-grade 2 anchor bolts Anchor Bolts with. bolts made.of~a stronger material (ASTM A-325)'
| |
| * Installation of'an ATWS- ATWS Event Tree MP3-88-008 Millstone Unit 3 AMSAC System Mitigating System Actuation
| |
| ~
| |
| Circuitry (AMSAC). This system provides an alternate means'of tripping the turbine g
| |
| and actuating auxiliary.
| |
| * feedwater flow apart from the RPS.
| |
| Human. Reliability MP3-88-009 Modification to- Modifications-on Main Control Model.for Recircu-the Cold Leg l Board #2 to reflect the steps lation Fault Tree-Recirculation in procedure.EOP ES-1.3 titled, Array. '" Transfer to Cold Leg
| |
| ~ Recirculation" i.
| |
| l 04LW7X.0BD ..
| |
| | |
| 145 l Table 6-1
| |
| . SAMPLE PLANT IMPROVEMENTS WITH HIGH'PRA BASIS I Title Description status PDCR 3-86-126 Replacement of SAE-Grade 2 Installed- I EDG Lube Oil Cooler anchor bolts with bolts ,
| |
| Anchor Bolts made of stronger material (ASTM A-325). Improves soismic PRA. )
| |
| PDCR 3-88-008 ATWS mitigation-system to Installed l AMSAC provide alternate means of l turbine trip and actuation l of AFW. :
| |
| PDCR 3-88-009 Changes to Main Control Installed Cold Leg Recircula- Board to reflect " Transfer tion Array to Cold Leg Recircuhtion EOP." Improves hur.ar *.eli-ability.
| |
| RHR Autoclosure Eliminates a major con- PRA analysis 1' Interlock Removal tributor'to RHR system completed.
| |
| unavailability during To be imple- l shutdown and provides alarm mented during to reduce potential for third interfacing system LOCA. refueling outage.
| |
| Loss of RHR in mid- Improves monitoring of RCS Installed i loop as a Response level, temperature, and RHR l to GL 87-12-and system performance during l GL 88-17 reduced inventory conditions.
| |
| AOP 3560 Loss of Provides once-through- Implemented Service Water cooling of charging pumps for loss of service water event, and reduces poten-tial for loss of all RCP seal cooling.
| |
| Accumulator Check Provides part stroke test Implemented Valves Surveillance of accumulator check valves every refueling interval based on PRA recommendation.
| |
| C41.W7X.08D
| |
| | |
| i.
| |
| i 146 Table 6-1 SAMPLE PLANT: IMPROVEMENTS. WITH HIGH PDA BASIS (Continued)
| |
| TR-Z.1-Response to Steps added to' procedure to Implement High' Containment ensure sufficient water in ed Pressure containment recirculation pumps. Minimizes potential consequences of incere instrument tube rupture with failure of.all quench-spray (S 2 C sequence in WASH-1400).
| |
| Station Blackout Addition of third. diesel Conceptual ;
| |
| .. Rtasolution generator for safe shutdown design loads. stage. To be imple-mented in-accordance-with SBO require- 1 ments, j
| |
| 'l i
| |
| l l
| |
| J ,
| |
| c a x.os:
| |
| | |
| I Y 147 l
| |
| ._ Table 6-2 l
| |
| ADDRESSMENT OF.SICNIFICANT PRA FINDINGS Finding Addressed-By Station Blackout is 1. Station Blackout' sequence re-major contributor to analyzed.-
| |
| public risk.
| |
| : 2. AC-independent' containment spray evaluated (found'not necessary).
| |
| : 3. Recovery of Offsite Power l
| |
| ~
| |
| steps in procedure is a I priority for operator !
| |
| $. training.
| |
| : 4. Procedure in place for severe weather conditions, including hurricane.
| |
| : 5. Air-cooled diesel generator. I to be added.
| |
| : 6. Numerous activities institut-ed in response to Station Blackout rule. _1 Interfacing Systems 1. Interfacing systems LOCA .
| |
| LOCA is major contri- sequences re-analyzed.
| |
| butor to.public risk.
| |
| : 2. Third MOV in RHR suction 1 lines credited.
| |
| : 3. Alarm for open valve to be' q added as part of RHR auto - -
| |
| closure interlock removal project.
| |
| : 4. 1988 Emergency Exercise involved PSS - identified ISLOCA in RHR pump suction j '
| |
| line.
| |
| : 5. RHR system walkdown in ESF y building performed to deter-mine characteristic of poten- j tial releases.
| |
| Auxiliary Feedwater 1. AFW system, recovery of main and-Feed and Bleed feedwater, and primary feed Failures are in many and bleed procedure assigned I accident sequences. high priority for operator training.
| |
| ,04LW7X.06D w
| |
| | |
| g y ,
| |
| 148 Failure of' containment l '. Design change'for cold leg sump recirculation is recirculation array has been-found in' dominant implemented.
| |
| sequences. 1
| |
| : 2. Transfer-to sump.
| |
| recirculation steps in. j emergency operating procedure H is a priority for operator training.
| |
| : 3. Service water to containment I recirculation cooler MOVs is j a priority for maintenance l activities.
| |
| Seismic-induced.sta- 1. Anchor bolts replaced with J tion blackout is major bolts of stronger material.
| |
| risk contributor; dom-inated by diesel gen-erator oil cooler an-chor bolt failure.
| |
| : 2. AC-independent containment spray of high g-level capa-city evaluated and found not necessary.
| |
| Failure of safety in- 1. Part stroke testing every jection accumulators refueling outage effected.
| |
| -is! major contributor to core melt frequency. <
| |
| ' Loss of 125 V vital DC 1. Loss of-DC test performed power-identified'as-before commercial operation- -
| |
| major' contributor to.
| |
| core melt-frequency (initial PSS submit- ,
| |
| tal)~.
| |
| Loss-of 120 V vital AC 1. " Note / Precaution" added to power identified as operating procedures to alert major. contributor to operator of loss of Emergency core melt frequency Diesel sequencer on loss of (initial PSS submit- vital AC.
| |
| 1 tal).
| |
| : 2. Centrol room simulator model corrected based on PRA in-sight regarding plant response.
| |
| : 3. Credit for manual recovery taken in PSS for certain support states.
| |
| c4anx.oso '
| |
| i
| |
| | |
| 5 y 1
| |
| 149 .
| |
| I i
| |
| " Dry"-lower reactor 1. Risk' impact'of flooded reac-
| |
| , cavity results:in sub- tor cavity design evaluated.
| |
| stantial~H 2- Prior to commercial
| |
| * generation. operation. , ;
| |
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| . _ . - - - .2 _-___:-----_.____--_____-____-_-_----___---______---_-_____----__-
| |
| | |
| i' _ ,
| |
| OPERATOR ELEASE PELEASE SEQUENCE NUMBEl i ESF' LOAD SERV 4CE ENTRY POWT NO LOSS OF' EMERGENCY - WATER RECOVERS MODE MODE AC BUSES ' ' CABINETS SEQUENCER PROB.
| |
| OFFSITE AVAILABl.E TRAINS ESFEGLS POWER avast /. ate AVAILABLE AVAILABLE EGLS SW RECOVERY LOOP EP .' ESF ,
| |
| E.P. 1.00E,00 3.02E-04 SS6 37 Transfer to page 2 of 2 d 'd2EaN 34E47 SS6 as 4 DOE-dM 1'fJ3E47 I SS6 39
| |
| 'd7E47 1.14E-10 SS7 40 2A2E47 SS6 41 9.30E-04 4.40E-04 1.24E-10 SS7 42 267 SS6 43 9.30E44 "4.42E 1.25E-10 SS7 44 4.76E-00 SS6 45A 1.50E-05 ' . ' ## ~ g 4A1E-11 SS7 458 -
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| SS6 46
| |
| -Q 152E-07
| |
| " 1.55E-10 SS7 47 ~
| |
| 1.16E-03 325E-10 SS6' 48A 1.00E+ 00 ~ 9.30E44 1M-02 328E-12 SS7 488 3.52E-07 SS6 49 l4.42E& 1.58E-10 SS7 50 1.16E 1 3.25E-10 SS6 51A 19.30E44 ~
| |
| 100E-02 3.20E-12 SS7 stb 4.92E-09 SS7 52A $
| |
| 1.64E-05 100E @ 4.97E-11 SS8 52B 1.5GE-06 SS6 63
| |
| " 4##" 7.01E-10 SS7 54
| |
| 'I '#'" 1.48E-00 SS7 55 e 1 ASE-00 SM 56 ~
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| " 8.16E-13 CS7 ,57 521E-03 1.16Em s 1.73E-12 SS7 4 58 3.07E-04 '9.30E-04 1.16E-03 1M@ SS7 M it.64E-05 2.61E-11 SS8 60 '
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| 1.5GE-06 SS6 61
| |
| " 4 7.04E-10 SS7 62 1
| |
| '9 30E& - - -
| |
| 1.48E-00 SS7 63 1 ASE-00 SS6 64
| |
| .4.42E& - 8.19E-13 SS7 66 521E-03 . 1.16E-03 s
| |
| # SS7 66 8 1.73E-12 1.16E-03 1 A6E-00 SS7 67 1.64E-05 2A1E-11 SS8 68 5.14E-08 SS7 69
| |
| # 5.98E SS7 70 1.68E-04 5.90E-11 SS7 71 1.16E-03 1 646 05 8.43E-13 SS8 72 SUPPORT STATE page 1 of 2) EVENT TREE 6/04/90 Figure 3.1-1: Support System Event Tree for Non Support System Related Initiators
| |
| | |
| NO LOSS OF EMERGENCY ESF LOAD SERVICE OPERATOR' RELEASE RELEASE 'SEQUENCE NURSEf OFFSITE AC BUSES CABINETS . SEQUENCER WATER RECXNERS MODE MODE POWER AVAILABLE AVAILABLE AVAILABLE . TRAINS ESF/EGLS PROS AVAILABLE LOOP EP ESF EGLS SW RECOVERY l
| |
| \.
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| 9.96E-01 SS1' 1 ,
| |
| 2.45i:-05 2.48E-05 SS2 2 l 2.41lE-05 SS2 3 2.48E-C5 1.91(E-09 1.90E-09 SS3 4 822E-04 SS2 5 2.49E-05 2.05E-08 SS2 6- 1 824E-04 2.49E-05 2.05E48 SS3 7 1.91 E-09 SS3 8 1.57E-12 NM SS2 9 >
| |
| 2.49E-05 2.05E SS2 to 8 24E-04 2.49E-05 2.05E48 SS2 11 1.91fE-09 1.57E-12 SS3 12 1.55E-05 SS2 13A 1.00E-02 SS3 138 1.56E-07 1.57E-05 2.49E-05 3.90E-10 SS3 14 2.49E-05 3.90E-10 SS3 15 1.91 E-09 2.99E-14 SS3 16 -
| |
| 2.49E-05 1.16E-03 SS2 17 3 2.89E-08 SS2 18 2.49E-05 2.89E-08 SS3 19 1.91 E-09 2.22E-12 SS3 20 1.16E-03 9.47E-07 SS2 21A 1.00E-02 SS3 21B 9.57E-09 8 24E-04 2.49E-05 2.38E-11 SS3 22 2.49E-05 2.38E-11 SS3 23 1.91 E-09 1.83E-15 SS3 24 ;
| |
| 1.16E-03 SS2 25 i 2.49E-05 2.89E-08 SS3 26 ,
| |
| 2.49E-05 2.89E48 SS2 27 1.91 E-09 28 'I 222E-12 SS3 !
| |
| 1.16E-OS 9.47E47 SS2 29A
| |
| :100E 9.57E-09 SS3 298 -
| |
| 8 24E-04 2.49E-05 2.38E-11 SS3 30 2.49E-05 2.38E-11 SS3 31 1.91 E-09 1.83E-15 SS3 32 1.62E-05 SS2 33A ri 00E-02 1.63E-07 SS4 338 1.64 E-05 l2.49E-05 4.07E-10 SS4 34 - t
| |
| ;249E-05 4.07E-10 SS4 35 I t.91 E-09 3.12E-14 SS4 36 m/90 Figure 3,1-1 (Contd.) ,
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| ' t PEAK: GROUND ACCELERATION RESULTS FOR MILLSTONE.- i 0.01 O
| |
| . C -EPRI NP4395 D'(Ref '317)? j A - NUREG/CR 4855 (Ref. 316) _ '' N
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| . O DAMES & MOORE (Ref. 314)
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| 155 Over:ll Technic:1 Pr ject Management Ver6 cation of Test, Maintenance and Operating Assumptions Northeast Utilities Service Co.
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| Northeast Nuclear Energy Co. e Probabilistic Riek Assessment 6
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| o Unit 3 Operations 2S sts e 2 Shift Supervisors -
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| o Millstone Unit 3 Project Office -
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| e Unit 3 Startup Testing Group 1 Project Engineer o 2 Engineers o Radiological Assessment Branch 1 Engineer o Generation Civil Engineering 2 Engineers a
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| 1 Engineer 3 o Milstone Unit 3 Project Office 1 Project Engineer Degraded Core Phenomenology o Probabilistic Risk Assessment 8 Engineers Fauske Assoc.. 2 Technicians
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| ; o Core and Containment Analysis 1 Engineer 4 Engineers o Offshore Power Systems (Jacksonsonville, Fla.)
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| 7 Engineers
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| - Core Concrete Interactions o Reliability Engineering Michael A. Co radini Assoc 2 Engineers
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| o Millstone Unit 3 Project Office Stress Analysis of Containment Hatches I Oih W. J. Woolev Co. 7 Engineers o Environmental Division a Engineers 1 Scientist o StNetural Division 7 Engineers FIGURE 5.1 1 PROJECT ORGANIZATION
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| APPENDIX A Front-Line Event Trees 4
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| APPENDIX B !
| |
| i i.
| |
| System and Event Tree l Success Criteria l
| |
| i
| |
| {
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| I 5
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| l f I' i
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| TABITC B.1 8
| |
| SYSTEll SUCCESS CRITERIA i $YSTEM EVENT TEES SUCCESS CRITERIA OPERATOR ACTIONS REPEMNCIES RWI T18E NFERENCES 1
| |
| Aus f15ary 1) All transients Feed any 2 of 4 steam Confirm operetfen of I) Condensate water Bhsst start within 1,3,4,6, and 7 .
| |
| ; Feeduster except steam jfne generators with any system storage tank 30 elastes and l AF-1 breats aumfItary feeduster 2) Electric pouer ren for 24 hours
| |
| ' 2) Incore fostrument pump costilaation and 3) Actuotfen signal tube rupture secondary cooling 4) gente steam system
| |
| : 3) Smell LOCA l
| |
| l AF-2 Il Steam generato'r Feed any 2 of 3 Confirm operetten of 1) Condensate water Run for 24 heers 3.3.4,6 and 7 tube rupture unfaulted stems gene- system storage tank
| |
| : 2) Steam ifne break retors with any ausf- 2) Electric power transients llary feeduster pump 3) Actuation signal combinetten and 4) feels steam systee secondary cooling AF-3 I) Loss of offstte Feed any 2 of 4 steam Confirm operetten of 1) Condensate water Run for 24 heers 1,3.4.6 and 7 Y power-support generators with any system following storage tank
| |
| - state 7 combinetton of the pouer recovery 2) Electric power motor driven ausfilary 3) Actuation signal
| |
| * feeduster pumps and 4) femin steam system secondary cooling Contefament 1) Lar9e LOCA 1 out of 2 rectremia_ 1) Realfgn system 1) Electric power Run for 24 heers 1.3.4 and 15 Low pressure 2) flettum LOCA tien pumps dedicated 2) Ensere service 2) Service water
| |
| : 3) Actuation signal Recirculation 3) Small LOCA for ECCS rectrculation water coolf ag to .
| |
| Cooling delfverin5 flew 3) Conf *rm operation el Water in sump R-1 to the reactor coolant of system system and service l water cooling to the l containment cooler in I the operational recf r-colation train .
| |
| Contafseent 1) All event trees I out of 2 recircula- 1) Realign system II Electric peser Ron for 21 heers - 1.3.4, and 15 High pressure except large LOCA tion pumps dedicated 2) Ensere service 21 Service water Direct Nigh pres.
| |
| Rec f rculat1on for ECC5 recf rculation water coolf af to 3) Actuation signal sure Injettien l
| |
| Cooling delfverfag flow cooler 4) Meter in sum , runs for the first
| |
| ! R-2 to the section of the 3) Conf f ra operatfor 5) Continued operation 3 heers I
| |
| h1gh pressere fnjectfon of system of the hlgh pressore pumps and service der injection systee for cooling to the contain- 21 hours ,
| |
| sent cooler in the !
| |
| operational recf rcula-tion train 0
| |
| , e,- - . x e - +~ - , w , - - .. -- - n,, ,v w e- ., - - es a wm, -
| |
| | |
| TABLE B.1 (t'ont*d)
| |
| SYSTEN 5etCE55 CRITERIA
| |
| ~
| |
| EEPERIENCIES AWN TIME NFEE NCES SUCCESS CRITERIA frEART6R ACTIONS 5f57EM EMNT TREES I) Electric power sun for 24 hours ~ I,3,4 and 35 Containment 1) All event trees I eet of 2 recircola- 3) Ensere service tien pumps ,*=dicated meter coeltog to 2) Service seter 5 pray Necir- to cooler .. 3) Acteetion signal colation for continued Wafe- 4) Water la sump.
| |
| R.3 seet sprey rectrch?a- 2) Confire system
| |
| ( tien delfverlag flom operetten to the contef aseet
| |
| ' spray recirculatlea spray aerales and se -
| |
| ofte ester cooline to the contef ament .soler
| |
| . in the operatf sel rectrculatte* spray trafn g
| |
| : 1) Electric poner Sependent en I.3 and 33 Charging Pump 1) All event trees I chorging pep celing 1) Coeffre operatfee che'T ing penny a pump supplyf ag cooltag of system 2) Service seter run tfee
| |
| : Cooling 3) Acteettee sfp1
| |
| " for each chargfag pump -
| |
| in operation
| |
| : 1) Coeffre operatlee 1) Electric power Dependent en 3.3 and 13 Safety injec- 1) All event trees I safety fajectlen safety injectfee pump coef f ag peny say- of systee 2) Service fester tion Pump 31 Actoetien signal pump run tfee Cool 1og plying cool 1mg for each safety fejection pesy la operation men for 4 heers 1,3 and 8 1 w h spray pump Il Confire operation 1) Electric peuer Quench 5 prey 1) All event trees 2) Acteettee signet 05 delfverfag flom to one of systee w h spray heeder 3) NW5T 2 eet of 4 high pres- 3) Coeffre operetten 1) Electric pouer nun for 3 heers- 3.3.4 and 9 High Pressere 1) Large LOCA 2) Chorgfag and/or sure safety fejection of sestem Safety injec-pumps deliverfag flom safety fejection tion pump coolfag HP-1 to the reacter coolant 3) Acteetion sfgnol
| |
| . system
| |
| : 4) WWST -
| |
| Run feltfally for 1 through 4 e Il Coeffre operation 1) Electric power g through 32 HP-2 I) All event trees I eet of 4 high pres.
| |
| of system 2) Chorging and/or 3 heers during encept large LOCA sore safety Snyxtfee _fejectfen pies an pemips delfvering flom safety fejectfen yes ,coollag additformel 21 hoors to the reactor coolant 31 Acteatfen signol derfag rectrcelation system e) WWST
| |
| | |
| I li11 ii 8 '
| |
| TABLE B.1 (Cont'd)
| |
| Sv5 tem SutCE55 Cn:Tra:A SPERATOR ACTies5 Kpte8EstfE5 RW 7pE EFfE8KES SYSTEM EWNT TEES SECESS CRITERIA
| |
| : 1) Confirm operation 1) Electric poner nun for 3 hours 1,3,4 and 9 tow pressure 1) All LOCA event trees 1 est of 2 low pres-Safety Injec- sere safety injection of system 2) Service mater pumps deliverf g flow 3) Actuatfan signal tion to the reactor coolant 4) RWST Lp
| |
| * system I) Confire operetton moce met appilcable 1.3.4 and g Accumulators 1) tarve LOCA Successful efscharge of ACC 2) Medium LOCA '3 accumulators into of system 3 intact reactor coel-ant loops
| |
| : 1) Conffra operetten II Electric power Den for 24 hours I,3 and le Service Water 1) All event trees 1) Two service mater trates in operation of system 2) Water searce
| |
| : 2) One service mater train in operation
| |
| : 1) Confire operation 1) Switchyard Asaf1 Ale for I,3 and 5 Electrical 1) All event trees 1) Two E5F buses 24 heets poner energfred of systen 2) Diesel generators and acteatfen signal
| |
| : 2) One E5F bus
| |
| ?
| |
| 4.a energfmed
| |
| : 1) Chf11ed eeter systs met appilcable I and 3 Refuellag 1) All event trees Available with feven. None Water Storage tory egm1 to or greater than technical specIff-Tank TK c.atfen Confire closere 1) 5fgnal actuation met appiscable I 3.4 and 6 Main Steam 1) Steam Ifne break Isolatten of faulted Isolation inside contef ament steam generator - clos.
| |
| MS-3 ere of either 1 out of I snain steam isoletten valve (MSIV) er closure of 3 out of 3 MSIVs Coeffre c1esare 1) Signal actuatten met applica.te 1.3.4 and 6 MS-2 1) Steae Ifne break Isolatten of break-outside containment closure of 3 out of 4 Mstys Confirm trfp 1) Signal acteatlen Ret appifCable 1.3 and 4 Reactor pro- 1) All event trees Generation of reactor tection System except reactor trip trfp signal and f aser-RT-1 and the large and tion of reactor control sedfum LOCA event rods
| |
| . trees
| |
| | |
| TABLE B.1-(Cont'd)'
| |
| SYSTEM SWCCESS CalTERIA EEPEIBENCIES SUN Tut AEMEENET5 SUCCESS CRITERIA WERETOR ACil0h5 SYSTEN EVENT TMES None not applicable 3,3 and 4 I
| |
| Insertion of reactor Confire trip RT-2 !) Seactor trip control rods Isot appitcable 1,3 and 4
| |
| : 1) Anticipated tron- Operator.Intiated Ituuselly inttfate None RT-3 reactor trip sients without reactor trip with scram fasertfon of reactor control rods within one ofnete upon fellere of an automatic reactor trip or signal Confire trfp mene met applicable 1.3.4 and 16 Turbine Trfp 3) Anticipated tran- Closure of all turbine sients without step or control selves scram
| |
| . . . _ _ _ k
| |
| | |
| TACLE B.2 I
| |
| EVENT TEE SUCCESS CRITERIA PLANT FUETIONS OR EVENT TEE SYSTEMS EQUIED OPERATOR ACTION 5 EQUIPENT REQUIED Large LOCA 1) Low pressure safety 1) Realign containment 1) Safety injection injection (LP) recirculation spray actuation signal
| |
| : 2) Low pressure recirculation system to RCS .-2) RWST cooling (R-1) recirculation 3) Accumulators
| |
| : 1) High pressure safety 1) Realign containment 1) Safety injection injection (HP-1) recirculation spray actuation signal
| |
| : 2) Low pressure recirculation system to RCS 2) RWST cooling (R-1) recirculation 3) Accumulators i
| |
| en Medium LOCA 1) High pressure safety 1) Realign containment 1) Safety injection
| |
| & injection (HP-2) recirculation spray actuation signal
| |
| : 2) High pressure recirculation system to RCS 2) RWST cooling (R-2) recirculation 3) Accumulators
| |
| : 1) Low pressure safety injection 1) Depressurize primary 1) Safety injection actua-
| |
| : 2) Auxiliary feedwater (AF-1) system using auxi- tion signal ;
| |
| : 3) Low pressure recirculation 11ary feedwater and 2) RWST i cooling (R-1) steam relief valves 3) Accumulators j
| |
| : 2) Realign containment 4) Steam reifef valves recirculation spray system to RCS ,
| |
| recirculation !
| |
| t Small LOCA 1) High pressure safety 1) Controlled primary 1) Reactor trip (RT-1) injection (HP-2) depressurization 2) Safety injection actua- ;
| |
| tion signal
| |
| : 2) Auxiliary feedwater (AF-1) l
| |
| : 3) Refueling water storage i
| |
| tank Note 1 The table only gives the success criteria in terms of front line event tree systems; support systems are not shown. !
| |
| I !
| |
| . t 4 !
| |
| | |
| TABLE B.2 (Cont'd)
| |
| EENT TEE SUCCESS CRITERIA i
| |
| ! PLANT FUNCTIONS OR EENT TEE SYSTEMS EQUIED OPERATOR ACTIONS EQUIPE NT E QUIK O i
| |
| Small LOCA 1) High pressure safety 1) Realign containment 1) Reactor trip (RT-1) t (con't) injection (W-2) recirculation spray 2) Safety injection actua-
| |
| : 2) Auxiliary feeduater (AF-1) system to RCS tion signal
| |
| : 3) High pressure recirculation recirculation 3) Refueling water storage cooling (R-2) tank
| |
| : 1) High pressure safety 1) Initiate bleed and 1) Reactor trip (RT-1) 1 injection (HP-2) feed cooling 2) Safety injection actua-
| |
| : 2) High pressure recirculation 2) Realign containment tion signal
| |
| , cooling (R-2) recirculation spray 3) Refuelin5 water storage l w system to RCS tank j
| |
| & recirculation 4) Pressurizer PORVs 1 1) Auxiliary feedwater (AF-1) 1) Depressurize primary 1) Reactor trip (RT-1) i 2) Low pressure safety injec- system using auxili- 2) Safety injection actua- ,
| |
| tion (LP) ary feeduater and tion signal
| |
| ; 3) Low pressure recirculation steam relief valves 3) Refueling water storage ,
| |
| cooling (R-1) 2) Realign containment tank recirculation spray 4) Steam relief valves i
| |
| system to RCS i i recirculation -
| |
| i l Steam Generator 1) High pressure safety None 1) Reactor trip (RT-1)
| |
| ; Tube Rupture injection (HP-2) 2) Safety injection actua- ,
| |
| 1 2) Auxiliary feedwater (AF-2) tion signal
| |
| : 3) Refueling water storage j i tank
| |
| ,.c p - _,.m .
| |
| #y ~
| |
| %; w- ew-v-e- , e-=
| |
| * e, e spe , =e - - - ~ v-+- .-w..=---%. -w e +-<_ . . . ~- -e.a-w --- *..m=m+ -m-ee nmw
| |
| | |
| 2 TABLE B.2 (Cont'd)
| |
| ; EENT TEE SUCCESS CRITERIA i
| |
| i PLANT FIRICTIONS OR l EENT TEE SYSTEMS EQUIED OPERATOR ACTI0liS EQUIPENT EQUIED ,
| |
| Steam Generator 1) High pressure safety 1) Initiate bleed and 1) Reactor trip (RT-1).
| |
| Tube Rutpure injection (HP-2) feed cooling 2) Safety injection actua- l (con't) 2) High pressure recirculation 2) Realign containment -
| |
| tion signal cooling (R-2) recirculation spray 3) Refueling teater storage i system to RCS tank
| |
| - recirculation 4) Pressurizer PORVs t
| |
| : 1) Auxiliary feedwater (AF-2) 1) Initiate secondary 1) Reactor trip (RT-1) ,
| |
| depressurization to 2) Safety injection actua-terminate break flow tion signal
| |
| : 3) Steam relief valves l
| |
| : 1) Auxiliary feedwater 1) Depressurize primary I) Reactor trip (RT-1)
| |
| : 2) 1.ow pressure safety system using auxill- 2) Safety injection actua- .
| |
| ; injection (LP) ary feedwater and tion signal steam relief valves 3) Refueling water storage j tank
| |
| : 4) -Steam tw11ef valves Steamline 1) Auxiliary feedwater (AF-2) Norr 1) Reactor trip (RT-1) .
| |
| Break Inside 2) Safety injection actua-Containment tion signal
| |
| : 3) Main steam isolation i (MS-1) l t
| |
| : 1) High pressure safety 1) Initiate bleed and 1) Reactor trip (RT-1) !
| |
| injection (HP-2) feed cooling 2) Safety injection actua- l
| |
| : 2) High pressure recirculation 2) Realign containment tion signal (R-2) recirculation . spray 3) Refueling teater storage system to RCS tank recirculation 4) Pressurizer PORVs .
| |
| l l I l .. .
| |
| l l ,. - - . - .. . - . .
| |
| | |
| i TABLE B.2 (Cont'd)
| |
| EVENT TEE SUCCESS CRITERIA l
| |
| PLANT FIRICTIONS OR EVENT TEE SYSTEMS KQUIED GPE MTOR ACTIONS EQUIPK N KQUIED
| |
| )
| |
| Steamline 1) Aux 111ary feeduster (AF-2) Ilone 1) Reactor trip (E-1) f 2) Safety injectfen actes-l Break Outside Containment
| |
| - tien signal l
| |
| : 3) Main steam isoletten
| |
| ! (MS-1) i
| |
| : 1) High pressure safety 1) Initiate bleed and 1) Reactor trip ( E -1)
| |
| ; injection (HP-2) feed cooling 2) Salety injection actos-
| |
| : 2) High pressure recirculation 2) Realign containment tion signal (R-2) rectrculation spray 3) Refueling water storage 4
| |
| m system to IICS tank l E recirculation 4) Pressurizer PORVs I Loss of Reactor 1) Auxiliary feedwater (AF-1) Ilone 1) Reactor trip (RT-1)
| |
| Coolant Flow
| |
| : 1) Higli pressu" -
| |
| sty 1) Initiate bleed and 1) Reacter trip (RT-1) l feed cooling 2) Refueling water storage f ajectfon (L *
| |
| {
| |
| ! 2) H?2 h pressure rw.!rtulation 2) Real1gn costalasent tank I (R-2) recIrtulation spray 3) Pressurfzer PORVs
| |
| ? system to RCS h recirculation toss of Main 1) Auxiliary feedwater (AF-1) 10ene 1) Reactor trip (RT-1) l Feedwater t
| |
| ; I) Migh pressure safety 1) Initiate bleed and 1) Reactor trip (E-1) 1 injection (IF-2) . feed cooling 2) Refuelfag water storage t
| |
| : 2) Hfgh pressure recfnulat1on 2) Real1gn ;;ontaf ament tank i (R-2) -rec 1rculation spray 3) Presserfzer PORVs l i
| |
| system to RCS recirculation l
| |
| l'
| |
| | |
| TABt.E B.2 (Cont'd)
| |
| EENT TEE SUCCESS CRITERIA PLANT F11NCTIONS OR OPENLTOR ACTIONS _
| |
| EQUIPK NT K QUIE D EENT TEE SYSTEMS EQUIED
| |
| : 1) Reactor trip (RT-1)
| |
| : 1) Auxiliary feedwater (AF-1) None .
| |
| Prfrary to Secondary Power Mismatch
| |
| : 1) Reactor trip (RT-1)
| |
| : 1) High pressure safety 1) Initiate bleed and 2) Refueling teater storage feed cooling injection (HP-2) 2) Realign containusent tank
| |
| : 2) High pressure recirculation recf rculatfon spray - 3) Pressurfzer POWS (R-2) system to RCS recirculation
| |
| : 1) Reactor trip (RT-1)
| |
| : 1) Auxiliary feedwater (AF-1) None
| |
| $ Turbine Trip
| |
| : 1) Reactor trip (RT-1)
| |
| : 1) High pressure safety 1) Inftfate bleed and 2) Refueling esoter storage feed cooling injection (W-2) 2) Realign containment tank
| |
| : 2) High pressure recirculation recirculation spray 3) Pressurizer PORVs (R-2) system to RCS
| |
| ~
| |
| recirculation
| |
| : 1) Reactor trip (RT-2)
| |
| : 1) Auxiliary feedwater (AF-1) None Reactor Trip
| |
| : 1) Reactor trip (RT-1)
| |
| : 1) High pressure safety 1) Initiate bleed and 2) Refueling teater storage feed cooling injection (W-2) 2) Realign containment tank
| |
| : 2) High pressure recircula- recirtulation spray 3) Pressurizer PORVs tion (R-2) system to DCS recirculation
| |
| | |
| TABLE B.2 (cont'd)-
| |
| i EVENT TEE SUCCESS CRITERIA j
| |
| h PLANf FUNCTIONS OR EVENT TEE SYSTEMS REQUIED GPERATOR ACTIONS EQUIPE NT KQUIE D i
| |
| Core Power 1) Auxiliary feedwater (AF-1) None 1) Reactor trip (RT-1)
| |
| Excursion
| |
| : 1) High pressure safety 1) Initiate bleed and 1) Reactor trip (RT-1) injection (HP-2) feed cooling 2) Refueling water storage
| |
| : 2) High pressure recirculation 2) Realign containment tank (R-2) recirrulation spray 3) Pressurizer PORVs system to RCS recirculation y Spurious 1) Auxiliary feedwater (AF-1) None .
| |
| : 1) Reactor trip (RT-1)
| |
| ; g Safety Injection
| |
| : 1) High pressure safety 1) Initiate bleed and 1) Reactor trip (RT-1) injection (HP-2) feed cooling 2) Refueling water storage l 2) Realign containment tank
| |
| : 2) High pressure recirrulation (R-2) recirculation spray 3) Pressurizer PORVs l
| |
| j system to RCS recirrulation Loss of Offsite 1) Auxiliary feedwater (AF-1) None 1) Reactor trip (RT-1) g Power-Support
| |
| : States 5, 6 and 8 1) High pressure safety 1) Initiate bleed and 1) Reactor trip (RT-1) l injection (HP-2) feed cooling 2) Refueling water storage
| |
| : t. 2) High pressure recirculation 2) Realign containment tank i (R-2) recirrulation. spray 3) Pressurizer PORVs l
| |
| L system to RCS recirculation -
| |
| Loss of Offsite 1) Auxiliary feedwater None 1) Reactor trip (RT-1) l Power-Support (AF-1 or AF-3) 2) Power recovert
| |
| [ State 7 3) No reactor coolant i
| |
| punp seal LOCA ,,
| |
| | |
| TABLE D.2 (cout.'d)-
| |
| EVEN TEE SUCCESS CRITERIA
| |
| ^
| |
| PLANT FIRETIONS OR SYSTEftS EQUIED GPERATOR ACTIONS EQUIPPENT EQUIED EE NT T EE
| |
| : 1) Auxiliary feeduater (AF-1 1) Initiate high pressure 1) Reactor trip (RT-1)
| |
| Loss of Offsite safety injection due 2) Power recovery Power-Support or AF-3) 3) Refueling'woter storage State 7 (con't) 2) High pressure safety to reactor coolant injection (W -2) pump seal LOCA tank
| |
| : 3) High pressure rectrrulation 2) Realign containment recirculation spray (R-2)
| |
| ^
| |
| systems to RCS recirculation
| |
| : 1) Initiate bleed and 1) Reactor trip ( E-1)
| |
| : 1) High pressure safety 2) Power recovery injection (N7-2) ferJ cooling
| |
| : 2) Realign containment 3) Refueling water storage i' 2) High pressure rectreulation tank recirculation spray C (R-2) system to RCS 4) Pressurfzer.PORVs rec 1rculat1on
| |
| : 1) Controlled primary 1) Reactor trfp (RT-1)
| |
| Incore Instru- 1) Hfgh pressure safety 2) Safety 1mjection actus-ment Tube injection (HP-2) depressurfzation
| |
| : 2) Auxi1fary feedwater (AF-1) tfen sfgnal Rupture 3) Refueling water storage tank
| |
| : 1) Realign containment 1) Reactor trip (RT-1)
| |
| : 1) High pressure safety 2) Safety injection actua-
| |
| - injection (HP-2) recirculation spray tion signal
| |
| : 2) Auxf11ery feedseter (AF-1) system to RCS
| |
| : 3) Quench spray recirculation 3) Refueling meter storage
| |
| : 4) High pressure recintulation tank rec 1rtulation cooling (R-2) 69
| |
| ._.__._n_
| |
| | |
| 9 TABt.E B.2 (Cont'd)
| |
| EVENT TEE SUCCESS CRITERIA PLA(FINICTIONS OR GPERATOR ACTIONS EQUIPMENT E QUIE D EVENT TEE SYSTEMS EQUIED
| |
| : 1) Delay actuation of I) Reactor trip (RT 1)
| |
| Incore 1) High pressure safety 2) Safety injection actua-injection (W-2) containment recircula- tion signal Instrument tion until sufficient Tube Rupture 2) Auxiliary feedwater (AF-1) sump water level exists 3) Refueling water storage (con't) 3) High pressure recirculation 2) Realign containment tank cooling (R-2) recirculation spray system to RCS recirculation I) Reactor trip (RT-1)
| |
| : 1) High pressure safety 1) Initiate bleed and 2) Safety injection actua-injection (HP-2) feed cooling
| |
| , 2) Realign containment tion signal
| |
| : 2) Quench spray recirculation spray 3) Refueling water storage J.,
| |
| " 3) High pressure recirculation tank recirculation cooling (R-2) system to RCS recirculation 4) Pressurizer PORVs
| |
| : 1) Reactor trip (RT-1)
| |
| : 1) High pressure safety 1) Initiate bleed and 2) Safety injection actua-injection (HP-2) feed cooling
| |
| : 2) Delay actuation of tion signal
| |
| : 2) High pressure recirtulation 3) Refueling water storage cooling (R-2) containment recircula-tion until sufficient tank samp water level exists . 4) Pressurizer PORVs
| |
| : 3) Realign containment recirculation spray system to RCS recirculation
| |
| : 1) Depressurize primary 1) Reactor trip (RT-1) -
| |
| : 1) Auxiliary feedseter (AF-1) 2) Safety injection actua-
| |
| : 2) Low pressure safety system using aux 111-ary feedwater and tion signal injection (LP) 3) Refueling water storage
| |
| : 3) Quench spray- steam relief valves
| |
| '2) Realign containment tank
| |
| : 4) Low pressure recirtulation 4) Steam relief valves cooling (R-1) spray system to RCS recirculation ..
| |
| | |
| TABLE B.2 (Cont'd) i EVENT TEE SUCCESS CRITERIA PLANT FIRICTIONS OR :
| |
| EVENT TEE SYSTEMS EQUIED OPElWLTOR ACTIONS EQUIPENT KQUIED l Incore Instrument 1) Auxiliary feedwater (AF-1) 1) Depressurize primary 1) Reactor trip (RT-)) '
| |
| Tube Rupture 2) Low pressure safety system using auxiliary 2) Safety injection actua-(con't) injection (LP) feeduster and steam tion signe' i
| |
| : 3) Low pressure recirtulation relief valves 3) Refueling water storage i cooling (R-1) 2) Delay containment tank ;
| |
| recirculation until 4) Steam relief valves !
| |
| sufficient sump water j level exists 1
| |
| : 3) Realign containment !
| |
| recirculation spray ,
| |
| i
| |
| , system to RCS i
| |
| : j. ,L recirculation u t l Anticipated 1) Auxiliary feedwater (HF-1) None 1) Reactor trip (RT-3)
| |
| Transients W1thout Scram
| |
| : 1) High pressure safety 1) Initiate bleed and 1) Reactor trip (RT-3) injection (HP-2) feed cooling 2) Refueling water storage ;
| |
| l- 2) High pressure recirculation 2) Realign containment tank (R-2) recirculation spray 3) Pressurizer PORVs system to RCS recirculation
| |
| ; - 1) Auxiliary feedwater (E-1) 1) Initiate emergency 1) Turbine trip on power
| |
| : 2) High pressure safety boration level < 25 percent j injection (PL < 3 pertent) i 2) Pow d level < 25 per-cent (PL < 25 percent) !
| |
| or ATWS pr~ essure relief
| |
| , (PR) '
| |
| : 3) Emergency boration
| |
| : 4) Refueling water storage tank ;
| |
| I
| |
| | |
| - __=
| |
| TABLE B.2 (Cont'd)
| |
| EENT TEE SECESS CRITERIA PLANT FIRICTIONS OR
| |
| ~ SYSTEMS EQUIED GPERATOR ACTIONS EQUIPENT KQUIED EVENT TEE
| |
| : 1) Initiate emergency 1) Turbine trip (TT) or Anticipated 1) High pressure safety ATNS pressere reifef Transients Wfth- fa.jection (HP-2) boration and bleed
| |
| : 2) High pressure recirculation and feed cooling- -(PR) out Scram 2) Power levei < 25 per-(R-2) 2) Realign containment (con' t) rec 1rculation spray cent system to RCS 3) Refueling water storage recirculation tank
| |
| : 4) Pressurizer PORVs i'
| |
| I i
| |
| i ee
| |
| | |
| ii APPENDIX C Support State Event Trees for Support System Related initiators
| |
| * g i i
| |
| 5
| |
| | |
| i 4
| |
| E B
| |
| s 0
| |
| 9 t /
| |
| - 4 U O N d 6
| |
| E C
| |
| N EL R
| |
| E O A8 A8 A8A8 -
| |
| E S 78,4444444444445555555555M901234567890:2 333 0123455678890112234567 56666666666777 O W
| |
| P E E S T I
| |
| AE S ED 66,S676T67676767677867767778677677787778 F L
| |
| EO RA d SS3SSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSS SS3SSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSSS F
| |
| O F
| |
| O E 8366699684779 S
| |
| A . 40,7474757376837685736669SM0000000000400 14 4000044000000004440444 S
| |
| S EES O L0O E0R EE,EEEEEEEEEEEEEEEEEEEEEEMEEEEEEEEEEEEeE 33,1808657570757E71290347 0.5 1.3 8.0.6.6.0.5.6.9. ,.7 45, 449034724485$6 L RBP 0 9 4 ,7.1.0 1.0.5.5 1 0 0.0.1 0.0.0.6.6.1.3 8.0.6394941115111S111152462S69524625681112 N E
| |
| V I
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| G RSDS OREL E E
| |
| TERG AW A R T
| |
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| 4 4
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| |
| | |
| i l'
| |
| l i-NO LOSS OF EMERGENCY ESF LOAD SERVICE OPERATOR ELEASE MLEASE SEQUENCE Ml2 M4 ,
| |
| ENTRY PONT MODE OFFSITE AC SUSES CABINETS SEQUENCER WATER RECOVERS tsODE POWER AVAE.ABLE AVAEABLE AVAILAEE.E TRAINS v2 FAILED PROS-AVAE.ABLE ESFEGLS E_P. LOOP EP ESF EGLS SW OA ;
| |
| SAGE 41 SS2 2 i
| |
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| |
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| |
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| 1J6E47 SS7 62 l9~30E-04 2.8E47 SS7 63 3.54E47 SS6 64 1.16E-03 .4 42EM 1.5eE-10 SS7 66 329E-10 SS7 as l i 1.16E-03 3.54E47 SS7 67 i l 3 07E-04 1 ~64E-05 5.01E-OS SSs se 1 -86 SS7 60 ;
| |
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| |
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| |
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| |
| SS EVENT TREE GfvEN LOSS OF 1 SW TRAM GOW90 1
| |
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| |
| | |
| ENTRY PONT NO LOSS OF EMERGENCY ESF LOAD SERVICE OPERATOR RELEASE RELEASE SEQUENCE NURABEI l OFFSITE AC BUSES CAENNETS SEQUENCER WATER RECOVERS RIODE RADOE POWER AVAILABLE AVAILABLE AVAR.ABLE TRAINS 1/2 FAILED PROS.
| |
| AVAILABLE ESFEGLS E_P. LOOP EP ESF EGLS SW OA 9.98E 01 SS2 25 2.49E-05 SS3 26 2.48E-05 2ME 2.48E-05 SS2 27 1.91E-09 1.91E-09 SS3 28 y
| |
| 8.15E-04 SS2 29A H 00E W 823E-06 SS3 298 2.49EE 2.05E-08 SS3 30 8 24E-04 2M-05 2.05E-08 SS3 31 191E-09 157E-12 SS3 32
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| | |
| i ENTRY POWT NO LOSS OF E9ERGENCY ESF LOAD SEfWDCE OPERATOR RELEASE RELEASE SEQUENCE fetmAEf OFFSITE AC SUSES CA8pdETS SEQUENCER IMATER RECOVERS hsODE te00E POWER AVAEABLE AVAILABLE AVAILAELE TRmseS 1/2 FAILED PROS.
| |
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| |
| 3 07E 04 164E-05 i 5 01E4m SSe es j
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| 1 GOEC6 SS7 ao '
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| j 5 21E 03 1W43 1 SeE88 SS7 70 t .16E43 186E4e SS7 71 t 64E45 2 62E-I t SSS 72 ,
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| i - SS EVENT TREE GIVEN LOSS OF 120VAC-3 OR 4 emees ;
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| l' 1
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| \
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| I i
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| i l
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| 4 i
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| 1 1
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| l 1
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| 1 APPENDIX D i I
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| 1 I
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| Human Error Probabilities i 1
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| l l
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| l 1
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| l I
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| i I
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| -1 I
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| l l'
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| i, 6
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| i
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| ___._._...___._.._..w-me,.
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| _m,w..,. e ww -w--w--r-a-+-,---se we y v v-'--*-1
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| | |
| TABLE D.1 HISWI ERfl0R Pfl08ABII.ITIES FOR OPERATOR ACTIONS IN EVENT TREES T1se Failure Human Error Operator Applicable Probablitty Available Type Action Event Trees or Analysts 30 C 5 x 10-8 OA-1 ET03 ET16 2.3 x lo-*
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| 20 C OA-l' ET02 1 x 10-'
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| 30 C ET03, ET16 OA-2 30 C 1 x 10-8 ET03 ET04, ETOS. ET06 ET16 OA-3 30 0 1 x 10-8 OA-4 ET04 5 x 10-8 o to C 0A-5 ET04 OA-6 ET05 1 x 10-8
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| - 30 C Support States I and 5 1 x 10-8 60 0 Support States 2, 3, 4, 6 OA-6' ET06 ET13 1 x 10-8 20 C Support States I and 5 30 0 1 x 10-8 Support States 2, 3, 4, 6 1 x 10-*
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| 30 C OA-7 ET07 - ET21 (ET14A) 1 x 10-8 30 C OA-7' ET148 1 x 10-8
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| >60. C OA-8 I2) ET22 C 1 x 10-'
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| 10 OA-8,(2) ET22 1 x 10-'
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| 10 0 OA-9( 3) ET16
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| TABLE D.1L(Cont'd)
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| HUMAN ERROR PROBASILITIES FOR OPERATOR ACTIONS IN EVENT TREES Tfee Fa1 lure Nunen Error Operator Applicable Type Prebebflity Event Trees or Analysts Available
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| _Act' m I C 1 x 10-8 RT-3'#I ET22 1 C 1 x 10-8
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| -RT-4 ET22 60 C 2 x 10-'
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| R-1 ET01-ET04, ET15 60 C 1 x 10-*
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| R-2 III ET02-ET16 ET22 ,
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| 60 C 1 x 10-8 05' ET148 30 C 1 x 10-8 ESF ESF Recovery, Section 2.2.6 o NA C 1 x 10-8 A SI SI Recovery, Section 2.2.3.4 '
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| 30 C I a 10-8 SBI Consequential'S8I, Section 2.2.3.5 30 0 1 x 10-8 500 Consequential 580, Section 2.2.3.5 10 C 5 x 10-8 52 Consequential $2,.Section 2.2.3.5 NA P 1-m 10-8 SEQ Fire Analysts, Section 2.5 Recovery Analysis, Section 3.0 NA C 1 x 10-*
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| HP-2 NA C 1 x 10-8 OA-3 Recovery Analysis, Section 3.0 Recovery Analysis, Section 3.0 . 60 C 1 x 10-'
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| AFR OAF ET03 ET04. ET07-ET13. ET16 1 x 10-8 30 C ET20-ET21
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| - P. 4 x le-'
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| R-1 ET01-ET03, ET16 E102-ET16, ET17. ET18 ET20-ET22
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| - P 2 x 10-'
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| R-2
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| .,,..,,,...,m_,,., , - . , . , , , , , , , , , . , _ , . , _ , . ,
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| ..e, y...ip,y. -p,. ,.;,_ , ,, ,. .,.. , , , , .
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| ! h 5
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| b . . . . - .
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| m -O O O O @
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| m a m e m
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| {,
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| W. ,
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| l MM M MM i
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| e e e oo r bb I
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| M W , i W g
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| * & & AA f
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| - - R
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| * W .
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| a me W
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| m 1 I I
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| - I u . ;
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| ( > e
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| +
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| i g <
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| l ?u I.
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| V g i
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| ** m l' d m W
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| ; me o k
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| ' 4
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| [J j.' -
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| en se e W r tA M I
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| > a e e 5 k ,
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| l' 4 A C @
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| b L & m e l
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| * W -
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| .W e
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| * b2>
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| & g g e W bL b I I > e e & W >e s' W s== Wm i
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| * ed e M e I 6 e a m me N m N N gW '
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| O M O C C N l; 4 b3 3 bbb i
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| I O
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| - ed C ^
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| : 4. O @
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| w i
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| L
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| * 5
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| +8 LA. M N CC l ,w Q D-3
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| .n
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| )
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| Notes:
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| 1 (1) Controlled primary depressurization.,,0A-2, is used in the Small LOCA and Incoce Instrument Tube Rupture event trees. This action would conserve -
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| RWST inventory by avoiding or overriding a Containment Depressurization Accident (CDA) signal. The best estimate time available for this activ-ity is at least 30 minutes and therefore, the NREP screening value of 0.01 for the HEP is used. If OA-2 fails, the operator would have about three hours before receiving an alam to initiate RCS recirculation and I then at least an hour to actually initiate RCS recirculation to maintain j primary system inventory. These operator ac+. ions are not independent I actions. Therefore, for these two cases, the operator error portion of j the high pressure recirculation event, R-2, is calculated as follows:
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| o The owrall time frame is long; at least four hours.
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| o The overall probability of success is high, even if OA-2 fails,. '
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| due to the time available for the operator to recognize RCS :
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| recirculation will be necessary, redundant alams to announce the- j need for RCS recirtulation, and the time available after the
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| .l alam for the operator to initiate RCS recirculation. Therefore, the NREP screening value of 1x10-4 is used, t
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| o The HEP conditional probability of failure of R-2, given failure
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| ( at 0A-2 is: $
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| l 10~4 g-2 HEP (R-2/0A-2) ,
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| l 10-2
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| ''o All other values of R-2 are based on a conservative 60 minute time frame and have a HEP = 0.001. .
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| l (2) ' The AT.WS event tree considers the operator actions involved in achieving (
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| ! emergency shutdown given failure of the RPS to achieve automatic ;
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| D-4 ,
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| I shutdown. This is deemed to be a reliable operator action since the con- l ditions are easy to recognize and reactor trip is a major item of concern )
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| for the operator and is emphasized in training. The ATWS event tree is ,
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| strur.tured to distinguish between ATWS events that require emergency -
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| boration in the short time frame and those that require boration in the ;
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| long tie; frame. If the ATWS event tree node PL (initial reactor power ;
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| level less than 25 percent or the moderator temperature coefficient more negative than -5 pcm/*F) succeeds, then emergency boration is not I required within the short time frame (i.e., less than 10 minutes). In !
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| these instances, since a longer time frame exists for the operator to !'
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| start emergency boration (> 60 minutes) no dependence between RT-3, manual trip to drive the control rods in within one minute and 0A-8 is modeled and a HEP of 0.01 is assigned. j o- Thus, the operator actions of interest are OA-8 and 0A-8', emergency ;
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| boration plus primary bleed and feed cooling, coupled with f ailure of l RT-3 and PL., Although the time frame for combined ai:tions of RT-3 and-0A-8 or 0A-8' is 10 minutes, an overall HEP of 0.01 is assigned to reflect the assessed reliability of the operator in maintaining plant i safety functions. Dependence is assumed between emergency boration and ;
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| manual reactor trip such that emergency boration is only considertd on failure of manual reactor trip. A HEP of 0.1 is conservatively assigned to the conditional probability of operator failure to initiate emergency
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| ]
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| boration (OA-8 or 0A-8') given that the control rods did not insert into the core even though the operator had made the decision to shutdown manually. For the case where the operator fails in his decision to manually shutdown the reactor, a conditional probability of failure of 1.0 is given to the decision to initiate emergency boration.- f l' (3) The NREP HEP value of 0.5 for 10 minutes is abdified to 0.1 to account i for the fact that the operator has at least thirty minutes to identify the location of a small LOCA and verify containment sump level prior to reaching OA-9, delay containment recirculation (Section 2.2.3.2).
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| D-b L
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| --__4._-,.._ , . . _ _ , . _ _ _ _ _ _ -__ - ",._________i._____._._____.__.___________.______^___.__
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| r a a
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| : (4)1 Theifireanalys'isconsidersthepossibilityofhumanerror'in !
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| 1 following written. procedures for transfer of con, trol'of safeguards ]
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| . systems from.the control room to the auxiliary shutdown panel -
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| . necessitated by fire damage in the control room, instrument rack .
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| roem,' or cable spreading room. The rocedure for transfer of r I
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| control and operation from the auxiliary shutdown panel will be practiced on a regular basis by the operations' personnel. The.NREP if
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| * screening value for human errors occurring within a procedural :
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| i :
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| . framework where recovery is possible at the point of erroneous q a
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| 6
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| [' action is used to estimate the HEP for this analysis.
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| :f (5) The-plant and systems analysis includes the operator actions for.
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| L+ >
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| recovery; failures of both ESF cabinets, both EGLS cabinets, or one 4 b of each and failure of manual SI. Failure of the ESF cabinets, Y
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| ) EGLS cabinets, or a combination of each are easy to recognize and T require.the operator to follow his procedures and manually start '
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| a l <
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| , the affected-systems. The NREP screening value of 0.01 for 30- 4 minute time, frames is used. ;
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| p ..
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| : i. -f
| |
| ( th? .ase of: operator failure to follow-up on failure of automatic
| |
| ; 15 similar to the ESF/EGLS case above; it is easy to recognize, and the. operator is. required to follow his procedures and manually L ' initiate'SI. The time available for the operator-to recognize the {
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| 1 p 3-l.
| |
| event and take corrective action ranges from a few minutes (very j few, low: probability cases; e.g. , large LOCA) to half an hour or
| |
| 'r
| |
| 'more -(vast majority, more probable cases; e.g,, small LOCA and steam line break). An HEP of 0.1 is assigned to reflect the range and variability of time available to the operator, i i.
| |
| (6) Unlike the other procedural errors, the procedural HEPs for recovery l of main feedwater are included in appropriate places in the fault
| |
| [
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| tree. The HEP listed in Table D d is a total of each of the a-
| |
| . . individual HEPs included in the fault tree.
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| i i
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| D-6
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| % ____n . - .. . .
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| -~
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| g:--
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| W .
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| TABLE D.2
| |
| ^
| |
| THUfRN ERROR PROBA8ILITIES FOR FAULT TREE ANALYSIS HUNul ERROR RATE Type - - Operator ESF .EP. ..
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| EP .;
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| of Error Error - System Per Demand Variance Ceausents.' .
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| : 1. Omissfon ' Failure to restore a manual All ESF 1.0x10 1.0x10-8 See Table 2-0.11I !
| |
| ' valve-to normal position Systems "
| |
| i after test or maintenance act.
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| . 2. Omission ~ Failure to restore a' motor- All ESF 1x10-5 1.2x10-10 Valve status displayed at C8 driven pump-or an air or . Systems- valve switch and procedures motor operated value to~ with short list of checkoff - ;
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| normal position after test or provisions (< 10 special maintenance act. - instructionsT. EP also applies to restoration of a )
| |
| y pump C8 to normal position j' u after test or maintenance.- _
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| 3.
| |
| ~
| |
| Omission Failure to restore an alarmed . All ESF '1x10-5 1.2x10-10 Valve status displayed at C8 '
| |
| motor-driven pump or. an Sy. items valve switch, valve out-of -
| |
| af r'or motor operated valve position annunciator at C8,
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| . to normal position after test and procedures with short or maintenance act. list of checkoff provisions
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| .(< 10 special instructions). !
| |
| IfP also applies to restor-- ;
| |
| ation of a pump 08 to normal-
| |
| ~
| |
| l after test or maintenance- i that has similar status dis-play.
| |
| n.,-. . ..n.. .g .,.n. rv_.,. .,v--, ,. , . , , , . a. , - . . . . , s.n .+ , _ _ , __ _ , _ _ _ _ _ _ _ , _ _ _ _ _ _ , , _ , , . .,
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| | |
| TABLE D.2 (Cont'd)
| |
| Operator ESF HEP HEP Type of Error E rror System Per Demand Variance Comments
| |
| : 4. Procedural Error of omission / commission All ESF *1.0x10-3 1.1x10-6 Written procedures that spe-Error /With in operation of air-or Systems cify post-acci knt valve Recovery motor-operated valve required position are in use. Three for accident mitigation. operators are monitoring accident mitigation actions.
| |
| : 5. Procedural Error or omission / commission All ESF *1.0x10-3 1.1x10-6 Written procedures that spe-Error /With in operation of motor- e: Systems cify fluid flow requirements Recovery turbine-driven pump required are in use. Three operators for accident mitigation. are monitoring accident mitigation actions.
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| m
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| * Data Sourte
| |
| '!EP - U.S. Nuclear Regulatory Commission, " National Reliability Evaluation Program (NEP) Procedures Guide,"
| |
| NUEG/CR-2815, BNL-NUEG-51559, Review Draft, June 21, 1982.
| |
| | |
| TABLE Do3'
| |
| - Type of Error No. Desc ripti on 1 Failure to restore a manual valve after test and maintenance Assumpti ons:
| |
| o Level 1 tagging o Independent valve line-up perfomed af ter restoration o Checklist. required (1 10 items)
| |
| THERP Tree: a A
| |
| .999 .001 e
| |
| A
| |
| ,3 3
| |
| 5 .F 2 3 LTask Description
| |
| * HEP (Pounds)
| |
| A L = Error. of amission = .001 ( .0005 to .005) in using checklist -
| |
| Table 20-20 No.1 A' = Checker f ails to = .1 (.05 to .5) detect error A Table 20-16, No. 2 Quantification:
| |
| 1 F3 ,~A
| |
| * A' = (.001) (.1) = 1 x 10-4 Recommended Value: 1 x 10-4
| |
| *NUREG/CR-1278, Draft Report, October 1980, i
| |
| D-9}}
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