ML20216J379
| ML20216J379 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 03/10/1998 |
| From: | Haag C, Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20216J351 | List: |
| References | |
| NUDOCS 9803230430 | |
| Download: ML20216J379 (550) | |
Text
{{#Wiki_filter:- A AP600 DOCUMENT COVER SHEET
# TDC: IDS: 1 S Form 58202G(5/94)(t:\xxxx.wpf:tx)
AP600 CENTRAL FILE USE ONLY:
/ . 0058.FRM RFS#: RFS ITEM #:
V ' AP600 DOCUMENT NO. REVISION NO. ASSIGNED TO GWGLO22 11 Page 1 of 2 ALTERNATE DOCUMENT NUMBER: WORK BREAKDOWN #: 3.2.4 DESIGN AGENT ORGANIZATION: Westinghouse TITLE: AP600 Probabilistic Risk Assessment ATTACHMENTS: DCP #/REV. INCORPORATED IN THIS DOCUMENT REVISION: CALCULATION / ANALYSIS
REFERENCE:
ELECTRONIC FILENAME ELECTRONIC FILE FORMAT ELECTRONIC FILE DESCRIPTION (C) WESTINGHOUSE ELECTRIC COMPANY 1998 0 WESTINGHOUSE PROPRIETARY CLASS 2 This document contains information proprietary to Westinghouse Electric Company, a division of CBS Cc poration; it is submitted in confidence and is to be used solely for the purpose for which it is fumished and retumed upon request. This document and such informatP.a is not to be reproduced, transmitted, disclosed or used otherwise in whole or in part without prior written authortzation of Westinghouse Electric Company, subject to the legends contained hereof. O WESTINGHOUSE PRCPRIETARY CLASS 2C This document is the property of and contains Proprietary Information owned oy Westinghouse Electric Company and/or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document in strict accordance with the terms and conditions of the agreement u'vjer which it was provided to you.
@ WESTINGHOUSE CLASS 3 (NON PROPRIETARY)
COMPLETE 1 IF WORK PERFORMED UNDER DESIGN CERTIFICATION gg COMPLETE 2 IF WORK PERFORMED UNDER FOAKE. 1 O DOE DESIGN CERTIFICATION PROGRAM - GOVERNMENT LIMITED RIGHTS STATEMENT [See page 2) Copyright statement: A license is reserved to the U.S. Govemment under contract DE AC03-90SF18495.
@ DOE CONTRACT DELIVERABLES (DELIVERED DATA)
Subject to specified exceptions, disclosure of this data is restricted until September 30,1995 or Design Certification under DOE contract DE-ACO3-90SFtB495, whichever is later. EPRI CONFIDENTIAL: NOTICE: 1E203 4 5 CATEGORY: A N B C D E F 2 O ARC FOAKE PROGRAM - ARC LIMITED RIGHTS STATEMENT ISee page 2) Copyright statement: A license is reserved to the U.S. Govemment under contract DE-FCO2-NE34267 and subcontract ARC-93-3-SC-001. O ARC CONTRACT DELIVERABLES (CONTRACT DATA) Subject to specsfied exceptions, disclosure of this data is restr6cted under ARC Subcontract ARC-93 3-SC-001. ORIGINATOR SIGNATUR DA E ,
**Q AP600 RESPONSIBLE MAhAGER SIG
<PY Jf98 .
URI: ' APPROVAL DATE B. A. McIntyre w y y gf Ammal of the responsible ir.sr.i.ger signehes that document is complete, all reqGred reviews are complete, electronse file is attached and document is ro6 eased for use. 9003230430 990316 PDR ADOCK 05200003 - E PDR ,
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AP600 DOCUMENT COVER SHEET Page 2 Form 58202G(5/94) - LIMITED RIGHTS STATEMENTS DOE GOVERNMENT UMITED RIGHTS STATEMENT (A) These data are submitted with hmited rights under govemrnent contract No. DE AC03-90SF18495. These data may be reproduced and used by the govemment with the express hmitat on that they will not, without wntten permission of the contractor, be used for purposes of manufacturer nor disclosed outside the govemment; except that the govemment may disclose these data outside the govemment for the following purposes, if any, provided that the govemment makes such disclosure subject to prohibition against further use and disdosure: (1) This ' Proprietary Data' may be disclosed for evaluation purposes under the restrictions above. (11) The 'Propnetary Data" may be disdosed to the Electnc Power Research Institute (EPRI), electric utility representatives and their direct consultants, exduding direct commercial competitors, and the DOE National Laboratones under the prohibitions and restrictions above. (B) This notice shall be marked on any reproduction of these data, in whole or in part. ARC UMITED RIGHTS STATEMENT: This proprietary data, fumished under Subcontract Number ARC-93-3-SC-001 with ARC may be duphcated and used by the govemment and ARC, subject to the hmitations of Artido H-17.F. of that subcontract, with the express hmitations that the propnetary data may not be disclosed outside the govemment or ARC, or ARC's Class 1 & 3 members or EPRI or be used for purposes of manufacture without prior permission of the Subcontractor, except that further disclosure or use may be made solely for the following purposes: This proprietary data may be disclosed to other than commercial competitors of Subcontractor for evaluation purposes of this subcontract under the restnction Inst the proprietary data be retained in confidence and not be further disdosed, and subject to the terms of a non-disclosure agreement between the Subcontractor and that organization, exduding DOE and its contractors. DEFINITIONS CONTRACT /DEUVERED DATA - Consists of docurnents (e.g. specifications, drawings, reports) which are generated under the DOE or ARC contracts which contain no background proprietary data. EPRI CONFIDENTIALITY / OBLIGATION NOTICES NOTICE 1: The data in this document is subject to no confidentiality obligations. NOTICE 2: The data in this document is proprietary and confidential to Westinghouse Electric Company and/or its Contractors. It is forwarded to recipient under an obligation of Confidence and Trust for hmited purposes only. Any use, disdosure to unauthorized persons, or copying of this document or parts thereof is prohibited except as agreed to in advance by the Electric Power Research Institute (EPRI) and Westinghouse Electnc Company. Recipient of this data has a duty to inquire of EPRI and/or Westinghouse as to the uses of the information contained herein that are permitted. NOTICE 3: The data in this document is proprietary and confidential to Westinghouse Company and/or its Contractors. It is forwarded to recipient under an obligation of Confidence and Trust for use only in evaluation tasks specifically authorized by the Electnc Pcwer Research Institute (EPRI). Any use, disclosure to unauthorized persons, or copying this document or parts thereof is prohibited except as agreed to in advance by EPRI and Westinghouse Electric Company. Recipient of this data has a duty to inquire of EPRI and/or Westinghouse as to the uses of the information contained herein that are permitted. This document and any copies or excerpts thereof that may have been generated are to be retumed to Westinghouse, directly or through EPRI, when requested to do so. J NOTICE 4: The data in this document is proprietary and confidential to Westinghouse Electnc Company and/or its Contractors. ft is being revealed in confidence and trust only to Employees of EPRI and to certain contractors of EPRI for hmited evaluation tasks authonzed by EPRI. Any use, disclosure to unauthorized persons, or copying of this document or parts thereof is prohibited. This Document and any copies or ' excerpts thereof that may have been generated are to be retumed to Westinghouse, directly or through EPRI, when requested to do so. NOTICE 5: The data in this document is propnetary and confidential to Westinghouse Electric Company and/or its Contractors. Access to this data is given in Confidence and Trust only at Westinghouse fadhties for hmited evaluation tasks assigned by EPRI. Any use, disclosure to unauthonzed persons, or copying of this document or parts thereof is prohibited. Neither this document nor any excerpts therefrom are to be removed from Westinghouse facilities. EPRI CONFIDENTIALITY / OBLIGATION CATEGORIES CATEGORY "A'-(See Delivered Data) Consists of CONTRACTOR Foreground Data that is contained in an issued reported. CATEGORY T -(See Dehver(d Data) Consists of CONTRACTOR Foreground Data that is not contained in an issued report, except for computer programs. CATEGORY 't"- Consists of CONTRACTOR Background Data except for computer programs. CATEGORY 'D"- Consists of computer programs developed in the course of performing the Work. CATEGORY T- Consists of computer programs developed prior to the Effective Date or after the Effective Date but outside tho scope of the Work. CATEGORY 7*- Consists of administrative plans and administrative reports. l 9'
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a=ansm: List of ENectiva Fcges f /N U .. AP600 PRA List of Effective Pages - Volume 1 Page Revision Page Revision Page Revision. Page Revision
- 1. 8 2-9 8 2-32 7 2-55 7 1-2 8 2-10 7 2-33 7 2-56 7 1-3 8 2-11 7 2-34 7 2-57 7 1 -4 8 2-12 7 2-35 7 2-58 7 1-5 8 2-13 7 2-36 7 2-59 7 1-6 8 2-14 7 2-37 7 2-60 7 1-7 8 2-15 7 2-38 7 2-61 7 1-8 8 2-16 7 2-39 7 2A-1 7 1-9 8 2 17 7 2-40 7 2A-2 7 1-10 8 2-18 7 2-41 7 2A-3 7
(~'3 1-11 8 2-19 7 2-42 7 2A-4 7 z'; m l-12 8 2-20 8 2-43 7 2A-5 7 1-13 8 2-21 7 2-44 7 2A-6 7 1-14 8 2-22 7 2-45 7 2A 7 7 2-23 7 2-46 7 2A-8 7 2-1 7 2-24 7 2-47 7 2A-9 7 2-2 7 2-25 7 2-48 7 2A-10 7 2-3 7 2-26 7 2-49 7 2A-11 7 2-4 7 2-27 7 2-50 7 2A-12 7 2-5 8 2-28 7 2-51 7 2A-13 7 2-6 8 2-29 7 2-52 7 2A-14 7 2-7 8 2-30 7 2-53 7 2A-15 7 2-8 8 2-31 7 2-54 8 2A-16 7
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mans j List cf Effectiva Pages O AP600 PRA List of Effective Pages - Volume 1 Page Revision Page Revision Page Revision Page Revision 2A 17 7 2A-38 7 3-10 7 5-1 7 2A-18 7 2A-39 7 3-11 7 5-2 7 2A-19 7 2A-40 7 3-12 7 5-3 7 2A-20 7 2A-41 7 3-13 7 5-4 7 2A-21 7 2A-42 7 3-14 7 5-5 7 2A-22 7 2A-43 7 3-15 7 5-6 7 2A-23 8 2A-44 7 3-16 7 5-7 7 2A-24 7 2A-45 7 5-8 7 2A-25 7 2A-46 7 4-1 7 5-9 7 through 4-147 2A-26 7 2A-47 7 5-10 7 2A-27 7 4A-1 7 5-11 7 2A-28 7 4A-2 7 5-12 7 2A-29 7 3-1 7 4A-3 7 5-13 7 2A-30 7 3-2 7 4A-4 7 5-14 7 2A 31 7 3-3 7 4A-5 7 5-15 7 2A-32 7 3-4 7 4A-6 7 5-16 7 2A-33 7 3-5 7 4A-7 7 5-17 7 2A-34 7 3-6 7 4A-8 7 5-18 7 2A-35 7 3-7 7 4A-9 7 5-19 7 2A-36 7 3-8 9 5-20 7 2A-37 7 3-9 7 5-21 7 9 hla h 998 6 W85tingh0USe oNpolNwaWy.llkra-loe.wpf.lb 2
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U AP600 FRA List of Effective Pages - Volume 1 Page Revision Page Revision Page Revision Page Revision 5-22 7 6-8 9 7-17 2 7-38 2 5-23 7 6-9 7 7-18 2 7-39 2 through 6-44 5 24 7 6-45 9 7-19 2 7-40 2 5-25 7 6-46 7 7-20 2 7-41 2 through 6-147 5-26 7 7-21 2 7-42 2 5-27 7 7-1 2 7-22 2 5-28 7 7-2 2 7-23 2 8-1 7 5-29 7 7-3 2 7-24 2 8-2 7 f3 5-30 7 7-4 2 7-25 2 8-3 7 ( 5-31 7 7-5 2 7-26 2 8-4 7 1 5-32 7 7-6 2 7-27 2 8-5 7 5-33 7 7-7 2 7-28 2 8-6 7 5-34 7 78 2 7-29 2 8-7 7 7-9 2 7-30 2 8-8 7 6-1 7 7-10 2 7-31 2 8-9 7 6-2 7 7-11 2 7-32 2 8 10 7 6-3 7 7-12 2 7 33 2 8-11 7 6-4 7 7-13 2 7-34 2 8-12 7 6-5 7 7-14 2 7-35 2 8-13 7 6-6 7 7-15 2 7-36 2 8-14 7 6-7 7 7-16 2 7-37 2 8 15 7 _n
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List cf Effective Pages e AP600 PRA List of Effective Pages - Volume 1 Page Revision Page Revision Page Revision Page Revision 8-16 7 9-14 7 10-10 7 11-19 7 8 17 7 9-15 7 10-11 7 11-20 7 8 18 7 9-16 7 10-12 7 11-21 7 8 19 7 9-17 8 10-13 7 11-22 7 8-20 7 9-18 8 11-23 7 8-21 7 9-19 8 11-24 7 8-22 7 9-20 8 11-1 7 11-25 7 8 23 7 9-21 8 11-2 7 11-26 7 8-25 7 9-22 8 11-3 7 11-27 7 9-23 8 11-4 7 11-28 7 9-24 8 11-5 7 11-29 7 9-1 7 9-25 8 11-6 7 11-30 7 9-2 7 9-27 8 11-7 7 11-31 7 9-3 7 11-8 7 11-32 7 9-4 7 11-9 7 11-33 7 9-5 7 10-1 7 11-10 7 11-34 7 9-6 7 10-2 7 11-11 7 11-35 7 9-7 8 10-3 7 11-12 7 11-36 7 9-8 8 10-4 7 11-13 7 11-37 7 9-9 8 10-5 7 11-14 7 11-38 7 9-10 7 10-6 7 11-15 7 11-39 7 9-11 7 10-7 7 11-16 7 11-40 7 9-12 7 10-8 7 11-17 7 11-41 7 9-13 7 10-9 7 11-18 7 11-42 7 l Revision: 11 i March 1998 ENEL.
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List of Effective Pages () AP600 PRA List of Effective Pages - Volume 1 Page Revision Page Revislor. Page Revision Page Revision 11-43 7 12 21 7 13-5 2 14-15 2 11-44 7 12 22 8 13-6 2 14-16 2 12-23 8 13-7 2 14-17 3 12-1 7 12-24 8 13-8 2 14-18 2 12-2 7 12-25 7 13-9 2 14-19 2 12-3 7 12-26 7 13-10 2 14-20 2 12-4 7 12-27 7 13-11 2 14-21 2 12-5 7 12-28 7 13-12 2 14-22 2 12-6 7 12-29 7 14-23 2 12-7 7 12-30 7 14 1 2 14-25 2
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1 O i AP600 PRA i List of Effective Pages - Volume 1 Page Revision Page Revision Page Revision Page Revision 14-42 3 15-11 7 16-6 3 17-21 7 14-43 3 15-12 7 16-7 2 14-44 3 15 13 7 16-8 2 18-1 2 14-45 3 15-14 7 18-2 2 14-47 3 15-15 7 17-1 7 18-3 2 14-49 3 15-16 7 17-2 7 18-4 2 14-51 3 15-17 7 17-3 7 18-5 2 14-53 3 15-18 7 17-4 7 18-6 2 14-55 3 15-19 7 17-5 7 18-7 2 14-57 3 15-20 7 17-6 7 18-8 2 14-59 3 15-21 7 17-7 7 18-9 2 14-61 3 15-22 7 17-8 7 18-10 2 15-23 7 17-9 7 18-11 2 15-1 7 15-24 7 17-10 7 18-12 2 15-2 7 15-25 7 17-11 7 18-13 2 15-3 7 15-27 7 17-12 7 15-4 7 15-29 7 17-13 7 19-1 7 15-5 7 17-14 7 19-2 7 15-6 7 16-1 2 17-15 7 19-3 7 15-7 7 16-2 2 17-16 7 19-4 7 15-8 7 16-3 2 17-17 7 19-5 7 15-9 7 16-4 2 17-18 7 19-6 7 15-10 7 16-5 3 17-19 7 19-7 7 Revision: 11 O March 1998 Wd W Westingh00SS onip0lWrev_1151oe.wpf:Ib 6
List of Effective Pages ,/~ b AP600 PRA List of Effective Pages - Volume 1 Page Revision Page Revision Page Revisica Page Revision 19-8 7 21-23 2 21-46 7 19-9 7 21-1 2 21-24 2 21-47 7 19-10 7 21-2 2 21-25 2 21-48 7 19-11 7 21 3 2 21-26 2 21-49 7 19-12 7 21-4 2 21-27 2 21-50 7 19-13 7 21-5 7 21-28 2 21-51 7 19-14 7 21-6 7 21-29 2 21-52 7 15 7 21-7 3 21-30 2 21-53 7 19-17 7 21-8 3 21-31 2 21-54 7 21-9 2 21-32 2 21-55 7 m 20-1 2 21-10 2 21-33 7 21 56 7 20-2 2 21-11 2 21-34 7 21-57 7 20-3 2 21-12 2 21-35 7 21-58 7 20-4 2 21-13 2 21-36 7 21-59 7 20-5 2 21-14 2 21-37 7 21-61 7 l 20-6 2 21 15 2 21-38 7 21-63 7 , l 20-7 2 21-16 2 21-39 7 20-8 2 21 17 2 21-40 7 22 1 2 l 20-9 2 21 18 2 21-41 7 22-2 2 20-10 2 21 19 2 21-42 7 22-3 2 20-11 2 21-20 2 21-43 7 22-4 2 20-13 2 21 21 2 21-44 7 22-5 2 20-15 2 21 22 2 21-45 7 22-6 2 ENEL Revision: 11 MDM muh March 1998 7 oNp01MWv.llW-tw.wptib l 1
i List of Effectiv2 Pages e AP600 PRA List of Effective Pages Volume 1 Page Revision Page Revision Page Revision Page Revision 22-7 2 22-30 7 22-55 7 23-21 2 22-8 2 22-31 7 22-57 7 23-22 2 22 9 2 22-32 7 23-23 7 22-10 2 22 33 7 23-1 2 23-24 7 22-11 2 22-34 7 23-2 2 23-25 7 22-12 2 22-35 7 23-3 8 23-26 7 22-13 2 22 36 7 23-4 2 23-27 7 22 14 2 22-37 7 23-5 2 23-28 7 22-15 2 22 38 7 23-6 2 23-29 7 22-16 2 22 39 7 23-7 2 23-30 7 22-17 2 22-40 7 23-8 2 23-31 7 22-18 2 22-41 7 23-9 2 23-32 7 22-19 2 22-42 7 23-10 2 23-33 7 22-20 2 22-43 7 23-11 2 23-34 7 22-21 2 22-44 7 23-12 2 23-35 7 22-22 2 22-45 7 23-13 2 23-36 7 22-23 2 22-46 7 23-14 2 23-37 7 22-24 2 22-47 7 23-15 2 23-38 7 22-25 2 22-48 7 23-16 2 23-39 7 22-26 2 2249 7 23-17 2 23-40 7 22-27 2 22-50 7 23-18 2 23-41 7 22-28 2 22-51 7 23-19 2 23-42 7 22-29 7 22-53 7 23-20 2 23-43 7 Revision: 11 O March 1998 o:\ip0lWa\rev.llipra-loe.wpf:Ib 8 hhw 3 Westingh0Use
List of Effectiv2 Pages \_/ AP600 PRA List of Effective Pages - Volume 1 Page Revision Page Revision Page Revision Page Revision 23-45 7 24-22 2 23-47 7 24-23 2 23-48 7 24-24 2 23-49 7 24-25 2 23-50 7 24-26 2 24-27 2 24-1 2 24-28 2 24-2 2 24-29 2 24-3 2 24-30 2 24-4 2 24-31 2 /~N 24-5 2 24-32 2 24-6 2 24-7 2 24-9 2 j l 24-11 2 ' 24-13 2 24-15 2 24-16 2 24-17 2 24-18 2 24-19 2 24 4'0 2 24-21 2 ,.n .
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E3 List of Effective Pcges O AP600 PRA List of Eff'ective Pages - Volume 2 Page Revision Page Revision Page Revision Page Revision 25-1 2 26-1 7 26-21 10 27-12 7 25-2 2 26-2 10 26-22 10 27-13 7 25-3 2 26-3 10 26-23 10 25-4 2 26-4 10 26-24 10 28-1 7 through 28-140 25-5 2 26-5 10 26-25 10 25-6 2 26-6 10 26-26 10 25-7 2 26-7 10 26-27 10 29-1 7 25-8 2 26-8 10 26-28 7 29-2 l7 through 26-232' 25-9 2 26-9 10 29-3 7 25-10 2 26-10 10 27-1 7 29-4 7 25-11 2 26-11 10 27-2 7 29-5 7 25-12 2 26-12 10 27-3 7 29-6 7 25-13 2 26-13 10 27-4 7 29-7 7 25-14 2 26-14 10 27-5 7 29-8 7 25-15 2 26-15 10 27-6 7 29-9 7 25-16 2 26-16 10 27-7 7 29-10 7 25-17 2 26-17 10 27-8 7 29-11 7 25-18 2 26-18 10 27-9 7 29-12 7 25-19 2 26-19 10 27-10 7 29-13 7 26-20 10 27-11 7 29-14 7 O h1a 1998 oNpoIWwv.1151oe.wpt.ib W W Westkigh00S8 10
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-v AP600 PRA List of Effective Pages - Volume 2 Page Revision Page Revision Page Revision Page Revision 29-15 7 30A-11 2 31-14 7 32-18 7 29-16 7 30A-12 2 31-15 7 32-19 7 29-17 7 30A-13 2 31-16 7 32-20 10 29 18 7 30A-14 2 31-17 7 32-21 7 29-19 7 30A-15 2 32-22 7 29-20 7 30A-16 2 32-1 7 32-23 7 29-21 7 30A-17 2 32-2 7 32-24 7 29-22 7 30A-18 2 32-3 7 32-25 7 29-23 7 32-4 7 32-26 7 31-1 7 32-5 7 32-27 7 rw 7 31-2 7 32-6 7 32-28 7 (j. 30-1 through 30-115 31 3 7 32-7 7 32-29 7 30A-1 2 31-4 7 32-8 7 32-30 7 30A 2 2 31-5 7 32-9 7 32-31 7 30A-3 2 31-6 7 32-10 7 32 32 7 30A-4 2 31-7 7 32-11 7 32-33 7 30A-5 2 31-8 7 32-12 7 32-34 7 j l'
30A-6 2 31-9 7 32-13 10 32-35 7 30A-7 2 31-10 7 32-14 7 32-36 7 30A-8 2 31-11 7 32-15 7 32-37 7 30A-9 2 31-12 7 32 16 7 32-38 7 30A-10 2 31-13 7 32-17 7 32-39 7 i
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List cf Effectiv2 Pages e AP600 PRA List of Effective Pages Volume 2 Page Revision Page Revision Page Revision Page Revision 32-40 7 35-1 8 35-20 8 36-1 8 3241 7 35-2 8 35-21 8 36-2 8 32-42 7 35-3 8 35-22 8 36-3 8 32-43 7 35-4 8 35-23 8 36-4 8 32-44 7 35-5 8 35-24 8 36-5 8 32-45 7 35-6 8 35-25 8 36-6 9 32-46 7 35-7 8 35-26 8 36-7 10 32-47 7 35-8 8 35-27 8 36-8 8 35-9 8 35-28 8 36-9 8 33-1 7 35-10 8 35-29 10 36-10 $ through 33-17 33-18 8 35-11 8 35-30 8 33-19 7 35-12 8 through 33-66 35-13 8 34-1 8 35-14 10 l through 34-487 35-15 8 35-16 8 35-17 8 35-18 8 35 19 8
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List of Effective Pages ,O V , AP600 PIM List of E&ctive Pages Volume 3 Page Revision Page Revision P. age Revision Page Revision . 37 1 8 39-6 8 39-25 8 42-1 8 37-2 8 39-7 8 42-2 8 37-3 10 39-8 8 40-1 8 42-3 8 37-4 8 39-9 8 40-2 11 42-4 8 37-5 10 39-10 8 40-3 8 42 5 8 39 11 8 40-4 11 42-6 8 38 1 8 39-12 8 40-5 8 42-7 8 38 2 8 39-13 8 42-8 8 38-3 8 39-14 11 41-1 8 42-9 8 38-4 10 39-15 8 41-2 8 42-10 8 /O 38-5 10 39-16 8 41-3 8 42-11 8 38-6 10 39-17 8 41-4 8 42-12 8 38-7 10 39-18 8 41-5 10 42-13 8 39-19 8 41-6 8 42-14 8 through 41-123 39-1 8 39-20 8 39-2 8 39-21 8 41 A 1 8 43-1 8 through through 41 A-375 43-162 39-3 11 39-22 8 41B-1 8 through 41B-120 39-4 8 39-23 8 __ 39-5 10 39-24 8 ENE Revision: 11 3 Westinghouse mmt, March 1998 13 onipoivr=wv 115ra-loe.wpf:Ib
e-: =- 5 List of Effective Pages er O\ AP600 PRA List of Effective Pages - Volume 3 Page Revision Page Revision Page Revision Page Revision 44-1 5 49-15 8 44-2 8 46-1 8 49-16 8 44-3 8 49-17 8 44-4 8 47-1 8 49-18 8 44-5 8 49-19 9 44-6 8 48-1 8 49-20 8 through 49-47 44-7 8 49-48 11 44-8 8 49-1 8 44-9 8 49 2 8 44-10 8 49-3 8 44-11 8 49-4 8 44-12 9 49-5 8 49-6 8 45-1 8 49-7 8 45-2 8 49-8 9 45-3 8 49-9 8 45-4 8 49-10 8 45-5 8 49-11 8 45-6 8 49-12 8 45-7 9 49-13 8 45-8 8 49-14 8 through 45-136 O Revision: 11 ENEL March 1998 ' a -- W Westingh0USB oNp01\prahv.,Ilipra-lotypf.lb 14
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AP600 PRA List of Effective Pages - Volume 4 Page Revision Page Revision Page Revision Page Revision 50-1 8 54-10 8 54C-1 11 56-1 8 through through and through 50-76 54-32 54C-4 56-11 54-33 9 50A- 1 11 54-34 8 55-1 9 56-12 5 through through through through 50A-3 54-39 55-6 56-111 54-40 9 55-7 10 51-1 8 54-41 8 55-8 9 through through through 51-21 54-97 55-72 54-98 9 55-73 10 g 52-1 8 54-99 9 55-74 9 57-1 9 i f through through through
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AP600 PRA List of Effective Pages - Volume 4 , Page Revision Page Revision Page Revision Page Revision 59 1 8 59-104 8 D-1 11 through through through 59-36 59-203 D-108 59-37 9 59-204 11 59-38 9 59-205 8 59 39 8 59-206 8 59-40 8 59-207 11 through 59-236 59-41 9 59-42 8 A-1 2 through through 59-68 A-296 , 59-69 11 59-70 8 through 59-74 59-75 11 B-1 11 through througii 59-82 B-36 59-83 8 through 59-86 59-87 9 59-88 8 C-1 9 through th:ough 59-100 C-10
$9-101 9 C-11 11 59-102 9 59-103 11 O
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i O i ) V TABLE OF CONTENTS (Cont.) Section 11111 EBER CHAPTERS 46 THROUGH 48 DELETED CHAPTER 49 OFFSITE DOSE EVALUATION 49.1 Introduction ............................................. . 49-1 49.2 Conformance with Regulatory Requirements . . . . . . . . . . . . . . . . . . . . . . . . . 49-1 49.3 Assu mptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 9-2 49.4 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 9-2 49.5 Dose Evaluation Results and Discussions . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-6 49.6 Quantification of Site Risk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-7 49.7 Risk Quantification Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-7 49.8 References ................................................ 49-8 CHAlrTER 50 IMPORTANCE AND SENSITIVITY ANALYSIS 50.1 Introduction ...............................................50-1 50.2 Importance Analyses for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 501 50.2.1 Initiating Event Importances (Case 1) . . . . . . . . . . . . . . . . . . . . . . 5 0-2 50.2.2 Common Cause Failure Importances (Case 2) . . . . . . . . . . . . . .. . 50-3 p 50.2.3 Human Error Importances (Case 3) . . . . . . . . . . . . . . . . . . . . . . . . 50-5 V 50.2.4 Component Importances (Case 4) . . . . . . . . . . . . . . . . . . . . . . . . . . 50-6 50.3 System Importances for Core Damage .............................50-7 50.4 Human Error Sensitivity Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-9 50.4.1 Set Human Error Probabilities to 1.0 (Failure) in Core Damage Results (Case 25) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-9 50.4.2 Set Human Error Probabilities to 0.0 (Success) in Core Damage Results (Case 26) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50- 10 ; 50.4.3 Assess Importance of Increasing Human Error Probabilities by a Factor of 10 (Case 27) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-10 50.5 Other Sensitivity Analyses for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . 50-11 1 50.5.1 Diesel Generator Mission Time (Case 28) . . . . . . . . . . . . . . . . . . 50- 1 1 50.5.2 Impact of Passive System Check Valves on Core Damage Frequency (Case 29) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-12 50.5.3 Instmmentation and Control Cutoff Probability (Case 30) . . . . . . . . 50-12 50.5.4 Containment Rectreulation After Safety Injection Line Break Event (Case 3 1 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0- 12 50.5.5 Quantification Truncation Probability (Case 32) . . . . . . . . . . . . . . . 50-13 50.5.6 Sensitivity to ADS Stage 4 Success Criteria (Case 33) . . . . . . . . . . 50-13 50.5.7 Squib Valve Failure Probability (Case 34) . . . . . . . . . . . . . . . . . . . 50-13 50.5.8 Circuit Breaker Failure Probability (Case 35) . . . . . . . . . . . . . . . . . 50-14 50.5.9 End-State Importances (Case 36) . . . . . . . . . . . . . . . . . . . . . . . . . 5 0- 14 m) ( N/ r.rm. April l xxxl oAap600$ra\rev.9 pre-toc.wpf.lb i 1
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O TABLE OF CONTENTS (Cont.) Section Title Page 50.6 Sensitivity and Importance Analyses For Large Release Frequency . . . . . . . . 50- 15 50.6.1 Importance Analyses For Large Release Frequency . . . . . . . . . . . . 50- 15 50.6.2 Sensitivity Analyses For Large Release Frequency . . . . . . . . . . . . . 50-21
$0.7 Sensitivity Analysis for Offsite Dose Risk . . . . . . . . ....... . . . . . . . . . 50-22 50.8 Results Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-23 Attachment SOA ATWS PRA Sensitivity Case . .. ........ ....... ...... 50A-1 CHAPTER 51 UNCERTAINTY ANALYSIS 51.1 Introduction ............................. ......... ....... 51-1 51.2 Methodology . . . . . . . . . . . . . ................. ..... ......... 51-1 51.3 Summary of Results . . . . . . . . . . . .............. . ............ . 51-3 51.4 Sensitivity Studies for the Uncertainty Calculations ........ . . . . . . . . . . . 51 -4 51.4.1 Uncenainty in the Cutoff Frequency . . . . . . . . . . . ........ ... 51-4 51.4.2 Uncertainty in the Number of Cutsets Sampled . . . . . . . . . . . . . . . . 51-4 51.4.3 Uncertainty in the Mean Failure Probability for Basic Events . . . . . . . 51-4 51.4.4 Sensitivity to the Random Number Input for Sampling . . . . . . . . . . .51-5 51.5 Re ferences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... ........... 51-6 l
CHAPTER 52 RTNSS - FOCUSED PRA SENSITIVITY STUDY 52.1 Focused PRA Sensitivity Study Analysis Method . . . . . . . . . . . . . . . . . . . . . . 52-1 52.1.1 Core Damage Frequency Calculation . . . . . . . . . . . . . . . . . . . . . 5 2 -2 52.1.2 Release Frequency Calculation . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2-5 52.2 At-Power Focused PRA Sensitivity Study . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2-5 52.2.1 At-Power Focused PRA Sensitivity Study Core Damage Frequency Quantification . ....... ..... ............... 52-6 52.2.2 At-Power Focused PRA Sensitivity Study Release Frequency Quantification .. ............. ...... . . . . 52-11 52.3 Shutdown Focused PRA Sensitivity Study . . .......................52-16 52.3.1 Shutdown Focused PRA Sensitivity Study Core Damage Quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2- 1 6 52.3.2 Shutdown Focused PRA Sensitivity Study Release Frequency Calculation .... .............. . , . . . . 52-19 52.4 Focused PRA Sensitivity Fire Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2-21 52.5 Focused PRA Sensitivity Study Flooding Analysis . . . . . . . . . . . . . . . . . . . . 52-21 52.5.1 At-Power Focused PRA Sensitivity Study Flooding Scenarios . . . . . 52 21 52.5.2 Shutdown Focused PRA Sensitivity Study Flooding Scenarios . . . . . 52-22 52.5.3 Focused PRA Sensitivity Study Flooding Analysis Results S u m mary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2-23 52.6 Focused PRA Sensitivity Study Results and Conclusions . . ... . ....... 52-23 52.7 References .................................... . .. ..... 52-24 O Revision: 11 ENEL March 1998 wxas T Westinghouse 04ra\revj!4ra-toc.wpf:lb xxxii w.
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-CHAPTER 53 DELETED CHAITER 54 LOW-POWER AND SHUTDOWN RISK ASSESSMENT 54.1 Introduction .... ........................................54-1 54.2 Initiating Events ............................................ 54-1 54.2.1 Identification . . . . . . . .... .................... ..... 54-2 54.2.2 Events Modeled . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-2 54.2.3 Shutdown Phases Summary Description . . . . . . . . . . . . . . . ..... 54-3 54.2.4 Initiating Events for Operaung Modes . . . . . . . . . . . . . . . . . . . . . . . 54-4 54.2.5 Actuanng Signals and Systems Available ................... 54-15 54.2.6 Scenarios for Detailed Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . 54-16 54.1.7 Summary of Initiating Events Analyzed . . . . . . . . . . . . . . . . . . . . 54-21 54.3 Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-22 54.3.1 Shutdown Frequency . . . . . . . . . . . . . . . . . . . . ..... . .... 54-22 j 54.3.2 Mission Times . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-25 54.4 Event Tree Development . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-28 54.4.1 Event Tree LOSP-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-30 *q r 54.4.2 Event Tree RNS-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4-3 3 V
54.4.3 Event Tree CCW-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-3 3 54.4.4 Event Tree LOCA-PR-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-33 54.4.5 Event Tree LOCA-V24-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-34 54.4.6 Event Tree RCS-OD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-36 54.4.7 Event Tree LOSP-D . ................................54-38 54.4.8 Event Tree RNS-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.9 Event Tree CCW-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.10 Event Tree LOCA-V24-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.11 Boron Dilution Events (Reactivity Events) . . . . . . . . . . . . . . . . . . . 54-41 54.4.12 Boron Dilution Events Due to Chemical and Volume Control System Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-45 54.4.13 Endstates Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-48 54.5 Fault Tree Models for Shutdown and Low-Power Events . . . . . . . . . . . . . . . 54 48 54.5.1 Insuumentation and Control Modeling for Shutdown (Level 1) .... 54-48 54.5.2 Instrumentation and Control Modeling for Shutdown (Level 2) .... 54-51 54.6 Success Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-51 54.6.1 MAAP4 Code Analysis for Shutdown Success Criteria . . . . . . . . . . 54-52 ! 54.6.2 ~ MAAP4 Pammeter File . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-52 54.6.3 MAAP4 Input Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-54
'54.6.4 Definition of MAAP4 Cases From Event Trws . . . . . . . . . . . . . . . 54-55 54.6.5 Results From MAAP4 Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . 54-57 n r 54.7 Common Cause Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-57 i
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w = 1M11 O TABLE OF CONTENTS (Cont.) Section Title }_' age 54.8 Human Reliability Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-57 54.8.1 Operator Actions Calculated . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-5 8 54.8.2 Conditional Human Error Probabilities . . . . . . . . . . . . ...... . 54-64 54.9 Fault Tree Quantification ....................... ..... ...... . 54-64 54.10 Level 1 Core Damage Frequency Quantification . . . . . . . . . . . . . . . . . .... 54-67 54.10.1 Core Damage Quantification Method . . . . . . . . . . . . . . . . . . . . . 54-68 54.10.2 Quantification Inputs . . . . . . .......................... 54-69 54.10.3 Level 1 Shutdown Core Damage Frequency Results . . . . . . . . . . . . 54-70 54.11 Shutdown and Low-Power Release Category Quantification . . . . . . . . . . . . . . 54-71 54.12 Shutdown Assessment Importance and Sensitivity Analyses . . . . . . . . . . . . . . 54-71 54.12.1 importance Analyses for Core Damage at Shutdown . . . . . . . . . . . . 54-72 54.12.2 Other Sensitivity Analyses for Shutdown Core Damage . . . . . . . . . 54-77 54.13 Summary of Shutdown Izvel 1 Results ........................... 54-81 54.14 Re ferences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........ 54-87 Attachment 54A Design Change Impact on Low-Power and Shutdown Risk Assessment ............... ................... 54A-1 Attachment 54B Surge Line Flooding Effect on Low-Power and Shutdown , Risk Assessment .. ............. ......... ....... 54B-1 Attachment 54C Effect of Modifications to Safe / Cold Shutdown PRA . . . . . . .. 54C-1 CHAPTER 55 SEISMIC MARGIN ANALYSIS . . . . . . . . .. ........... . . 55-1 55.1 Introduction ........................ ...................... 55-1 55.2 Calculation of HCLPF Values . ... ...... .. ............ ... . 55-2 55.2.1 Seismic Margin HCLPF Methodology . ... ............ ... 55-2 55.2.2 Calculation of HCLPF Values . . . . . . . . . . . . . . . . . . . . . ...... 55-2 55.3 Seismic Margin Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5- 12 55.3.1 SMA Model and Assumptions . . . . . . . . . . . . . . . . . . . . ...... 55-14 55.3.2 Seismic Initiating Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-16 55.3.3 Initiating Event Category HCLPFs . . . . ................. . 55-17 55.3.4 Event Tree Models . . . . . . . . . . . . . . .................. . 55-20 55.3.5 Fault Tree Modeling and Quantification . . . . . . . . . . . . . . ..... 55-28 55.3.6 Seismic Event Core Damage Sequence Evaluation . . . . . ....... 55-36 55.3.7 Containment Performance Model . . . . . . . . . . . . . . . . . . . . .... 55-37 55.4 Calculation of Sequence and Plant HCLPF . . . . . . . . ............... . 55-38 55.4.1 HCLPFs for Easic Events . . . . . . . . . . . . . . . . .... . . . . . . 5 5-39 55.4.2 Calculation of Initiating Event HCLPFs . . . . . . . . . . . . . . . . . . . 55-39 i 55.4.3 Calculation of System Fault Tree HCLPFs ....... .......... 55-39 55.4.4 Calculation of Sequence HCLPFs . . ........ ............ 55-40 55.4.5 Calculation of Plant HCLPF . . . . . . . . . ..... . . . . . . . . . . . . 5 5-4 3 55.4.6 Large Release HCLPF . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-43 Revision: ' 11 e ENEL March 1998 mg_ [ WestighollS8 o$prsWy.,mpra-toc.wpf:Ib xxxiv
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\ l TABLE OF CONTENTS (Cont.) Section Title _Page 55.5 Sensitivity Analyses . . . . . . . . . . . . . .. ... ..... ... . .... ' 5-46 5 55.5.1 Robust Fuel and Core Assembly . .. ..... .... .. .. . 55-48 55.5.2 Credit for Operator Actions . .. . ......... . . . 55-49 55.5.3 Less Credit for Operator Actions in LOSP Event at 0.09g ... . . . 55-52 55.5.4 72-Hour Mission Time . . .... ..... ... . . . . . . . 5 5-53 55.5.5 Containment Isolation - Smaller Size Valves . . . . . . . . . . . . . 55-56 55.5.6 Steam Generator Tube Rupture Success Criteria . . . . . .. .. .. 55-57 55.5.7 Steam Line Break Suecess Criteria . ..... . ... . . . . . . 55-5 8 55.5.8 Seismic Interaction Between Turbine and Auxiliary Buildings . .. 55-59 55.6 SMA Results and Insights . . . . . . . .. ...... ... . .. . 55-63 55.9.6 AP600 SMA Results ... . ... ........ .. .. .. . 55-63 55.9.6 AP600 SMA Insights . . ... . .. .... ........ .... 55-68 55.7 References ..... ..... .. ....... ... . ........ . .. . 55-70 Attachment 55A System HCLPF Calculations . . .. .. ... . ...... .. 55A-1 Attachment SSB Sequence HCLPF Calculations . ... .. .. . ... . . 55B-1 n Attachment 55C Seismic Margin Analysis HCLPF Sensitivity Study . ... 55C-1 CHAPTER 56 PRA INTERNAL FLOODING ANALYSIS 56.1 Introduction . ... .... .... ..... .............. .. . 56-1 56.1.1 Definitions . ..... . ..... .. . . ..... . . . . 56-1 56.2 Methodoloi,y . ........... .. . .. .............. ..... . . 56-1 56.2.1 Summary of Methodology . . . . . ............ .. . . . 56- 1 56.2.2 information Collection . . . . . . ... .. .... ... .... .... 56-2 56.2.3 Initial Screening Assessment ... ......... ..... . . . . . 56-3 56.2.4 Detailed Screening Assessment . . ........... .. . . . . . 56-4 56.2.5 Identification of Flood-Induced Initiating Events .... . .... . . 56-6 56.2.6 Initiating Event Frequencies ............. .. ......... . 56-7 56.3 Assumptions . . . . . . . . . . . . . . . . . . . . ................ . . . . . 5 6-7 56.3.1 General Flooding Analysis Assumptions and Engineering Judgments . 56-7 56.3.2 AP600-Specific Assumptions . . . . . . . . . . . . .. . . . . . . . . . 5 6-9 56.4 Information Collection . . . . . . . . . .. ... . . . ... .. .. 56-11 56.4.1 PRA-Modeled Equipment and Locations . . . . . . . . . . . .... . . 56-11 56.4.2 Identification of Areas for Flooding Evaluation .... ... .. .. 56-11 56.5 At-Power Operations . ........ .. . .. ............. .. . . . 56-12 56.5.1 Initial Screening Assessment ... ....... . ....... . . 56-12 56.5.2 Detailed Screening Assessment ............. ...... . . . . . 56-12 56.5.3 Identification of Flood-Induced Initiating Events .... . .... . 56-28 56.5.4 Calculation of Flood-Induced Initiating Event Frequencies . .... 56-32 56.5.5 Quantification of At-Power Flood-Induced Events ......... . . 56-39 m a
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O TABLE OF CONTENTS (Cont.) Section I!!)e P. age 56.6 Shutdown Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-41 56.6.1 Detailed Screening Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6-41 56.6.2 Identification of Flood. Induced Initiating Events . . . . . . . . . . . . . . 56-42 56.6.3 Calculation of Flood. Induced Initiating Event Frequencies . . . . . . . . 56-43 56.6.4 Shutdown Quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 48 56.7 Seismically Induced Flooding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 51 56.8 Flooding Hazards During Refueling Outages . . . . . . . . . . . . . . . . . . . . . . . . 56-52 56.9 Flooding Sensitivity Study .... ...............................56-52 56.9.1 Flooding Human Error Probabilities Sensitivity Study . . . . . . . . . . . 56 52 56.10 Summary of Findings . . . . . . . ................................5653 CHAFTER 57 INTERNAL FIRE AN#1YSIS 57.1 Introduction ............................................ . 57-1 57.2 Qualitative Analysis Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 2 57.3 Quantitative Methodology of Fire Area Frequency . . . . . . . . . . . . . . . . . . . . . 57 6 57.3.1 Fire Frequency Calculations . . . . . . . . . . . . . . . . . . . . . . . .. . . 57-6 57.3.2 Fire Damage Category Quantification . . . . . . . . . . . . . . . . . . . . . . . 57-7 57.4 Core Damage Quantification Methodology . . . . . . . . . . . . . . . . . . . . . . . . . 5 7 10 57.5 Fire Analysis Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-12 57.5.1 Qualitative Analysis Assumptions and Other Modeling Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 12 57.5.2 Quantification Assumptions And Modeling Considerations . . . . . . . 57-14 57.6 At. Power Qualitative Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-17 57.7 At-Power Quantitative Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 19 57.7.1 Fire Ignition Frequencies for Qunntitative Analysis . . . . . . . . . . . . 57-19 57.7.2 Fire Damage Category Quantification . . . . . . . . . . . . . . . . . . . . . . 5719 57.7.3 Individual Area PRA Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 19 1 57.8 Control Room Fire Analysis - Power Operation . . . . . . . . . . . . . . . . . . . . . 5 7 22 57.8.1 Description of the Control Room and AsWA Fire Protection . . . 57 22 57.8.2 Alternate Shutdown Capability . . . . . . . . . . . . . . . . . . . . . . . . . . 57 23 57.8.3 Fire Hazard Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 7-24 57.8.4 AP600 Control Room Fire Evaluation . . . . . . . . . . . . . . . . . . . . . . 57-25 57.8.5 Fire Scenario Identification and Frequency Determination . . . . . . . . , 57 29 57.8.6 Control Room Fire Scenario Quantification and Results . . . . . . . . . 57 32 1 57.9 Shutdown Fire Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-33 l 57.9.1 Fire Ignition Frequencies during Shutdown Modes of Operation . . . . 57-33 { 57.9.2 Fire Damage Category Quantification . . . . . . . . . . . . . . . . . . . . . . 57-34 i 57.9.3 Individual Area PRA Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 34 57.9.4 Fire Analysis for Safe Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . 57-34 57.9.5 Fire Analysis for Mid-Loop Cperation . . . . . . . . . . . . . . . . . . . . . 5 7-41 e Revision: 9 April 11,1997 W85tiligh0llS8 oAap60oprs\rev 9Wtoe.wpf.lb xxxvi
I i g\ bl TABLE OF CONTENTS (Cont.) Section Title _P_ age 57.10 Summary and Conclusions . . . . . . . . . .. . ....... .. ... . . . . 57-42 57.10.1 At-Power Analysis . . . . ...... ... .. .. . .. . 57-42 57.10.2 Shutdown Fire Analysis . . ..... .... .. ... .... . 57 45 57.10.3 Conclusions . . . . . . . . . .. ..... ................. .. 57-47 57.11 References .. . .... ..... .. . ... ... . .. . ...... . 57-48 ATTACHMENT 57A DEFINITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57A-1 ATTACHMENT 57B DESIGN CHANGE EFFECT ON INTERNAL FIRE ANALYSIS . . 57B-1 CHAPTER 58 WINDS, FLOODS, AND OTHER EXTERNAL EVENTS 58.1 Introduction ..... . .................. .... ....... ... . 58-1 58.2 Extemal Events Analysis . .......... .......... ... .. 58-1 58.2.1 Severe Winds and Tomadoes . . . . ..... ........ . . . . . . . . . 5 8- 1 58.2.2 External Floods .. ... ........ ........ . ... . . . . . 5 8-2 58.2.3 Tr=nsportation and Nearby Facility Accidents .. .. ... . . . 58-2 58.3 Conclusion . . . . . . .. ......... ...... ......... . . ... 58-3 58.4 References ... . . .......... . .. .. . .. .. ..... . 58-3 m CHAPTER 59 PRA RESULTS AND INSIGHTS (V) 59.1 Introduction . . ....... ..... .. ... .. . . . . 59-1 59.2 Use of PRA in the Design Process . .. .. . ..... . ........ . 59-3 59.2.1 Stage 1 - Use of PRA During the Early Design Stage . . . . .... 59-4 j
$9.2.2 Stage 2 - Preliminary PRA . . . . . ....... ... .. . . 59-5 l 59.2.3 Stage 3 - AP600 PRA Submittal to NRC (1992) .. . . . . . 59-7 59.2.4 Stage 4 - PRA Revision 1 (1994) . . . ... .. . . ... .. . . 59-8 59.2.5 Stage 5 - PRA Revisions 2-8 (1995-1996) . . ...... ...... . 59-8 i 59.3 Core Damage Frequency from Internal Initiating Events at Power . . . . . . . . 5 9- 10 !
59.3.1 Dominant Core Damage Sequences .... . . .... . . . 59-12 l 59.3.2 Component Importances for At-Power Core Damage Frequency . . 59-44 ) 59.3.3 System Importances for At-Power Core Damage . . . .... . 59-44 ' 59.3.4 System Failure Probabilities for At-Power Core Damage . . . . . . . 5 9-4 5 59.3.5 Common Cause Failure Importances for At-Power Core Damage . 59-45 59.3.6 Human Error Importances for At-Power Core Damage ... . 59-45 59.3.7 Accident Class Importances . . . .. . . . . . ........... . . . . . . . 5 9-4 7 59.3.8 Sensitivity Analyses Summary for At-Power Core Damage . . . . . . 59-47 59.3.9 Summary of Important Level 1 At-Power Results . . . . . . .. . . 59-48 59.4 Large Release Frequency for Internal Initiating Events at Power . . ... .. . 59-51 59.4.1 Dominant Large Release Frequency Sequences ......... . . . . 59-52 59.4.2 Sensitivity Analyses for Containment Response . . . . ... ... .. 59-72 l l l
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O TABLE OF CONTENTS (Cont.) Section Title Page 59.4.3 Comparison of Initiating Event Importances for Core Damage Frequency and Large Release Frequency . . . . . . . . . . ...... 59-72 59.4.4 Summary of Important Level 2 At-Power Results . . . . . . . . . . . . . 59-73 59.5 Core Damage and Severe Release Frequency from Events at S hutdo wn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 59-75 59.5.1 Summary of Shutdown Level 1 Results . . . . . . . . . . . . . . . . . . . . . 59-75 59.5.2 Large Release Frequency for Shutdown and Low-Power Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 9- 81 59.5.3 Shutdown Resub Summary . . ............. . . . . . . . . . 59-82 59.6 Results from Intemal Flooding, Intemal Fire, and Seismic Margin Analyses . . . . . . . ........ ....... ........ ............... 59-82 59.6.1 Results of Intemal Flooding Assessment .. . . ...... .. 59-82 59.6.2 Results of 7'temal Fire Assessment ...................... 59-83 59.6.3 Results of seismic Margin Analysis . . . . . . . . .............. 59-87 59.7 Plant Dose Risk from Release of Fission Products . . . . ....... ..... 59-87 59.8 Overall Plant Risk Results . . . . . . . . . . ........... .. ... . . . . . . 59-88 59.9 Plant Features Important to Reducing Risk . .................. ... . 59-89 59.9.1 keactor Design .......... . .... .. ...... ... .... 59-90 59.9.2 Systems Design . . . . . . . . . . . . ............ .. ........ 59-91 59.9.3 Instrumentation and Control Design . . . . . . ..... . . . . . . . 59-94 59.9.4 Plant Layout .. . . .. ...... ... .... . .... 59-95 59.9.5 Plant Structures . . . . . .. ... .... ... . ....... . 59-96 59.9.6 Containment Design . . . .. .. ............. ....... . 59-96 59.10 PRA Input to the Design Certification Process . .. ............ .. 59-101 59.10.1 PRA Input to Reliability Assurance Program . . . . . . . . . 59-102 59.10.2 PRA Input to Certified Design Material ........... . . . . . . 59-102 59.10.3 PRA Input to the Technical Specifications . . .... ..... . .. 59-102 59.10.4 PRA Input to MMI/ Human Factors / Emergency Response Guidelines . . . . . . . ...... ......... . .. .... . . . . 59-102 59.10.5 Summary of PRA-Based Insights .... ..... .. .. ... 59-103 59.10.6 Combined License Information . . . . . . . . . ...... . . . . . . . . 59-103 APPENDIX A MAAP4 ANALYSIS TO SUPPORT SUCCESS CRITERIA .. ... .. A-1 APPENDIX B EX-VESSEL SEVERE ACCIDENT PHENOMENA ..... ... .... . . . . B- 1 APPENDIX C DESIGN CHANGES THAT OCCURRED AFTER THE PRA ANALYSES WERE COMPLETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . C- 1 APPENDIX D EQUIPMENT SURVIVABILITY ASSESSMENT . . . . . ... . . . . . . . . . . . . D- 1 O Revision: 11 ENEL [ W85tingh00S8 March 1998 mg_ oNravev liwa. toe.wpr.ib xxxviii
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G LIST OF TABLES (Cont.) Table No. Title P., age 57-25 Quantitative Summary - Control Room Fires Dunng Safe Shutdown ... ................ .. ................ .. 57-138 57-26 Summary of Qualitative Evaluation Results - Mid-Loop Operation . . . . . . . . . . . . . . . . . . . . ... ....... ... . . 57-139 57-27 Summary of Quantitative Results - Mid-Loop Operation . . . . . . . . . . . . . 57-146 57-28 Quantitative Summary - Control Room Fires During Mid-Loop Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 57-154 57-29 Safe Shutdown Containment Single-Hot-Shnrt LOCA Sensitivity . . . . . . . . 57-155 59-1 Contnbution of Initiating Events to Core Damage . . . . . . . . . . . . .. . 59-104 59-2 Conditional Core Damage Probability of Initiatmg Events . . . . . . . . . . . . . 59-105 59-3 Internal Initiating Events at Power Dominant Core Damage Sequences . .. 59-106 59-4 Sequence 1 - Safety Injection Line Break Dominant Cutsets (SI-LB-02) . . . . 59-108 59-5 Sequence 2 - Intermediate LOCA Dominant Cutsets (NLOCA-03) ... ... 59-113 59-6 Sequence 3 - Large LOCA Dominant Cutsets (LLOCA-06) . . . . . -. . . . . . . 59-118 59-7 Sequence 4 - Large LOCA Dominant Cutsets (LLOCA-03) . . . . . . . . . . . 59-124 59-8 Sequence 5 - Reactor Vessel Rupture Cutset (RV-RP-02) . . . . . . . . . 59-130 em 59-9 Sequence 6 - Large LOCA Dominant Cutsets (LLOCA-11) . . ..... . . 59-131 ( ) 59-10 Sequence / - ATWS Dominant Cutsets (ATWS-28) . . .. ....... . .. 59-133 59-11 Sequence 8 - Medium LOCA Dominant Cutsets (MLOCA-03) .. .. 59-141 59-12 Sequence 9 - ATWS Dominant Cm: K (ATWS-13) . . . . .... . .. .. 59-146 59-13 Sequence 10 - Intermediate LOCA Do.ninant Cutsets (NLOCA-04) . .. 59-151 59-14 Sequence 11 - Safety Injection Line Break Dominant Cutsets (SI-LB 03) . 59-156 59-15 Sequence 12 - Small LOCA Dominant Cutts (SLOCA-03) . . ........ . 59-160 59-16 Sequence 13 - Core Makeup Tank Line Break Dominant Cutsets (CMTLB-03) 59-165 59-17 Sequence 14 - Steam Generator Tube Rupture Dominant Cutsets (SGTR-07) . 59-170 59-18 Sequence 15 - Steam Generator Tube Rupture Dominant Cutsets (SGTR-23) . 59-171 59-19 Sequence 16 - Large LOCA Dominant Cutsets (LLOCA-02) . . . . . . . . . . . . 59-177 59-20 Sequence 17 - Large LOCA Dominant Cutsets (LLOCA-05) . . . . . . . ... 59-183 59-21 Sequence 18 - Consequential SGTR Dominant Cutsets (SGTRC-03) . . . . . 59-189 59-22 Sequence 19 - Intermediate LOCA Dominant Cutsets (NLOCA-16) . . . .. 59-195 59-23 Typical System Failure Probabilities, Showing Higher Reliabilities for Safety Systems ................... ..... .... ....... . 59-201 59-24 Dominant CET Sequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 9-202 59-25 Comparison of Initiating Event Contribution to Core Damage and Large Release Frequencies ................... ......... .. .. 59-203 59-26 Summary of AP600 PRA Results . . . . . . . . . . . . . ........... . . . . . 59-204 59-27 Comparison of AP600 PRA Results to Risk Goals . . . . . . . . . . . . . . . . . . . 59-205 59-28 Site Boundary Dose Risk at 24 Hours ............ ............ . 59-206 59 AP600 PRA-Based Insights . . . . . . . . . . . . .. ................. . 59-207 !q i J ENE Revision: 11 UDI88 Eta:6. March 1998 liX o:\ap6ompra\rev.11pa-lot.wpf:Ib
namens== O LIST OF FIGURES Figure No. Title Page 2-1 Core Damage Logic Diagram for Internal Initiators . . . . . . . . . .... . . . . 2-5 8 2-2 Core Damage Logic Diagram for Internal Initiators . . . . . . . . . . . . . . . . . . . 2-59 2-3 Core Damage Logic Diagram for Internal Initiators ............... .. . 2-60 2-4 Core Damage Logic Diagram for Internal Initiators . . . . . . . ...... ..... 2-61 4-1 Large Loss-of-Coolant Accident Event Tree ... .......... . . . . . . . . . 4- 120 4-2 Medium Loss-of-Coolant Accident Event Tree . . . . . . . . . .... . . . . . . . . 4-121 4-3 Core Makeup Tank Line Break Event Tree . . . . . . . . . . . . . . . . . . . .. 4-122 4-4 Direct Vessel Injection Line Break Event Tree . . .. . .... . ... . . . 4-123 4-5 Intermediate Loss-of-Coolant Accident Event Tree ............ . . . . . . 4-124 4-6 Small Loss-of-Coolant Accident Event Tree .. ........ ... ..... . 4-125 4-7 Reactor Coolant System Leak Event Tree . .. ................. . . 4-126 4-8 Passive Residual Heat Removat Tube Rupture Event Tree . ... ... ... . 4-127 4-9 Steam Generator Tube Rupture Event Tree . . . . ........... ... . . 4-128 4-10 Reactor Vessel Rupture Event Tree ....... ........ .... . .... . 4-130 4-11 Interfacing Systems Loss-of. Coolant Accident Event Tree . .. . . ... . 4-131 4-12 Transients with Main Feedwater Event Tree .............. ......... 4-132 4-13 Transients with Loss of Reactor Coolant System Event Tree . . ..... . . 4-133 4-14 Transients with Loss of Main Feedwater Event Tree ....... . ........ 4-134 4-15 Transients with Core Power Excursion Event Tree . . . .. ... ... . . 4-135 4-16 Loss of Component Cooling Water System / Service Water System Event Tree . . . . ................. .. ..... .. . .... . 4-136 4-17 Loss of Main Feedwater Event Tree . ... ... ... ... ..... . . . 4-137 4-18 Loss of Condenser Event Tree . . . . . . ... .. .. ... .. . . 4-138 4-19 Loss of Compressed Air Event Tree . . ........... . .. . . 4-139 4-20 Loss of Offsite Power Event Tree ... .... . ... .. ....... . 4-140 4-21 Main Steam Line Break Downstream of Main Steam Isolation Valves Event Tree . . . . ........... . .. ...... .... . .. . 4-141 4-22 Main Steam Line Break Upstream of Main Steam Isolation Valves Event Tree .... ........ .. ... .... .... . . . . . . . 4-142 4-23 Stuck-Open Secondary. Side Safety Valve Event Tree . .. ... ... . . . 4-143 4-24 Anticipated Transient Without Scram Precursor without Main Feedwater Event Tree . . .. ... ..... ............. . . . . 4-144 4-25 Anticipated Transient Without Scram Precursor with Injection Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . . . . . . . . . 4- 14 6 4-26 Anticipated Transient Without Scram Precursor Transients with Main Feedwater Event Tree . . . . . . . . . . . . ............... . . . . . 4-147 O Revision: 9 April 11,1997 o \np60oprawvfprasof wpf.Ib lx h_ T Westifghouse
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%/ I'IST OF FIGURES (Cont.) Fisrure No. Ihlt Eggt 54 4 LOCA/RNS Pipe Rupture Dunng Hot / Cold Shutdown (RCS Filled) Even t Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-3 02 54-5 LOCA/RNS-V024 Opens During Hot / Cold Shutdown (RCS Filled) Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-303 54-6 Overdrauung of Reactor Coolant System During Dramdown to Mid-Loop . . . 54-304 54-7 Loss of Offsite Power (RCS Dramed) Event Tree ................... 54-305 54-8 Loss of RNS Initiator (RCS Dramed) Event Tree . . . . . . . . . . . . . . . . . . . . 54-306 54-9 Loss of CCW/SW Initiator (RCS Dramed) Event Tree . . . . . . . . . . . . . . . . 54-307 54-10 LOCA/RNS V024 Opens (RCS Dramed) Event Tree . . . . . . . . . . . . . . . . . 54-308 54-11 Accumulator Injection (Dilution Scenario) Event Tree . . . . . . . . . . . . . . . . . 54-309 54-12 Shutdown Transient Case SD1B2 RCS Pressure vs. Time . . . . . . . . . . . . . . 54-310 54-13 Shutdown Transient Case SD1B2 Mass Flow Rate vs. Time . . . . . . . . . . . . 54-311 54-14 Shutdown RNS Break Case SD3A (3500 gpm) . . . . . . . . . . . . . . . . . . . . . 54-312 54-15 Shutdown RNS Break Case SD3A2 (2000 gpm) . . . . . . . . . . . . . . . . . . . . 54-313 54-16 Shutdown RNS Break Case SD3A3 (1000 gpm) . . . . . . . . . . . . . . . . . . . . 54-314 54-17 Shutdown Plant Damage State Substate Event Tree for LP-ADS . . . . . . . . . 54-315 54-18 Shutdown Plant Damage State Substate Event Tree for LP-1 A . . . . . . . . . . . 54-316 54-19 Shutdown Plant Damage State Substate Event Tree for LP-3D . . . . . . . . . . . 54-317 54-20 Shutdown Plant Damage State Substate Event Tree for LP-3BR . . . . . . . . . . 54-318 54-21 Shutdown Plant Damage State Substate Event Tree for LP-3BE . . . . . . . . . 54-319 55-1 Seismic Initiating Event Hierarchy Tree . . . . . . . . . . . . . . . . . . . . . . . . . . 55-105 55-2 EQ-STRUC Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-106 55-3 EQ-RVFA Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-108 55-4 EQ-LLOCA Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . 55-109 55-5 EQ SLOCA Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-110 55-6 EQ-ATWS Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-111 55-7 EQ-STRUC Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-112 55-8 EQ.RVFA Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-113 55-9 EQ-LLOCA Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-114 55-10 EQ-SLOCA Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-115 55-11 EQ-ATWS Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-116 55-12 EQ-LOSP Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-117 55-13 EQLOSP Event Tree (for 0.5g level earthquake) . . . . . . . . . . . . . . . . . . . 5 5-118 J 55-14 EQAC2AB Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5- 1 19 55-15 EQ-XCIC Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5- 120 55-16 EQ-XADMA Fauh Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-121 55-17 EQ-XIW2A Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-122 55-18 EQ-RECIR Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-123 55-19 EQ.CM2SL Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-124 55-20 EQ-ADA Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-125 O- 55-21 EQ IW2AB Fauk Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-126 V IM S- luxi oW*v W*#1b AprH
LIST OF FIGURES (Cont.) e Firure No. Title Eage 55-22 EQ-PRHR Fault Tree . . . . . . . . . . . . . . ........ . . . . . . . . . . . . . . 55-127 55-23 EQ-PRESU Fault Tree . . . . . . ................... . . . . . . . . . . . 55-128 55-24 EQ-PMS Fault Tree . . . . . . . . . . ..... .... .. . . . . . . . . . . . . . 55-129 55-25 EQ-DC Fault Tree . . . . . . . . . ............... ... . . . . . . . . . . . 55-130 55 26 Class IE de Power Block Diagram . . . . . . . . . . . . . . . . . . . .. . . . . . . . 55-131 55-27 Containment Evaluation Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-132 55-28 EQ-STRUC Event Sequences . . . . . . . . . . . . . . . . .. ............ . 55-133 55-29 EQ-RVFA Event Sequences . . ......................... . . . . . 55 134 55-30 EQ-LLOCA Event Sequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 135 55-31 EQ-SLOCA Event Sequences . . . ... ................... . 55 136 55-32 EQ-SGTR Event Sequences ... .............. .. .......... . 55 137
'55 33 EQ-SLB Event Sequences . . .................. .......... . 55 138 55-34 EQ-ATWS Event Sequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 139 55-35 EQ-LOSP Event Sequences (for 0.5g level earthquakes) . . . . . . . . . . . . . . . 55-140 56-1 Flood Zones and Barriers Plan at 66'-6" . . . . . . . . . . ......... . . . 56-93 56-2 Flood Zones and Barriers Plan at 82'-6" . .. .. .... .... ... . 56-95 56-3 Flood Zones and Barriers Plan at 96'-6" . . . . . . . . . . . . . . . ... . . . . 56-97 56-4 Flood Zones and Barriers Plan at 100*-0" & 107'-2" . . . . . . .. .. 56-99 56-5 Flood Zones and Barriers Plan at 117*-6" . ....................... 56-101 56-6 Flood Zones and Barriers Plan at 135'-3" . . . . . . . . . . . . . . . . . . . . . . . . 5 6- 103 56-7 Flood Zones and Barriers Plan at 160'-6" & 153'-0" . . . . . ...... .. . 56-105 56-8 Floot Zones and Barriers Plan at 160'-6" & 180'-0" . . . . .... . . . . . . . 56-107 56-9 8-in. Fire Main Rupture at-Power Event Tree .......... .. .... . 56-109 56-10 8-in. Fire Main Rupture during Hot / Cold Shutdown Event Tree . . . . . . . . . . 56-110 56-11 8-in. Fire Main Rupture during RCS Drained Conditions Event Tree .. .. 56-111 57-1 Fire Progression Event Tree for 1200 AF 01 Fire Area . . . . . .. . . 57-156 59-1 Contribution of Initiating Events to Core Damage . . ..... ....... .. 59-233 59-2 Contribution of Initiating Events to Large Release Frequency and Core Damage Frequency . . . . . . . . ... . .. .. .. . .. .. 59-234 59-3 Total Plant CDF/LRF . . . . .. .... . .. .... .. .. ... .... . 59-235 59-4 24-Hour Site Boundary Dose Cumulative Frequency Distribution . . . . . . . . . 59-236 Revision: 11 O
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.p I (J TABLE OF CONTENTS (Cont.) Section I!. tit f. Bat CHAPTERS 46 THROUGH 48 DELETED CHAPI'ER 49 OFFSITE DOSE EVALUATION 49.1 Introduction ...............................................49-1 49.2 Conformance with Regulatory Requirements . . . . . . . . . . . . . . . . . . . . . . . . . 49-1 49.3 Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-2 49.4 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 9-2 49.5 Dose Evaluation Results and Discussions . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-6 49.6 Quantification of Site Risk . . . . . . . . . . . .........................49-7 49.7 Risk Quantification Results . . . . . . . . . ...........................49-7 49.8 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-8 CHAPTER 50 IMPORTANCE AND SENSITIVITY ANALYSIS 50.1 Introduction ...................................... . . . . . . . . 50-1 50.2 Importance Analyses for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-1, 50.2.1 Initiating Event Importances (Case 1) . . . . . . . . . . . . . . . . . . . . . . . 50-2 50.2.2 Common Cause Failure Importances (Case 2) . . . . . . . . . . . . . . . . . 50-3 p 50.2.3 Human Error Importances (Case 3) . . . . . . . . . . . . . . . . . . . . . . . . . 50-5 V 50.3 50.2.4 Component Importances (Case 4) . . . . . . . . . . . . . . . . . . . . . . . . . . 50-6 System Importances for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-7 50.4 Human Error Sensitivity Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-9 50.4.1 Set Human Error Probabilities to 1.0 (Failure) in Core Damage Results (Case 25) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-9 50.4.2 Set Human Error Probabilities to 0.0 (Success) in Core Damage Results (Case 26) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0- 10 50.4.3 Assess Importance of Increasing Human Error Probabilities by a Factor of 10 (Case 27) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-10 50.5 Other Sensitivity Analyses for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . 50-11 50.5.1 Diesel Generator Mission Time (Case 28) . . . . . . . . . . . . . . . . . . 50- 1 1 50.5.2 Impact of Passive System Check Valves on Core Damage Frequency (Case 29) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-12 50.5.3 Instrumentation and Control Cutoff Probability (Case 30) . . . . . . . . 50-12 50.5.4 Containment Recirculation After Safety Injection Line Break Event (Case 31 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50- 12 50.5.5 Quantification Truncation Probability (Case 32) . . . . . . . . . . . . . . . 50-13 50.5.6 Sensitivity to ADS Stage 4 Success Criteria (Case 33) . . . . . . . . . . 50-13 50.5.7 Squib Valve Failure Probability (Case 34) . . . . . . . . . . . . . . . . . . . 50-13 50.5.8 Circuit Breaker Failure Probability (Case 35) . . . . . . . . . . . . . . . . . 50-14 50.5.9 . End-State Importances (Case 36) . . . . . . . . . . . . . . . . . . . . . . . . . 5 0- 14
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I O TABLE OF CONTENTS (Cont.) Section Title Lage 50.6 Sensitivity and Importance Analyses For Large Release Frequency . . . . . . . 50-15 50.6.1 Importance Analyses For Large Release Frequency . . . . . . . . . . . 50- 15 50.6.2 Sensitivity Analyses For Large Release Frequency . . . . . . . . . . . 5 0-21 50.7 Sensitivity Analysis for Offsite Dose Risk . . . . . . . . . . . . . . . . . . . . . . . . . . 50-22 50.8 Results Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0-23 Attachment 50A ATWS PRA Sensitivity Case . . . . . . . . .......... ........ 50A-1 CHAPTER 51 UNCERTAINTY ANALYSIS 51.1 Introduction ... .......... . . ............ . .. . . . . . . . . . 51 - 1 51.2 Methodology . . . . . . . . .. .............. .... .. . . . . . . . . . . 51 - 1 51.3 Summary of Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........ ... 51-3 51.4 Sensitivity Studies for the Uncertainty Calculations . . . . . . . ....... . . 51-4 51.4.1 Urtertainty in the Cutoff Frequency . ....... .. ... . . . . . . 51 -4 51.4.2 Uncertainty in the Number of Cutsets Sampled . ... . ........ 51-4 51.4.3 Uncertainty in the Mean Failure Probability for Basic Events . . . . . . 51 -4 51.4.4 Sensitivity to the Random Number Input for Sampling . . . . . . . . . . . 51-5 51.5 References .......... ................ ....... . .. . .. 51-6 CHAPTER 52 RTNSS - FOCUSED PRA SENSITIVITY STUDY 52.1 Focused PRA Sensitivity Study Analysis Method . . .... ... . . . . . . . . . 5 2- 1 52.1.1 Core Damage Frequency Calculation . . . . . . . . . . . . . . . . . . 5 2-2 52.1.2 Release Frequency Calculation . . . . . . . . . . . . . .......... 52-5 52.2 At-Power Focused PRA Sensitivity Study . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2-5 52.2.1 At-Power Focused PRA Sensitivity Study Core Damage Frequency Quantification ... ..... ...... . . .......... 52-6 52.2.2 At-Power Focused PRA Sensitivity Study Release Frequency Quantification .. ............. ...... . . . . . 52-11 52.3 Shutdown Focused PRA Sensitivity Study . . . . . . . .. . .. ... . 52 16 52.3.1 Shutdown Focused PRA Sensitivity Study Core Damage Quantification . . . . . .......... ... ........ .. 52-16 52.3.2 Shutdown Focused PRA Sensitivity Study Release Frequency Calculation . ............................ . 52-19 52.4 Focused PRA Sensitivity Fire Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2-21 52.5 Focused PRA Sensitivity Study Flooding Analysis . . . ...... . . . . . . . 52-21 52.5.1 At-Power Focused PRA Sensitivity Study Flooding Scenarios . .. 52-21 52.5.2 Shutdown Focused PRA Sensitivity Study Flooding Scenarios . . . . . 52-22 52.5.3 Focused PRA Sensitivity Study Flooding Analysis Results S u m mary . . . . . . . . . . . . . . . . . . . . . . . . . .. . ... ... 52-23 52.6 Focused PRA Sensitivity Study Results and Conclusions . . . . ..... . ... 52-23 52.7 References . . . . . . . . . . . . . ... .... . ........ . . . . . . . . 52-24 O Revision: 11 ENEL March 1998 wy4h 3 Westinghouse o:WaWy llWa-toc.wptib XXXii
1151 v TABLE OF CONTENTS (Cont.) i i S.es.t,.on Title Egge CHAPTER 53 DELETED CHAPTER 54 LOW-POWER AND SHUTDOWN RISK ASSESSMENT 54.1 Introduction . .............................................54-1 54.2 Initiating Events .................. .............. ... .... 54-1 54.2.1 Identification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 54-2 54.2.2 Events Modeled . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-2 54.2.3 Shutdown Phases Summary Description . . . . . . . . . . . . . . . . . . . . . 54-3 54.2.4 Initiating Events for Operanng Modes . . . . . . . . . . . . . . . . . . . . . . . 54-4 54.2.5 Actuanng Signals and Systems Available . . . . . . . . . . . . . . . . . . . 54- 15 54.2.6 Scenarios for Detailed Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . 54-16 54.2.7 Summary of Initiating Events Analyzed . . . . . . . . . . . . . . . .... 54-21 54.3 Data............................................... . . . . 54-22 54.3.1 Shutdown Frequency . . . . ..................... ... 54-22 54.3.2 Mission Times . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-25 54.4 Event Tree Development . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-28 54.4.1 Event Tree LOSP-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4-3 0 54.4.2 Event Tree RNS-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-3 3 V Event Tree CCW-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-3 3 54.4.3 54.4.4 Event Tree LOCA-PR-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-33 54.4.5 Event Tree LOCA-V24-ND . . . . . . ...................... 54-34 54.4.6 Event Tree RCS-OD . . . . . . . . . . . . .............. . . . . . 54-36 54.4.7 Event Tree LOSP-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-3 8 54.4.8 Event Tree RNS-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.9 Event Tree CCW-D . . . . . . ........................... 54-40 54.4.10 Event Tree LOCA-V24-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.11 Boron Dilution Events (Reactivity Events) . . . . . . . . . . . . . . . . . . . 54-41 54.4.12 Boron Dilution Events Due to Chemical and Volume Control System Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-45 54.4.13 FndearM Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-48 54.5 Fault Tree Models for Shutdown and Low Power Events . . . . . . . . . . . . . . . 54-48 54.5.1 Insuumentation and Control Modeling for Shutdown (Level 1) . . . . 54-48 54.5.2 Insuumentation and Control Modeling for Shutdown (Level 2) .... 54-51 54.6 Success Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-51 54.6.1 MAAP4 Code Analysis for Shutdown Success Criteria . . . . . . . . . . 54-52 54.6.2 MAAP4 Parameter File . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-52 54.6.3 MAAP4 Input Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-54 54.6.4 Definition of MAAP4 Cases From Event Trees . . . . . . . . . . . . . . . 54-55 54.6.5 Results From MAAP4 Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . 54-57 54.7 Common Cause Analysis .....................................54-57 a s g Revision: 9 O s'JsL%. April 11,1997 uun o w w **
TABLE OF CONTENTS (Cont.) e Section Title Eage 54.8 Human Reliability Analysis . . . . . ........ ..................... 54-57 54.8.1 Operator Actions Calculated . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-58 54.8.2 Conditional Human Error Probabilities . . . . . . . . . . . .. ...... 54-64 54.9 Fault Tree Quantification ..................................... 54-64 54.10 Level 1 Core Damage Frequency Quantification . . . . . . . . . ........... 54-67 54.10.1 Core Damage Quantification Method . . . . . . . . . . . . . . . . . . . . . . 54-68 54.10.2 Quantification Inputs . . . . . . . . . . . ....... . . . . . . . . . . . . . 54-69 54.10.3 Level 1 Shutdown Core Damage Frequency Results . . . . . . . . . . . . 54-70 54.11 Shutdown and Low-Power Release Category Quantification . . . .... . . . . . 54-71 54.12 Shutdown Assessment Importance and Sensitivity Analyses . . . . . . . . . . . . . 54-71 54.12 1 Importance Analyses for Core Damage at Shutdown . . . . . . . . . . . 54-72 54.12.P Other Sensitivity Analyses for Shutdown Core Damage .. . . . . . . 54-77 54.13 Summary of Shutdown Level 1 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . 54- 81 54.14 References ............................................. . 54-87 Attachment 54A Design Change Impact on Low-Power and Shutdown Risk Assessment ...... ............................. 54A-1 Atuichment 54B Surge Line Flooding Effect on Low-Power and Shutdown Risk Assessment ...... ........ .... .... .. ...... 54B-1 Attachment 54C Effect of Modifications to Safe / Cold Shutdown PRA . . . . . ... 54C-1 CHAPTER 55 SEISMIC MARGIN ANALYSIS ............ ..... . ... 55-1 55.1 Intoduedon .. .......... .. .. .. ............... ... . .. 55-1 55.2 Calculation of HCLPF Values . . . . . . ..... ... . .. . ... ..... 55-2 55.2.1 Seismic Margin HCLPF Methodology . . . ... ..... 55-2 55.2.2 Calculation of HCLPF Values . . . ...... .... . ....... 55-2 55.3 Seismic Margin Model . . . . . . . . . . . . . . . . . . . . .......... . . . . 55-12 55.3.1 SMA Model and Assumptions . . . .... . ..... . .. . . 55-14 55.3.2 Seismic Initiating Events . . . . . . . . . ... ... . . . . . . . . . . . . 55 16 55.3.3 Initiating Event Category HCLPFs . . . . .. . .... .... . . . 55-17 55.3.4 Event Tree Models . . . . ......... ........ .. . ... . 55-20 55.3.5 Fault Tree Modeling and Quantification .... ...... . ... 55 28 55.3.6 Seismic Event Core Damage Sequence Evaluation . . . . . . . . . . . . . 55-36 55.3.7 Containment Performance Model . . . . . . . . . . . . . . . . . . . . . . . . . 55-37 55.4 Calculation of Sequence and Plant HCLPF . . . . . . . .... ..... . . . . 55-38 55.4.1 HCLPFs for Basic Events . . . . . . . . . . . . ............... . 55-39 55.4.2 Calculation of Initiating Event HCLPFs . . . . . . . . . . . . . . . . . . . 55-39 55.4.3 Calculation of System Fault Tree HCLPFs . . . . . . .... . . . . . . 55-39 55.4.4 Calculation of Sequence HCLPFs . . . . . . . . . . . . . . . . . . . . . . . . . 55-40 55.4.5 Calculation of Plant HCLPF . . . . . . . . ....... ......... . 55-43 55.4.6 Large Release HCLPF . . . . . . . . . . . . . . . . . . . . ..... . . . . . 55-43 Revision: 11 ENEL e March 1998 mjg T Westingh00Se oNwahv.Ilira-toc.wpf.lb XXXiv
( ) v TABLE OF CONTENTS (Cont.) Section Title Eage 55.5 Sensitivity Analyses . . ............... . ...... . . . .... .. 55-46 55.5.1 Robust Fuel and Core AssemSly ..... .. ....... ...... . 55-48 55.5.2 Credit for Operator Actions . . . . . ..... . . . . ..... . . . . . 55-49 55.5.3 Less Credit for Operator Actions in LOSP Event at 0.09g ...... . 55-52 55.5.4 72-Hour Mission Time . . . . . . . . . . .. . . . . . . . . . . . . . . . . . 55-5 3 55.5.5 Containment Isolation - Smaller Size Valves . . . ... . ..... . . . 55-56 55.5.6 Steam Generator Tube Rupture Success Criteria ........... .. 55-57 55.5.7 Steam Line Break Success Criteria . ...... . .. ... . . . . . . . . 55-5 8 55.5.8 Seismic Interaction Between Turbine and Auxiliary Buildings . . . . . 55-59 55.6 SMA Results and Insights . . . . . . . . . . . . . .. .. . . . . . ... .... 55-63 55.9.6 AP600 SMA Results . .... .. .. . ......... . . . . . 55-63 55.9.6 AP600 SMA Insights .. .. .. .. . . ... . . . . . . . . . . 55-68 55.7 References .... ............ .... . .. .. .. . .......... 55-70 Attachment 55A System HCLPF Calculations . . . . . . . .. . . .... .. 55A-1 Attachment 55B Sequence HCLPF Calculations . . ..... .. .. . .. ... . 55B-1 Attachment 55C Seismic Margin Analysis HCLPF Sensitivity Study . . . . . . 55C-1 (n) CHAPTER 56 PRA INTERNAL FLOODING ANALYSIS 56.1 Introduction .. . ... .. . ... . . . .. . .. . . .... . . . 56-1 56.1.1 Definitions . . . . ... ...... .. .. . . . ... ... 56-1 56.2 Methodology . . . .. ... ... .. .. . . . . . . ... . . . 56-1 56.2.1 Summary of Methodology . . . . ..... .... . . .. . .. . 56-1 56.2.2 Information Collection . . . .. . . . . . . .. . ... . . . 56-2 56.2.3 Initial Screening Assessment .... . . .. . . . .. . 56-3 56.2.4 Detailed Screening Assessment . . . .... . . . 56-4 56.2.5 Identification of Flood-Induced Initiating Events .. .. ... . . . 56-6 56.2.6 Initiating Event Frequencies .... . ... . . . . . .. .. . 56-7 56.3 Assumptions . . . . . . . . . . . . . . ........ .... . ....... .... . 56-7 56.3.1 General Flooding Analysis Assumptions and Engineering Judgments . 56-7 56.3.2 AP600-Specific Assumptions . .... . .......... . .. ... . 56-9 56.4 Infonnation Collection . ... . .............. ................ 56-11 56.4.1 PRA-Modeled Equipment and Locations . . ............. . 56-11 56.4.2 Identification of Areas for Flooding Evaluation . . . . . . . . . . . . . . . 56-11 56.5 At-Power Operations . . . . . . . . ..... . ........ .. . . . . . . . . 56-12 56.5.1 Initial Screening Assessment . . . . . . ..... ... . . ... .. . 56-12 56.5.2 Detailed Screening Assessment .... . .................. 56-12 56.5.3 Identification of Flood-Induced Initiating Events . . . . . ... .. . 56-28 56.5.4 Calculation of Flood-Induced Initiating Event Frequencies . . . . 56-32 56.5.5 Quantification of At-Power Flood-Induced Events . . . . .. . 56-39 _p~ I i LJ krme. I hI xxxy oVaWy, I Upra-toc.wpf.1b i L
111111 O TABLE OF CONTENTS (Cont.) Sectico I! tit East 56.6 Shutdown Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-41 56.6.1 Detailed Screening Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6-41 56.6.2 Identification of Flood Induced Initiating Events . . . . . . . . . . . . . . 56-42 56.6.3 Calculation of Flood. Induced Initiating Event Frequencies . . . . . . . . 56-43 56.6.4 Shutdown Quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-48 56.7 Seismically Induced Flooding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-51 56.8 Flooding Hazards During Refueling Outages . . . . . . . . . . . . . . . . . . . . . . . . 56-52 56.9 Flooding Sensitivity Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-52 56.9.1 Flooding Human Error Probabilities Sensitivity Study . . . . . . . . . . . 56-52 56.10 Summary of Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6-53 CHAPTER 57 INTERNAL FIRE ANALYSIS 57.1 Introduction ............ ................. ................ 57-1 57.2 Qualitative Analysis Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-2 57.3 Quantitative Methodology of Fire Area Frequency . . . . . . . . . . . . . . . . . . . . . 57 6 57.3.1 Fire Frequency Calculations .............................57-d 57.3.2 Fire Damage Category Quantification . . . . . . . . . . . . . . . . . . . . . . . 57-7 57.4 Core Damage Quantification Methodology . . . . . . . . . . . . . . . . . . . . . . . . 5 7- 10 57.5 Fire Analysis Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-12 57.5.1 Qualitative Analysis Assumptions and Other Modeling Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57- 12 57.5.2 Quantification Assumptions And Modeling Considerations .... .. 57-14 57.6 At. Power Qualitative Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-17 57.7 At. Power Quantitative Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-19 57.7.1 Fire Ignition Frequencies for Quantitative Analysis . . . . . . . . . . . . . 57-19 57.7.2 Fire Damage Category Quantification . . . . . . . . . . . . . . . . . . . . . . 57-19 57.7.3 Individual Area PRA Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-19 57.8 Control Room Fire Analysis - Power Operation . . . . . . . . . . . . . . . . . . . . . . 57-22 57.8.1 Description of the Cc etrol Room and Associated Fire Protection . . . 57-22 57.8.2 Alternate Shutdown Capability . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-23 57.8.3 Fire Hazard Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-24 57.8.4 AP600 Control Room Fire Evaluation . . . . . . . . . . . . . . . . . . . . . 57-25 57.8.5 Fire Scenario Identification and Frequency Determination . . . . . . . . 57-29 57.8.6 Control Room Fire Scenario Quantification and Results . . . . . . . . . 57-32
-57.9 Shutdown Fire Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 7-3 3 57.9.1 Fire Ignition Frequencies during Shutdown Modes of Operation . . . . 57-33 57.9.2 Fire Damage Category Quantification . . . . . . . . . . . . . . . . . . . . . . 57-34 57.9.3 Individual Area PRA Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-34 57.9.4 Fire Analysis for Safe Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . 57 34 57.9.5 Fire Analysis for Mid-Loop Operation . . . . . . . . . . . . . . . . . . . . . 57-41 O
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RE q \ ) Qt TABLE OF CONTENTS (Cont.) Section Title Eage 57.10 Summary and Conclusions . . . ... ............... . . . . . . . . . . 57-42 57.10.1 At. Power Analysis ....... ......... .... ......... . 57-42 57.10.2 Shutdown Fire Analysis .. ...... ...... .. ..... . . . . 57-45 57.10.3 Conclusions . . . . . ............. .. ...... .. . . . . . . 57-47 57.11 References . ......... ... .......... . .. .. . . . . . . . . . . 5 7-4 8
' ATTACHMENT 57A DEFINITIONS . ............. ...... .......... . 57A-1 ATTACHMENT 57B DESIGN CHANGE EFFECT ON INTERNAL FIRE ANALYSIS . . 57B-1 CHAPTER 58 WINDS, FLOODS, AND OTHER EXTERNAL EVENTS 58.1 Introduction ........... ................ . ...............58-1 58.2 External Events Analysis ............. ..... . . . .... . . . 58-1 58.2.1 Severe Winds and Tornadoes . . . . . . . .... . ....... . . . . 58-1 58.2.2 External Floods . . . . . . . . . .... .. . .... .. . ...... . 58-2 58.2.3 Transportation and Nearby Facility Accidents .... . . .. .. . . 58-2 58.3 Conclusion . ......... . . .. ... .... ..... . . .. . 58-3 58.4 References . .... ... . .................... ..... . . . . . . 5 8-3 m
( CHAPTER 59 PRA RESULTS AND INSIGHTS V) 59.1 Introduction ...... .... ... ... .. .... . . . . .59-1 59.2 Use of PRA in the Design Process ........... .... .. . . . 59-3 59.2.1 Stage 1 - Use of PRA During the Early Design Stage .. ... .. 59-4 59.2.2 Stage 2 - Preliminary PRA . . . . . . . . . .. .......... .. . 59-5 59.2.3 Stage 3 - AP600 PRA Submittal to NRC (1992) . ... . .... 59-7 59.2.4 Stage 4 - PRA Revision 1 (1994) . . . . ...... ..... ... . 59-8 59.2.5 Stage 5 - PRA Revisions 2-8 (1995-1996).. . ... . . . . . . . . . . . 5 9-8 59.3 Core Damage Frequency from Internal Initiating Events at Power ......... 59-10 59.3.1 Dominant Core Damage Sequences ....... ..... ... .. . 59-12 59.3.2 Component Importances for At-Power Core Damage Frequency . . . 59-44 59.3.3 System Importances fcr At-Power Core Damage . . . ... . . 59-44 59.3.4 System Failure Probabilities for At-Power Core Damage .... .. 59-45 59.3.5 Common Cause Failure Importances for At-Power Core Damage .. 59-45 59.3.6 Human Error Importances for At-Power Core Damage . ..... . 59-45 59.3.7 Accident Class Importances . . . . . . . . . . . ............. .. 59-47 59.3.8 Sensitivity Analyses Summary for At-Power Core Damage . . . . . . 59-47 59.3.9 Summary of Important Level 1 At-Power Results . . . . . . . . . . . . . 5 9-4 8 l 59.4 Large Release Frequency for Internal Initiating Events at Power . . . . . . . . . . . 59-51 59.4.1 Dominant Large Release Frequency Sequences . . . . . . . . . . . . . . . 5 9 -5 2 59.4.2 Sensitivity Analyses for Containment Response . . ..... .. .. 59-72 ,c 1
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a = a: O TABLE OF CONTENTS (Cont.) Section Title M 59.4.3 Comparison of Initiating Event Importances for Core Damage Frequency and Large Release Frequency . ........... . . . . . . 59-72 59.4.4 Summary of Imponant Level 2 At-Power Results . . . . . . . . . . . . . . 59-73 59.5 Core Damage and Severe Release Frequency from Events at Shutdown . . . . . .................. .. ........... . . . . 59-75 59.5.1 Summary of Shutdown Level 1 Results . . . . . . . . . . . . . . . . . . . . 59-75 59.5.2 Large Release Frequency for Shutdown and Low-Power Events . . . . .................... ........... . ... 59-81 59.5.3 Shutdown Results Summary . . . . ....................... 59-82 59.6 Results from Internal Flooding. Internal Fire, and Seismic Margin An aly ses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . 59-82 59.6.1 Results of Intemal Flooding Assessment . . . . . . . . . .. ...... 59-82 59.6.2 Results of Intemal Fire Assessment . .. .. .. ...... ..... 59-83 59.6.3 Results of Seismic Margin Analysis . . . ....... ....... 59-87 59.7 Plant Dose Risk from Release of Fission Products . ........... .... . 59-87 59.8 Overall Plant Risk Results . . . . . . . . . . . . . . . . ............. . . . . . . 59-88 59.9 Plant Features Imponant to Reducing Risk ...... .... .. . . . . . . . . . 59-89 59.9.1 Reactor Design . . . . . . .... .......... . . .... . . 59-90 59.9.2 Systems Design . . . . . . . . . . . . . . . .................. . 59-91 59.9.3 Instrumentation and Control Design . . . . ..... . . .. ... 59-94 59.9.4 Plant Layout . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59-95 59.9.5 Plant Structures . . . . . . . . . . . . . ........... . . . 59-96 59.9.6 Containment Design . . . . . . ..... . ... .. .. . ... 59-96 59.10 PRA Input to the Design Cenification Process . . . . . . ............... 59-101 59.10.1 PRA Input to Reliability Assurance Program . . . . . . . ....... 59-102 59.10.2 PRA Input to Certified Design Material ......... . . . . . . . . 59-102 59.10.3 PRA Input to the Technical Specifications . . . . . . . . . . . . . . . 59-102 59.10.4 PRA Input to MMI/ Human Factors / Emergency Response Guidelines . . . . . . . . . . . . . . . . ... ........ .. ... 59-102 59.10.5 Summary of PRA-Based Insights . . . . . . ....... . . . . . . . . 59-103 59.10.6 Combined License Information .................. .. . 59 103 APPENDIX A MAAP4 ANALYSIS TO SUPPORT SUCCESS CRITERIA . . . . . . . . . . . . . A- 1 APPENDIX B EX-VESSEL SEVERE ACCIDENT PHENOMENA ........ . ... . . . B-1 APPENDIX C DESIGN CHANGES THAT OCCURRED AFTER THE PRA ANALYSES WERE COMPLETED . . . . . . . . . . . . . . . . . . .....................C-1 APPENDIX D EQUIPMENT SURVIVABILITY ASSESSMENT . . . . . . . . . . . . . . . . . . . . . D- 1 i i O Revision: 11 ENEL March 1998 'a'a h
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i m I V q u- LIST OF TABLES (Cont.) Table No. Title Page 57-25 Quantitative Summary - Control Room Fires During Safe Shutdown . . . . . . . . .... ........ ................. 57-138 57-26 Summary of Qualitative Evaluation Results - Mid-Loop Operation . . .. ....... .......... . . . . . . . . . . . . . . 57-139 57-27 Summary of Quantitative Results - Mid-Loop Operation .... . . . . . . . . . 57- 146 57-28 Quantitative Summary - Control Room Fires During Mid-Loop Operation . . . . ..................... ......... .. 57-154 57-29 Safe Shutdown Containment Single-Hot-Short LOCA Sensitivity . . . . . . . . . 57-155 59-1 Contribution of Initiating Events to Core Damage . ....... . ... . . 59-104 59-2 Conditional Core Damage Probability of Initiating Events . . . . .. . ... 59-105 59-3 Internal Initiating Events at Power Dominant Core Damage Sequences . . . . 59-106 59-4 Sequence 1 - Safety Injection Line Break Dominant Cutsets (SI-LB-02) . . . . 59-108 59-5 Sequence 2 - Intermediate LOCA Dominant Cutsets (NLOCA-03) . . . . . .. 59-113 59-6 Sequence 3 - Large LOCA Dominant Cutsets (LLOCA-06) . . . ... .. . 59-118 59-7 Sequence 4 - Large LOCA Dominant Cutsets (LLOCA-03) . . . . . . . . . . . . 59-124 59-8 Sequence 5 - Reactor Vessel Rupture Cutset (RV-RP-02) . . . . . . . . . . . . . . 59-130
,m - 59-9 Sequence 6 - Large LOCA Dominant Cutsets (LLOCA-11) . . . ....... 59-131 59-10 Sequence 7 - ATWS Dominant Cutsets (ATWS-28) . . 59-133
() 59-11 Sequence 8 - Medium LOCA Dominant Cutsets (MLOCA-03) .... ...
. 59-141 59-12 Sequence 9 - ATWS Dominant Cutsets (ATWS-13) . . . . . . . . . . ...... 59-146 59-13 Sequence 10 - Intermediate LOCA Dominant Cutsets (NLOCA-04) . .. . 59-151 59-14 Sequence 11 - Safety Injection Line Break Dominant Cutsets (SI-LB-03) .. 59-156 59-15 Sequence 12 - Small LOCA Dominant Cutsets (SLOCA-03) . . . ...... . 59-160 59-16 Sequence 13 - Core Makeup Tank Line Break Dominant Cutsets (CMTLB-03) 59-165 59-17 Sequence 14 - Steam Generator Tube Rupture Dominant Cutsets (SGTR-07) . 59-170 59-18 Sequence 15 - Steam Generator Tube Rupture Dominant Cutsets (SGTR-23) . 59-171 59-19 Sequence 16 - Large LOCA Dominant Cutsets (LLOCA-02) . . . . . . . . . . . . 59-177 59-20 Sequence 17 - Large LOCA Dominant Cutsets (LLOCA-05) . . . . . . . . . . . . 59-183 59-21 Sequence 18 - Consequential SGTR Dominant Cutsets (SGTRC-03) . ..... 59-189 59-22 Sequence 19 - Intermediate LOCA Dominant Cutsets (NLOCA-16) . .. . 59-195 59-23 Typical System Failure Probabilities, Showing Higher Reliabilities for Safety Systems . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 59-201 59-24 Dominant CET Sequences . ...... ... . .. ....... . .. .. 59-202 59-25 Comparison of Initiating Event Contribution to Core Damage and La7e Release Frequencies ......... ... . ............ ... .. 59-203 59-26 Summary of AP600 PRA Results . . . . . . . . . . . . . .............. . 59-204 59-27 Comparison of AP600 PRA Results to Risk Goals . ... . . . . . . . . . . . . . 59-205 59-28 Site Boundary Dose Risk at 24 Hours ....... . . . . . . . . . . . . . . . . . . . 59-206 59-29 AP600 PRA-Based Insights . . . . . . . . . . .. ........... .. .. 59-207 g
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W LIST OF FIGURES e Fkure No. Ti.tle Eage 2-1 Core Damage Logic Diagram for Internal Initiators ...... . . . . . . . . . . . 2 -5 8 2-2 Core Damage Logic Diagram for Internal Initiators ..... .... .... .. 2-59 2-3 Core Damage Logic Diagram for Intemal Initiators ......... . . . . . . . . . 2-60 2-4 Core Damage Logic Diagram for Intemal Initiators . . . . . . . . . . . . . . .. .. 2-61 4-1 Large Loss-of-Coolant-Accident Event Tree ....... ..... . . . . . . . . . 4- 120 4-2 Medium Loss-of-Coolant Accident Event Tree . . . . . .. . . . . . . . . . . . . . . 4- 121 4-3 Core Makeup Tank Line Break Event Tree . . . . . . . . . . . . . . . . . . . . . . . . 4- 122 4-4 Direct Vessel Injection Line Break Event Tree .. . ....... . . . . . . . . 4-123 4-5 Intermediate Loss-of-Coolant Accident Event Tree . . . . . . . . . . .... ... . 4-124 4-6 Small Loss-of-Coolant Accident Event Tree . ................... 4-125 4-7 Reactor Coolant System Leak Event Tree . . .......................4-126 4-8 Passive Residual Ileat Removal Tube Rupture Event Tree . . . . . . . . . . . . . . 4-127 4-9 Steam Generator Tube Rupture Event Tree . . . . . . . ........ . . . . . 4-128 4-10 Reactor Vessel Rupture Event Tree .. ... . .......... .. . . . . . 4-130 4-11 Interfacing Systems Loss-of-Coolant Accident Event Tree . . . . . . . . . . . . . . 4-131 4-12 Transients with Main Feedwater Event Tree . .................. .. 4-132 4-13 Transients with Loss of Reactor Coolant System Event Tree . . . . . . . . . . . . 4-133 , 4-14 Transients with Loss of Main Feedwater Event Tree . . . . . . . . . . .. . .. 4-134 4-15 Transients with Core Power Excursion Event Tree . . . . . . . . . . ...... . 4-135 4-16 Loss of Component Cooling Water System / Service Water System Event Tree . . . ........... .. ... ............. ....... .. 4-136 4-17 Loss of Main Feedwater Event Tree . . . ........... .... ....... . 4-137 4-18 Loss of Condenser Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 4-138 4-19 Loss of Compressed Air Event Tree . . .. ........ .............. 4-139 4-20 Loss of Offsite Power Event Tree ......... .... ....... ... ... . 4-140 I 4-21 Main Steam Line Break Downstream of Main Steam Isolation Valves Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . ................ 4-141 4 22 Main Steam Line Break Upstream of Main Steam Isolation i Valves Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-142 l 4-23 Stuck-Open Secondary-Side Safety Valve Event Tree . . . .. ...... . . . . 4-143 j 4 24 Anticipated Transient Without Scram Precursor without Main I Feedwater Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . . . . 4-144 4-25 Anticipated Transient Without Scram Precursor with Injection E ve nt Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-146 4-26 Anticipated Transient Without Scram Precursor Transients with Main Feedwater Event Tree . . . . . . . . .......... ....... .. . .. 4-147 i O: Revision: -.9 April 11,1997 _ [ WSSthgh0llSe o \ap600;prawv.9pra-lof utf:1b lx
l 1, OST OF FIGURES (Cont.) Finnre No. IWg Eggs 54-4 LOCA/RNS Pipe Rupture Durms Hot / Cold Shutdown (RCS Filled) Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-302 54-5 . LOCA/RNS-V024 Opens During Hot / Cold Shutdown (RCS Filled) Event Tres . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-303 54-6 Overdraining of Reactor Coolant System During Dnandown to Mid-Loop . . . 54-304 54-7 Loss of Offsite Power (RCS Dramed) Event Tree .... ............. 54-305 54-8 Loss of RNS Initiator (RCS Drained) Event Tree . . . . . . . . . . . . . . . . . . . . 54-306 54-9 Loss of CCW/SW Initiator (RCS Drained) Event Tree . . . . . . . . . . . . . . . . 54-307
; 54-10 LOCA/RNS V024 Opens (RCS Druned) Event Tres . . . . . . . . . . . . . . . . . 54 308 54-11 Accumulator Injection (Dilution Scenario) Event Tree . . . . . . . . . . . . . . . . . 54-309 54-12 Shutdown Transient Case SD1B2 RCS Pressme vs. Time . . . . . . . . . . . . . . 54-310 54-13 Shutdown Transient Case SD1B2 Mass Flow Rate vs. Time . . . . . . . . . . . . 54-511 54-14 Shutdown RNS Break Case SD3A (3500 gpm) . . . . . . . . . . . . . . . . . . . . . 54-312 54-15 Shutdown RNS Break Case SD3A2 (2000 sym) . . . . . . . . . . . . . . . . . . . . 54-313 54-16 Shutdown RNS Break Case SD3A3 (1000 gym) . . . . . . . . . . . . . . . . . . . . 54-314 54-17 Shutdown Plant Damage State Substate Event Tree for LP-ADS . . . . . . . . . 54-315 54-18 Shutdown Plant Damage State Substans Event Tree for LP-1A . . . . . . . . . . . 54-316 54-19 Shutdown Plant Damage State Substaes Event Tree for LP-3D . . . . . . . . . . . 54-317
( 54-20 Shutdown Plant Damage State Substate Event Tree for LP-3BR . . . . . . . . . . 54-318 54 21 Shutdown Plant Damage State Substate Event Tree for LP-3BE . . . . . . . . . . 54-319 55-1 Seismic Initiating Event L. J.y Tres . . . . . . . . . . . . . . . . . . . . . . . . . . 55-105 55-2 EQSTRUC Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-106 55-3 EQRVFA Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 108 55-4 EQLLOCA Initianng Event Famit Tree . . . . . . . . . . . . . . . . . . . . . . . . . . 55-109 55 5 EQSLOCA Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-110 55-6 EQATWS Initiating Event Fault Tres . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-111 55 EQSTRUC Event Tres . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-112 55-8 EQRVFA Event Tres . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-113 55-9 EQLLOCA Event Tres . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-114 55-10 EQSLOCA Event Tres . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-115 55-11 EQATWS Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-116 l 55-12 EQLOSP Event Tres . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-117 55-13 EQLOSP Event Tree (for 0.5g level eenhquake) . . . . . . . . . . . . . . . . . . . . 55-118 55 14 EQAC2AB Fank Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-119 ' 55-15 ~ EQXCIC Fauk Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-120 55 EQXADMA Fank Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-121 55-17 EQXIW2A Fank Tres . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-122 55-18 EQRECIR Fauk Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-123 55-19 EQ.CM2SL Fauk Tres . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-124 55-20 . EQADA Fauk Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-125 [;- 55-21 EQIW2AB Fauk Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-126 .s - Revision: 9 H ED April 11,1997 i Imi .wsomm e w . wor
W e, LIST OF FIGURES (Cont.) Figure No. Title hage 55-22 EQ-PRHR Fault Tree . . . . . . . . .............. .......... . . . 55-127 55-23 EQ-PRESU Fault Tree . . . . . . . . ........ ............. . 55-128 55-24 EQ-PMS Fault Tre e . . . . . . . . . . . . ...... .. ....... . . . . . . . 55-129 l 55-25 EQ.DC Fault Tree . . . . . ................... . . . . . . . . . . . . . . . 55- 130 55-26 Class IE de Power Block Diagram . . .... ..................... 55-131 55-27 Containment Evaluation Model ... ...........................55-132 55 28 EQ-STRUC Event Sequences . . .. .......... . . . . . . . . . . . . . . . 5 5-13 3 55-29 EQ-RVFA Event Sequences . . . . . . . . .................... ... . 55-134 55-30 EQ-LLOCA Event Sequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-135 55-31 EQ-SLOCA Event Sequences . . . . . . . . . . . . . .. ... . . . . . . . . . . 55-136 55 32 EQ-SGTR Event Sequences . ..................... ..... ... 55-137 55 33 EQ-SLB Event Sequences . . . . . . . . . . . . . . . . . .... ....... ... 55-138 55-34 EQ-ATWS Event Sequences . . ........... .... .. . . . . . . . . . . 5 5-13 9 55-35 EQ-LOSP Event Sequences (for 0.5g level earthquakes) . ......... ... 55-140 56-1 Flood Zones and Barriers Plan at 66'-6" . . ...... .... .. .. . . . 56-93 56-2 Flood Zones and Barriers Plan at 82'-6" . . ...... ......... .. ... . 56-95 56-3 Flood Zones and Barriers Plan at 96'-6" . . . . . . . . . . . . . . . . . . . . . . ... . 56-97 56-4 Flood Zones and Barriers Plan at 100'-0" & 107'-2" . . . . . . ..... . 56-99 56-5 Flood Zones and Barriers Plan at 117'-6" ........ .. ........ 56-101 56-6 Flood Zones and Barriers Plan at 135'-3" . . . . . . .............. .. 56-103 56-7 Flood Zones and Barriers Plan at 160'-6" & 153'-0" . . ......... 56-105 56-8 Flood Zones and Barriers Plan at 160'-6" & 180'-0" . . . .. .. .. .. . 56-107 56-9 8-in. Fire Main Rupture at-Power Event Tree ....... . ...... . . . . 56-109 56-10 8-in. Fire Main Rupture during Hot / Cold Shutdown Event Tree . . . . . . .. . 56-110 56-11 8-in. Fire Main Rupture during RCS Drained Conditions Event Tree ... . 56-111 57-1 Fire Progression Event Tree for 1200 AF 01 Fire Area . . . . . . . . . . . .. . 57-156 59-1 Contribution of Initiating Events to Core Damage .. . .. .. . . . . . 59-233 59 2 Contribution of Initiating Events to Large Release Frequency and Core Damage Frequency . ...... ........... ... ......... 59-234 59-3 Total Plant CDF/LRF . . . . . . .. ............ ..... ...... . . 59-235 59-4 24-Hour Site Boundary Dose Cumulative Frequency Distribution . . . . . . . . 59-236 Revision: 11 e March 1998 3 Westiligh00Se owemva*v.is w w
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f n (v) TABLE OF CONTENTS (Cont.) Section Tit.le
. Ease CHAPTERS 46 THROUGH 48 DELETED CHAPTER 49 OFFSITE DOSE EVALUATION 49.1 Introduction ........................................ . . . . . . 49-1 49.2 Conformance with Regulatory Requirements . . . . . . . . . . . . . . . . . . . . . . . 4 9- 1 49.3 Assu mptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-2 49.4 Method ology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 9-2 49.5 Dose Evaluation Results and Discussions . . . . . ..................... 49-6 49.6 Quantification of Site Risk . . . . . . . . ............................49-7 49.7 Risk Quantification Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 9-7 49.8 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 9- 8 CHAPTER 50 IMPORTANCE AND SENSITIVTTY ANALYSIS ,
50.1 Introduction ...............................................50-1 j 50.2 Importance Analyses for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0- 1 50.2.1 Initiating Event Importances (Case 1) . . . . ...... . . . . . . . . . . . 50-2' 50.2.2 Common Cause Failure Importances (Case 2) . . . . . . . . . . . . . . . . . 5 0-3 m 50.2.3 Human Error Importances (Case 3) . . . . . . . . . . . . . . . . . . . . . . . . . 50-5
) 50.2.4 Component Importances (Case 4) . . . . . . . . . . . . . . . . . . . . . . . . . 50-6 50.3 System Importances for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-7 50.4 Human Error Sensitivity Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-9 50.4.1 Set Human Error Probabilities to 1.0 (Failure) in Core Damage Results (Case 25) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-9 50.4.2 Set Human Error Probabilities to 0.0 (Success) in Core Damage Results (Case 26) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0 10 50.4.3 Assess Imponance of Increasing Human Error Probabilities by a Factor of 10 (Case 27) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-10 50.5 Other Sensitivity Analyses for Core Damage . . . . . . . . . . . . . . . . . . . . . . . 50- 1 1 50.5.1 Diesel Generator Mission Time (Case 28) . . . . . . . . . . . . . . . . . . 50- 1 1 50.5.2 Impact of Passive System Check Valves on Core Damage Frequency (Case 29) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50- 12 50.5.3 Instrumentation and Control Cutoff Probability (Case 30) . . . . . . . . 50-12 ;
50.5.4 Containment Recirculation After Safety Injection Line Break Event (Case 31 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-12 50.5.5 Quantification Truncation Probability (Case 32) . . . . . . . . . . . . . . . 50-13 50.5.6 Sensitivity to ADS Stage 4 Success Criteria (Case 33) . . . . . . . . . . 50-13 ) 50.5.7 Squib Valve Failure Probability (Case 34) . . . . . . . . . . . . . . . . . . . 50-13 50.5.8 Circuit Breaker Failure Probability (Case 3 5) . . . . . . . . . . . . . . . . . 50-14 50.5.9 End-State Imponances (Case 36) . . . . . . . . . . . . . . . . . . . . . . . . . 5 0- 14 A
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TABLE OF CONTENTS (Cont.) e Section Title _Page 50.6 Sensitivity and Importance Analyses For Large Release Frequency . . . . . . . . 50-15 50.6.1 Importance Analyses For Large Release Frequency . . . . . . . . . . . . 50-15 50.6.2 Sensitivity Analyses For Large Release Frequency . . . . . . . . . . . . . 50-21 50.7 Sensitivity Analysis for Offsite Dose Risk . . . . . . . . . . . .......... .. 50-22 50.8 Results Summary . . . . . . . . ............ . . . . . . . . . . . . . . . . . . . . 50-23 Attachment 50A ATWS PRA Sensitivity Case . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50A-1 CtIAITER 51 UNCERTAINTY ANALYSIS 51.1 Introduction ..................... .......... . ...... .... 51-1 51.2 Methodology . . . . . . . . . . . .. ......... .... ..... ... . .. 51-1 51.3 Summary of Results . . . . . . . . . . . . ...... ....... .............. 51-3 51.4 Sensitivity Studies for the Uncenainty Calculations ......... . .. ..... 51-4 51.4.1 Uncertainty in the Cutoff Frequency . . . . . . . . . . ... . .. 51-4 51.4.2 Uncertainty in the Number of Cutsets Sampled . . . . ... .. .. .51-4 51.4.3 Uncenainty in the Mean Failure Probability for Basic Events . . . . . . . 51-4 51.4.4 Sensitivity to the Random Number Input for Sampling . . . . . . . . .51-5 51.5 References . . . . . . . . . . . . . . . . . . . . . . . .......... .. ...... . 51-6 CHAITER 52 RTNSS - FOCUSED PRA SENSITIVITY STUDY 52.1 Focused PRA Sensitivity Study Analysis Method . . . . . ..... . . . . . . . 52- 1 52.1.1 Core Damage Frequency Calculation . . . .. ..... . . 52-2 52.1.2 Release Frequency Calculation . . . . . . . . . . . . . . . . . 5 2 -5 52.2 At-Power Focused PRA Sensitivity Study . . . ........... ..... . . . 52-5 52.2.1 At-Power Focused PRA Sensitivity Study Core Damage Frequency Quantification .............. .. .. ...... . . 52-6 52.2.2 At-Power Focused PRA Sensitivity Study Release Frequency Quantification ............ . . .. . . 52-11 52.3 Shutdown Focused PRA Sensitivity Study .. .... . ...... . . . . 52-16 52.3.1 Shutdown Focused PRA Sensitivity Study Core Damage Quantification . . ... .. ... . .... . . . . . . 52-16 52.3.2 Shutdown Focused PRA Sensitivity Study Release Frequency Calculation . .......... .. ... . ...... . . . 52-19 52.4 Focused PRA Sensitivity Fire Analysis ......... ................ . 52-21 52.5 Focused PRA Sensitivity Study Flooding Analysis . . . . . . . . ... . . . . . . . 52-21 52.5.1 At-Power Focused PRA Sensitivity Study Flooding Scenarios . .. . 52-21 52.5.2 Shutdown Focused PRA Sensitivity Study Flooding Scenarios . . . . . 52-22 52.5.3 Focused PRA Sensitivity Study Flooding Analysis Results Summary .......................... ......... ... 52-23 52.6 Focused PRA Sensitivity Study Results and Conclusions . . . ... . . . . . . 52-23 52.7 References . . ........... .. . . . . . . . . 52-24 e = L i' 0%raWv.,llipra-toc.wpf.lb xxxii m_ w St-i
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TABLE OF CONTENTS (Cont.) Section Title P. age 1 CHAPTER 53 DELETED CHAPTER 54 LOW-POWER AND SHUTDOWN RISK ASSESSMENT 54.1 Introduction ....... ............... ...... ......... . . . 54-1 54.2 Initiating Events . ................ .. ........... ... ..... 54-1 54.2.1 Identification . . . . ............. ........ . . ... 54-2 54.2.2 Events Modeled . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. 54-2 54.2.3 Shutdown Phases Summary Description . . . . .. . . . . . . . . . . . . . . . 54-3 54.2.4 Initiating Events for Operaung Modes . . . . . . . . . . . . . . . . . . . . . . 54-4 54.2.5 Actuating Signals and Systems Available . . . . . . . . . . . . . . . . . . 54-15 54.2.6 Scenarios for Detailed Analysis . . . . . . . . . . . . . . . . . . . . . . . . . 54- 16 54.2.7 Summary of Initiating Events Analyzed . . . . . . . . . . . . . . . . 54-21 54.3 Data................................ ............... .. 54-22 54.3.1 Shutdown Frequency . . ............................ 54-22 54.3.2 Mission Times . . . . . . . . . . . . .... ......... . . . . . . . . . . 54-25 54.4 Event Tree Development . . . . . . . . . ......................... .. 54-28 ; p 54.4.1 Event Tree LOSP-ND . . . . . . ... ............... ... .. 54-30 V 54.4.2 54.4.3 Event Tree RNS-ND .. ........... ... ......... Event Tree CCW-ND . . ........ ...........
... 54-33
. . . . . . . . . 54-3 3 54.4.4 Event Tree LOCA-PR-ND . . . . . . ...................... 54-33 54.4.5 Event Tree LOCA-V24-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-34 54.4.6 Event Tree RCS-OD . . . . . . . ... ................. .... 54-36 54.4.7 Event Tree LOSP-D . . .......... . . . . . . . . . . . . . . . . . . 54-3 8 54.4.8 Event Tree RNS D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.9 Event Tree CCW-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.10 Event Tree LOCA-V24-D . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 54 40 54.4.11 Boron Dilution Events (Reactivity Events) . . . . . . . . . . . . . . . . . . . 54-41 54.4.12 Boron Dilution Events Due to Chemical and Volume Control System Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-45 54.4.13 Endstates Summny . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-4 8 54.5 Fault Tree Models for Shutdown and Low-Power Events . . . . . . . . . . .... 54-48 54.5.1 Instrumentation and Control Modeling for Shutdown (Level 1) ... 54 48 54.5.2 Instrumentation and Control Modeling for Shutdown (Level 2) .... 54-51 54.6 Success Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-51 54.6.1 MAAP4 Code Analysis for Shutdown Success Criteria . . . . . . . . . . 54-52 54.6.2 MAAP4 Parameter File . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-52 54.6.3 MAAP4 Input Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-54 54.6.4 Definition of MAAP4 Cases From Event Trees . . . . . . . . . . .... 54-55 ,
54.6.5 Results From MAAP4 Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . 54-57 ) g_ 54.7 Common Cause Analysis . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . ..... 54-57 , i .
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TABIE OF CONTENTS (Cont.) Section Title _P_gg i 54.8 Human Reliability Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-57 54.8.1 Operator Actions Calculated . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 58 54.8.2 Conditional Human Error Probabilities . . .................. 54-64 54.9 Fault Tree Quantification ..................................... 54-64 54.10 Level 1 Core Damage Frequency Quantification . . . . . . . . . . . . . . . . . . . . . . 54-67 54.10.1 Core Damage Quantification Method . ......... .......... 14-68 54.10.2 Quantification Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-69 54.10.3 Level 1 Shutdown Core Damage Frequency Results . . . . . . . . . . . 54-70 54.11 Shutdown and Low-Power Release Category Quantification . . . . . . . . . . . . . . 54-71 54.12 Shutdown Assessment Importance and Sensitivity Analyses . . . . . . . . . . . . . 54-71 54.12.1 Importance Analyses for Com Damage at Shutdown . . . . . . . . . . . . 54-72 54.12.2 Other Sensitivity Analyses for Shutdown Core Damage ......... 54-77 i 54.13 Summary of Shutdown Level 1 Results ........................... 54-81 l 54.14 References ................................................ 54-87 Attachment 54A Design Change Impact on Low-Power and Shutdown Risk Assessment ........... .... ................... 54A-1 Attachment 54B Surge Line Flooding Effect on Low-Power and Shutdown Risk Assessment ........ .......... ................ 54B-1 Attachment 54C Effect of Modifications to Safe / Cold Shutdown PRA . . . . . . ... 54C-1 CHAPTER 55 SEISMIC MARGIN ANALYSIS . . . . . . ..... ....... .... .55-1 55.1 Introduction .......... ...... .. ..... .. ... ... .. .... 55-1 55.2 Calculation of HCLPF Values . . . ............. .............. . 55-2 55.2.1 Seismic Margin HCLPF Methodology ........... ... ...... 55-2 55.2.2 Calculation of HCLPF Values . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-2 55.3 Seismic Margin Model . . . . . ......... ....... .... ....... .. 55-12 55.3.1 SMA Model and Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5- 14 55.3.2 Seismic Initiating Events . . . .. . ... ......... ....... 55-16 ) 55.3.3 Initiating Event Category HCLPFs . . . . . . . ... . . . . . . . . . 55-17 I 1 55.3.4 Event Tree Models . . . . . . . . ............... ... . .... 55-20 i l 55.3.5 Fault Tree Modeling and Quantification . ... . . . . . . . . . . . . . . 5 5-2 8 l 55.3.6 Seismic Event Core Damage Sequence Evaluation . . . . . . . . . . . 55-36 55.3.7 Containment Performance Model . . . . . . . . . . . . . . . . . . . . . . . . . 55-37 55.4 Calculation of Sequence and Plant HCLPF . . . . . . . . . . . . . . . . . . . . . . .. 55-38 55.4.1 HCLPFs for Basic Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-3 9 55.4.2 Calculation of Initiating Eveat HCLPFs . . . . . . . . . . . . . . . . ... 55-39 55.4.3 Calculation of System Fai < Tree HCLPFs . . . . . . . . . . . . . . . .. 55-39 55.4.4 Calculation of Sequence HCLPFs . . . . . . . . . . . . . . . . . . ...... 55-40 55.4.5 Calculation of Plant HCLPF . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-4 3 55.4.6 Large Release HCLPF . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-4 3 O Revision: 11 ENEL March 1998 ca h T Westinghouse o$prairevJ lipra-toc.wpf;ib xxxiv
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Section Title Page 55.5 Sensitivity Analyses . . . . . ....................... ....... . . . 55-46 55.5.1 Robust Fuel and Core Assembly . . . . . . . . . . . . . . . . . . . . . . . . . 55-48 55.5.2 Credit for Operator Actions . . ...................... . . . 55-49 55.5.3 Less Credit for Operator Actions in LOSP Event at 0.09g . . . . . 55-52 55.5.4 72-Hour Mission Time . . . . . . . . . . ... .................5553 55.5.5 Containment Isolation - Smaller Size Valves . . ..... . . . . . . . . 55-56 55.5.6 Steam Generator Tube Rupture Success Criteria . . . . . . . . . . . . . . . 55-57 55.5.7 Steam Line Break Success Criteria . . . . . . . . . . . . . . . . . . . . . . . 5 5-5 8 55.5.8 Seismic Interaction Between Turbine and Auxiliary Buildings . . . . . 55-59 55.6 SMA Results and Insights . . . ......................... . . . . . . 55-63 55.9.6 AP600 SMA Results ......... ...................... 55-63 55.9.6 AP600 SMA Insights . . . . . .... ..... ............... . 55-68 55.7 References ... .............. ..... ......... ..... . . . . . . 55-70 Attachment 55A System HCLPF Calculations . ..... ........ . . . . . . . . . . 55 A- 1 Attachment SSB Sequence HCLPF Calculations . . . .................. ... 55B-1 p Attachment 55C Seismic Margin Analysis HCLPF Sensitivity Study ..... . ... 55C-1 ( l
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CHAPTER 56 PRA INTERNAL FLOODING ANALYSIS 56.1 Introduction ........ ................ ............ .. ... . 56-1 56.1.1 Definitions . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . 56- 1 56.2 Methodology . . ......................... ...... ..... . . . . . 56-1 56.2.1 Summary of Methodology . . . . . ...... ........... ..... . 56-1 56.2.2 Information Collection .......... . ............. ..... 56-2 56.2.3 Initial Screening Assessment . . ....... . . . . . . . . . . . . . . . . . . 5 6-3 56.2.4 Detailed Screening Assessment . .. . .. .... . . . . . . . . . 5 6-4 56.2.5 Identification of Flood-Induced Initiating Events ... ......... . 56-6 56.2 6 Initiating Event Frequencies . ............ .. ... ... . . 56-7 56.3 Assumptions . . ................. ................ ..... . . 56-7 56.3.1 General Flooding Analysis Assumptions and Engineering Judgments . 56-7 56.3.2 AP600-Specific Assumptions ....... .............. ... 56-9 56.4 Information Collection . . . . . . . . .......... ....... . . . . . . . . . . . 56- 1 1 56.4.1 PRA.Modeled Equipment and Locations . . . . . . . . . . . . . . . .. . . . . 56-11 56.4.2 Identification of Areas for Flooding Evaluation . . . . . . . . . . . . . . 5 6- 1 1 56.5 At-Power Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6- 12 56.5.1 Initial Screening Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-12 56.5.2 Detailed Screening Assessment ................ . . . . . . . 56- 12 56.5.3 Identification of Flood. Induced Initiating Events ........... .. 56-28 56.5.4 Calculation of Flood. Induced Initiating Event Frequencies . . . . . . . . 56-32 56.5.5 Quantification of At-Power Flood-Induced Events . . .. . . . . . . . . 56-39 y
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O TABLE OF CONTENTS (Cont.) Seedon lille P.3ge 56.6 S hutdown Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-41 56.6.1 Detailed Screening Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6-41 56.6.2 Identification of Flood-Induced Initiating Events . . . . . . . . . . . . . . 56-42 56.6.3 Calculation of Flood. Induced Initiating Event Frequencies . . . . . . . . 56-43 56.6.4 Shutdown Quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 48 56.7 Seismically Induced Flooding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 51 56.8 Flooding Hazards During Refueling Outages . . . . . . . . . . . . . . . . . . . . . . . 56 52 56.9 Floodmg Sensitivity Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 52 56.9.1 Flooding Human Error Probabilities Sensitivity Study . . . . . . . . . . . 56 52 56.10 Summary of Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6-53 CHAPTER 57 INTERNAL FIRE ANALYSIS 57.1 Introduction .............................. ................ 57-1 57.2 Qualitative Analysis Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 2 57.3 Quantitative Methodology of Fire Area Frequency . . . . . . . . . . . . . . . . . . . . . 57 6 57.3.1 Fire Frequency Calculations ............................. 57-6 57.3.2 Fire Damage Category Quantification . . . . . ................. 57-7 57.4 Core Damage Quantification Methodology . . . . . . . . . ... ....... . .. 57 10 57.5 Fire Analysis Assumptions . . . . . ............................. . 57-12 57.5.1 Qualitative Analysis Assumptions and Other Modeling Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 7- 12 57.5.2 Quantification Assumptions And Modeling Considerations ...... . 57-14 57.6 At-Power Qualitative Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-17 57.7 At. Power Quantitative Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-19 57.7.1 Fire Ignition Frequencies for Quantitative Analysis . . . . . . . . . . . . . 57-19 57.7.2 Fire Damage Category Quantification . . . . . . . . . . . . . . . . . . . . . . 57-19 57.7.3 Individual Area PRA Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-19 57.8 Control Room Fire Analysis - Power Operation . . . . . . . . . . . . . . . . . . . . . . 57 22 57.8.1 Description of the Control Room and Associated Fire Protection . . . 57-22 57.8.2 Alternate Shutdown Capability . . . . . . . . . . . . . . ............ 57-23 57.8.3 Fire Hazard Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 24 57.8.4 AP600 Control Room Fire Evaluation . . . . . . . . . . . . . . . . . . . . . . 57 25 57.8.5 Fire Scenario Identification and Frequency Determination . . . . . . . . 57 29 57.8.6 Control Room Fire Scenario Quantification and Results ......... 57-32 57.9 Shutdown Fire Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 33 57.9.1 Fire Ignition Frequencies during Shutdown Modes of Operation . . . . 57-33 57.9.2 Fire Damage Category Quantification . . . . . . ............... 57-34 57.9.3 Individual Area PRA Analysis . . . . . . . . . . . . . . . . . . ....... . 57-34 57.9.4 Fire Analysis for Safe Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . 57 34 57.9.5 Fire Analysis for Mid-Loop Operation . . . . . . . . . . . . . . . . . . . . 5 7-41 th oMp600;praWv.9ptoc.wpf,1b xxxvi m_ w See-. l I _ 1
(- U TABLE OF CONTENTS (Cont.) Section Title Eage 57.10 Summary and Conclusions .. ... . .. ....... .. .... .. . . . . . 57-42 57.10.1 At-Power Analysis . ........ .... .. .. .. . ... . 57-42 57.10.2 Shutdown Fire Analysis . . . . . . . . . . . . . . . . . ............. 57-45 57.10.3 Conclusiora . . . . . . . . . . . . . . . . .... .... . . . . . . . 57-47 57.11 References ..... .... ...... ..... ... . . . . . . . . 57-48 ATTACHMENT 57A DEFINITIONS ............. ... .... . ....... 57A-1 ATTACHMENT 57B DESIGN CHANGE EFFECT ON INTERNAL FIRE ANALYSIS 57B-1 CHAPTER 58 WINDS, FLOODS, AND OTHER EXTERNAL EVENTS 58.1 Introduction .. . .. . ..... ........ ... . . . . . . . . . . . . . 5 8- 1 58.2 External Events Analysis ..... ............... ............. . . 58-1 58.2.1 Severe Winds and Tomadoes . . . . . . .... ..... ...... 58-1 58.2.2 External Floods ........... .... . . .. .. ... . 58-2 58.2.3 Transportation and Nearby Facility Accidents ........ .... . 58-2 58.3 Conclusion . . . . . . . . . . .... .... ............. ..... .... 58-3 58.4 References .............. .... ... .. . . .. . . ....... 58-3 m CHAPTER 59 PRA RESULTS AND INSIGHTS (}
' 59.1 Introduction . ............... . . . ........ . . . 59-1 59.2 Use of PRA in the Design Process . . . .. .. .. ... . .. .. . . 59-3 59.2.1 Stage 1 - Use of PRA During the Early Design Stage . . .. .. . . 59-4 59.2.2 Stage 2 - Preliminary PRA . .......... ... ... ....... 59-5 59.2.3 Stage 3 - AF600 PRA Submittal to NRC (1992) . .. . .... .. 59-7 59.2.4 Stage 4 - PRA Revision 1 (1994) ......... . . .. . . . . 59-8 59.2.5 Stage 5 - PRA Revisions 2-8 (1995-1996) ... . .. . . .... . 59-8 59.3 Core Damage Frequency from Intemal Initiating Events at Power . . . . ... . 59-10 59.3.1 Dominant Core Damage Sequences .. .. ...... ... ... . 59-12 1 59.3.2 Component Importances for At-Power Core Damage Frequency . . . . 59-44 59.3.3 System Importances for At-Power Core Damage . ........ . 59-44 59.3.4 System Failure Probabilities for At-Power Core Damage . . . . . . . . . 59-45 59.3.5 Common Cause Failure Importances for At-Power Core Damage .. 59-45 59.3.6 Human Error Importances for At-Power Core Damage . . . . . . . . 59-45 59.3.7 Accident Class Importances . ..... ... . . . . . . . . . . . . . 5 9-47 59.3.8 Sensitivity Analyses Summary for At-Power Core Damage . . . . . . 59-47 59.3.9 Summary of Important Level 1 At-Power Results . . . . . .. .. . 59-48 j
$9.4 Large Release Frequency for Intemal Initiating Events at Power . . . . . . . . . 5 9-51 59.4.1 Dominant Large Release Frequency Sequences . . . . . . . . . . . . . . . 5 9-5 2 59.4.2 Sensitivity Analyses for Containment Response . . . . . . . . . . . . . 5 9-72 m
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0 TABLE OF CONTENTS (Cont.) Section Title Eage 59.4.3 Comparison of Initiating Event Importances for Core Damage Frequency and Large Release Frequency . . . . . . ........ . . 59-72 59.4.4 Summary of Important Level 2 At-Power Results . . . . . .. . . . . . 59-73 59.5 Core Damage and Severe Release Frequency from Events at Shutdown . . . . . . . . . . . . . .... . . . . . .... . . . . . . . . . 5 9-75 59.5.1 Summary of Shutdown Level 1 Results . . . . ....... . . . . 59-75 59.5.2 Large Release Frequency for Shutdown and Low-Power Events .......... .......... ...... ..... ..... . 59-81 59.5.3 Shutdown Results Summary . .... ................. .. 59-82 59.6 Results from Internal Flooding, Internal Fire, and Seismic Margin A n aly ses . . . . . . . . . . . . . . . . . . . . . . . . . .. .. ......... . .. 59-82 59.6.1 Results of Intemal Flooding Assessment .. ... .... ..... 59-82 59.6.2 Results of Intemal Fire Assessment ....... . ...... ..... 59-83 59.6.3 Results of Seismic Margin Analysis ... ......... .... .. 59-87 59.7 Plant Dose Risk from Release of Fission Products . . . . . . . . . . . . . ... 59-87 59.8 Overall Plant Risk Results . . . . ........ ..... ....... . . . . 59-88 59.9 Plant Features Important to Reducing Risk . . . .................. .. 59-89 59.9.1 Reactor Design ............. ...... . . .. . ...... 59-90 59.9.2 Systems Design . . .. .... .... . . .. .... .. . 59-91 59.9.3 Instrumentation and Control Design . . . .. ......... .... 59-94 59.9.4 Plant Layout .. ... .. ... . ... .. .... .. . . 59-95 59.9.5 Plant Structures . .. . . ... . .. .. ... .... . . 59-96 59.9.6 Containment Design . . . . . . . . .. ... ...... .. ..... 59-96 59.10 PRA Input to the Design Certification Process . . . ... .. .. ...... 59-101 59.10.1 PRA Input to Reliability Assurance Program .. .......... . 59-102 59.10.2 PRA Input to Cenified Design Material . . ...... .... .. 59-102 59.10.3 PRA Input to the Technical Specifications . . . . . ...... .... 59-102 59.10.4 PRA Input to MMI/ Human Factors / Emergency Response Guidelines . . . . . . . . . . ....... . .... .. . ..... . 59-102 59.10.5 Summary of PRA-Based Insights . ...... .... . .. . 59-103 59.10.6 Combined License Information . .... ... ...... ... 59-103 APPENDIX A MAAP4 ANALYSIS TO SUPPORT SUCCESS CRITERIA . . ..........A-1 APPENDIX B EX-VESSEL SEVERE ACCIDENT PHENOMENA .... ....... . .. B-1 APPENDIX C DESIGN CHANGES THAT OCCURRED AFTER THE PRA ANALYSES WERE COMPLETED . . . . . . . . ....... ............... . . . C- 1 APPENDIX D EQUIPMENT SURVIVABILITY ASSESSMENT . ....... . .. . . . D- 1 O Revision: 11 ENEL March 1998 mur- 3 Westinghouse onprairev.llWtoc,wpf.lb XXXViii
iO) v LIST OF TABLES (Cont.) Table No. Title hge 57-25 Quantitative Summary - Control Room Fires During Safe Shutdown .............. . ... . ...... ...... ...... 57-138 57-26 Summary of Qualitative Evaluation Results - Mid-Loop Operation . .. ........ ............. . .. ...... 57-139 57-27 Summary of Quantitative Results - Mid-Loop Operation ........ ...... 57-146 57-28 Quantitative Summary - Control Room Fires During Mid-Loop Operation . .... ....... .................. .... 57-154 57-29 Safe Shutdown Containment Single-Hot-Short LOCA Sensitivity . . . . . . . . 57-155 59-1 Contribution of Initiating Events to Core Damage ........ .......... 59-104 59-2 Conditional Core Damage Probability of Initiating Events . . . . ... ... . 59-105 59-3 Internal Initiating Events at Power Dominant Core Damage Sequences ... 59-106 59-4 Sequence 1 - Safety Injection Line Break Dominant Cutsets (SI-LB-02) . . . 59-108 59-5 Sequence 2 - Intermediate LOCA Dominant Cutsets (NLOCA-03) . . . .. . 59-113 59-6 Sequence 3 - Large LOCA Dominant Cutsets (LLOCA-06) . . . . . . .... . 59-118 59-7 Sequence 4 - Large LOCA Dominant Cutsets (LLOCA-03) . ... . . . . 59-124 59-8 Sequence 5 - Reactor Vessel Rupture Cutset (RV-RP-02) . . . . . . . .. . . 59-130 59-9 Sequence 6 - Large LOCA Dominant Cutsets (LLOCA-11) .... 59-131 73 . . ( ) 59-10 Sequence 7 - ATWS Dominant Cutsets (A'IWS-28) ........ .. .. . 59-133 59-11 Sequence 8 - Medium LOCA Dominant Cutsets (MLOCA-03) . ....... 59-141 59-12 Sequence 9 - ATWS Dominant Cutsets (ATWS-13) . . . . ...... .... . 59-146 59 13 Sequence 10 - Intermediate LOCA Dominant Cutsets (NLOCA-04) 59-151 59-14 Sequence 11 - Safety Injection Line Break Dominant Cutsets (SI-LB 03) . . 59-156 59-15 Sequence 12 - Small LOCA Dominant Cutsets (SLOCA-03) . .. . ... 59-160 59-16 Sequence 13 - Core Makeup Tank Line Break Dominant Cutsets (CMTLB-03) 59-165 59-17 Sequence 14 - Steam Generator Tube Rupture Dominant Cutsets (SGTR-07) . 59-170 59-18 Sequence 15 - Steam Generator Tube Rupture Dominant Cutsets (SGTR-23) 59-171 59-19 Segunce 16 - Large LOCA Dominant Cutsets (LLOCA-02) . . ... .. . 59-177 59-20 Sequence 17 - Large LOCA Dominant Cutsets (LLOCA-05) . . . . . . .. 59-183 59-21 Sequence 18 - Consequential SGTR Dominant Cutsets (SGTRC-03) . . . . . . . 59-189 59-22 Sequence 19 - Intermediate LOCA Dominant Cutsets (NLOCA-16) . . ..... 59-195 59 23 Typical System Failure Probabilitie;, Showing Higher Reliabilities for Safety Systems . . . ................... ...... . ...... . 59-201 59-24 Dominant CET Sequences . . .... .. ....... ... ..... 59-202 59-25 Comparison of Initiating Event Contribution to Core Damage and Large Release Frequencies ............. .................. . 59-203 59-26 Summary of AP600 PRA Results . .... .. ........ ...... .. . 59-204 59-27 Comparison of AP600 PRA Results to Risk GorJs . . . . . . . . . . .. ..... 59-205 59-28 Site Boundary Dose Risk at 24 Hours .... . ...... .. .. . .. 59-206 59-29 AP600 PRA-Based Insights . . . . . . . . . ........ ...... . ...... 59-207 L) ENE Revision: 11 UN88 WA:&- March 1998 lix o:\ap600\praNrev_I t'pra-lot.mpf:1 b
LIST OF FIGURES e fimeJp2 Title Eage 2-1 Core Damage Logic Diagram for Intemal Initiators ....... .... . . 2-58 2-2 Core Damage Logic Diagram for Intemal Initiators . . . .. .. ....... . 2-59 2-3 Core Damage Logic Diagram for Intemal Initiators . . . ... ............ 2-60 2-4 Core Damage Logic Diagram for Intemal Initiators .. ....... ...... .. 2-61 4-1 Large Loss-of-Coolant-Accident Event Tree ........ ............ .. 4 120 4-2 Medium Loss-of-Coolant Accident Event Tree . .. ... .. ........ . 4-121 4-3 Core Makeup Tank Line Break Event Tree . . . . . . ................ . 4-122 4-4 Direct Vessel In;cction Line Break Event Tree . . . .. . .......... . . 4-123 4-5 Intermediate Loss-of-Coolant Accident Event Tree . ............. . . . . 4-124 4-6 Small Loss-of-Coolant Accident Event Tree .......... . ...... .. . 4-125 4-7 Reactor Coolant System Leak Event Tree . ....... . ........ ... 4-126 4-8 Passive Residual Heat Removal Tube Rupture Event Tree . . . . . . . . . . . . 4-127 4-9 Steam Generator Tube Rupmre Event Tree . . . . . . ........ .. . 4-128 4-10 Reactor Vessel Rupture Event Tree ....... . . . . . . . . . . . . . . . . . 4- 130 4-11 Interfacing Systems Loss-of-Coolant Accident Event Tree . ............. 4-131 4-12 Transients with Main Feedwater Event Tree ... ....... .... . . 4-132 4-13 Transients with Loss of Reactor Coolant System Evert Tree . ..... .... 4-133 4-14 Transients with Loss of Main Feedwater Event Tree . . . .. ...... . . 4-134 4-15 Transients with Core Power Excursion Event Tree .............. ... 4-135 4-16 Loss of Component Cooling Water System / Service Water System Event Tree . . . . . . . . . . . . . . . . . . . . . .............. . ....... 4-136 4-17 Loss of Main Feedwater Event Tree . . . . .... . .... ........... 4-137 4-18 Loss of Condenser Event Tree . . . . . .. .... . ................ 4-138 4-19 Loss of Compressed Air Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . .. 4-139 440 Loss of Offsite Power Event Tree ........... .. . . . . . . . . . . . . . . . . 4- 140 4-21 Main Steam Line Break Downstream of Main Steam Isolation Valves Event Tree . . . . . . ......... ...... ................ . 4-141 4-22 Main Steam Line Break Upstream of Main Steam Isolation Valves Event Tree . . ... .. .. ..... ....... .. ........ 4-142 4-23 Stuck-Open Secondary-Side Safety Valve Event Tree ............. .. . 4-143 4-24 Anticipated Transient Without Scram Precursor without Main Feedwater Event Tree . . . ...... . ..... .. .... . . .. 4-144 4-25 Anticipated Transient Without Scram Precursor with Injection Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . . 4-146 4-26 Anticipated Transient Without Scram Precursor Transients with Main Feedwater Event Tree . . . . . . . ... ...... ............. .. 4-147 O Revision: 9 April 11,1997 o \np600% pram.9\pra lof*?f .!b [x f W85fi!)gh00S6 w____ . _ _ _ _ _ _ - _ _ _ - - _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ - _ _ _ - _ - _ _ - _ _ _ _ _ _ . _ _ _ - - _ _ _ - _ _ _ _ - _ - -
9 O b LIST OF FIGURES (Cont.) Firure No. Ijigg East 54-4 LOCA/RNS Pipe Rupture D.tring Hot / Cold Shttdown (RCS Filled) Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-3 02 54-5 LOCA/RNS-V024 Opens During Hot / Cold Shutdown (RCS Filled) Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-303 54-6 Overdrairung of Reactor Coolant System During Draindown to Mid-Loop . . . 54-304 54-7 Loss of Offsite Power (RCS Drained) Event Tree ................... 54-305 54-8 Loss of RNS Initiator (RCS Drained) Event Tree . . . . . . . . . . . . . . . . . . . . 54-306 54-9 Loss of CCW/SW Initiator (RCS Dramed) Event Tree . . . . . . . . . . . . . . . . 54-307 54-10 LOCA/RNS V024 Opens (RCS Drained) Event Tree . . . . . . . . . . . . . . . . . 54-308 54-11 Accumulator Injection (Dilution Scenario) Event Tree . . . . . . . . . . . . . . . . . 54-309 54-12 Shutdown Transient Case SDIB2 RCS Pressure vs. Tirne . . . . . . . . . . . . . . 54-310 54-13 Shutdown Transient Case SD1B2 Mass Flow Rs's vs. Tirne . . . . . . . . . . . . 54-311 54-14 Shutdown RNS Break Case SD3A (3500 gpm) . . . . . . . . . . . . . . . . . . . . . 54-312 54-15 Shutdown RNS Break Case SD3A2 (2000 gpm) . . . . . . . . . . . . . . . . . . . . 54-313 54-16 Shutdown RNS Break Case SD3A3 (1000 gpm) . . . . . . . . . . . . . . . . . . . . 54-314 54-17 Shutdown Plant Damage State Substate Event Tree for LP-ADS . . . . . . . . . 54-315 54-18 Shutdown Plant Damage State Substate Event Tree for LP-1 A . . . . . . . . . . . 54-316 54-19 Shutdown Plant Damage State Substate Event Tree for LP-3D . . . . . . . . . . . 54-317 b] 54-20 54-21 Shutdown Plant Damage State Substate Event Tree for LP-3BR , . . . . . . . . . 54-318 Shutdown Plant Damage State Substate Event Tree for LP-3BE . . . . . . . . . . 54-319 55-1 Seismic Initiating Event Hierarchy Tree . . . . . . . . . . . . . . . . . . . . . . . . . . 55-105 55-2 EQSTRUC Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-106 55-3 EQRVFA Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-108 55-4 EQLLOCA Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . 55-109 55-5 EQSLOCA Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-110 55-6 EQATWS Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-111 55-7 EQSTRUC Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-112 55-8 EQRVFA Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-113 55-9 EQLLOCA Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-114 i 55-10 EQSLOCA Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-115 l 55-11 EQATWS Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-116 i 55-12 EQLOSP Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-117 I 55-13 EQLOSP Event Tree (for 0.5g level canhquake) . . . . . . . . . . . . . . . . . . . . 55-118 l 55-14 EQAC2AB Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-119 55-15 EQXCIC Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-120 55-16 EQXADMA Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-121 55-17 EQXIW2A Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-122 55-18 EQRECIR Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-123 55-19 EQ-CM2SL Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-124 55-20 EQ-ADA Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-125
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(J x. 55-21 EQ1W2AB Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-126 Revision: 9 W== Egg CERW April 11,1997 1xxx o;Www.ep -int.wpt:tto Li---______- _ _
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LIST OF FIGURES (Cont.) e Firure No. Title P. age 55-22 EQ-PRHR Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-127 55-23 EQ-PRES U Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-128 55-24 EQ.PMS Fault Tree . . . . . . . . . . . . ....................... . 55-129 55-25 EQ-DC Fault Tree . . . . . . . . . . . .... . ................... .. 55-130 55-26 Class IE de Power Block Diagram . . . ......... . . . . . . . . . . . . . . . . 55- 131 55-27 Containment Evaluation Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-132 55 28 EQ-STRUC Event Sequences ..... ..........................55-133 55-29 EQ-RVFA Event Sequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 55 134 55-30 EQ-LLOCA Event Sequences . ... ............. . . . . . . . . . . . 5 5 13 5 55-31 EQ-SLOCA Event Sequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-136 55-32 EQ-SGTR Event Squences .... . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5 13 7 55-33 EQ-SLB Event Sequences . . . ....................... . . . . . . . . 55-138 55-34 EQ-ATWS Event Sequences . . . .............................55-139 55-35 EQ-LOSP Event Sequences (for 0.5g level earthquakes) . . . . ... ...... 55-140 56-1 Eood Zones and Barriers Plan at 66'-6" . . . . . . . . . . . . . . . . . . . . . . 5 6-9 3 56-2 Flood Zones and Barriers Plan at 82'-6" . . ............ . . . . . . . . . 5 6-95 56-3 Flood Zones and Barriers Plan at 96'-6" . ................ ...... . 56-97 56-4 Flood Zones and Baniers Plan at 100'-0" & 107'-2" . . . . . . . . . . . . . . . . . . . 56-99 56-5 Flood Zones and Barriers Plan at 117'-6" ......... .... ...... . . 56-101 56-6 Flood Zones and BarTiers Plan at 135'-3" ....................... . 56-103 56-7 Flood Zones and Barriers Plan at 160' 6" & 153'-0" . . .... .. . ... . . 56-105 56-8 Flood Zones and Barriers Plan at 160'-6" & 180*-0" . . . . ..... .. .. . . 56-107 56-9 8-in. Fire Main Rupture at-Power Event Tree .............. ..... . . 56-109 56-10 8-in. Fire Main Rupture during Hot / Cold Shutdown Event Tree . . . . . . . . . 56-110 56-11 8-in. Fire Main Rupture during RCS Drained Conditions Event Tree . . . . . . 56-111 57-1 Fire Progression Event Tire fer 1200 AF 01 Fire Area . . . . . . . . . . . . . . . 57-156 59-1 Contribution of Initiating Events to Core Damage ........... . . . . . . . 59-233 I 59-2 Contribution of Initiating Events to Large Release Frequency and Core Damage Frequency . . . . .. ............. . ....... 59-234 59-3 Total Plant CDF/LRF . . . . . . .. . ...... ....... ........ 59-235 l 59-4 24-Hour Site Boundary Dose Cumulative Frequency Distribution . . . . . . . 59-236 l l j Revision: 11 e March 1998 ow+n e ne.u w wpra 1xxxii hw [ WestiflghollSe
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- 39. In-Vessel Retention of Molten Core Debris
). < 4 %J the completion of Reference 39-1, confirm the heat transfer assumptions at Rayleigh numbers beyond 10l ! 1 l
l The full-scale, slice-geometry ULPU testing, shown in Figure 39-3 (Reference 39-1), j investigates the critical heat flux on the external surface of the lower head of the reactor vessel. The test provides full-height water elevation capability to investigate the effects of varied water height and subcooling. The test determines critical heat fluxes at the various azimuthal locations on the lower head extemal surface. Advanced ULPU Configuratien III testing provides data for prototypical reactor vessel steel material with surface preparations to Westinghouse specifications and external cooling water flow restrictions to model the effect of reactor vessel reflective insulation. This test is also used to provide oscillatory pressure data for the reactor vessel insulation design. The ROAAM analysis also investigates transient aspects of core relocation to the lower head and development of the steady-state heat transfer system described above. Investigations of lower head vessel failure due to jet impingement (Reference 39-1) and in-vessel steam i explosion (References 39-2,39-6,39-7, and 39-8) have been performed and it is concluded that these phenomena will not fail the vessel. Investigations of the transient development of { molten pool conditions conclude that the steady-state heat fluxes bound the transient l conditions. Therefore, vessel failure prior to the development of the natural circulating pool and extemal cooling is physically unreasonable. (ov) The results of in-vessel retention ROAAM analysis have been peer-reviewed by an international panel of 17 experts in the fields of severe accident progression, heat transfer, thermal-hydraulics, and structural mechanics. The conclusion that vessel failure is physically unreasonable under thermal-hydraulic conditions of in-vessel retention is considered to be resolved and is credited in the AP600 PRA, provided that the sequence meets the criteria outlined above. Based on the results of the ROAAM testing and analysis, vessel failure is concluded to be physically unreasonable in the AP600 PRA provided the following conditions are met:
- The reactor coolant system is depressurized.
)
l The vessel is submerged above the top of the molten debris pool. { .
- Reactor vesse! reflective insulation allows the ingress of water at the bottom and egress of steam at the top. ;
I The reactor vessel extemal surface conditions do not preclude the wetting phenomena I identified as the cooling mechanism in the ULPU testing. Each of these items is discussed below. g)s Revision: 11 W UNE
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ENEL Wa h March 1998 39-3 ovawn11wc394:wis9s I l I 9
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- 39. In Vessel Retention of Molten Core Debris 1 39.3 Reactor Coolant System Depressurization e
l Reactor coolant system pressure is determined at node DP (Chapter 36) on the containment i event tree. Sequences that proceed downstream from node DP are depressurized to less than i 150 psi in the reactor coolant system. Based on the structural analysis of the reactor vessel I in Reference 39-2, the maximum static pressure that the damaged vessel can withstand is I approximately 400 psi. Therefore, sufficient reactor coolant system depressurization is I attained at all sequences considered for in-vessel reten; ion. I 39.4 Reactor Cavity Flooding (Node IR) i Reactor cavity flooding is considered at node IR on the containment event tree. Cavity I flooding is accomplished through either operator action or through the progression of the i accident. The operator floods the cavity by opening a motor-operated valve and a squib valve I in each of the recirculation lines between the in-containment refueling water storage tank and I the containment recirculation sump, as shown in Figure 39-4. De water floods the I containment by flowing out of the recirculation sumps, filling the floodable region, shown in I Figure 39-5, of the containment to the 107' 2" elevation, shown in Figure 39-6. I Some accident sequences flood the cavity automatically, because core damage occurs as a i result of failure to recirculate water to the core once it is drained from the in-containment I refueling water storage tank, or injection occurs but is not sufficient or soon enough to prevent I core damage. His section discusses the evaluation of successful cavity flooding to prevent I vessel failure at node IR for each of the AP600 accident classes. I 39.4.1 Success Criteria l The question that is considered at node IR to determine success or failure of cavity flooding I for in-vessel retention is: 1 Does the in-containment refueling water storage tank water flood the reactor cavity sufficiently I to submerge the reactor vessel above the elevation of the molten core debris in the lower I head? I Two criteria are us ,d to determine success at node IR: 1) the reactor vessel hemisphere must I be submerged prior to relocation of core debris to the lower head; and 2) the water level must I be above the top of the in-vessel debris bed prior to full relocation of debris to the lower i head. I 39.4.1.1' Water Elevations Success Criteria l Based on the criteria above, success is determined by the timing of water flooding in the I containment with respect to core melt relocation timing in the reactor vessel. He timing and I the uncertainties in the timing of core relocation to the lower head are discussed in I Appendix P of Reference 391, 'Ihe expected time of initial debris relocation to the lo'ver l I Revision: 8 j September 30,1996 6 Westinghouse j %V0mpraWv.s\wc39.wpf:ltso92296 39-4 l 1 L- - - - -
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- 39. In. Vessel Retention of Molten Core Debris I
i 'Ihe results of the analyses showed that the insulation was able to meet each of the defined I functional requirements. The design of the reactor vessel insulation provides an engineered I pathway for water-cooling the vessel and for venting steam from the reactor cavity. Design I changes to the insulation were completcd to ensure that the stress and deflection requirements I were met to provide adequate pathways for ingress of water and venting of steam. I l 39.6 Reactor Vessel External Surface Treatment l Based on the reactor vessel system design specification. the only treatment of the extemal I surface of the reactor vessel is a protective paint applied by the manufacturer prior to I shipping. The paint protects the vessel carbon steel surface during ', hipping and storage. I Removal of external surface paint, or any other treatment of the extemal reactor vessel ' I surface, is not expected to occur. l The ULPU testing includes tests using prototypical steel with pnint applied according to l Westinghouse paint application specifications. The aged paint surface actually increased the I wetability of the vessel external surface and increased the critical heat flux. In the PRA, it I is assumed that no extemal surface treatment of the reactor vessel impairs heat removal from I the vessel external surface. I 39.7 Reactor Vessel Failure (Node VF) l 39.7.1 Node VF Success Criteria 1 The question considered at node VF to determme success or failure of rer.ctor vessel integrity I is:
- i Is the core debris maintamed inside the reactor vessel?
i I Success is credited at node VF if debris is maintamed in the reactor vessel and relocation to l l the containment is prevented. Based on the ROAAM analysis of in-vessel retention. an intact I reactor vessel remains intact if the reactor coolant system is depressurized (success at node l DP) and the reactor vessel is adequately submerged (success at node IR). However, in I accident class 3C, the vessel rupture initiating event, the vessel is failed prior to core damage I and relocation. In this case, success is credited if vessel failure does not allow debris j l relocation to the cavity. ' I Success criteria are as follows: I For all accident classes except 3C, success of node DP and node IR results in success I at node VF. l - For accident class 3C, success at node DP and node IR, and maintaining the debris I inside the faulted reactor vessel, result in success at node VF. v ENE Revision: 8 EN raarb September 30,1996 39-13 = W *v 8 W ti m 296
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4 39. In-Vessel Retention of Molten Core Debris O For all accident classes except 3C (vessel rupture initiating event), maintaining the debris in the vessel is ensured by ve.ssel integrity (success at nodes IR and DP). In accident class 3C, the vessel is failed below the intact core as a result of the initiating event. Since vessel rupture produces core damage, regardless of system availability, the failure of ADS and gravity injection has negligible frequency in accident class 3C. Core damage is caused by the inability to reflood the core until the reactor cavity is filled. AP600 has the unique cavity flooding feature that, once the cavity is filled up to the break, water can reflood back into the vessel as the containment compartments fill to arrest core damage before full core relocation. Only a limited amount of debris is likely to relocate to the lower head. The most likely failure for the reactor vessel initiating event is a local failure above the top of the lower head hemisphere at the beltline of the vessel. This location has the highest fluence and brittleness from exposure. Debris relocated into the lower head is guaranteed to be water cooled in the vessel. Therefore, for accident class 3C, a scalar failure probability value of 0.1 for debris relocation is assigned to node VF. A sensitivity to this value is investigated and discussed in Chapter 43, 39.8 Summary . The fault trees and scalar values linked for nodes IR and VF are summarized in Tables 39-2 and 39-3, respectively. 39.9 References 39-1 Theofanous, T.G., et al., "In-Vessel Coolability and Retention of a Core Melt," DOE /ID-10460, July 1995. 39-2 Theofanous, T.G., et al., " Lower Head Integrity Under In-Vessel Steam Explosion Loads," l DOE /ID-10541, June 1996. 39-3 Theofanous, T.G., "On the Proper Formulation of Safety Goals and Assessment of Safety Margins for Rare and High-Consequence Hazards," Reliability Engineering & Systems Safety, Summer 1996. 39-4 Theofanous, T.G., et al., "The Probability of Mark-I Containment Failure by Melt Attack on the Liner," NUREG/CR-6025, November 1993. 39-5 Theofanous, T.G., et al., "The Probability of Containment Failure by Direct Containment Heating in Zion," NUREG/CR-6075. December 1994. I 39-6 Theofanous, T.G., et al., " Premixing of Steam Explosions: PM-ALPHA Verification Studies," l DOE /ID-10504, September 1996. I i 1 39-7 Theofanous, T.G., et al., " Propagation of Steam Explosions: ESPROSE.m Verification i Studies," DOE /ID-10503, August 1996. I l 39-8 Theofanous, T.G., Volume 1 " Appendices E, F, and G to DOE /ID-10541," and Volume 2 - l l " Addenda to DOE /ID-10541. -10503, -10504," October 1997, and Volume 3 " Addenda to ! l DOE /ID-10503,10504," December 1997. O1 Revision: 11 March 1998 W Westingh0USB , o \pra\rev Il\sec39 wpf.Ib-o30298 39-14
- 40. Passio Containment Cooling FM =-
O I CHAPTER 40 l PASSIVE CONTAINMENT COOLING l In this Chapter, the probability of failure of the passive containment cooling event tree node i PC in the containment event tree (CET) is calculated. Figures 40-1 and 40-2 show the passive I containment cooling system (PCS) and containment. l Analysis has been performed to demonstrate that containment cooling by air only maintains I the containment pressure at less than 80 psig. The analysis, initially performed with the i MAAP4 computer code and benchmarked with the W_ GOTHIC computer code using MAAP4 I mass and energy predictions, predicts a maximum containment pressure of approximately 80 I psig. The analysis included the following conservative assumptions: I
- 14-inch cold leg break l
- 1979 decay heat + 2a l - 15.7 psia initial containment pressure I 120'F initial contamment air temperature l
- 102 percent core power I - Outside air temperature of 100'F l
- Passive containment cooling system water failed upon demand G
V l i Figure 40-3 shows the containment pressure prediction for the MAAP4 and WGOTHIC analyses. I At a pressure of 80 psig, the conditional contamment failure probability' is less than 5.0E-4 I (Chapter 42). While passive containment cooling system failure in the Level 2 quantification I is assumed to result in contamment failure, the failure of passive contamment cooling system I water being delivered to the containment shell is not considered to be a containment failure i mechanism as the contamment cooling with only air limits the containment pressurization i sufficiently that the probability of containment failure is very low, less than 5.0E-4. I nerefore, for consideration of the passive contamment cooling system system in the Level l 2 quantification, there is only one failure mode that is assumed: plugging of the drams at the I floor of the annulus around the contamment shell. I If the passive contamment cooling system provides water to cool the containme:nt surface, the I plugging of the drams at the floor of the annulus around the contamment shell would I constitute a failure mode. Then, the water will eventually rise above the 1.6-meter level, I blocking the air inlets into the annulus from the baffle. His will reduce the heat removal I 'The containment failure probability assumes a shell temperature of 331*F, the saturation
,- I temperature corresponding to Service Level C Failure Pressure for the contamment (90 psig).
? V tur- Septem r 30 40-1 m:Pr.;Jtev_ssaac40.wpf:1b 092296
- 40. Passiva Containme:t Cooling O
from the annulus. Although this may not necessarily constitute failure of the containment heat removal, it is pessimistically assumed to do so. l There are two 100-percent drains in the vertical wall of the Shield Building. Surveillance of I these drains is performed every two years. One drain is sufficient to prevent overflowing of the passive containment cooling system to block the air inlet. The probability, q(PC), of failure of the PC event tree node due to the drain plugging is considered to be a rare event due to the following considerations:
- The annulus is shielded from random accumulation of debris that may potentially plug the drains.
! = The drains are located on the Shield Building vertical wall above the annulus floor, and I have screens to prevent small animals from entering the drains.
There are no data on this failure mode. Even if it is assumed that both drains will plug once during the plant lifetime, the failure probability would be SE-04 per week, for a 40-year plant lifetime. I Based on the rarity of this failure mode and engineering judgement, a failure probability of
.0001 is assigned to the PC node, given that a core damage event has occurred.
q(PC) = 1E-04. i O Revision: 11 51 arch 1998 oWaWviI\sec40.wpf Ib4121898 40-2 h [ Westinghouse k -.
- 40. e iv. co:t.i. : cooii.,
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- Westi@ouse o:W\revilwc40.wpf.lM121898 404
- 49. Offsite Dose Evaluation i
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- 49. Offsite Dose Evaluation e
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- k. - - . . . < , -
c ( w TABLE OF CONTENTS (Cont.) Section Title Eage CHAPTERS 46 THROUGH 48 DELETED CHAPTER 49 OFFSITE DOSE EVALUATION 49.1 Introduction ...............................................49-1 49.2 Conformance with Regulatory Requirements . . . . . . . . . . . . . . . . . . . . . . . . . 49-1 49.3 Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-2 49.4 M ethodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 9-2 49.5 Dose Evaluation Results and Discussions . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-6 49.6 Quantification of Site Risk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-7 49.7 Risk Quantification Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-7 49.8 References . . . . . . . . . . . . . .................................. 49-8 CHAPTER 50 IMPORTANCE AND SENSITIVITY ANALYSIS 50.1 Introduction ..............................................50-1 50.2 Importance Analyses for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-1, 50.2.1 Initiating Event Importances (Case 1) . . . . . . . . . . . . . . . . . . . . . . . 50 2 50.2.2 Common Cause Failure Importances (Case 2) . . . . . . . . . . . . . . . . . 50 3 A 50.2.3 Human Error Importances (Case 3) . . . . . . . . . . . . . . . . . . . . . . . . . 50-5 Cl 50.3 50.2.4 Component Importances (Case 4) . . . . . . . . . . . . . . . . . . . . . . . . . . 50-6 System Importances for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-7 50.4 Human Error Sensitivity Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-9 50.4.1 Set Human Error Probabilities to 1.0 (Failure) in Core Damage Results (Case 25) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-9 50.4.2 Set Human Error Probabilities to 0.0 (Success) in Core Damage Results (Case 26) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50- 10 50.4.3 Assess Importance of Increasing Human Error Probabilities by a Factor of 10 (Case 27) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-10 50.5 Other Sensitivity Analyses for Core Damage . . . . . . . . . . . . . . . . . . . . . . . . 50-11 50.5.1 Diesel Generator Mission Time (Case 28) . . . . . . . . . . . . . . . . . . 50- 1 1 50.5.2 Impact of Passive System Check Valves on Core Damage Frequency (Case 29) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50- 12 50.5.3 Instrumentation and Control Cutoff Probability (Case 30) . . . . . . . . 50-12 50.5.4 Containment Recirculation After Safety Injection Line Break j Event (Case 31) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50-12 ; 50.5.5 Quantification Truncation Probability (Case 32) . . . . . . . . . . . . . . . 50-13 50.5.6 Sensitivity to ADS Stage 4 Success Criteria (Case 33) . . . . . . . . . . 50-13 50.5.7 Squib Valve Failure Probability (Case 34) . . . . . . . . . . . . . . . . . . . 50-13 50.5.8 Circuit Breaker Failure Probability (Case 35) . . . . . . . . . . . . . . . . . 50-14 50.5.9 End-State Importances (Case 36) . . . . . . . . . . . . . . . . . . . . . . . . 50-14 s g
/ \ \ )
LJ ENE Revision: 9 DE Et2N, April 11,1997 XXXi 0:WWW 9\ pre-toc.wpf.lb
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TABLE OF CONTENTS (Cont.) Section Title Page 50.6 Sensitivity and Importance Analyses For Large Release Frequency . . . . . . . . 50- 15 50.6.1 Importance Analyses For Large Release Frequency ... . . . . . . . 50- 15 50.6.2 Sensitivity Analyses For Large Release Frequency . . . ........ . 50-21 50.7 Sensitivity Analysis for Offsite Dose Risk . . . . . . . . . . . . . . . . ......... 50-22 50.8 Results S u mmary . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . ... . . . 50-23 Attachment SOA ATWS PRA Sensitivity Case . . . . . ............ ....... 50A-1 CHAPTER 51 UNCERTAINTY ANALYSIS 51.1 Introduction ........................... ... . . .. ....... 51-1 51.2 Methodology . . . . . . . . . . ........ . .... ... ..... . ....... 51-1 51.3 Summary of Results . . . . . . ................... . . . ....... 51-3 51.4 Sensitivity Studies for the Uncertainty Calculations .... .... .. . .... 51-4 51.4.1 Uncertainty in the Cutoff Frequency . . . . . ...... .. ....... 51-4 51.4.2 Uncertainty in the Number of Cutsets Sampled ............... 51-4 51.4.3 Uncertainty in the Mean Failure Probability for Basic Events . . . . . . 51-4 51.4.4 Sensitivity to the Random Number Input for Sampling . . ....... 51-5 51.5 References . . . ............ ... .. ... . . . . . . .. 51-6 CHAPTER 52 RTNSS - FOCUSED PRA SENSITIVITY STUDY 52.1 Focused PRA Sensitivity Study Analysis Method ... . ... . . . . . . . . . 52-1 52.1.1 Core Damage Frequency Calculation .. . ................ . 52-2 52.1.2 Release Frequency Calculation . . . . . . ........ .. ...... 52-5 52.2 At-Power Focused PRA Sensitivity Study . . . . . ............. . ... 52-5 52.2.1 At-Power Focused PRA Sensitivity Study Core Damage Frequency Quantification ............ ... ...... . ... 52-6 52.2.2 At-Power Focused PRA Sensitivity Study Release Frequency Quantification . .. ............ .. . .. . . 52-11 52.3 Shutdown Focused PRA Sensitivity Study . . . ....... .... . .. .. 52-16 52.3.1 Shutdown Focused PRA Sensitivity Study Core Damage Quantification . . . . . . . . . . ......... . . .. 52-16 52.3.2 Shutdown Focused PRA Sensitivity Study Release Frequency Calculation ........ .... ....... ..... . 52-19 52.4 Focused PRA Sensitivity Fire Analysis . . . . . . . . . ...... ........... 52-21 52.5 Focused PRA Sensitivity Study Flooding Analysis . ........... .... 52-21 52.5 1 At-Power Focused PRA Sensitivity Study Flooding Scenarios . . . . . 52-21 52.5.2 Shutdown Focused PRA Sensitivity Study Flooding Scenarios . . . 52-22 52.5.3 Focused PRA Sensitivity Study Flooding Analysis Results Summary . . . . . . . . ...... ........... . .. ..... .. 52-23 52.6 Focused PRA Sensitivity Study Results and Conclusions . . . . . . . . . .... 52-23 52.7 References ............ .......... .......... . . .. .... 52-24 O Revision: 11 ENEL March 1998 mm. 3 Westinghouse o:\prairev.lliptw. toc.wpf;lb XXXii
1 i U TABLE OF CONTENTS (Cont-) Secdon Title P,gge CHAPTER 53 DELETED CHAPTER 54 LOW-POWER AND SHUTDOWN RISK ASSESSMENT 54.1 Introduction .... ........ ......... ....................... 54-1 54.2 Initiating Events ................ ........................... 54-1 54.2.1 Identification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. 54-2 54.2.2 Events Modeled . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-2 54.2.3 Shutdown Phases Summary Description . . . . . . . . . . . . . . ... .. 54-3 54.2.4 Initiating Events for Operanng Modes . . . . . . . . . . . . . . . . . . . . . 54-4 54.2.5 Actuating Signals and Systems Available . . . . . . . . . . . . . . . . . . 54- 15 54.2.6 Scenarios for Detailed Analysis . . . . . .............. . . . . . 54-16 54.2.7 Summary of Initiating Events Analyzed . . . . . . . . . . . . . . . . ... 54-21 54.3 D ata . . . . . . . . . . . . . . . . . . . . . ............................. 54-22 54.3.1 Shutdown Frequency . . ........... ........... .... 54-22 54.3.2 Mission Times . . . . . . . . . . . ..................... ... . 54-25 54.4 Event Tree Development . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-28 54.4.1 Event Tree LOSP-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4-3 0 (n #
) 54.4.2 54.4.3 Event Tree RNS-ND .. . .......................
Event Tree CCW-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-33
. . . . 54-33 J j
54.4.4 Event Tree LOCA-PR-ND ..................... . . . . . . 54-3 3 j 54.4.5 Event Tree LOCA-V24-ND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-34 54.4.6 Event Tree RCS-OD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-36 54.4.7 Event Tree LOSP-D . . ............................... 54-38 54.4.8 Event Tree RNS-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.9 Event Tree CCW-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.10 Event Tree LOCA-V24-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-40 54.4.11 Boron Dilution Events (Reactivity Events) . . . . . . . . . . . . . . . . . . . 54-41 54.4.12 Boron Dilution Events Due to Chemical and Volume Control System Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-45 54.4.13 Endstates Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-48 54.5 Fault Tree Models for Shutdown and Low-Power Events . . . . . . . . . . . . . . . 54-48 54.5.1 Insuumentation and Control Modeling for Shutdown (Levd 1) .... 54-48 54.5.2 Instrumentation and Control Modeling for Shutdown (Level 2) .... 54-51 54.6 Success Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-51 J 54.6.1 MAAP4 Code Analysis for Shutdown Success Criteria . . . . . . . . . . 54-52 54.6.2 MAAP4 Parameter File . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-52 54.6.3 MAAP4 Input Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-54 54.6.4 Definition of MAAP4 Cases From Event Trees . . . . . . . . . . . . . . . 54-55 54.6.5 Results From MAAP4 Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . 54-57 54.7 Common Cause Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-57
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TABLE OF CONTENTS (Cont.) Section .T. itle Page 54.8 Human Reliability Analysis . .. ..... ... .. ......... . . . . . 54-57 54.8.1 Operator Actions Calculated . . . . .. ... .... . . . . . . . . . . 54-5 8 54.8.2 Conditional Human Error Probabilities .... .............. 54-64 54.9 Fault Tree Quantification .. ............. ............ . . 54-64 54.10 Level 1 Core Damage Frequency Quantification . . .. . ........ . . . . 54-67 54.10.1 Core Damage Quantification Method . . . . ......... ..... 54-68 54.10.2. Quantification Inputs . . . . . . . .................. ..... 54-69 54.10.3 Level 1 Shutdown Core Damage Frequency Results . . . . . ..... . 54-70 54.11 Shutdown and Low-Power Release Category Quantification . . . . . . . . . . . . . 54-71 54.12 Shutdown Assessment Importance and Sensitivity Analyses . . . . . . . . . . . . . 54-71 54.12.1 Importance Analyses for Core Damage at Shutdown . . . . . . . . . . . 54-72 54.12.2 Other Sensitivity Analyses for Shutdown Core Damage . .. . . . . 54-77 54.13 Summary of Shutdown Level 1 Results ... . ...... ... . . . . . . . . . . 54-81 54.14 References ........ ...... ............. .. .......... 54-87 Attachment 54A Design Change Impact on Low-Power and Shutdown Risk Assessment .................................... 54A-1 Attachment 54B Surge Line Flooding Effect on Low-Power and Shutdown Risk Assessment ...... ........ .. . . . . .. . 54B-1 l Attachment 54C Effect of Modifications to Safe / Cold Shutdown PRA .... ... 54C-1 l CHAPTER 55 SEISMIC MARGIN ANALYSIS . . ... .. ........ .. . 55-1 55.1 Introduction ... ..... ..... .. . . .. .. ...... ... 55-1 l 55.2 Calculation of HCLPF Values . . . .. . .... .. .... .. .. . 55-2 l 55.2.1 Seismic Margin HCLPF Methodology ... .. .. .. .. ... . 55-2 55.2.2 Calculation of HCLPF Values . . . . .. . .. ... ... .55-2 55.3 Seismic Margin Model ... ..... . .... ........ ... . ... 55-12 55.3.1 SMA Model and Assumptions . . . . . .... . .... . . . . . . . 55-14 55.3.2 Seismic Initiating Events . . . . . . . . . . . . .......... .. . . . 55-16 55.3.3 Initiating Event Category HCLPFs . . ... . ... .... .... . 55-17 55.3.4 Event Tree Models . . . . . . . . . . . . ..... .... ... .. . 55-20 i 55.3.5 Fault Tree Modeling and Quantification . . . . . . . . . ..... .... 55-28 l 55.3.6 Seismic Event Core Damage Sequence Evaluation . . ...... . . 55-36 55.3.7 Containment Performance Model . . . . . . . . ........ . ..... 55-37 55.4 Calculation of Sequence and Plant HCLPF .... ............ .. .. . 55-38 55.4.1 HCLPFs for Basic Events . . . . .. . . .. . . . . . . . . . . . . 5 5-39 55.4.2 Calculation of Initiating Event HCLPFs . .. . . . . . . . . . . . . . 5 5-3 9 55.4.3 Calculation of System Fault Tree HCLPFs . . . . . . . . . . . . .. 55-39 55.4.4 Calculation of Sequence HCLPFs . . . . . . .......... ..... 55-40 55.4.5 Calculation of Plant HCLPF . . . . . . . .... ........... . 55-43 l 55.4.6 Large Release HCLPF . . . . . . . . . .. .. .......... .. . 55-43 Revision: 11 O
' ENEL W85tingh00S8 March 1998 W=i.,,
o$pra\rev llipra-toc.wpfdb XXXiv
1 { C n i v TABLE OF CONTENTS (Cont.) Section Title P. age 55.5 Sensitivity Analyses . . . ......................... .......... . 55-46 55.5.1 Robust Fuel and Core Assembly . . . . . . . . . . . . . . . . . . . . . . . . 5 5 -4 8 55.5.2 Credit for Operator Actions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-49 55.5.3 Less Credit for Operator Actions in LOSP Event at 0.09g . . . . . . . . 55-52 55.5.4 72-Hour Mission Time . . . . . . . . . . ............. . . . . . 55-53 55.5.5 Containment Isolation - Smaller Size Valves . . . . . . . . . . . . . . . . 55-56 55.5.6 Steam Generator Tube Rupture Success Criteria . . . . . . . . . . . . . . . 55-57 55.5.7 Steam Line Break Success Criteria . . . . . . . . . . . . . . . . . . . . . . . . 55 58 55.5.8 Seismic Interaction Between Turbine and Auxiliary Buildings . . . . . 55-59 55.6 SMA Results and Insights .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-63 55.9.6 AP600 SMA Results ................................. 55-63 55.9.6 AP600 SMA Insights . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-68 55.7 References ............................................55-70 Attachment 55A System HCLPF Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55A-1 Attachment SSB Sequence HCLPF Calculations . . . . . . . . . . ........ ....... 55B 1 o Attachment SSC Seismic Margin Analysis HCLPF Sensitivity Study ........... 55C-1 i
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CHAPTER 56 PRA INTERNAL FLOODING ANALYSIS , 56.1 Introduction ........ .. ......... ..................... .. 56 1 56.1.1 Definitions ....... ................. ............ . . 56-1 l 56.2 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......... ...... 56-1 j 56.2.1 Summary of Methodology . . . . . . . . . . . . . . . . ......... . . . . 56-1 j 56.2.2 Information Collection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-2 ' 56.2.3 Initial Screening Assessment . . . . . . . . . . . . . . . . . . . . . . . .. . . 56-3 56.2.4 Detailed Screening Assessment ...........................56-4 56.2.5 Identification of Flood. Induced Initiating Events .............. . 56-6 I 56.2.6 Initiating Event Frequencies ........ . . . . . . . . . . . . . . . . . . . . 5 6-7 56.3 Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6-7 56.3.1 General Flooding Analysis Assumptions and Engineering Judgments . 56-7 56.3.2 AP600-Specific Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-9 56.4 Information Collection . . . . ..................................56-11 56.4.1 PRA-Modeled Equipment and Locations . . . . . . . . . . . . . . . . . . . 56-11 56.4.2 Identification of Areas for Flooding Evaluation . . . . . . . . . . . . . . 56- 1 1
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56.5 At-Power Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-12 ;
$6.5.1 Initial Screening Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6- 1 2 1 56.5.2 Detailed Screening Assessment .................. . . . . . . . 56-12 56.5.3 Identification of Flood. Induced Initiating Events . . . . . . . . . . . . . . 56-28 56.5.4 Calculation of Flood-Induced Initiating Event Lequencies . . . . . . . . 56-32 56.5.5 Quantification of At-Power Flood-Induced Events ...... . . .. 56-39 .
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HR O TABLE OF CONTENTS (Cont.) Section I.itig bag 56.6 Shutdown Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6-41 56.6.1 Detailed Screening Assessment .......... . . . . . . . . . . . . . . . 5 6-41 56.6.2 Identification of Flood Induced Initiating Events . . . . . . . . . . . . . . 56-42 56.6.3 Calculation of Flood-Induced Initiating Event Frequencies . . . . . . . . 56-43 56.6.4 Shutdown Quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-48 56.7 Seismically Induced Flooding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56-51 56.8 Flooding Hazards During Refueling Outages . . . . . . . . . . . . . . . . . . . . . . . . 56 52 56.9 Flooding Sensitivity Study ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6-5 2 56.9.1 Flooding Human Error Probabilities Sensitivity Study . . . . . . . . . . . 56-52 56.10 Summary of Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 53 CHAPTER 57 INTERNAL FIRE ANALYSIS 57.1 Introduction ...................... ....... ................57-1 57.2 Qualitative Analysis Methodology . . .... ....... ................57-2 57.3 Quantitative Methodology of Fire Area Frequency . . . . . . . . . . . . . . . . . . .. 57-6 57.3.1 Fire Frequency Calculations .......................... . . 57-6 57.3.2 Fire Damage Category Quantification . . . . . . ... . . . . . . . . . . . . 57-7 57.4 Core Damage Quantification Methodology . . . . . ....... . . . . . . . . . . . 5 7- 10 57.5 Fire Analysis Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 7 12 57.5.1 Qualite.tive Analysis Assumptions and Other Modeling Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57- 12 57.5.2 Quantification Assumptions And Modeling Considerations . . . . . . . 57-14 57.6 At. Power Qualitative Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 7- 17 57.7 At-Power Quantitative Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 7- 19 57.7.1 Fire Ignition Frequencies for Quantitative Analysis . . . . . . . . . . . . . 57-19 57.7.2 Fire Damage Category Quantification . . . . . . . . . . . . . . . . . . . . . 5 7- 19 57.7.3 Individual Area PRA Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-19 ) 57.8 Control Room Fire Analysis - Power Operation . . . . . . . . . . . . . . . . . . . . . . 57 22 ) 57.8.1 Description of the Control Room and Associated Fire Protection . . . 57 22 57.8.2 Alternate Shutdown Capability . . . . . . . . . . . . . . . . . . . . . . . . . . 57 23 57.8.3 Fire Hazard Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 24 57.8.4 AP600 Control Room Fire Evaluation . . . . . . . . . . . . . . . . . . . . . . 57-25 I 57.8.5 Fire Scenario Identification and Frequency Determination . . . . . . . . 57 29 ) 57.8.6 Control Room Fire Scenario Quantification and Results ..... ... 57-32 57.9 Shutdown Fire Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-33 57.9.1 Fire Ignition Frequencies during Shutdown Modes of Operation . . . . 57-33 57.9.2 Fire Damage Category Quantification . . . . . . . . . . . . . . . . . . . . . . 57 34 , 57.9.3 Individual Area PRA Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . 57-34 i 57.9.4 Fire Analysis for Safe Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . 57-34 l 57.9.5 Fire Analysis for Mid-Loop Operation . . . . . . . . . . . . . . . . . . . . . 57-41 l 9i Revision: 9 April 11,1997. [ Westinghoure o wepawvppm4acupt.lb uxvi ! i l J _.__J
iBi A .b TABLE OF CONTENTS (Cont.) Section Title bage 57.10 Summary and Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 7-4 2 57.10.1 At-Power Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 7-4 2 57.10.2 Shutdown Fire Analysis . . . . . . . . . . . . . . . . . .... .... . . . 57-45 57.10.3 Conclusions . . . . ................ .. ............. . 57-47 57.11 References . . . . . . . . .............. .......... . . . . . . . . . . . . 5 7-4 8 ATTACHMENT 57A DEFINITIONS ................................. .. 57A-1 ATTACHMENT 57B DESIGN CHANGE EFFECT ON INTERNAL FIRE ANALYSIS . . 57B-1 CHAPTER 58 WINDS, FLOODS, AND OTHER EXTERNAL EVENTS 58.1 Introduction ...............................................58-1 58.2 External Events Analysis ................. ....................58-1 58.2.1 Severe Winds and Tomadoes . . . . . . . . . . . . . ............... 58-1 58.2.2 Extemal Floods . . . . . ....................... ... . . . . 5 8-2 58.2.3 Transportation and Nearby Facility Accidents . . . . . . . . . . . . . . 5 8-2 58.3 Conclusion . ..................... ... .. ... ......... .. 58-3 58.4 Re ferences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58-3 m /) V CHAPTER 59 PRA PESULTS AND INSIGHTS 59.1 Introduction ......... ................................. . . . 59-1 59.2 Use of PRA in the Design Process ... ....... ............. .. . . 59-3 59.2.1 Stage 1 - Use of PRA During the Early Design Stage . . . . . . . . . . . 59-4 I 59.2.2 Stage 2 - Preliminary PRA . . . . . . . . . . . . . . . .......... . . 59-5 59.2.3 Stage 3 - AP600 PRA Submittal to NRC (1992) . . . . . . . . . . . . . . . 5 9-7 59.2.4 Stage 4 - PRA Revision 1 (1994) . . . . . . . . . . . . . . . . . . . . . . . . . . 59-8 3 59.2.5 Stage 5 - r'RA Revisions 2-8 (1995-1996) . . . . . . . . . . . . . . . . . . . . 59-8 l 59.3 Core Damage Frequency from Intemal Initiating Events at Power . . . . . . .. 59-10 59.3.1 Dominant Core Damage Sequences .......................59-12 59.3.2 Component Importances for At-Power Core Damage Frequency . . . 59-44 59.3.3 System Importances for At-Power Core Damage . ....... . . . . 59-44 59.3.4 System Failure Probabilities for At-Power Core Damage . . . . . . . . . 59-45 59.3.5 Common Cause Failure Importances for At-Power Core Damage . . . 59-45 59.3.6 Human Error Importances for At-Power Core Damage . . . . . . . . . . 59-45 59.3.7 Accident Class Importances . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59-47 59.3.8 Sensitivity Analyses Summary for At-Power Core Damage -. . . . . . 59-47 59.3.9 Summary of Important Level 1 At-Power Results . . . . . . . . . . . . . . 59-48 i 59.4 Large Release Frequency for Internal Initiating Events at Power . . . . . . . . . . . 59-51 ; 59.4.1 Dominant Large Release Frequency Sequences ............. . 59-52 59.4.2 Sensitivity / nalyses for Containment Response . . . . . . . . . . . . . 59-72 l 1 ,- m l ENE Revision: 11 DD88 CW.W March 1998 XxXyji oVab.llWa-toc *ptib J
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TABLE OF CONTENTS (Cont.) 0 Section Title .P,gite 59.4.3 Comparison of Initiating Event Importances for Core Damage Frequency and Large Release Frequency . . . . . . . . . . . . . . . . . . . . 59-72 59.4.4 Summary of Important Level 2 At-Power Results . . . . . . . . . . . . . 59-73 59.5 Core Damage and Severe Release Frequency from Events at Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . ............. . . . . 59-75 59.5.1 Summary of Shutdown Level 1 Results . . . . . . . . . . . . . . . . . . . . . 59-75 59.5.2 Large Release Frequency for Shutdown and Low-Power Events . . . . ............ ... ........... .. . . . . 59-81 59.5.3 Shutdown Results Summary . . . . . ...... ......... . . . . 59-82 59.6 Results from Intemal Flooding. Internal Fire, and Seismic Margin Analyses . . . . . . . . . . . . . . . . . . .... ............. ....... . . . 59-82 59.6.1 Results of Internal Flooding Assessment . . . . . . . . . . . . . . . . . . . 59-82 59.6.2 Results ofInternal Fire Assessment . .. . ................59-83 59.6.3 Results of Seismic Margin Analysis . . . . . . . . . . . . . ........ 59-87 59.7 Plant Dose Risk from Release of Fission Products . . .... .. .. 59-87 59.8 Overall Plant Risk Results . . . . . . . . . . . . . . ........... .. ... .. 59-88 59.9 Plant Features Important to Reducing Risk .. . ... ..... . . . . . . . . . 5 9-89 59.9.1 Reactor Design . . . . . . . . . . . . . . . . . . ....... ... . . . . 59-90 59.9.2 Systems Design . ........ .... . .. .. . . . . . . . . 5 9-91 59.9.3 Instrumentation and Control Design . . . . ...... ... . . . . . . 59-94 59.9.4 Plant Layout . . . . . .. ..... .... .... .. . . . . . . . . 5 9-95 59.9.5 Plant Structures . . . . . . . . ...... ..... . .. . . . . . . 59-96 59.9.6 Conte!nment Design . . . . . .... ............... . . . 59-96 59.10 PRA Input to the Design Certification Process . . . . ..... ... . 59-101 59.10.1 PRA Input to Reliability Assurance Program . . . . . . .. . 59-102 59.1u.2 PRA Input to Certified Design Material .. . .. . . . . . . 59-102 59.10.3 PRA Input to the Technical Specifications . . . . . . . . . . ... . . 59-102 59.10.4 PRA Input to MMI/ Human Factors / Emergency Response Guidelines . . . . ...... .... ........ .. .. . . . . . . 59-102 59.10.5 Summary of PRA-Based Insights . . . . . . . . . . . . . . . . . . . . . . 59-103 59.10.6 Combined License Information ........ .. .. . ... 59-103 APPENDIX A MAAP4 ANALYSIS TO SUPPORT SUCCESS CRITERIA . . .. . .. . . . A-1 APPENDIX B EX VESSEL SEVERE ACCIDENT PHENOMENA . . ... . .. . . B-1 APPENDIX C DESIGN CHANGES THAT OCCURRED AFTER THE PRA ANALYSES WERE COMPLETED . . . . . . . . . . . . ........ ...... ... . . . . . . . C- 1 APPENDIX D EQUIPMENT SURVIVABILITY ASSESSMENT . . . . . . . . . . . . .. . . . . . . D- 1 O A as h 1998 L, 3 Westiligh00Se l c:Wa\rev.llWtoc.wpf Ib xxxviii
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LIST OF TABLES (Cont.) j i Table No. Title P_ age 57-25 Quantitative Summary - Control Room Fires During Safe Shutdown ........ ............. ..................... 57-138 57-26 Summary of Qualitative Evaluation Results - Mid-Loop Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57-139 57 27 Summary of Quantitative Results - Mid-Loop Operation . . . . . . . . . . . . . . 57- 146 28 Quantitative Summary Control Room Fires During Mid-Loop Operation . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . 57-154 57-29 Safe Shutdown Containment Single-Hot-Short LOCA Sensitivity . . . . . . 57-155 59-1 Contribution of laitiating Events to Core Damage . . . . . . . . . . . . . . . . . . 59-104 59-2 Conditional Core Damage Probability of Initiating Events . . . . . . . . . . . . . . 59-105 59-3 Intemal Initiating Events at Power Dominant Core Damage Sequences . . . . . 59-106 59-4 Sequence 1 - Safety Injection Line Breat Dominant Cutsets (SI-LB-02) . . . . 59-108 59-5 Sequence 2 - Intermediate LOCA Dominant Cutsets (NLOCA-03) . . . . . . . . 59-113 59-6 Sequence 3 - Large LOCA Dominant Cutsets (LLOCA-06) . . . . . . . . . . . 59-118 59-7 Sequence 4 - Large LOCA Dominant Cutsets (LLOCA-03) . . ....... .. 59-124 59-8 Sequence 5 - Reactor Vessel Rupture Cutset (RV-RP-02) . . . . . . . . . . . . . 59-130 7 59-9 Sequence 6 - Large LOCA Dominant Cutsets (LLOCA-11) . . . . . . . . . . . . 59-131 ( i 59-10 Sequence 7 - ATWS Dominant Cutsets (ATWS-28) . . . . . . . . . . . . . . . . . . 59-133
'd 59-11 Sequence 8 - Medium LOCA Dominant Cutsets (MLOCA-03) . . . . . . . . . . 59-141 59-12 Sequence 9 - ATWS Dominant Cutsets (ATWS-13) . . . . . . . . . . . . . . .. . . . 59-146 59-13 Sequence 10 - Intermediate LOCA Dominant Cutsets (NLOCA-04) . . . ... 59-151 59-14 Sequence 11 - Safety Injection Line Break Dominant Cutsets (SI-LB-03) . . . 59-156 59-15 Sequence 12 - Small LOCA Dominant Cutsets (SLOCA-03) . . . . . . . . . . . 59-160 59-16 Sequence 13 - Core Makeup Tank Line Break Dominant Cutsets (CMTLB-03) 59-165 59-17 Sequence 14 - Steam Generator Tube Rupture Dominant Cutsets (SGTR-07) . 59-170 59-18 Sequence 15 - Steam Generator Tube Rupture Dominant Cutsets (SGTR-23) . 59-171 59 19 Sequence 16 - Large LOCA Dominant Cutsets (LLOCA-02) . . . . . . . . . . 59-177 59-20 Sequence 17 - Large LOCA Dominant Cutsets (LLOCA-05) . . . . . . . . . . . . 59-183 59-21 Sequence 18 - Consequential SGTR Dominant Cutsets (SGTRC-03) . . . . . . . 59-189 59-22 Sequence 19 - Intermediate LOCA Dominant Cutsets (NLOCA-16) . . . . . . . 59-195 59-23 Typical System Failure Probabilities, Showing Higher Reliabilities for Safety Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59-201 59-24 Dominant CET Sequences . . . . . . .............................. 59-202 59-25 Comparison of Initiating Event Contribution to Core Damage and Large Release Frequencies ............ ....................... 59-203 59-26 Summary of AP600 PRA Results . . . . . . . . . . . ......... . . . . . . . . . 59-204 59 27_ Comparison of AP600 PRA Results to Risk Goals . . . . . . . . . . . . . . . . . . 59-205 59-28 Site Boundary Dose Risk at 24 Hours . .......................... 59-206 59-29 AP600 PRA Based Insights . . . . . . . . . . . . . . . ....... . . . . . . . . . . . 5 9-207
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LIST OF FIGURES e Firure No. Title E, age 2-1 Core Damage Logic Diagram for Internal Initiators . . . . . . . . . . . . . . . . . . . 2 -5 8 2-2 Core Damage Logic Diagram for Internal Initiators . . ............ .... 2-59 2-3 Core Damage Logic Diagram for Internal Initiators . . . . . . . . . . . . . . . . ... 2-60 2-4 Core Damage Logic Diagram for Internal Initiators . . . . . . . . ........ . . 2-61 4-1 Large Loss-of-Coolant-Accident Event Tree .................. . . . 4-120 4-2 Medium Loss-of-Coolant Accident Event Tree . . . . . . . . . ............ . 4-121 4-3 Core Makeup Tank Line Break Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . 4-122 4-4 Direct VesselInjection Line Break Event Tree . ....... . . . . . . . . 4-123 4-5 Intermediate Loss-of-Coolant Accident Event Tree . . ............ ... 4-124 4-6 Small Loss-of-Coolant Accident Event Tree .. ......... . . . . . . . . . 4-125 4-7 Reactor Coolant System Leak Event Tree ........... ... ..... . . . , 4-126 4 53 Passive Residual Heat Removal Tube Rupture Event Tree . ............ 4-127 4-9 Steam Generator Tube Rupture Event Tree . . . . ....... . . . . . . . . . . 4-128 l 4-10 Reactor Vessel Rupture Event Tree ......... ..... ......... . . . 4-130 l 4-11 Interfacing Systems Loss-of-Coolant Accident Event Tree . . . . . . ..... . 4-131 l 4-12 Transients with Main Feedwater Event Tree .... . . . . . . . . . . . . . . . . . . . 4- 13 2 ! 4-13 Transients with Loss of Reactor Coolant System Event Tree . . ..... . . . . 4-133 1 4-14 Transients with Loss of Main Feedwater Event Tree . . .. .. ...... ... 4-134 4-15 Transients with Core Power Excursion Event Tree . ................ 4-135 4-16 Loss of Component Coolir.g Water SystenVService Water System Event Tree . . . . . . . . . . . . . . . . . . . . . . . .. ............. . . 4-136 4-17 Loss of Main Feedwater Event Tree . ....... .. ... . . . . . . . . . . . . 4- 137 4-18 Loss of Condenser Event Tree . . . . . ....... . . .. . . . . . . . . . . 4- 13 8 4-19 Loss of Compressed Air Event Tree . . ..... ...... ... ...... . 4-139 4-20 Loss of Offsite Power Event Tree ......... ....... ........ . 4-140 ! 4 21 Main Steam Line Break Downstream of Main Steam Isolanon l Valves Event Tree . .. ........ ...... . ............. . .. 4-141 l 4-22 Main Steam Line Break Upstream of Main Steam Isolation Valves Event Tree ... ................. ......... .... . . 4-142 4-23 Stuck-Open Secondary-Side Safety Valve Event Tree . . . . . . ...... .. . 4-143 4-24 Anticipated Transient Without Scram Precursor without Main Feedwater Event Tree . . .. ....... . . ............ . . 4-144 l 4-25 Anticipated Transient Without Scram Precursor with Injection Evem Tree . . . . . . . . ....................................4-146 4-26 Anticipated Transient Without Scram Precursor Transients with Main Feedwater Event Tree . . . . . . . . . . ..................... ... 4-147 O Revision: 9 ENEl. April 11,1997 wn:::h,, M$0m o \apsprevev.9ps lor wpr.ib 1x
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LIST OF FIGURES (Coat.) Firure No. Ihlt P.ARE 54-4 LOCA/RNS Pipe Rupture During Hot / Cold Shutdown (RCS Filled) Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-302 54-5 LOCA/RNS-V024 Opens During Hot / Cold Shutdown (RCS Filled) Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54-303 54-6 Overdraznmg of Reactor Coolant System During Draindown to Mid. Loop . . . 54-304 54-7 Loss of Offsite Power (RCS Dramed) Event Tree . . . . . . . . . . . . . . . . . . . 54- 305 54-8 Loss of RNS Initiator (RCS Dramed) Event Tre. . . . . . . . . . . . . . . . . . . . . 54-306 54-9 Loss of CCW/SW Initiator (RCS Dramed) Event Tree . . . . . . . . . . . . . . . . 54-307 54-10 LOCA/RNS V024 Opens (RCS Drained) Event Tree . . . . . . . . . . . . . . . . . 54-308 54-11 Accumulator Injection (Dilution Scenario) Event Tree . . . . . . . . . . . . . . . . . 54-309 54-12 Shutdown Transient Case SDIB2 RCS Pressure vs. Tune . . . . . . . . . . . . . . 54-310 54-13 Shutdowr. Transient Case SDIB2 Mass Flow Rate vs. Time . . . . . . . . . . . . 54-311 54-14 Shutdown RNS Break Case SD3A (3500 gpm) . . . . . . . . . . . . . . . . . . . . . 54-312 54-15 Shutdown RNS Break Case SD3A2 (2000 gpm) . . . . . . . . . . . . . . . . . . . . 54-313 54-16 Shutdown RNS Break Case SD3A3 (1000 gpm) . . . . . . . . . . . . . . . . . . . . 54-314 54-17 Shutdown Plant Damage State Substate Event Tree for LP-ADS . . . . . . . . . 54-315 54-18 Shutdown Plant Damage State Substate Event Tree for LP-1 A . . . . . . . . . . . 54-316 54-19 Shutdown Plant Damage State Substate Event Tree for LP-3D . . . . . . . . . . . 54-317 Shutdown Plant Damage State Substate Event Tree for LP-3BR . . . . . . . . . . 54-318 p 54-20 d 54-21 Shutdown Plar.: Damage State Substate Event Tree for LP-3BE . . . . . . . . . . 54-319 55-1 Seismic Initiating Event Hierarchy Tree . . . . . . . . . . . . . . . . . . . . . . . . . . 55-105 55-2 EQ.STRUC Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-106 55-3 EQ-RVFA initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-108 55-4 EQLLOCA Initiating Event Fault Tree .......................... 55-109 55-5 EQ-SLOCA Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-110 55-6 EQ-NIWS Initiating Event Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-111 55 7 EQ.STRUC Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-112 55-8 EQ-RVFA Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-113 55-9 EQ-LLOCA Event Dee . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-114 55-10 EQ.SLOCA Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-115 55-11 EQA*IWS Event Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-116 55-12 EQ-LOSP Event Tree . . . . . . . . . . . . - . . . . . . . . . . . . . . . . . . . . . . . . . . 55-117 55-13 EQ-LOSP Event Tree (for 0.5g level canhquake) . . . . . . . . . . . . . . . . . . . . 55-118 55-14 EQ-AC2AB Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-119 55-15 EQ XCIC Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-120 55-16 EQ-XADMA Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-121 55-17 EQ XIW2A Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-122 55-18 EQ-RECIR Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-123 55 19 EQ-CM2SL Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55-124 55-20 EQ-ADA Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-125 55-21 EQ-IW2AB Fault Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5-126
\ %J g Revision: 9 IEtas April 11,1997 tmi . % %. w ,ca L _______
m: ne w .g LIST OF FIGURES (Cont.) O
. Fleure No. Title P_ age 55-22 EQ-PRHR Fault Tree . . . . .... ....................... . . . . . 55-127 55-23 EQ-PRESU Fault Tree . . . . . . ...............................55128 55-24 EQ-PMS Fault Tree . . . . . . . ...... . . . . . . . . . . . . . . . . . . . . . . . . . 5 5- 129 55-25 EQ-DC Fault Tree . . . . . . . . . . . ... .............. . . . . . . . . . . 5 5-130 55-26 Class IE dc Power Block Diagram . . . . . . . . . . . . . . . . . . . .. . . . . . . . 55-131 55-27 Containment Evaluation Model . . . . . . . . ......................55-132 55-28 EQ-STRUC Event Sequences . . . . . ...................... ... 55-133 55 29 EQ.RVFA Event Sequences . . . . . . .................. . . . . . . . . 55-134 55-30 EQ-LLOCA Event Sequences . . . . . . . . ........... ..... .. ... 55-135 55-31 EQ-SLOCA Event Sequences . . . ......... . . . . . . . . . . . . . . . 5 5- 136 55 32 EQ-SGTR Event Sequences ...................... . . . . . . . . . 55-137 55-33 EQ-SLB Event Sequences . . . ... .................. . . . . . . . . 55-138 55-34 EQ-ATWS Event Sequences . . . . .... . . . . . . . . . . . . . . . . . . . . . . . . 5 5- 139 55-35 EQ-LOSP Event Sequences (for 0.5g level earthquakes) . . . . . . . . . . . . . . . 55-140 56-1 Flood Zones and Barriers Plan at 66'-6" . . . . ......... . . . . . . . . . . . . 5 6-93 56-2 Flood Zones and Barriers Plan at 82*-6" . . . . . . . . . . . . . . . . . . . . . . . . . 5 6-9 5 56-3 Flood Zones and Barriers Plan at 96'-6" . . . . . . . . . . . . . . . . . . . . . ..... 56-97 56-4 Flood Zones and Barriers Plan at 100'-0" & 107'-2" . . . . . . . . . . . . . . . . 5 6-99 56-5 Flood Zones and Barriers Plan at i17'-6" . . . . . . . . . . . . . . . . . . . . 56- 101 56-6 Flood Zones and Barriers Plan at 135'-3" . . . . . . . . . . . . . . . . . . . . . . .. 56-103 56-7 Flood Zones and Barriers Plan at 160'-6" & 153'-0" . . . . . . . . . . . . . . . . 56-105 56-8 Flood Zones and Barriers Plan at 160'-6" & 180'-0" . . . ........ . . . . 56-107 56-9 8-in. Fire Main Rupture at-Power Event Tree ........ ..... . . . . . . . 56-109 56-10 8-in. Fire Main Rupture during Hot / Cold Shutdown Event Tree . ..... .. 56-110 56-11 8-in. Fire Main Rupture during RCS Drained Conditions Event Tree .. .. . 56-111 57-1 Fire Progression Event Tree for 1200 AF 01 Fire Art.a . .... . . . . . . . 57-156 59-1 Contribution of Initiating Events to Core Damage ...... . .. . . . . . . 59-233 59-2 Contribution of Initiating Events to Large Release Frequency and Core Damage Frequency . . .. .. ......... . .. . .... .. 59-234 59-3 Total Plant CDF/LRF , . . . .. ... .. ...... ... ... . . . . . 59-235 59-4 24-Hour Site Boundary Dose Cumulative Frequency Distribution . . . . . . . . . 59-236 Revislan: 11 O
March 1998 owuw==. uw.u-pr*
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50A. ATWS PRA Sensitivity Case f (3
\j' I ATTACHMENT 50A I
I ATWS PRA SENSITIVITY CASE I I In the PRA, when anticipated transient without scram (ATWS) occurs due to failure to insert I the control rods following a demand, the ATWS event trees are modeled to have at least I short-term automatic reactor coolant system (RCS) cooling with passive residual heat removal I (PRHR) or startup feedwater (when main feedwater is unavailable). If short-term cooling is I available, then the RCS pressure control is addressed. This control requires that the i pressurizer safety valves operate and the plant is not in the unfavorable exposure time (UET) I ofits fuel cycle. After both of these conditions are successful, namely short-term cooling and I pressure control of the RCS, then the next phase of the ATWS mitigation is addressed. , I l A sensitivity analysis was performed on the ATWS success criteria and the UET value, in I the sensitivity case, ATWS event trees are modeled to have at least short-term automatic RCS I cooling with only PRHR (when main feedwater is unavailable) and also requires operation of I the core makeup tanks (CMTs). A UET of 3 percent is assumed, based on LOFTRAN l analyses (Reference 50A-1) of ATWS events. I I The changes made for this sensitivity analysis include the following: I (g d t i I
- a. The UET is defined as the percentage of full-power core life where the moderator temperature coefficient is not bounded by that assumed in the LOFFRAN analyses I where acceptable results are shown (i.e., RCS pressure less than 3200 psig following ;
I an ATWS). For this sensitivity case, the UET is 3 percent. j i 6 I b. No credit is taken in an ATWS event for the operation of the startup feedwater system. I I c. The ADALT success criteria associated with the DP node of the containment event tree I for end state 3A is moved to the PRHR2 node in the ATWS event tree. ADALT I success criteria requires CMT injection (which also needs successful stopping of the I reactor coolant pumps), PRHR operation, and no consequential steam generator tube I rupture (SGTR). l l d. With item c implemented, the DP node in containment event tree for end state 3A is I assigned a failure probability of 1.0, which means that all 3A sequences go to I containment bypass and contribute to the large release frequency. This is an overly I conservative assumption because it ignores the possibility that the passive safety I systems and passive containment cooling system could mitigate the consequences of the i l ATWS, or at least prevent containment bypass. I I 'Ihe AP600 core damage frequency (CDP) and large release frequency (LRF) results for this l ATWS sensitivity case are summarized in Table 50A-1. The conclusion from the sensitivity I I case is that the change to the core damage frequency and large release frequency is negligible, 73 Revisioru 11 EMM M EL mtumb March 1998 50A-1 ob r ;e.iar:50awpf:1b ! I
"iE 50A. ATWS PRA S:nsitivity Ccss I and the impact on containment effectiveness is minimal (e.g., an absolute decrease of O .
I 1 percent). I I Reference I l 50A-1 B.A. McIntyre, Westinghouse, to T.R. Quay, NRC, "AP600 Response to FSER Open items," I Westinghouse letter DCPiNRCl227, dated January 23,1998. O i i l Revision: 11 O l March 1998 _ T Westinghouse obem;,,uithwpt:Ib 50A-2 f
m-50A. ATWS PRA Senzitivity Case (q
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l Table 50A-1 l l ATWS SENSITIVITY CASE RESULTS l Sensitivity Case Base Case i ATWS CDF 5.4E-09/yr 1.0E-08 /yr l AP600 CDF (internal events at 1.65E-07/yr 1.7E-07/yr i power) I % Contribution of ATWS to CDF 3.3 % 5.9 % 1 3A Plant Damage Class Frequency 5.4E-09/yr 1.0E-08/yr I (from ATWS) l LRF from 3A Plant Damage Class 5.4E-09/yr 4.2E-09/yr l Total LRF (internal events at power) 1.9E-08/yr 1.8E-08/yr I % Contribution of ATWS to LRF 2.8 % 2.3 % 7 kj s l Containment Effectiveness (Ceff) 88.2 % 89.2 % I Conditional Containment Failure 11.8 % 10.8 % I Probability (CCFP) i A k ]
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54A. Design Change hopact on Low Power and Shutdown Risk Assessment t v) i Having made changes to the basic events described above, the quantification output file is i recalculated. The resulting core damage frequency is 3.4E-08 events per year, which is I approximately 60 percent of the baseline core damage frequency of 5.5E-08. I I Unscheduled Maintenance Unavailability I l Based on the NRC request to assess the impact of unscheduled maintenance on the IRWST l during drained condition, unavailability of one IRWST train due to unscheduled maintenance i is incorporated into the model. Incorporating the train maintenance unavailability was I accomplished by setting basic event IRWMOD05S to 3.0E-03; the probability of a safety-i related system being in maintenance. I I Module IRWMOD05S previously represented failure of the MOV in the IRWST Train A to I open due to mechanical failure. As stated above, due to the design change, this failure mode i no longer exists. Consistent with the new design configuration, module IRWMOD06S I (associated with mechanical failure of the MOV in IRWST Train B)is set to zero. I I The train maintenance unavailability is then added to the new cutset file obtained from the ] I recalculation in the previous section. The resulting core damage frequency remains at 3.4E-08 1 l events per year.
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I g) t I l ne top 500 dominant cutsets from the quantification output file reflecting the revised basic event probabilities associated with the new IRWST configuration and maintenance l unavailability are shown in Table 54A-1. As stated above, the core damage frequency is I shown to be 3.4E-08 events per year. The associated component importance file is shown in I Table 54A-2. l l Single IRWST Train Operation Sensitivity I I ne current revision of Technical Specification 3.5.6 allows one of the IRWST injection trains I to be out of service during reduced inventory conditions. This Technical Specification i provision is to allow for unscheduled repairs of components in any one of the IRWST I injection trains during drained conditions. His provision is not intended to allow plant i personnel '.o schedule or plan maintenance activities to be conducted during reduced inventory I conditions, and, therefore, should not be misconstmed as such. I i However, a sensitivity study is performed to estimate the core damage frequency if one train i of the IRWST is failed. For this conservative estimate of the new IRWST design I configuration, one IRWST train is assumed to fail during filled and drained shutdown 1 l conditions. That is, it is assumed that the IRWST has only one injection line by essentially I setting the other train failure probability equal to 1.0. i l A G ENE Revision: 9 l UDM WaW April 11,1997 54A-5 mWA9*t54WP f:Ib _--_a
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54A. Design Change impact on Low-Pcw:r and Shutdowa Risk Assessment This sensitivity is performed by modifying the following basic events in the quantification e output file: IWACV122AO = 1.0; IWACV123AO = 5.8E-04; IWACV124AO = 1.0; IWACV125AO = 5.8E-04; IWBCV122AO = 1.75E-03; IWBCV123AO = 5.8E-04; I IWBCV124AO = 1.75E-03; IWBCV125AO = 5.8E-04;IWX-CV-AO = 3.1E-05;IWX-1 MV-GO! = 3.8E-05; IRWMOD05S = 0; and IRWMOD06S = 0. Requantification of the cutset file with these changes provides a core damage frequency of l 2.0E-07 events per year. Since this evaluation is a sensitivity study of the main exercise in assessing the reliability of the new IRWST design, the output file for this case is not included in this report. This estimate of the core damage frequency at shutdown is believed to be conservative for the following reasons:
= One train of the IRWST is assumed to fail during both drained and non-drained conditions.
l l
- Approximately 61 percent of the co,e damage frequency of 2.0E-07 events per year is attributed to dominant failures that include failure of the IRWST injection line strainer (i.e., the top two cutsets total 1.2E-07, and these cutsets include basic event i IWB-PLUG).
1
= Re IRWST water is considered to be clean water, suitable for injection into the RCS.
- Re IRWST is covered to preclude material that could clog the strainer from entering the tank.
- De IRWST strainer is assigned a conservative failure probability of 2.4E-04 in Revision 6 of the PRA. The plugging of the IRWST strainer is expected to be a much less probable event than reflected in a failure probability of 2.4E-04.
Therefore, failure of the IRWST line strainer is more dominant than would be expected for failure of a strainer in a clean water system. If a more realistic failure probability is used for strainer plugging, the resulting core damage frequency of the model having a single IRWST train configuration should be much closer to the core damage frequency of 5.5E-08 events per year. His is obtained from the baseline shutdown model of PRA, Revision 6. 54A.2 IRWST Recirculation Paths Changed from 10-inch and 4-inch Lines to 6-inch Lines This design change has no impact on the shutdown core damage frequency calculation, since the recirculation function is not credited in the Level 1 shutdown PRA. (See note in Section 54A.3 below.) O Revision: 11 March 1998 W Westilighouse c4rs'Rev_1I\attS4a wpf Ib-021898 54A-6
54B. Surge Line Flooding Effect on Low P:w;r and Shutdowa Risk Assessment
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( ) 1 l ATTACHMENT 54B l l SURGE LINE FLOODING EFFECT ON LOW-POWER AND SHUTDOWN RISK l ASSESSMENT l I 54B.1 Assessment of Shutdown PRA Including Effect for Surge Line Flooding I l An examination of the surge line flooding concems was conducted on the AP600 design to I determine the capability of the IRWST gravity injection function during reduced inventory I conditions if normal residual heat removal (RNS) cooling is lost. l l 'Ihe results of that examination indicated that, if RNS cooling is lost during reduced inventory j l conditions with the reactor coolant system open, a vent path through the ADS 4th stage is ' I required to preclude the occurrence of surge line flooding and thereby not affect gravity 1 injection. j i i I i Therefore, to address surge line flooding, Westinghouse has included the requirement for I l-out-of-4 ADS stage 4 valves to be opened, in the success efiteria for events during reduced I inventory conditions. This success criterion for 1-out-of-4 ADS stage 4 valves to be opened I is determined by analyses documented in Section 4.8 of the AP600 Shutdown Evaluation i l Report, WCAP-14837, (Reference 54B-1). Technical Specification 3.4.14 allows for 2 of the A l 4 ADS stage 4 valves to be out of service during reduced inventory conditions. This technical () I l specification provision is to allow for unscheduled repairs of as many as two ADS stage 4 valves during drained conditions. I I With the opening of ADS fourth stage, the RCS depressurizes within 24 hours, and the I containment sump recirculation function is required. Thus, given the requirement for
! operation of the ADS during reduced inventory conditions, the containment sump recirculatio" I function is also included in the reduced inventory accident seq.iences when both the ADS and I graxity injection are successful. The success criterion for containment sump recirculation I requires operation of 1-out-of-2 trains; this success criterion is consistent with the basic l success criteria for operating systems during reduced inventory (mode 5) conditions when I decay heat level is expected to be very low. The affected reduced inventory accident I sequences show success of containment sump recirculation leading to the "OK" end-state, and I failure of containment sump recirculation leading to core damage end-state "LP-3BL" or l "LPCBP."
l l The baseline and focused shutdown PRA as reported in Revision 6 is requantified o include I the effects of the additional ADS success criteria for surge line flooding and containment I sump recirculation success criteria for accident mitigation. This requantification includes 1 Levels 1 and 2 of the PRA. The results of this evaluation are used to derive the shutdown I PRA insights, as opposed to the results reported in shutdown PRA Revision 6. 77 ( )
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ENE Revision: 11 Mif$0ilS8 mt=&, March 1998 54B-I ov*evi nwt54b wpr Ib
in 4 1 54B. Srrgi Line Flooding Effect on Low P:wer cnd Shutdown Risk Assessmelt I All shutdown drained cases are affected by the surge line flooding issue. These cases are as O i follows: { l i IEV-RCSOD RCS overdraining when entering mid-loop condition i IEV-LOSPD Loss of offsite power during mid-loop operation
! IEV-RNSD Loss of RNS during mid-loop operation I IEV-CCWD Loss of CCS/SWS during mid-loop operation ! IEV-LOCA24D LOCA through RNS-V024 during mid-loop operation i
I The method of evaluating the effect of these success criteria on the shutdown PRA involves I the following process: I l a. Include ADS stage 4 and containment sump recirculation functions in the shutdown i event trees for drained conditions. The revised event trees are provided in l Figures 54B-1 through 54B-5. I I b. Construct two ADS fault trees: ADASD for transient and loss of coolant accident cases I (i.e., with offsite power available), and ADASDP for the station blackout case. The i success criteria summary tables for these fault trees are provided in Tables 54B-1 and l 54B-2. I l c. Construct ADS support system fault trees for the fault tmes developed in item b. l d. Construct two containment sump recirculation fault trees: RECIRCS for transient and loss of coolant accident cases, and RECIRBS for the station blackout case. The success 9 I I criteria summary tables for these fault trees are provided in Tables 54B-3 and 54B-4. I I e. Construct containment sump recirculation suppo:t system fault trees for the fault trees I developed in item d. I I f. Construct initiating event sequences for the reduced inventory cases to include ADS and i containment sump recirculation success and/or failure, as appropriate. These sequences I used the fault trees for the IRWST, RNS, CCS, and SWS that are reported in the I shutdown PRA, Revision 6, without modification. I l g. Add success of the operator action to stop RCS drain-down in applicable sequences for I the RCS overdrain initiating eve.et (RCSOD); this is represented by basic event RHN-l MAN 04-SUCC. l l h. Group the non-drained initiating event sequences from shutdown PRA, Revision 6, with I the newly constructed initiating event sequences for the drained cases, discussed in I items f and g above. Quantify this group of sequences, using the same truncation limits I l as used for the Revision 6 shutdown PRA, to assess the effect of the changes in the I drained sequences on the baseline and focused shutdown core damage frequencier. Revision: 11 O ENEL
-~~ W WB5tingh00S8 March 1998 oNravev11wt54b wpr:Ib.02259s 54B-2 L
548. Surge Line Flooding Effect on Low P;w:r cnd Shutdow2 Risk Assessmelt = g i V i i. Use the Level 1 quantification results from the effort in item (h) to calculate the I baseline and focused shutdown large release frequencies. I I Note, the shutdown PRA, Revision 6, reflects the conservative modeling of RNS trees in i sequences associated with the loss of offsite power (LOSP) initiating event. This I conservatism consists of using an OR gate instead of an AND at the second level of the RNS l fault tree files RNP2, RNT2, RNP2D and RNT2D. By correcting of these fault tree files, the I baseline core damage frequency changes from 5.5E-08 to 4.7E-08 events per year; the I dominant core damage sequences with single RNS train failure from LOSP events are I eliminated by these corrections. The corrected RNS fault tree files are used in this evaluation I to resolve the surge line issue. l l 54B.1.1 Baseline PRA Shutdown Core Damage Frequency Results l I By incorporating the changes discussed above into the shutdown PRA model, the baseline I core damage frequency (CDF)..:hanges from 4.7E-08 to 9.0E-08 events per year; an increase I of 91 percent. However,it is d.portant to note that a CDF of 9.0E-08 is still very low. The l sequence results for this baseline case are shown in Table 54B-5, and the top 200 cutsets from I the quantification output file are shown in Table 54B-6. The associated component i importance file is shown in Table 54B-7. l l 54B.1.2 Focused PRA Shutdown Core Damage Frequency Results / 1 l O l By incorporating the changes discussed above into the model from the shutdown PRA, I Revision 6, the focused PRA sensitivity study core damage frequency changes from 4.lE-07 l to 5.9E-07 events per year; an increase of 44 percent. However, a CDF of 5.9E-07 is still l l very low. The sequence results for this focused case are shown in Table 54B-8, and the top i 200 cutsets from this quantification output file are shown in Table 54B-9. The associated I component importance fi!e is shown in Table 54B-10. l l 54B.2 Design Change Effect on Surge Line Cases ; I !' I 'Ihis section documents the effect of design changes on the baseline and focused PRA results l from the surge line cases addressed in the previous section (54B.1) of this attachment. The I cutset output files for the results in subsections 54B.l.1 and 54B.I.2 form the starting point I of these sensitivity studies. I I These particular design changes, made subsequent to preparing the design reponed in the I shutdown PRA, Revision 6, are as follows: I l 1. IRWST injection check valves maintaining the RCS pressure boundary changed to squib l valves; (i.e., check valves 123A/B and 125A/B are changed to squib valves 123A/B and 1 125A/B, respectively). These squib valves serve to isolate the IRWST and operate to l
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p l v l ENEL Revision: 11 ! [ WE3tiflgh00Se w xia: L n March 1998 )' 54B-3 oWVeviliarts46.wpr.ib
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!=" ~Hi 54B. Surgi Line Flooding Effect on Low P:wIr and Shutd:wn Risk Assessment ~
l provide IRWST injection when called upon. Note, IRWST motor-operated O\ I valves 121 A & B are no longer required to be closed during drained conditions. I l 2. IRWST recirculation paths changed from 10-inch and 4-inch lines to 6-inch lines in all l paths l l 3. IRWST motor-operated valves 1ISA & B and check valves 120A & B maintaining the l IRWST water level changed to squib valves l l 4. CCS valves to the RNS heat exchanger changed from manual to air-operated i ! 5. The capacity of the service water basin reduced; its duration changed from 24 hours to l 12 hours I l 6. Service water system valves V037A & B changed from air-operated to motor-operated l l 7. The number of PRHR heat exchangers changed from two to one i I 8. RNS check valves V015A & B on the DVI lines changed to stop check valves. I i This evaluation is an extension of the studies documented in Attachment 54A. The same basic l event changes, shown in Attachment 54A, are made in these studies for the base cases. l Additional basic event changes related to the 4th stage ADS components, containment sump l recirculation components, and human error probabilities (HEPs) are made to perform I sensitivity studies in addition to those in Attachment 54A. l I In these studies, the effect of the design changes are evaluated together by performing the I steps in Section 54A.9, but on the surge line-related cutset files. In other words, the I sensitivity studies that assess the effect of each design change (as documented in i Sections 54A.1 through 54A.8) are not performed in this exercise. I i Note, for the first design change discussed above, pertaining to the IRWST injection valves, I the four squib valves have new dependencies on I&C systems and electric power systems for I actuating signals. The protection and safety monitoring system along with Class IE de power i provide the safety-related signal to these valves, with the diverse actuation system and the I non-Class IE de power providing the nonsafety-related signal. Each squib valve receives a 1 safety-related actuation signal from a different division, providing a robust redundancy I configuration. Since the check valves, which did not need a signal to open, have been I changed to squib valves, the I&C and power dependencies present a new challenge to the 1 IRWST actuation model. However, as discussed below, the I&C and power dependencies are i not included in the sensitivity study models. I I
- Because of the robust redundancy of the safety-related signals, and the fact that the i nonsafety-related systems provide a backup to the safety-related signals, the I&C and I power dependencies will be dominated by common cause failures, with random failures Revision: 11 O
March 1998 oiva\revilwt54b.wpf.lb-o22598 54B-4 hhw 3 Westiflghouse
542. Surge Line Flooding Effect on Low Paw;r cad Shutdown Risk Assessment I 7 ,\
\'J I being negligible. The non-drained condition shutdown cases already include those I&C l and power common cause failures in the event sequences that include failure of the l ADS, and in fact, cutset reduction mies will prevent these common cause events from I occurring in the sequences involving failure of IRWST, since IRWST follows successful l ADS operation. Therefore, there is no impact on the design change sensitivity study J l for the non-drained cases due to the new dependency on I&C and power systems. ]
I l
- For the drained condition shutdown cases modeled in shutdown PRA revision 6, the i design for the IRWST required the two IRWST motor-operated valves upstream of the I check valves to be closed. To inject water into the RCS upon loss of core cooling, one I of these motor-operated valves had to open. These two motor-operated valves received I safety related signals and power from separate divisions. The design has since changed I (as noted above) to include four squib valves with a success criteria of one-out-of-four I to open, and the motor-operated valves are no longer required to be closed during i drained conditions. Since each of the four squib valves receives a separate safety-1 related I&C signal and a separate safety-related power :.ource, the redundancy scheme I has improved. Therefore, the design change impact sensitivity study for the drained I cases is conservative with respect to the associated system dependencies.
I l The sensitivity studies for this evaluation, consisting of the following cases, are documented I in the subsections that follow: 1 ( l Case 1 - Effect of design changes on the baseline surge line core damage frequency V l Case 2 - Effect of design changes on the focused PRA sensitivity study surge line core I damage frequency i Case 3 - Effect on case 1 core damage frequency with single IRWST train, its I corresponding recirculation train, and 2 ADS 4th stage valves operation l Case 4 - Effect on case I core damage frequency when all HEPs are set to 0.5 l Case 5 - Effect on case 3 core damage frequency when all llEPs are set to 0.5. l l l 54B.2.1 Case 1 Sensitivity Study - Effect on Baseline (Surge Line) Results I l The Case i sensitivity study estimates the effect of the design changes on the baseline ) I shutdown core damage frequency of 9.0E-08, discussed above in subsection 54B.1.1. l l This estimation is conducted by making changes to the basic events in the quantification 1 output file from the surge line Level 1 baseline case. The following steps were done: I I l a) Revise basic event probabilities associated with the new IRWST configuration and I unscheduled maintenance unavailability discussed in subsection 54A.1; IWX-MV-GOl l is set to 2.6E-05, IWX-CV-AO set to 3.0E-05, IRWMOD05S set to 3.0E-03, and l IRWMOD06S is set to 0. l m ENEL Revision: 11 UDM WMh. March 1998 ! 54B-5 c:4=3\miliattS4b wpf;1b l I
l
=ai 54B. Srrg2 Line Flooding Effect c2 L;w Pswir and Shutdowa Risk Assessment I b) Revise the basic event frequencies for CCS initiating events to reflect the changes O
I discussed in subsection 54A.4; IEV-CCWD is set to 6.02E-04, and IEV-CCWND set I to 3.75E-03. 1 I c) Revise the basic event frequencies for the loss of offsite power initiating events to i reflect the changes discussed in subsection 54A.5; IEV-LOSPD is set to 4.44E-03, and I IEV-LOSPND is set to 1.71E-02. l I Having modified the basic events to incorporate these design changes into le shutdown PRA, I the baseline quantification output file is recalculated to estimate the shutdown core damage I frequency. The resulting shutdown core damage frequency changes from 9.0E-08 events per l year to 1.0E-07 events per year. The top 200 cutsets from this quantification output file are i shown in Table 54B-11. The associated component importance file is nown in l Table 54B-12. I l 54B.2.2 Case 2 Sensitivity Study - Effect on Focused PRA (Surge Line) Results l l De Case 2 sensitivity study estimates the effect of the design changes on the focused PRA i sensitivity study core damage frequency of 5.9E-07, discussed above in subsection 54B.1.2. i i This estimation is conducted by making changes to the basic events in quantification output I file from the surge line Level 1 focused PRA case. The following steps were done: l a) Revise basic event probabilities associated with the new IRWST configuration and O i unscheduled maintenance unavailability discussed in subsection 54A.1; IWX-MV-GO1 I is set to 2.6E-05, IWX-CV-AO set to 3.0E-05, IRWMOD05S set to 3.0E-03, and i IRWMOD06S set to 0. 1 I b) Revise the basic event frequencies for CCS initiating events to reflect the changes I discussed in subsection 54A.4; IEV-CCWD is set to 6.02E-04, and IEV-CCWND set I to 3.75E-03. 1 I Note, the service water system basin capacity design change is not included in the focused 1 PRA sensitivity study because this failure does not affect the SWS-related initiating events, I and the focused PRA does not credit nonsafety-related systems in the event trees. I I Having modified the basic events to incorporate these design changes into the shutdown PRA, I the focused PRA quantification output file is recalculated to estimate the shutdown core I damage frequency. The resulting shutdown core damage frequency changes from 5.9E-07 I events per year to 5.6E-07 events per year. The top 200 cutsets from this quantification I output file are shown in Table 54-13.The associated component importance file is shown in l Table 54B-14. Revision: 11 O March 1998 b W W85tingh00Se o4raWvil\att$4b wpf.lb.022$98 $48 6
I 542. Surge Line Flooding Effect on Low P:w:r and Shutdown Risk Assessment B== g i I l 54B.2 3 Case 3 Sensitivity Study - Single IRWST Injection Train, Corresponding Sump i Recirculation Train & 2 ADS 4th Stage Operation I i AP600 Technical Specifications (TS) 3.5.8 allows one of the IRWST injection trains and one 1 of the containment sump recirculation trains to be out of service during reduced inventory I conditiens; and TS 3.4.14 allows two ADS 4th stage valves to be out of service during I reduced inventory conditions. These technical specification provisions are to allow for l unscheduled repairs of components in any one of the IRWST injection trains, any one of the I containment sump recirculation trains and any two ADS 4th stage valves during drained I conditions. These provisions are not intended to allow plant personnel to schedule or plan I maintenance activities to be conducted during reduced inventory conditions, and, therefoie, I should not be misconstrued as such. l l However, a sensitivity study is performed to estimate the core damage frequency if one train i of IRWST, the associated containment sump recirculation train and two ADS 4th stage valves i are failed at the same time. For this conservative estimate of the design configuration, one l IRWST train, the associated containment sump recirculation train, and two ADS 4th stage I valves are assumed to fail during filled and drained shutdown conditions. It is assumed that l IRWST train B is available, and train A is unavailable. Additionally, the redundant injection I path via valve RNS-V023 is assumed to be unavailable. That is, it is assumed that (a) the l IRWST has only one injection line by essentially setting the other IRWST train failure I probability and the failure probability of RNS-V023 injection path equal to 1.0; (b) the O l containment sump recirculation function is provided by only one train (corresponding to the V i available gravity injection line) by setting the other recirculation train failure probability equal i to 1.0; and (c) the ADS 4th stage has only two valves by propagating the common cause i failure of two 4th stage valves through the shutdown model. I I Note, the ADS 4th stage operation with only 2 valves is reflected in the model by changing I only the common cause of the 4 ADS 4th stage squib valves to common cause of 2 ADS 4th I stage squib valves. Changes.to the squib valves random failure probabilities were not I attempted since these changes would not cht.nge the result of this sensitivity study. Note also I that operation with one available train of containment sump recirculation is reflected by setting 1 the failure probability of the other train equal to 1.0, and changing the common cause of 4 l recirculation squib valves to common tause failure of 2 squib valves. I I This sensitivity is performed by modifying basic events in the quantification output file from l Case 1, discussed above in subsection 54B.2.1. The following basic events are modified as I shown: l l ADX-EV-SA = 3.8E-05; IWACV122AO = 1.0; IWACV123AO = 5 BE.04; I IWACV124AO = 1.0; IWACV125AO = 5.8E-04; IWBCV122AO = 1.75E-03; 1 IWBCV123AO = 5.8E-04; IWBCV124AO = 1.75E-03; IWBCV125AO = 5.8E-04; I IWX-CV-AO = 3.lE-05; IWX-MV-GOI = 3.8E-05; IRWMOD05S = 0; I IRWMOD06S = 0; RN23 MOD 5S = 1; IWX-EV4 SA = 3.8E-05; REA-PLUG = 1.0. [%
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ENEL R*'iSI ": 11 T Westinghouse w re:: w . March 1998 54B-7 owrevi natts4b.wpf:1b
i 3... ma E 54B. S:rg) Line Flooding Effect on Irw P:wer cnd Shutdown Risk Assessment j l Basic events RHN-MAN 05 and BEN-MAN 05C, representing operator failure to initiate O l injection via valve RNS-V023, are dropped from the quantification cutsets, since total i unavailability of this injection path is fully accounted for by setting the basic event i (RN23 MOD 5S) for hardware failure of this valve to 1.0. l l Subsection 29.4.6 of Chapter 29 discusses the various common cause failure combinations of I valves to be considered in calculating common cause failure of valves in the IRWST. For i example, common cause failure (CCF) of 2 IRWST squib valves is calculated by: 1 (1 combination of 2/4 + 2 combinations of 3/4 + 1 combination of 4/4) = 1.0E-05 + l 2(1.0E-06) + 2.6E-05 = 3.8E-05. This value is used for basic events IWX-MV-G01 and i ADX-EV-SA to represent the CCF of two squib valves in the IRWST and ADS, respectively, I for this sensitivity study. The common cause failure of two IRWST injection check valves i is calculated similarly by summing the CCF combination 2 of 2/4 and 4/4 check valves I (5.4E-07 + 3.0E-05 = 3.lE-05). The value of combinations of 3/4 check valves is not shown I since, as discussed in Chapter 29, CCF of 3/4 check valves is small compared to CCF of 4/4. I Common cause failure of 2 IRWST check valves is represented by basic event IWX-CV-AO l in this sensitivity study. Failure of the available recirculation train due to CCF of the I recirculation squib valves is assigned the failure probability of 3.8E-05, same as the value I assigned to CCF associated with the IRWST and ADS, above; this CCF probability of l 3.8E-05 for the recirculation squibs is assigned to basic event IWX-EV4-SA. Consistent with I the common cause evaluations in Chapter 29, CCF of the motor-operated valves in the recirculation lines is not applied to scenarios with a single available recirculation train. l I l Requantification of the cutset file with these changes provides a core damage frequency of g I 1.4E-06 events per year. However, a core damage frequency of 1.4E-06 is still quite low. l The top 200 cutsets from this quantification output file is shown in Table 54B-15. The I associated component importance file is shown in Table 54B-16. I l This estimate of the core damage frequency at shutdown is believed to be conservative for the I following reasons: l l
- One train of IRWST is assumed to fail during both drained and non-drained conditions l
l
- One train of containment sump recirculation is assumed to fail during both drained and I non-drained conditions l
l
- Two ADS 4th stage valves are assumed to fail during both drained and non-drained I conditions I
l
- The IRWST water is considered to be clean water, suitable for injection into the RCS I
l
- The IRWST is covered to preclude material that could clog the strainer from entering i the tank Revision: 11 O
March 1998 d W W85tilighotise o \ prs \reviliatt54b.wpf:Ib-On598 54B-S
54B. Surge Line Flooding Effect on Low-P;w;r cnd Shutdown Risk Assessment Y l
- The IRWST strainer is assigned a conservative failure probability of 2.4E 04 in the l PRA. The plugging of the IRWST strainer is expected to be a much less probable I event than reflected in a failure probability of 2.4E-04.
l l 548.2.4 Case 4 Sensitivity Study Setting HEPs to 0.5 in Case 1 Cutsets l l The Case 4 sensitivity study estimates the effect of the setting the human error probabilities I to 0.5 in the quantification output file from Case 1, discussed above in subsection 54B. 1. l The following basic events are set to 0.5: l l ADN-MAN 01, CCB-MAN 01, IWN-MAN 00. IWN-MAN 00C, LPM-MANOS, i REC-MANDAS, REC-MANDASC, REN-MAN 04, RHN-MANO3, RHN-MAN 04, i RHN-MAN 04-SUCC; RHN-MANOS, RHN-MAN 05C, PRN-MANO1. SWB-MAN 02, I and ZON-MAN 01. I l Requantification of the cutset file with these changes provides a core damage frequency of l 2.5E-06 events per year. However, a core damage frequency of 2.5E-06 is still quite low. I The top 200 cutsets from this quantification output file is shown in Table 54B-17. 'Ihe I associated component importance file is shown in Table 54B-18. l l 54B.2.5 Case 5 Sensitivity Study - Setting HEPs to 0 5 in Case 3 Cutsets l l The Case 5 sensitivity study estimates the effect of the setting the human error probabilities V l to 0.5 in the quantification output file from Case 3, discussed above in subsection 54B.2.3. I The following basic events are set to 0.5: I l ADN-MAN 01, CCB-MAN 01, IWN-MAN 00, IWN-MAN 00C, LPM-MANOS, l REC-MANDAS, REC-MANDASC, REN-MAN 04, RHN-MANO3, RHN-MAN 04, l RHN-MAN 04-SUCC; PRN-MANO1, SWB-MAN 02, and ZON-MANO1. l l Requantification of the cutset file with these changes provides a core damage frequency of 1 4.9E-06 events per year. However, a core damage frequency of 4.9E-06 is still quite low. I The top 200 cutsets from this quantification output file is shown in Table 54B-19. The I associated component importance file is shown in Table 54B-20. I l 54 B.',.6 Conclusion I i l The design changes were evaluated to determine their effect on the PRA results and insights. l The previous subsections provide details on the five quantitative sensitivity studies that I basically assess the effects of these design changes on the shutdown (surge line) baseline and I focused PRA core damage frequencies. The results of this evaluation are as follows: I ; I a) When the design changes are incorporated together, the baseline (surge line) shutdown l PRA cere damage frequency changes from 9.0E-08 to 1.0E-07 events per year, an 1 increase of only 11 percent. [ N}l ENE Revision: 11 3 Westingh0088 ';mm
. March 1998 54B-9 owvi nan 54b.wpr.ib
hie Ulf ni 54B. Surge Line Flooding Effect on Low P w:r end Shutdown Risk Assessment When the design changes are incorporated together, the focused (surge line) shutdown O l b) l PRA core damage frequency changes from 5.9E-07 to 5.6E-07 events per year; a l decrease of 6 percent. I I c) Although some top cutsets are reordered, the top 50 to 100 cutsets are more or less the I same. This ca" *.se seen when the cutsets for the baseline and focused shutdown studies I in Section 5* 1 are compared with the results from Cases 1 and 2 in Section 54B.2. I l d) Although basic events importance are changed and reordered in some cases, the most I important events are relatively the same, as shown by their "importance % decrease" l values in the related component importance tables. l l c) When the HEPs are set to 0.5 in the surge line baseline case, the core damage I frequency becomes 2.5E-06 events per year, which is about 28 times greater than the I base case core damage frequency of 9.0E-08 events per year. This increase in the base I case core damage frequency is somewhat significant even though a core damage l frequency of 2.5E-06 is still quite low. The result indicates that the operator actions I are important in maintaining a very low core damage frequency for internal events at I shutdown. I l f) Although the functions of the IRWST and the RNS and its support systems (CCS and l SWS) are still important to maintaining plant safety during shutdown, the function of l ADS becomes more important than reflected in Revision 6 of the shutdown PRA, since I the ADS 4th stage is now required to preclude the effects of surge line flooding and I thereby maintain plant safety. Similarly, centainment sump recirculation becomes an I important function for the mitigation of accidents during drained conditions. The I importance of these functions is due mainly to their required operability during reduced I inventory conditions, where initiating events dominate the plant shutdown risk. l l 54B.3 AP600 Shutdown Level 2 PRA Using PRA Revision 8 Severe Accident i Phenomenology l I In this section, the AP600 shutdown PRA large release frequency is calculated using the I containment event tree methodology from revision 8 of the AP600 at-power PRA. The I analysis uses the Level I shutdown PRA results reported in this attachment, section 54B.1, I which includes the surge line flooding and long-term cooling updates. The purpose of this I analysis is to quantify the L se release ft qui acy for the baseline and focused shutdown PRA l and to show the sensitivi y o, the calcula 6.s to the assumption that a diffusion flame at the l IRWST vent fails the cor, 'tinm# l i The major limitation with usu.g Level 1 PRA results from section 54B.1 is plant design I changes noted in section 54B.2 are eot included in the analysis. However, the effect of these I changes on the system failure probab lity is a second order effect and the overall magnitude i of the failure probability is judged to be approximately unchanged. The effect of the severe Revision: 11 O ENEL March 1998 '- W Wesdnl$00Se oNrsrevi\ar:54b.wpr.id.022598 54B-10
54B. Surge Line Flooding Effect ca L,w P w;r cud Shutdowa Risk Assessment (v ) l accident methodology change from that used in shutdown PRA revision 2 to that used in at-I power PRA revision 8 on the large release frequency is measured directly from this analysis. l. l 54B.3.1 Level 1/ Level 2 Interface l l The surge line flooding results of the shutdown PRA are binned into 6 accident classes which I correspond to similar accident classes from the at-power PRA: 1 I Accident Class LP-1 A No RCS depressurization i Accident Class LP-3D Partial RCS depressurization l Accident Class LP-3BR Full RCS depressurization, core damage caused by 1 insufficient or late core reflood l Accident Class LP-3BE Full RCS depressurization, failure of gravity injection i Accident Class LP-3BL Full RCS depressurization, failure of long-tenn core I cooling l Accident Class LPCBP Core damage initiated by conminment bypass I l The accident class frequencies are presented in Tables 54B-21 and 54B-22 for the baseline I and focused shutdown PRAs, respectively. I l 54B.3.2 Containment Event Trees I n l The containment event tree structures used for accident classes LP-3BR,' LP-3BE and LP-3D ( ) 'O I in the shutdown analysis are the same containment event trees used for accident classes 3BR, l 3BE and 3D in revision 8 of the at-power PRA. The end states of the containment event trees I are binned into the same release categories, as well: l l Release Category IC Intact containment l Release Category BP Containment bypass l Release Category CI Containment isolation failure I Release Category CFE Early containment failure (during or immediately after core i damage) l Release Category CFI Intermediate containment failure (less than 24 hours) l Release Category CFL Late containment failure (greater than 24 hours) l l Accident class LP-1 A fails depressurization by definition and is assumed to always induce a l steam generator tube rupture which bypasses the containment. Accident class LPCBP l bypasses the containment by definition. No recovery actions are credited for LP-1 A and l LPCBP, so no containment event trees are used to quantify the large release frequencies for i these accident classes. 'Ihe frequency for accident classes LP-1A and LPCBP are added I directly into release category BP. l I The containment event trees for the baseline PRA are presented in Figures 54B-6 through l 54B-9, and for the focused PRA in Figures 54B-10 through 54B-13. Figures 54B-14 and I 54B-15 present the containment event trees for LP-3D for the baseline and the focused PRAs, s ._/ ENEl. Revision: 11 W8d@ll88 mtn:& March 1998 54B-11 e4rairevlisatr54b.wpt:1b
g W E 54B. Surge Line Flooding Effect ca IAw P:w:r cnd Sh:tdow:2 Rtk Assessmert I respectively, modified for the sensitivity case which assumes containment failure if a diffusion O I flame can be produced at the IRWST vents. 1 I 54B3.2.1 System Node Failure Probabilities l l The failure probabilities assigned to the system nodes on the containment event trees are I calculated from the fault tree linking results of the surge line flooding case (section 54B.1). I Note that the fault tree top events used in the containment event tree (CET) quantification are I fault trees specifically modeled for shutdown conditions; not at-power fault trees that were I modified for the new CET calculations. These fault tree top events were previously used in I the Level 2 assessment of shutdown events documented in the Revision 6 PRA. In the 1 Revision 8 PRA, Table 54-7 and the success criteria tables contain the information regarding I the special assumptions made for the shutdown modes. Table 54-8 presents the system I unavailability status information with reference to the proper Technical Specification for that I system. I I The following CET nodes use fault tree linking results for assigning the CET failure I probabilities; I I Node DP: Depressurization i Node IS: Containment Isolation l Node IR: Reactor Cavity Flooding from IRWST l Node IG: Hydrogen Igniters i Tables 54B-23 through 54B-26 identify which of the accident classes use fault tree linking for the above I nodes and the fault tree top event name. I l The failure probabilities for each node are presented on the event tree figures. l l 54BJ.2.2 Phtnomena Node Failure Probabilities I i The base case failure probabilities assigned to the phenomena nodes on the containment event I trees are scalar values assigned from the revision 8 at-power containment event trees for the , I corresponding accident classes. The failure probabilities for each node are presented on the I event tree figures. I I The at-power phenomenological probability values are conservative with respect to values I expected at shutdown. Decay heat rates for shutdown events are substantially lower than at- l l power events. For in-vessel retention of molten core debris via external cooling of the reactor i vessel, the lower decay heat rate translates to a lower volumetric power density in the lower I head. The heat fluxes predicted for the at-power scenarios bound the heat fluxes expected at I shutdown and therefore the results are conservative. ) l I For hydrogen generation and combustion phenomena, containment failure is dominated by ) I carly detonation. The release location and the hydrogen generation rate are the important Revision: 11 O Ed March 1998 W Westinghotise ! ovawviiwt54b.wpr:1 w 3039: 54B-12
T l [ 7v- g:h . i% - ( 542. Sage line Hoodmg Effect on Iow.Pbw r and Shutdown Risk Assessment f # I i'
/7- l U l phenomena that affect detonation. Shutdown accidents are dominated by reduced inventory l i l cases which fail core cooling. The release location is to the steam generator compartments, I I which participate in containment mixing. Hydrogen release to the dead-ended compartments I is limited so at-power detonation probability bounds with respect to shutdown. Hydrogen i generation rates are also reasonably predicted. The core heat up rate at shutdown is slower I than at-power due to the lower decay heat, but once cladding oxidation begins, the heat of I reaction dominates the process and the progression is similar. If the oxidation reaction is not I steam limited, the hydrogen generation rate is similar to the at-power rates. If the reaction I is steam limited, the at-power generrtion rates bound the shutdown rates. Therefore, the I at-power phenomenological probability estimates are conservative for the shutdown analysis.
l l For the diffusion flame sensitivity, the accident class LP-3D containment event tree is i modified based on the assumption that a diffusion flame at the IRWST vents fails the l containment with a probability of 1. The other low pressure accident classes either fail the l containment by definition (LP-1A and LPCBP) or discharge the hydrogen through stage 4 l ADS lines to the steam generator compartment (LP-3BR and LP-3BE) where the containment I wall is shielded from the burning. I i 54B3.23 level 2 PRA Quantification Results l l The quantification of the containment event tree paths are presented on the tree figures. 1 Summaries of the baseline and focused PRA quantification results and dominant sequences
/,h I are presented in Tables 54B-21 and 54B-22, respectively. Summaries of the baseline and I focused PRA diffusion flame sensitivities and dominant sequences are presented in l Tables 54B-27 and 54B-28, respectively.
l I 54BJJ Shutdown level 2 PRA Conclusion l l The baseline PRA shutdown large release frequency is 1.5E-08 per reactor-year. The large i release frequency is dominated by reactor vessel failure due to the failure to flood the reactor I cavity which contributes 66 percent. Containment bypass due to steam geraerator tube rupture, I both as an initiating event and induced by high RCS pressure and temperature accounts for i approximately 24 percent of the large release. Ten percent of the large release frequency is I containment isolation failure. All other containment failure modes contribute negligibly to l l the large release frequency. He assumption that a diffusion flame at the IRWST vents fails
]s I the containment doubles the large release frequency. The diffusion flame failure sequence I accounts for more than 50 percent of the large release frequency in this case, however, its I frequency is on the order of 10'8 per reactor-year. ,
I I he focused PRA shutdown large release frequency is 33E-07 per reactor-year. The large I release frequency is dominated by reactor vessel failure due to failure to flood the reactor i cavity which contributes 82 percent. Early hydrogen detonation centributes 12 percent of the j l large release frequency. Late detonation contributes approximately 4.6 percent of the large ; I release frequency. The other containment failure mode together account for approximately j i I percent of the large release. By assuming a diffusion flame at the IRWST vents fails the n k) ENEL Revision: 11 M Rid E 88 cmlh March 1998 54B-13 osprairevinaits4b.wpt:ib l t ,
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- 54B. S:rg2 Linz Flooding Effect en Low P4w:r cnd Sh:tdown Risk Assessment I 1 i
l containment, the large release frequency increases by 20 percent. The frequency of the O i diffusion flame failure is 7.3E-08 per reactor-year and doesnt cause the large release I frequency to exceed the goal of IE-06 per reactor-year. l l 548.4 References l l 54B-1 "AP600 Shutdown Evaluation Report," WCAP-14837, Revision 0, March 1997. I I O Revision: 11 O March 1998 ovawviisarts4b.wpt:1b e.2598 - 54B-14 h6 3 Westinghouse
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54B. S:rge Line Flooding Effect on Low Power und Shutdown Risk Assessment E h (O l Table 54B-1 l l FAULT TREE "ADASD" l SUCCESS CRITERIA
SUMMARY
l Event description: At.tomatic depressurization system, autornatically actuated, fails to provide RCS vent l path following a loss of the residual heat removal capability during reduced inventory I shttdown conditions with the vessel head intact. For this scenario, offsite power is I available. I Success configuration: 1 out of 4 lines of stage 4 l System initial status: All squib valves of stage 4 closed l Mission time: 24 hot.rs I Components required Squib valve of stage 4 is required to open l to chr.nge status: l Initiating signals: PMS: automatic actuation on low hot leg water level signal l Operator actions: Manual actuation from the main control room when automatic actuation fails I - basic events: LPM-MANOS and ADN-MAN 01 performed through PMS I - basic event: REC-MANDAS performed through DAS
/ \
( L l Table 54B-2 I I FAULT TREE "ADASDP" I SUCCESS CRITERIA
SUMMARY
l Event description: Automatic depressurization system, automatically actuated, fails to provide RCS vent I path following a loss of the residual heat removal capability during reduced inventory I shutdown conditions with the vessel head intact. For this scenario, station blackout I has occurred. I Success configuration: 1 out of 4 lines of stage 4 I System initial status: ' All squib valves of stage 4 closed j l Mission time: 24 hours l Components required Squib valve of stage 4 is required to open I to change status: I Initiating signals: PMS: automatic actuation on low hot leg water level signal ; I Operator actions: Manual actuation from the main control room when automatic actuation fails I - basic events: LPM-MANOS and ADN MAN 01 performed through PMS I - basic event: REC-MANDAS performed through DAS 4 (3 1 -
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! 54B. Surg: Line Flooding Effect ca Low P wxr and Shutdown Risk Assessment O
l Table 54B-3 I I FAULT TREE "RECIRCS" SUCCESS CRITERIA
SUMMARY
l Event description: Failure of recirculation lines to deliver water from the I containment sump to the reactor coolant system following a l transient or LOCA during drained shutdown conditions, given I the success of IRWST/ gravity injection l Success configuration: Flow from containment sump (via opening of either the i MOV/ squib valve flowpath or the check valve / squib valve I flowpath) to the RCS via 1 out of 2 gravity I injection / recirculation lines l System initial status: Check valves (V122A/B), squib valves (Vil8A/B and l V120A/B), and motor-operated valves (Vil7A/B) on I recirculation lines are closed i Mission time: 24 hours l Components required to change status: Motor-operated valve and squib valve, or check valve and squib l valve in each recirculation train must open l Initiating signals: Low-3 IRWST water level and ADS signal l Operator actions: Manual opening of the recirculation motor-operated and squib l valves if they fail to automatically open; basic event l REN-MAN 02 performed through PMS l l Manual opening of recirculation squib valves (in the check I valve paths) if they fail to open automatically; I basic event: REC-MANDAS performed through DAS I l Manual actuation of sump recirculation if IRWST low-3 water i level signal fails; I - basic event REN-MAN 04 performed through PMS (primary I cue to the operator is: increasing containment sump level; I secondary cues is: IRWST injection check valves indicating I open operation) l l l Note: 1) The MOVs and the squib valves (in the MOV paths) are manually actuated through the PMS only. I l 2) The squib valves in the check valve paths are manually actuated through the PMS or DAS. Revision: 11 O March 1998 Nd W Westinghouse owwviiwts4hwr r ib422598 54B-16
m= 54B. Surge Line Flooding Effect on Low P;w:.r and Shutdown Risk Assessment 5
;a) 4 v _
l Table 54B-4 l l FAULT TREE "RECIRBS" SUCCESS CRITERIA
SUMMARY
l Event description: Failure of recirculation lines to deliver water from the I containment sump to'the reactor coolant system following a I station blackout, given the success of IRWST/ gravity injection I Success configuration: Flow from containment sump (via opening of either the l MOV/ squib valve flowpath or the check valve / squib valve
,1 flowpath) to the RCS via 1 out of 2 gravity I injection / recirculation lines l System initial status: Check valves (V122A/B), squ'b valves (Vil8A/B and l V120A/B), and motor-operatej valves (Vi17A/B) cn I recirculation lines are closed l Mission time: 24 hours l Components required to change status: Motor-operated valve and squib valve, or check valve and I squib valve in each recirculation train must open I initiating signals: Low-3 IRWST water level and ADS signal l Operator actions: Manual opening of the recirculation motor-operated and squib O I valves if they fail to automatically open; basic event
(,,/ l REN-MAN 02 performed through PMS I I Manual actuation of sump recirculation if IRWST low-3 water l level signal fails; I - basic event REN-MAN 04 performed through PMS (primary I cue to the operator is: increasing containment sump level; I secondary cues is: IRWST injection check valves indicating I open operation) l Note: 1) The MOVs and the squib valves (in the MOV paths) are manually actuated through the PMS only.
- 2) The squib valves in the check valve paths are manually actuated through the PMS.
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( ) I l Table 54B-7 (Sheet 1 of 12) I 4 I BASIC EVENT IMPORTANCE - BASELINE SHUTDOWN PRA I INCLUDING SURGE-LINE FLOODING EFFECTS SYSTEM UNAVAILABILITY (Q) = 8.970E-08 NUMDER OF BASIC EVENT 4 = 316 NUMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVENT BASIC EVENT (% DECREASE) CUTSETS UNAVAILABILITY PROBABILITY 1 IEV-CCWD 69.66 2442 6.2483E-08 4.2300E-04 2 IWX-MV-G01 32.89 266 2.9499E-08 5.5000E-05 3 ADX- EV-SA 17.93 290 1.6087E-08 3.0000E-05 4 INX-EV4-SA 15.53 210 1.3931E-08 2.6000E-05 5 IEV-RNGD 13.40 1207 1.2020E-08 8.1500E-05 5 IWX-FL-GP 7.41 168 6.6469E-09 1.2000E-05 7 REX-FL-CP 7.17 151 6.4316E-09 1.2000E-05 8 IEV-LOSPD 5.95 3158 5.3327E-09 1.4800E-03 9 OTH-R1 4.52 2938 4.0504E-09 4.2000E-01 10 IEV-RCSOD 3.66 564 3.2816E-09 4.4400E-06 11 IWX-CV-AO 3.49 395 3.1292E-09 1.5000E-04 12 IEV-LOCA24ND 3.06 121 2.7492E-09 1.6800E-05 33 CCX-SFTW 3.04 26 2.7251E-09 1.2000E-06 14 RHN-MAN 05C 2.89 27 2.5955E-09 1.5000E-01 15 RHN-MAN 04 2.29 134 2.0497E-09 5.5000E-02 16 CCX-ORY-SPX 2.17 112 1.9484E-09 3.6300E-06 17 IWN-MAN 00C 2,09 23 3.8777E-09 5.0000E-02 18 IEV-LOCA24D 1.86 546 1.6687E-09 1.1300E-05 19 CCX-XMTRX 1.57 97 1.4050E-09 2.6300E-06 20 IEV-CCWND 1.54 69 1.3841E-09 3.2100E-03
/N 21 RHN-MA:404-SUCC 1.52 468 1.3678E-09 9.4500E-01 !
I ) 22 SUC-RIS 1.43 220 1.2824E-09 5.8000E 01
\__/ 23 IWX-XMTR 1.20 102 1.0808E-09 2.0100E-04 24 REN-MAN 04 1.20 90 1.0758E-09 1.0000E-02 25 CCX-EP-SAM 1.14 229 1.0215E-09 8.6200E-06 26 REC-MANDASC 1.01 46 9.0394E-10 5.0600E-01 27 LPM-MAN 05 .96 37 8.6177E-10 6.8300E-04 28 IWN-MANDO .88 189 7.8572E-10 1.1500E-03 29 IWB-PLUG .73 443 6.5444E-10 2.4000E-04 30 CCX-BY-PN .72 363 6.5015E-10 4.7000E-05 31 REC-MANDAS .72 31 6.4723E-10 1.1600E-02 32 MDAS .64 75 5.7593E-10 1.0000E-02 33 IRWMOD055 .51 498 4.5342E-10 3.2200E-03 34 RN23 MOD 5S .50 966 4.4863E-10 2.2100E-03 35 ED1 MOD 03 .48 125 4.3214E-10 2.2020E-03 36 IEV-RNSND .46 46 4.1438E-10 9.6100E-04 37 CCX-ORY-SP .46 18 4.1251E-10 2.7700E-04 38 ZO1 MOD 01 .39 625 3.5041E-10 2.0200E-02 39 CCX-IV-XR .37 277 3.3078E-10 2.4000E-05 40 ZOX-PD-ES .36 113 3.1940E-10 2.0000E-03 41 CCX-XMTR195 .33 16 2.9956E-10 2.0100E-04 42 PL5 MOD 11 .33 77 2.9731E-10 2.0900E-03 43 REN-MAN 05 .33 833 2.9661E-10 1.6000E-03 44 IEV-LOSPND .25 40 2.2376E-10 8.1300E-03 45 202 MOD 01 .23 232 2.1058E-10 2.0200E-02 46 ECICB100VO .23 483 2.1052E-10 1.2300E-02 47 RNX-PM-FS .23 59 2.0572E-10 7.7000E-04 48 ECX-CB-GO .21 92 1.9126E-10 1.2000E-03 49 CCX+ INPUT-LOGIC .19 29 1.6716E-10 1.0300E-04 50 PL50301ASA .18 52 1.6440E-10 1.1600E-03 51 PL503 01B.c a. .18 52 1.6440E-10 1.1600E-03 52 SUC-R2S .18 9 1.6422E-10 7.6000E-01 53 RNX-KV-GO .18 52 1.6282E-10 6.1000E-04 54 SWBMOD02 .17 174 1.5621E-10 2.4400E-02 55 RNBMOD07S .17 184 1.5488E-10 1.5300E-02 56 RNAMOD068 .17 134 1.5199E-10 1.5300E-02 57 IEV-!OCAPRND .16 59 1.4228E-10 1.5100E-05 7
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Rethion: 11 W~ W85tingh00S8 - w March 1998 54B-43 'W"*54bwpr: b
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- 54B. Surge Line Floodi:g Effect on Lew P wer cnd Sh:tdown Risk Assessment h"= I l Table 54B-7 (Sheet 2 of 12)
O l I BASIC EVENT IMPORTANCE - BASELINE SHUTDOWN PRA INCLUDING SURGE-LINE FLOODING EFFECTS ! 1 SYSTEM UNAVAILABILITY (Q) = 8.970E-08 NUMBER OF BASIC EVENTS = 316 NUMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM EASIC EVENT BASIC EVENT (% DECREASE) CUTSETS UNAVAILABILITY FROBABILITY 58 CCX-PMAMOD1X .15 195 1.3392E-10 4.6300E-05 59 EC2CB200VO .14 186 1.2726E-10 1.2300E-02 60 ECX-CB-GC .13 74 1.1597E-10 7.3000E-04 i 61 PMAMOD11 .09 296 8.4017E-11 2.0900E-03 1 1 62 ED1 MOD 01 .08 27 7.0843E-11 5.0400E-04 63 ZOX-DG-DR .08 59 6.9537E-11 4.4000E-04 ! 64 CCBMOD015 .07 115 6.5398E-11 1.0400E-02 ! 65 SWAMOD07 .07 68 6.0739E-11 7.1200E-03 66 ADN-MAN 01 .07 58 6.0093E-11 4.9300E-04 67 OTH-R2 .07 29 5.9239E-11 2.4000E-01 68 IDDBSDDITM .07 411 5.8673E-11 3.0000E-04 69 IDDBSDSITM .07 411 5.8673E-11 3.0000E-04 70 IDDBSDK1TM .06 404 5.8169E-11 3.0000E-04 71 ED1 MOD 11 .06 59 5.0500E-11 3.1700E-04 72 ED1 MOD 113 .06 59 5.0500E-11 3.1700E-04 73 CCAMOD02 .05 66 4.6660E-11 7.1000E-03 74 PMA0301ASA .05 218 4.5109E-11 1.1600E-03 75 PMA0301BSA .05 218 4.5109E-11 1.1600E-03 76 EDiMOD13 .05 22 4.4450E-11 3.1700E-04 77 ZOX-DG-DS .05 54 4.4157E-11 2.8000E-04 78 IRWMOD06S .05 495 4.2876E-11 3.2200E-03 79 ED1 MOD 07 .05 22 4.2797E-11 3.0500E-04 80 IDBBSDDITM .05 181 4.1926E-li 3.0000E-04 81 IDBBSDK1TM .05 181 4.1926E-11 3.0000E-04 82 IDBBSDSITM .05 181 4.1926E-11 3.0000E-04 l 83 PXX-AV-LA .05 57 4.1919E-11 G.1000E-05 l 84 CCX-PMA030X .04 130 3.9816E-11 1. 3 8 f'O f -05 I 85 REA-PLUG .04 92 3.9394E-11 2.4000E-04
- 86 REB-PLUG .04 92 3.9394E-11 2.4000E-04 l 87 CCX-PL4 MOD 1 .04 25 3.7158E-11 1.419]E-04 88 IWA-PLUG .04 182 3.5711E-11 2.4000E-04 89 EC1REDG2GA .04 78 3.4214E-11 4.3600E-03 l 90 EC1REDG1GA .04 65 3.3986E-11 4.3600E-03 l 91 RNDEP0235A .04 322 3.2416E-11 1.7100E-04 92 CCX-FMAMOD1-SW .04 128 3.1888E-11 1.1000E-05 93 IDAMOD05 .03 354 3.0503E-11 5.1600E-04 l 94 ED2 MOD 03 .03 106 2.8237E-11 2.7000E-03 l 95 CCX-FMDMODI .03 298 2.6501E-11 1.4100E-04 96 CCX-PL403 .03 22 2.5485E-11 9.6900E-05 97 EC5EPMGBISA .03 14 2.3856E-11 1.7100E-04 98 IWACV122AO .03 178 2.2481E-11 8.7600E-03 99 IWACV123AO .03 178 2.2481E-11 8.7600E-03 100 IWACV124AO .03 178 2.2481E-11 8.7600E-03 101 IWACV125AO .03 178 2.2481E-11 8.7600E-03 102 IW10n170SPX .02 34 2.1416E-11 9.4600E-05 103 IW20R160SPX .02 34 2.1416E-11 9.4600E-05 104 ZOX-PD-ER .02 36 2.0190E-11 1.3000E-04 105 CCX-PL1 MOD 1 .02 14 1.9655E-11 1.4100E-04 106 CCX-PL7 MOD 1 .02 14 1.9655E-11 1.4100E-04 107 IDAMOD04 .02 340 1.8935E-11 3.1700E-04 108 EC1BS001TM .02 442 1.8675E-11 2.7000E-03 109 EC1BS012TM .02 442 1.8675E-11 2.7000E-03 110 EC1BS121TM .02 442 1.8675E-11 2.7000E-03 111 ED3 MOD 07 .02 34 1.8614E-11 3.0500E-04 112 ZO1 MOD 04 .02 129 1.8555E-11 1.2500E-03 113 PMAMOD11X .02 369 1.7906E-11 6.5300E-04 114 CCX-PKD030 .02 211 1.7752E-11 9.6900E-05 1.6898E-11 3.0000E-04 e
115 IDABSDSITM .02 261 a h 1998 . 6 W85tingh00S8 oNra\revll\att54b wpf:ltr022598 54B-44 _ _ _ _ . -.-~ .
54B. Surge Line Flooding Effect on Low P;wer and Shutdowa Risk Assessinent E g i !z : i Table 54B-7 (Sheet 3 of 12) 1 l BASIC EVENT IMPORTANCE - BASELINE SHUTDOWN PRA INCLUDING SURGE LINE FLOODING EFFECTS SYSTEM UNAVAILABILITY (Q) = 8.970E-08 NUMBER OF BASIC EVENTS = 316 NiJMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVENT BASIC EVENT (% DECREASE) CUTSETS UNAVAILABILITY PROBABILITY 116 ED3 MOD 03 .02 45 1.6662E-11 2.2020E-03 117 IW1TL170UFX .02 33 1.5467E-11 6.8400E-05 118 IW2TL160UFX .02 33 1.5467E-11 6.8400E-05 119 CCX-PL103 .02 12 1.3479E-11 9.6900E-05 120 CCX-PL703 .02 12 1.3479E-11 9.6900E-05 121 RNX-CV-GO .01 20 1.3367E-11 5.1000E-05 122 PL4MCD11 .01 68 1.2513E-11 2.0900E-03 123 ZO2M0004 .01 70 1.2037E-11 1.2500E-03 124 PLIMoD11 .01 57 1.1547E-11 2.0900E-03 125 PL7 MOD 11 .01 57 1.1547E-11 2.0900E-03 126 PL5XS00ASA .01 12 1,1113E-11 8.0000E-05 127 SWAMOD09P .01 40 9.6475E-12 1.8500E-03 128 SWDMOD09P .01 45 9.6353E-12 1.8500E-03 129 20X-BL-Es .01 20 9.0387E-12 6.0000E-05 130 CCX-FMDMoD4 .01 154 8.7220E-12 4.9800E-05 131 CCX-BY-PN1 .01 20 8.5918E-12 5.7000E-05 132 CCX-EF-SAMX .01 87 8.4646E-12 2.9400E-06 133 IRWMOD01 .01 53 8.2695E-12 1.2000E-02 134 2RWMOD03 .01 53 8.2695E-12 1.2000E-02 135 IRAEP121ASAX .01 63 7.9408E-12 5.8300E-05 tN 136 CCX-FMAMOD2X .01 22 7.8346E-12 8.5800E-05 t i 137 201DG001TM .01 74 7.4256E-12 4.6000E-02 \ ,/ 138 CCX-PMAMOD1 .01 87 7.3481E-12 1.4100E-04 139 ALL-IND-FAIL .01 31 7.1614E-12 1.0000E-06 140 PL40301ASA .01 44 6.3370E-12 1.1600E-03 141 PL40301BSA .01 39 6.1235E-12 1.1600E-03 142 PL10301ASA .01 38 5.9783E-12 1.1600E-03 143 PL70301ASA .01 38 5.9783E-12 1.1600E-03 144 PL10301BSA .01 36 5.8931E-12 1.1600E-03 145 PL70301BSA .01 36 5.8931E-12 1.1600E-03 146 AD2 MOD 01 .01 21 5.0246E-12 5.5400E-02 147 AD2 mod 02 .01 21 5.0246E-12 5.5400E-02 148 AD3 mod 03 .01 21 5.0246E-12 5.5400E-02 149 AD3 MOD 04 .01 21 5.0246E-12 5.5400E-02 150 CCX-PMA030 .00 59 4.4547E-12 9.6900E-05 151 PMA0301ASAX .00 176 4.3607E-12 1.6500E-04 152 IDCBSDDITH .00 183 4.3540E-12 3.0000E-04 153 IDCBSDSITM .00 183 4.3540E-12 3.0000E-04 154 PMA0301BSAX .00 157 4.2589E-12 1.6500E-04 155 RNX- PM- ER .00 17 4.1374E-12 1.6000E-05 156 CCX-VS-FA .00 5 4.0105E-12 3.8400E-05 157 REACV119GO .00 36 3.9688E-12 1.7500E-03 158 REBCV119GO .00 36 3.9688E-12 1.7500E-03 159 PMDMOD11 .00 235 3.9228E 12 2.0900E-03 160 IDCPSDK1TH .00 176 3.8508E-12 3.0000E-04 161 CCX-FM-ER .00 16 3.5961E-12 1.4000E-05 162 SWX-FM-ER .00 16 3.5961E-12 1.4000E-05 163 RHN-MANO3 .00 49 3.4266E-12 2.2600E-03 164 IWBCV122Ao .00 184 3.3498E-12 8.7600E-03 165 IWBCV123AO .00 184 3.3498E-12 8.7600E-03 166 IWBCV124Ao .00 184 3.3498E-12 8.7600E-03 167 IWBCV125A0 .00 184 3.3498E-12 8.7600E-03 168 CCX-IV-XR1 .00 11 3.3263E-12 2.4000E-05 169 IRWMOD10 .00 36 3.3142E-12 1.4600E-03 1TO IRWMOD12 .00 36 3.3142E-12 1.4600E-03 171 EC2BS002TM .00 208 3.29?7E-12 2.7000E-03 172 EC2BS022TM .00 208 3.2937E-12 2.7000E-03 1
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jrjyggt Itevision: 11 [ W85tingh0088 trix.h March 1998 54B-45 ovasrevi \ arts 4b.wpr.ib
54B. Surg) Line Flooding Effect c3 Low P:wer and Shutdown Risk Assessmert l Table 54B.7 (Sheet 4 of 12) e I l BASIC EVENT IMPORTANCE - BASELINE SIIUTDOWN FRA I INCLUDING SURGE LINE FLOODING EFFECTS SYSTEM UNAVAILABILITY (Q) = 8.970E-08 NUMBER OF BASIC EVENTS = 316 NUMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVENT BASIC EVENT (% DECREASE) CUTSETS UNAVAILABILITY PROBABILITY 173 EC2BS221TM .00 208 3.2937E-12 2.7000E-03 l'4 ED2 MOD 11 .00 39 3.1340E-12 3.1700E-04 175 CCX-PL4 MOD 1-SW .00 14 2.7743E-12 1.1000E-05 176 ED3 mod 01 .00 18 2.6048E-12 5.0400E-04 177 PMAM0012X .00 123 2.5440E-12 6.5300E-04 178 CCX-INPUT-LOGICX 00 12 2.3078E-12 2.5900E-05 179 PMAXS00ASAX .00 158 2.3072E-12 8.0000E-05 180 CCX-EP-SA .00 14 2.1763E-12 8.6200E-06 181 PL4 MOD 12 .00 30 2.1360E-12 2.0900E-03 182 EC1M0001 .00 31 2.0786E-12 2.0200E-04 183 PMAXS00ASA .00 28 1.9751E-12 8.0000E-05 184 IWCRS120BFA .00 23 1.8899E-12 8.7600E-04 185 IWDRS120AFA .00 23 1.8899E-12 8.7600E-04 186 IDDTD019RO .00 69 1.8874E-12 1.2000E-05 187 1DDTD020RQ .00 69 1.8874E-12 1.2000E-05 188 PMD0301ASA .00 157 1.8222E-12 1.1600E-03 l 189 CCX-FMDMOD1-SW .00 67 1.7268E-12 1.1000E-05 l 190 CCX-PMDMOD4-SW .00 67 1.7268E-12 1.1000E-05 l 191 PMD0301BSA .00 138 1.6428E-12 1.1600E-03 l 192 IWNTK001AF .00 5 1.6314E-12 2.4000E-06 I 193 IDBFD013RQ .00 20 1.6126E-12 1.2000E-05 194 IDBFD014RQ .00 20 1.6126E-12 1.2000E-05 l .00 1.4669E-12 1.1000E-05 l 195 CC.4-PL1 MODI-SW 8 ! 196 CCX-PL7 MOD 1-SW .00 8 1.4669E-12 1.1000E-05 197 EC2 MOD 01 .00 24 1.4182E-12 2.0200E-04 198 PMDMOD12 .00 97 1.3879E-12 2.0900E-03 199 CCX-LS-FA .00 32 1.3245E-12 5.3700E-06 200 ZON-MAN 01 .00 15 1.3117E-12 2.6700E-03 201 CCX-PMAMOD4 .00 41 1.2481E-12 4.9800E-05 202 RNAEPRNPSA .00 20 1.2281E-12 1.7100E-04 203 RNBEPRNPSA .00 20 1.2281E-12 1.7100E-04 204 IDAMoD08 .00 96 1.2181E-12 3.1700E-04 205 SWNTP001RI .00 10 1.1823E-12 5.2300E-03 l 206 PRAAV108LA .00 11 1.1753E-12 1.0900E-03 l 207 PRBAV108LA .00 11 1.1753E-12 1.0900E-03 208 SW7EPSBPASA .00 17 1.1633E-12 1.7100E-04 209 SWB-MAN 02 .00 8 1.1157E-12 1.6000E-03 210 PLIMOD12 .00 17 1.1082E-12 2.0900E-03 211 PL7M0012 .00 17 1.1082E-12 2.0900E-03 212 CCB-MAN 01 .00 13 1.0721E-12 1.0700E-03 213 CCX-IN-LOGIC-SW .00 8 9.7844E-13 1.1000E-05 214 CCX-FMAMOD2-SW .00 8 9.7844E-13 1.1000E-05 215 IRWMOD09 .00 18 8.5243E-13 1.4600E-03 216 IRWMOD11 .00 18 8.5243E-13 1.4600E-03 217 CC1EPSBPASA .00 15 8.3000E-13 1.7100E-04 218 CCNTF101RI .00 10 8.1191E-13 5.2300E-03 219 PMAMOD21X .00 14 6.0658E-13 1.1600E-03 220 ZO1 MOD 03 .00 17 7.7405E-13 1.0000E-04 221 PMDMOD41 .00 74 7.3741E-13 6.3500E-04 222 DAS .00 4 7,3730E-13 1.0000E-02 223 PL40302ASA .00 13 7.1831E-13 1.1600E-03 224 IDDBSDD1LF .00 37 7.0766E-13 4.8000E-06 225 IDDBSDK1LP .00 37 7.0766E-13 4.8000E-06 226 IDDBSDSILP .00 37 7.0766E-13 4.8000E-06 227 SW7EPSBPBSA .00 12 7.0599E-13 1.7100E-04 228 CCIEPSBPBSA .00 12 7.0577E-13 1.7100E-04 229 IRAEP121BSAX .00 66 6.9378E-13 5.8300E-05 230 IDBBSDD1LF .00 15 6.3971E-13 4.8000E-06 e
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542. Surge Line Flooding Effect on Low-P;wer and Shutdows Risk Assessment x ", l l Table 54B-7 (Sheet 5 of 12) l l BASIC EVENT IMPORTANCE - BASELINE SHUTDOWN PRA l INCLUDING SURGE-LINE FLOODING EFFECTS SYSTEM UNAVAILABILITY (Q) = 8.970E-08 NUMBER CF BASIC EVENTS = 316 NUMBER OF CUTSETS e 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVENT BASIC EVENT MDECREASE) CUTSETS UNAVAILABILITY PROBABILITY 231 IDBPSDKILF .00 15 6.3971E-13 4.8000E-06 232 IDEBSDS1LF .00 15 6.3971E-13 4.8000E-06 233 SW7EPCTFASA .00 10 6.2872E-13 1.7100E-04 234 SW7EPCTFBSA .00 10 6.2872E-13 1.7100E-04 235 CCX-AV-LA .00 5 6.2306E-13 6.1000E-05 236 ED1BSDS1LF .00 7 6.1418E-13 4.8000E-06 237 CCX-PMDEHO .00 35 5.9175E-13 4.0300E-06 238 ADX-MV-Go .00 4 5.8975E-13 1.1000E-03 239 PMD0302ASA .00 51 5.4645E-13 1.1600E-03 240 FMA0302ASAX .00 59 5.4567E-13 1.6500E-04 241 PRAAV108TM .00 9 5.3768E-13 5.0000E-04 242 PABAV108TM .00 9 5.3768E-13 5.0000E-04 2(3 IDDMOD33 .00 51 5.3380E-13 5.1600E-04 244 CMX VS-FA .00 7 5.2969E-13 3.8400E-05 245 CMX-cV-GO .00 4 5.2014E-13 5.1000E-05 24 fi IDDMOD32 .00 64 5.1326E-13 3.1700E-04 24'7 PL40302BSA .00 8 5.0482E-13 1.1600E-03 246 RC10R195SP .00 4 4.5386E-13 7.2200E-03 24W IWARS118BFA .00 11 4.4642E-13 8.7600E-04 250 IWBRS118AFA .00 11 4.4642E-13 8.7600E-04 (% 251 PMA0302BSAX .00 40 4.4387E-13 1.6500E-04 252 ZO2 MOD 03 .00 10 4.3199E-13 1.0000E-04
)
( 253 ZO2DG002TM .00 4 4.1556E-13 4,6000E-02 t x~j ' 254 PMD0302BSA .00 32 3.6704E-13 1.1600E-03 255 PL10302ASA .00 7 3.5960E-13 1.1600E-03 256 PL70302ASA .00 7 3.5960E-13 ?. 1600E-03 1 257 EC1 MOD 12 .00 67 3.3817E-13 4.8000E-05 258 RC1TL195UT .00 3 3.2829E-13 5.2300E-03 259 PL10302BSA .00 5 2.7434E-13 1.1600E-03 260 PL70302BSA .00 5 2.7434E-13 1.1600E-03 261 PL1 MOD 51 .00 5 2.6401E-13 8.7400E-04 262 CCX-FMAMOD4-SW .00 12 2.5979E-13 1.1000E-05 263 REAMOV11?TM .00 9 2.5291E-13 5.0000E-04 264 REBMOV117TM .00 9 2.5291E-13 5.0000E-04 265 ED3BSDS1TM .00 1 2.3204E-13 3.0000E-04 266 IDAMOD06 .00 48 2.2782C-13 4.3600E-05 l 2 6*r CCX-BC-SA .00 5 2.1194E-13 8.4000E-06 I 268 IDDM0038 .00 34 2.0051E-13 3.2700E-04 269 PLIXS00ASA .00 5 1.9484E-13 8.0000E-05 270 PL4XS00ASA .00 5 1.9484E-13 8.0000E-05 271 PL7XS00ASA .00 5 1.9484E-13 8.0000E-05 272 CCX-HE-AF .00 5 1.9040E-13 1.2000E-06 273 IDAMOD02 .00 31 1.7752E-13 2.7000E-03 274 IDAMOD03 .00 31 1.7752E-13 2.7000E-03 275 PLMMOD41 .00 4 1.6831E-13 6.3500E-04 276 PRN-1MN01 .00 4 1.5932E-13 4.0800E-04 277 IDABSDDITM .00 9 1.4433E-13 3.0000E-04 278 IDABSDK1TM .00 9 1.4433E-13 3.0000E-04 279 CCX-XMT"t .00 3 1.3800E-13 2.0100E-04 280 CCX-TRYSM .00 4 1.3323E-13 2.0100E-04 281 PMAMOD41 .00 34 1.3256E-13 6.3500E-04 282 PMDXS00ASA .00 32 1.2368E-13 8.0000E-05 283 IDCFD007RQ .00 18 1.1757t-13 ' 2000E-05 284 IDCFD00BRQ .00 18 1.1757E-13 1.2000E-05 285 PMDEH0AISA .00 28 1.0625E-13 8.0000E-05 286 PMAMOD22X .00 2 9.9809E-14 1.1600E-03 287 EC1 MOD 13 .00 3 9.6145E-14 4.8000E-05 i l ENEL Rnisim 11 [ W85tingh0088 Ir/ m bm. March 1998 54B-47 ovavn uwts.ib.wpt:ib
l 1 54B. Surge Line Floodi;g Effect en Low P:wIr rnd Shutdown Risk Assessment I Table 54B-7 (Sheet 6 of 12) l i BASIC EVENT IMPORTANCE - BASELINE SIIUTDOWN PRA INCLUDING SURGE-LINE FLOODING EFFECTS SYSTEM UNAVAILABILITY (Q) = 8.970E-08 FJMBER OF BASIC EVENTS = 316 NUMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVENT BASIC EVENT (% DECREASE) CUTSETS UNAVAILABILITY PROBABILITY 288 EC2 MOD 23 .00 3 9.6145E-14 4.8000E-05 289 EC2 MOD 22 .00 17 9.5368E-14 4.8000E-05 290 CCX-FMAEMOX .00 il 9.4591E-14 4.0300E-06 291 IDABSDS1LF .00 24 9.3592E-14 4.8000E-06 292 ZOX-BL-ER .00 2 8.6264E-14 1.3600E-06 293 ED2 mod 01 .00 2 8.5265E-14 5.0400E-04 294 20X-FL-GP .00 2 8.2379E-14 1.3000E-06 295 PRX-HR-ML .00 4 8.1157E-14 1.2000E-07 296 ECIM00121 .00 24 7.3386E-14 1.6800E-05 297 PMDMOD42 .00 12 6,8279E-14 6.3500E-04 298 IDCBSDD1LF .00 12 4.1966E-14 4.8000E-06 299 IDCBSDKILF .00 12 4.1966E-14 4.8000E-06 300 IDCBSDS1LF .00 12 4.1966E-14 4.8000E-06 301 IDAMoD07 .00 15 3.7303E-14 2.1900E-02 302 IDDMOD21 .00 6 3.6970E-14 1.9200E-04 303 PKDEHOA2SA .00 10 3.4972E-14 8.0000E-05 304 DGi-LOGIC .00 1 2.9976E-14 5.0000E-03 305 IDDMOD34 .00 10 2.1982E-14 4.3200E-05 306 ZJ3 mod 01 .00 8 1.3989E-14 8.4000E-05 307 ECIBS001LP .00 7 1.2768E-14 4.8000E-06 308 PHAEHOAISA? .00 6 1.0325E-14 8.0000E-05 309 PMAMoD42 .00 4 9.6589E-15 6.3500E-04 310 IWX-XMTRLW .00 2 5.7732E-15 9.2000E-09 311 IDDMOD35 .00 4 5.2180E-15 2.1900E-02 312 EC2 HOD 221 .00 3 4.6629E-15 1.6800E-05 313 IDDMOD22 .00 3 3.6637E-15 2.7000E-03 314 IDDMoD23 .00 3 3.6637E-15 2.7000E-03 315 PRCEP108SA .00 1 9.9920E-16 1.7100E-04 316 PRDEP108SA .00 1 9.9920E-16 1.7100E-04 l Revision: 11 O i March 1998 g [ Westingh00SS oNwaWvilWt54b.wpf Ib.022598 54B-48 I 1 J
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- 50. Surge Line Flooding Effect on LomPower and Shutdown Risk Assessment ~5 l
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\ l l Table 54B-7 (Sheet 7 Of 12) l I BASIC EVENT IMPORTANCE - BASELINE SHUTDOWN PRA I INCLUDING SURGE-LINE FLOODING EFFECTS SYSTEM UNAVAILABILITY (0) = 8.970E-08 NUMBER OF EASIC EVENTS e 316 NUMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVENT BASIC EVENT (% INCREASE) CUTSETS UNAVAILABILITY PROBABILITY 1 CCX-SFTW 2.531678E+06 26 2.2709E-03 1.2000E-06 2 IEV-PCSOD 823973. 564 7.3910E-04 4.4400E-06 3 IWX-FL-GP 617504. 168 5.5390E-04 1.2000E-05 4 CCX-ORY-SPX 598368. 112 5.3674E-04 3.6300E-06 5 IWX-MV-G01 597909. 266 5.3632E-04 5.5000E-05 6 ADX-EV-SA 597807. 290 5.3623E-04 3.0000E-05
'/ REX-FL-GP 597503. 151 5.3596E-04 1.2000E-05 8 IWX-EV4-SA 597329. 210 5.3581E-04 2.6000E-05 9 CCX-XMTRX 595577. 97 5.3423E-04 2.6300E-06 10 IEV-LOCA24ND 182432. 121 1.6364E-04 1.6800E-05 11 IEV-LOCA24D 164625. 546 1.4767E-04 1.1300E-05 12 IEV-CCWD 164605. 2442 1.4765E-04 4.2300E-04 13 IEV-RNSD 164412. 1207 1.4748E-04 8.1500E-05 14 CCX-EP-SAM 132114. 229 1.1851E-04 8.6200E-06 15 IWX-CV-AO 23253.0 395 2.0858E-05 1.5000E-04 16 CCX-BY-PN 15420.5 363 1.3832E-05 4.7000E-05 17 CCX-IV-XR 15364.5 277 1.3782E-05 2.4000E-05 18 IEV-LOCAPRND 10504.6 59 9.4226E-06 1.5100E-05 19 ALL-IND-FAIL 7983.69 31 7.1614E-06 1.0000E-06 20 IWX-XMTR 5993.31 102 5.3760E-06 2.0100E-04
/N 21 IEV-LOSPD 4010.98 3158 3.5979E-06 1.4800E-03
[ } 22 CCX-PKAMOD1-SW 3231.76 128 2.8989E-06 1.1000E-05
\s_ ,/ 23 CCX-PMAMOD1X 3224.37 195 2.8923E-06 4.6300E-05 24 CCX-PMA030X 3216.48 130 2.8852E-06 1.3800E-05 25 CCX-EP-SAMX 3209.69 87 2.8791E-06 2.9400E-06 26 IWB-PLUG 3039.21 443 2.7262E-06 2.4000E-04 27 CCX-INPUT-LOGIC 1809.10 29 1.62285-06 1.0300E-04 28 CCX-XMTR195 1661.17 16 1.4901E-06 2.0100E-04 29 CCX-ORY-SP 1659.74 18 1.4888E-06 2.7700E-04 30 LPM-MAN 05 1405.66 37 1.2609E-06 6.8300E-04 31 PXX-AV-LA 766.061 57 6.8716E-07 6.1000E-05 32 IWN-MANDO 760.813 189 6.8245E-07 1.1500E-03 33 IWNTK001AF 757.783 5 6,7973E-07 2.4000E-06 34 PRX-HR-ML 753.969 4 6.7631E-07 1.2000E-07 35 IWX-XMTRLW 699.573 2 6.2752E+07 9.2000E-09 36 ICV-RNSND 480.250 46 4.3079E-07 9.6100E-04 37 IEV-CCWND 479.163 69 4.2981E-07 3.2100E-03 38 RNX-PM-FS 297.611 59 2.6696E-07 7.7000E-04 39 RNX-KV-GO 297.378 52 2.6675E-07 6.1000E-04 40 CCX-PL4 MOD 1 293.748 25 2.6349E-07 1.4100E-04 41 CCX-PL403 293.179 22 2.6298E-07 9.6900E-05 42 RNX-CV-GO 292.178 20 2.6208E-07 5.1000E-05 43 RNX-FM-ER 288.272 17 2.5858E-07 1.6000E-05 44 CCX-FM-ER 286.357 16 2.5686E-07 1.4000E-05 45 SWX-PM-ER 286.357 16 2.5686E-07 1.4000E-05 46 CCX-EP-SA 281.454 14 2.5246E-07 8.6200E-06 47 CCX-PL4 MOD 1-SW 281.170 14 2.5221E-07 1.1000E-05 48 CCX-LS-FA 274.967 32 2.4665E-07 5.3700E-06 49 IW10R170SPX 252.354 34 2.2636E-07 9.4600E-05 50 IW20R160SPX 252.354 34 2.2636E-07 9.4600E-05 51 IW1TL170UFX 252.066 33 2.2610E-07 6.8400E-05 52 IW2TL160CFX 252.066 33 2.2610E-07 6.8400E-05 53 RN23 HOD 5s 225.808 966 2.0255E-07 2.2100E-03 54 ED1 MOD 03 218.302 125 1.9582E-07 2.2020E-03 55 IDDBSDDITM 217.967 411 1.9552E-07 3.0000E-04 56 IDDBSDSITM 217.967 411 1.9552E-07 3.0000E-04 57 IDDBSDX1TM 216.098 404 1.9384E-07 3.0000E-04 q
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Revision: 11 3 Westingh0088 w March 1998 54B.49 ov.wv11wts4b.wpf.ib
y ;= 3 54B. Surge Line Flooding Effect on Low P;w r and Shutdow2 Risk Assessment I Table 54B-7 (Sheet 8 of 12) O i I BASIC EVENT IMPORTANCE BASELINE SHUTDOWN PRA l INCLUDING SURGE LINE FLOODING EFFECTS l SYSTEM UNAVAILABILITY (Q) = 8.970E-08 NUMBER OF BASIC EVENTS = 316 NUMBER OF CUTSETS = 8252 WUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVENT BASIC EVENT (% INCREASE) CUTSETS UNAVAILABILITY PROBABILITY 58 MDEP023SA 211.297 322 1.8953E-07 1.7100E-04 59 CCX-FMDMoD1 209.503 298 1.8792E-07 1.4100E-04 60 RHN-MANOS 206.337 833 1.8508E-07 1.6000E-03 61 CCX-FMD030 204.221 211 1.8319E-07 9.6900E-05 62 CCX-PMDMOD4 195.242 154 1.7513E-07 4.9800E-05 63 REA-PLUG 182.947 92 1.6410E-07 2.4000E-04 64 REB-PLUG 182.947 92 1,6410E-07 2.4000E-04 65 Zox-PD-Es 177.680 113 1.5938E-07 2.0000E-03 66 ED1 MOD 11 177.544 59 1.5926E-07 3.1700E-04 67 ED1M00113 177.544 59 1.5926E-07 3.1700E-04 68 ECX-CB-GO 177.469 92 1.5919E-07 1.2000E-03 69 ECX-CB-GC 176.970 74 1.5874E-07 7.3000E-04 70 CCX-ME-AF 176.889 5 1.5867E-07 1.2000E-06 71 Z0X-DG-DR 176.108 59 1.5797E-07 4.4000E-04 72 toX-DG-DS 175.764 54 1.5766E-07 2.8000E-04 73 IDDFD019RQ 175.340 69 1.5728E-07 1.2000E-05 14 IDDFD020RQ 175.340 69 1.572BE-07 1.2000E-05 75 CCX- FMDMOD1- SW 175.010 67 1.5698E-07 1.1000E-05 76 CCX-PMDMOD4-SW 175.010 67 1.5698E-07 1.1000E-05 77 ZOX-PD-ER 173.118 36 1.5529E-07 1.3000E-04 78 CCX-DY-PN1 168.032 20 1.5072E-07 5.7000E-05 79 ZOX-BL-ES 167.932 20 1.5064E-07 6.0000E-05 80 IWA-PLUG 165.841 182 1.4876E-07 2.4000E-04 81 IDDBSDD1LP 164.356 37 1.4743E-07 4.8000E-06 82 IDDBSDK1LF 164.356 37 1.4743E-07 4.8000E-06 83 IDDBSDS1LF 164.356 37 1.4743E-07 4.8000E-06 84 CCX-FMDEMO 163.696 35 1.4684E-07 4.0300E-06 85 PL5 mod 11 158.256 77 1.4196E-07 2.0900E-03 86 PL50301ASA 157.818 52 1.4156E-07 1.1600E-03 87 PL50301BSA 157.818 52 1.4156E-07 1.1600E-03 88 EDIMoD01 156.624 27 1.4049E-07 5.0400E-04 89 IRWMOD05S 156.476 498 1.4036E-07 3.2200E-03 90 tD1 mod 07 156.384 22 1.4028E-07 3.0500E-04 91 ED1 MOD 13 156.271 22 1.4018E-07 3.1700E-04 92 IDBBSDDITM 155.754 181 1.3971E-07 3.0000E-04 93 IDBBSDK1TM 155.754 181 1.3971E-07 3.0000E-04 94 IDBBSDS1TM 155.754 181 1.3971E-07 3.0000E-04 95 EC5EPMGBISA 155.504 14 1.3949E-07 1.7100E-04 96 CCX-PLIMOD1 155.383 14 1.3938E-07 1.4100E-04 97 CCX-PL7 MOD 1 155.383 14 1.3938E-07 1 4100E-04 9E CCX-PL103 155.065 12 1.3909E-07 9.6900E-05 99 CCX-PL703 155.065 12 1.3909E-07 9.6900E-05 100 PL5XS00ASA 154.857 12 1.3891E-07 8.0000F-05 101 CCX-IV-KR1 154.508 11 1.3859E-07 2.4000E-05 102 IRAEP121ASAX 151.836 63 1.3620E-07 5.8300E-05 103 IDBFD013RQ 149.812 20 1.3438E-07 1.2000E-05 104 IDBFD014RQ 149.812 20 1.3438E-07 1.2000E-05 105 CCX-PL1 MOD 1-SW 148.669 8 1.3336E-07 1.1000E-05 106 CCX-PL7 mod 1-SW 148.669 8 1.3336E-07 1.1000E-05 107 IDBBSDD1LF 148.576 15 1.3327E-07 4.8000E-06 108 IDB3SDK1LF 148.576 15 1.3327E-07 4.8000E-06 109 IDBBSDSILF 148.576 15 1.3327E-07 4.8000E-06 110 ED1BSDS1LF 142.645 7 1.2795E-07 4.8000E-06 111 ADN-MAN 01 135.823 58 1.2183E-07 4.9300E-04 112 REN-MAN 04 118.731 90 1.0650E-01 1.0000E-02 113 CCX-VS-FA 116.427 5 1.0444E-07 3.8400E-05 114 CCX-FMAMOD2X 101.789 22 9.1305E-08 8.5800E-05 115 CCX-INPUT-LOGICX 99.3341 12 8.9103E-00 2.5900E-05 Revision: 11 O March 1998 ov*wlisaits4b wpr:1w2259s 54B-50
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541 Surge Line Flooding Effect on Low Power and Shutdown Risk Assessment n i 8 l Table 54B 7 (Sheet 9 of 12) l l BASIC EVENT IMPORTANCE . BASELINE SIIUTDOWN PRA I INCLUDING SURGE-LINE FLOODING EFFECTS SYSTEM UNAVAILABILITY (Q) = 8.970E-08 NUMBER OF BASIC EVENTS e 316 NUMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVENT BASIC EVENT MINCREASE) CUTSETS UNAVAILABILITY PROBABILITY 116 CCX-IN-LOGIC-SW 99.1616 8 8.8948E-08 1.1000E-05 117 CCX-PMAMOD2-SW 99.1616 8 8.894BE-08 1.2000E-05 118 20X-BL-ER 70.7129 2 6.3430E-08 1.3600E-06 119 ZOX-FL-GP 70.6443 2 6.3368E-08 1.3000E-06 120 ED3 MOD 07 68.0177 34 6.1012E-08 3.0500E-04 121 IDAMOD04 66.5690 340 5.9712E-08 3.1700E-04 122 IDAMOD05 65.8673 354 5.9083E-08 5.1600E-04 123 MDAS 63.5642 75 5.7017E-08 1.0000E-02 124 IDABSDS1TM 62.7745 261 5.6309E-08 3.0000E-04 125 REC-MANDAS 61.4879 31 5.5148E-08 1.1600E-02 126 CCX-FMAMOD1 58.0902 87 5.2107E-08 1.4100E-04 127 CCX-FMA030 51.2455 59 4.5967E-08 9.6900E-05 128 PMAMOD11 44.7218 296 4.0115E-08 2.0900E-03 129 PMA0301ASA 43.3020 218 3.8842E-08 1.1600E-03 130 PMA0301BSA 43.3020 218 3.8842E-08 1.1600E-03 131 IWN-MAN 00C 32.7720 23 3.5676E-08 5.0000E-02 132 RHN-MAN 04 39.2606 134 3.5217E-08 5.5000E-02 133 PMAXS00ASAX 32.1484 158 2.8837E-08 8.0000E-05 134 PMAMOD11X 30.5491 369 2.7403E-08 6.5300E-04 135 IEV-LOSPND 30.4340 40 2.7299E-08 8.1300E-03
/*g 136 PMA0301ASAX 29.4585 176 2.6424E-08 1.6500E-04 l 137 PMA0301BSAX 28.7708 157 2.5807E-08 1.6500E-04
\__,<) 138 CCX-BC-SA 2?.1281 5 2.5231E-08 8.4000E-06 139 CCX-FMAMOD4 27.9389 41 2.5061E-08 4.9800E-05 140 PMAXS00ASA 27.5213 28 2.4687E-08 8.0000E-05 141 CCX-FMAMOD4-SW 26.3291 12 2.3617E-08 1.1000E-05 142 CCX-FMAEH0X 26.1668 11 2.3472E-08 4.0300E-06 143 IDABSDS1LF 21.7371 24 1.949BE-08 4.8000E-06 144 201 MOD 01 18.9482 625 1.6997E-08 2.0200E-02 145 EC1CB100VO 18.8458 483 1.6905E-08 1.2300E 02 146 201 MOD 04 16.5282 129 1.4826E-08 1.2500E-03 147 RHN-MAN 05C 16.3973 27 1.4708E 08 1.5000E-01 148 IDCDSDDITM 16.1748 183 1.4509E-08 3.0000E 04 149 IDCBSDSITM 16.1748 18.3 1.4509E-08 3.0000E-04 150 CMX-VS-TA 15.3773 7 1.3793E-08 3.8400E-05 151 IRWMOD06S 14.7967 495 1.3273E-08 3.2200E-03 152 IDCBSDK1TM 14.3056 176 1.2832E-08 3.0000E-04 153 IRAEP121BSAX 13.2658 66 1.1899E-08 5.8300E-05 154 ED2 MOD 03 11.6274 106 1.0430E-08 2.7000E-03 155 EC1 MOD 01 11.4691 31 1.028BE-08 2.0200E-04 156 EC2CB200VO 11.3921 186 1.0219E-08 1.2300E-02 157 202 MOD 01 11.3869 232 1.0214E-08 2.0200E-02 158 CC? a AV-LA 11.3862 5 1.0213E-08 6.1000E-05 159 CMX-CV-Go 11.3693 4 1.019BE-08 5.1000E-05 160 RNBMOD07S 11.1124 184 9.9678E-09 1.5300E-02 161 ED2 MOD 11 11.0183 39 9.8835E-09 3.1700E-04 162 IDCFD007RG 10.9226 18 9.7976E-09 1.2000E-05 163 IDCFD008RQ 10.9226 18 9.7976E-09 1.2000E-05 164 RNAMOD06S 10.9053 134 9.7821E-09 1.5300E-02 165 ZO2 MOD 04 10.7222 70 9.6179E-09 1.2500E-03 166 IDCBSDD1LF 9.74688 12 8.7430E-09 4.8000E-06 167 IDCBSDK1LF P.74688 12 8.7430E-09 4.8000E-06 168 IDCBSDS1LF 9.74688 12 8.7430E-09 4.8000E *6 169 SWAM 0007 9.44254 68 8.4700E-09 7.1200E-03 170 EC1REDG2GA 8.71011 78 7.8130E-09 4.3600E-03 171 EC1REDG1GA B.65225 65 7.7611E-09 4.3600E-03 172 Zo1 MOD 03 8.62842 17 7.7397E-09 1.0000E-04 n
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*u "I 5411. Surge Line Flooding Effect a Low-PowLr and Shutdowa Risk Assessment i
! Table 54B.7 (Sheet 10 of 12) e l I HASIC EVENT IMPORTANCE . BASELINE SHUTDOWN PRA INCLUDING SURGE.LINE FLOODING EFFECTS SYSTEM UNAVAILABILITY (Q) = 8.970E-08 WUMBER OF BASIC EVENTS = 316 NUMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN EYSTEM BASIC EVENT BASIC TVENT (% INCREASE) CUTSETS UNAVAILABILITY PROBABILITY 173 ED3 MOD 03 8.41710 45 7.5501E-09 2.2020E-03 174 RNAEPRNPSA 8.00535 20 7.1808E-09 1.7100E-04 175 kNBEPRNPSA 8.00535 20 7.1808E-09 1.7100E-04 176 EC1 MOD 12 7.85390 67 7.0450E-09 4.8000E-05 177 EC2 MOD 01 7.82538 24 7.0194E-09 2.0200E-04 178 EC1BS001TM 7.68995 442 6.8979E-09 2.7000E-03 179 EC1B5012TM 7.68995 442 6.8979E-09 2.7000E-03 180 EC1BS121TM 7.68995 442 6.8979E-09 2.7000E-03 181 SW7EPSBPASA 7.56273 17 6.8017E-09 1.7100E-04 182 CCAMoD02 7.27445 66 6.5252E-09 7.1000E-03 183 SWBMOD02 6.96320 174 6.2460E-09 2.4400E-02 184 CCBMOD015 6.93742 115 6.2229E-09 1.0400E-02 185 PL4 MOD 11 6.66072 68 5.9747E-09 2.0900E-03 186 OTH-R1 6.23561 2938 5.5933E-09 4.2000E-01 187 PL1M0011 6.14634 57 5.5133E-09 2.0900E-03 188 PL7M0011 6.14634 57 5.5133E-09 2.0900E-03 189 PL40301ASA 6.08319 44 5.4566E-09 1.1600E-03 190 IDAMOD06 5.87885 48 5.2733E-09 4.3200E-05 191 PL40301BSA 5,87825 39 $.2728E-09 1.1600E-03 192 SWAMOD09P 5.80292 40 3.2052E-09 1.8500E-03 193 SWBMOD09P 5.79557 45 5.1986E-09 1.8500E-03 194 ED3 MOD 01 5.75881 18 5.1657E-09 5.0400E-04 195 PL10301ASA 5.73985 38 5.1478E-09 1.1600E-03 196 PL70301ASA 5.73885 38 5.1478E-09 1.1600E-03 197 PL10301BSA 5.65700 36 5.0743E-09 1.1600E-03 198 PL70301BSA 5.65700 36 5.0743E-09 1.1600E-03 199 CC1EPSBPASA 5.41024 15 4.8530E-09 1.7100E-04 200 ECIMOD121 4.86970 24 4.3681E-09 1.6800E-05 201 E02M0003 4.81543 10 4.3194E-09 1.0000E-04 202 SW7EPSBPBSA 4.60188 12 4.1279E-09 1.7100E-04 203 CCIEPSBPBSA 4.60044 12 4.1266E-09 1.7100E-04 204 PMAMOD12X 4.34032 123 3.8933E-09 6.5300E-04 205 IDAMODOB 4.28259 96 3.8415E-09 3.1700E-04 206 SW7EPCTTASA 4.09820 10 3.6761E-?9 1.7100E-04 207 SW7EPCTFBSA 4.09820 10 3.6761E-09 1.7100E-04 208 PKA0302ASAX 3.68625 59 3.3066E-09 1.6500E-04 209 PMA0302BSAX 2.99850 40 2.6897E-09 1.6500E-04 210 EC1BS001LF 2.96532 7 2.6599E-09 4.8000E-06 211 IWACV122AO 2.83596 178 2.5439E-09 8.7600E-03 212 IWACV123AO 2.83596 178 2.5439E-09 b.7600E-03 213 IWACv124A0 2.83596 178 2.5439E-09 8.7600E-03 214 IWACV125A0 2.83596 178 2.5439E-09 8.7600E-03 215 PLIXS00ASA 2.71500 5 2.4354E-09 8.0000E-05 216 PLexS00ASA 2.71500 5 2.4354E-09 8.0000E-05 217 PL7xS00ASA 2.71500 5 2.4354E-09 8.0000E-05 218 IRWMOD10 2.$2699 36 2.2667E-09 1.4600E-03 219 IRWMOD12 2.52699 36 2.2667E-09 1.4600E-03 220 REACv119GO 2.52389 36 2.2639E-09 1.7500E-03 221 REBCV119GO 2.52389 36 2.2639E-09 1.7500E-03 222 IWCRS120BFA 2.4030b 23 2.1556E-09 8.7600E-04 223 IWDRS120AFA 2.40308 23 2.1556E-09 8.7600E-04 22A F.C1 MOD 13 2.23292 3 2.0029E-09 4.8000E-05 225 EC2 MOD 23 2.23292 3 2.0029E-09 4.8000E-05 226 EC2 MOD 22 2.21487 17 1.9867E-09 4.8000E-05 227 PMDMOD11 2.08806 235 1.8730E-09 2.0900E-03 228 IDDMOD32 1.80445 64 1.6186E-09 3.1700E-04 229 PMD0301ASA 1.7492J 157 1.5690E-09 1.1600E-03 230 PMDXS00ASA 1.72337 32 1.5459E-09 8.0000E-05 Revision: 11 O ENEL March 1998 ch 3 Weninghouse o \prawvilinu54b wpf Ib.022598 $4B-52
54H. Surge Line Flooding Effect on Low-Power and Shutdown Risk Assessment
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s / l Table 54B-7 (Sheet i1 of 12) l l BASIC EVENT IMPORTANCE - BASELINE SHUTDOWN PRA INCLUDING SURGE-LINE FLOODING EFFECTS SYSTEM UNAVAILABILITY (Q) a 8.970E-08 NUMBER OF BASIC EVENTS e 316 NUMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVENT BASIC EVENT (% INCREASE) CUTSETS UNAVAILABILITY PROBABILITY 231 RHN-MANO3 1.68647 49 1.5128E-09 2.2600E-03 232 PMD0301BEA 1.57699 138 1.4146E-09 1.1600E-03 233 PMDEHOA1SA 1.48049 28 1.3280E-09 8.0000E-05 234 td2BS002TM 1.35629 208 1.2166E-09 2.7000E-03 235 EC2B5022TM 1.35629 208 1.2166E-09 2.7000E-03 236 EC2BS221TM 1.35629 208 1.2166E-09 2.7000E-03 237 PhDMOD41 1.29380 74 1.160$E-09 6.3500E-04 238 PRAAV108LA 1.20074 11 1.0771E-09 1.0900E-03 239 PRBAV108LA 1.20074 11 1.0771E-09 1.0900E-03 240 PRAAV108TM 1.19824 9 1.0748E-09 5.0000E-04 241 PRBAV1087M 1.19824 9 1.0748E-09 5.0000E-04 242 IDDMOD33 1.15268 51 1.0340E-09 5.1600E-04 243 PL4 MOD 12 1.13696 30 1.0199E-09 2.0900E-03 244 CC3-MAN 01 1.11586 13 1.0009E-09 1.0700E-03 245 SUC-RIS 1.03523 220 9.2860E-10 5.8000E-01 246 REC-MANDASC .983838 46 8.8250E-10 5.0600E-01 247 ED3BSDSITM .862010 1 7.7322E-10 3.0000E-04 248 SWB-MANO2 .776113 8 6.9617E-10 1.6000E-03 249 PMAMOD21X .774267 14 6.9452E-10 1.1600E-03 250 CCX-XMTR .765253 3 6.8643E-10 2.0100E-04
.O 251 IRWMOD01 .759035 53 6.8086E-10 1.2000E-02 l i 252 IRWMOD03 .759035 53 6.8086E-10 1.2000E-02 (j 253 CCX-TRNSM .738780 4 6.6269E-10 2.0100E-04 )
254 PMDMOD12 .738766 97 6.6267E-10 2.0900E-03 J 255 IDDMOD38 .704917 34 6.3231E-10 3.1700E-04 1 256 PL40302ASA .689540 13 6.1852E-10 1.1600E-03 l 257 IRWMOD09 .649947 18 5.8300E-10 1.4600E-03 l 258 IRWMOD11 .649947 18 5.8300E-10 1.4600E-03 259 ADX-HV-GO .597042 4 5.3555E-10 1.1000E-03 260 PL1 MOD 12 .589902 17 5.2914F-10 2.0900E-03 261 PL7 MOD 12 .589902 17 5.2914E-10 2.0900E-03 262 IWARS11BBFA .567632 11 5.0917E-10 8.7600E-04 263 IWBRS118AFA .567632 11 5.0917E-10 8.7600E-04 264 IDDMOD34 .567257 10 5.0883E-10 4.3200E-05 i 265 REAMOV1177M .563617 9 5.0556E-10 5.0000E-04 1 266 REBMOV117TM .563617 9 5.0556E-10 5.0000E-04 ' 267 ZON-MAN 01 .546234 15 4.8997E-10 2.6700E-03 268 IDABSDDITM .536178 9 4.8095E-10 3.0000E-04 269 IDABSDK1TM .536178 9 4.8095E-10 3.0000E-04 270 PMD0302ASA .524562 51 4.7053E-10 1.1600E-03 271 PMDEHOA2SA .487308 10 4.37125-10 8.0000E-05 272 PL40302BSA .484596 8 4.3468E-10 1.1600E-03 273 PRN-MAN 01 .435143 4 3.9032E-10 4.0800E-04 l 274 IWBCV122AO .422568 184 3.7904E-10 8.7600E-03 l 275 IWBCV123AO .422568 184 3.7904E-10 0.7600E-03 276 IWBCV124AO .422568 184 3.7904E-10 8.7600E-03 277 IWBCV125AO .422568 184 3.7904E-10 8.7600E-03 278 PMD0302BSA .352337 32 3.160$E-10 1.1600E-03 279 PL10302ASA .345196 7 3.0964E-10 1.1600E-03 280 PL70302ASA .345196 7 3.0964E-10 1.1600E-03 281 PL1 MOD 51 .336464 5 3.0181E-10 8.7400E-04 282 EC2 MOD 221 .309421 3 2.7755E-10 1.6800E-05 283 PLMMOD41 .295302 4 2.6489E-10 6.3500E-04 284 PL10302BSA .263347 5 2.3622E-10 1.1600E-03 285 PL70302BSA .263347 5 2.3622E-10 1.1600E-03 286 SWNTP001RI .250696 10 2.2487E-10 5.2300E-03 287 PMAwoD41 .232580 34 2.0862E-10 6.3500E-04
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ENEL Revision: 11 UDM Lat h . March 1998 54B 53 c:4rairevil\ arts 4b.wpf: b
54B. Surg 3 Line Flooding Effect en Low-P;w:r and Shutdows Risk AssessmeEt l Table 54B-7 (Sheet 12 of 12) e l l BASIC EVENT IMPORTANCE - BASELINE SIIUTDOWN PRA INCLUDING SURGE-LINE FLOODING EFTECTS SYSTEM UNAVAILABILITY (Q) = 8.970E-08 NUMBER OF BASIC EVENTS s 316 NUMBER OF CUTSETS = 8252 NUMBER DECREASE IMPORTANCE OF IN SYSTEM BASIC EVEFTP BASIC EVENT (% INCREASE) CUTSETS UNAVAILABILITY PROBABILITY 288 IDDM0021 .214623 6 1.9252E-10 1.9200E-04 289 OTH-R2 .209131 29 1.8759E-10 2.4000E-01 290 ED2 MOD 01 .188508 1 1.6909E-10 5.0400E-04 291 ZANMOD01 .185640 8 1.6652E-10 8.4000E-05 292 CCNTF101RI .172161 10 1.5443E-10 5.2300E-03 293 Zo1DG001TM .171684 74 1,5400E-10 4.6000E-02 294 PMAEHOAISAx .143872 6 1.2905E-10 8.00 M1 05 295 PMDMOD42 .119796 12 1.0746E-10 6.350; 04 296 PMAMOD22X 9.581091E-02 2 8.5942E-11 1.1600J. 03 297 AD2 MOD 01 9.551045E-02 21 8.5673E-11 5.54GOE-02 298 AD2 mod 02 9.551045E-02 21 8.5673E-11 5.5400E-02 299 AD3 MOD 03 9.551045E-02 21 8.5673E-11 5.5400E-02 300 AD3 MOD 04 9.551045E-02 21 8.5673E-11 5.5400E-02 301 RHN-MAN 04-SUCC 8.874552E-02 468 7.9605E-11 9.4500E-01 302 DAS 8.137406E-02 4 7.2993E-11 1.0000E-02 303 IDAMOD02 7.310176E-02 31 6.5572E-11 2.7000E-03 304 IDAMOD03 7.310176E-02 31 6.5572E-11 2.7000E-03 305 RC1TL195UF 6.9612B9E-02 3 6.2443E-11 5.2300E-03 306 RC10R195SP 6.957354E-02 4 6.2408E-11 7.2200E-03 307 SUC-R2S 5.781296E-02 9 5.1858E-11 7.6000E-01 308 PMAMOD42 1.694677E-02 4 1.5201E-11 6.3500E-04 309 ZO2DG002TH 9.607883E-03 4 8.6183E-12 4.6000E-02 310 DG1-LOGIC 6.650192E-03 1 5.9652E-12 5.0000E-03 311 PRCEP108SA 6.513126E-03 1 5.8423E-12 1.7100E-04 312 PRDEP108SA 6.513126E-03 1 5.8423E-12 1.7100E-04 313 IDAMOD07 1.857359E-03 15 * .6661E-12 2.1900E-02 314 IDDM0022 1.508667E-03 3 1.1533E-12 2.7000E-03 315 IDDMOD23 1.508667E-03 3 1.3533E-12 2.7000E-03 316 IDDMOD35 2.598091E-04 4 2.3305E-13 2.1900E-02 Revision: 11 e March 1998 % T Westinghouse o:WWvilWtS4b wpf;lt>.022598 $4B.54
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