ML20198P417
| ML20198P417 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 02/29/1996 |
| From: | Lindgren D, Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20198P397 | List: |
| References | |
| GW-GL-021-01, GW-GL-021-R05, GW-GL-21-1, GW-GL-21-R5, NUDOCS 9901070049 | |
| Download: ML20198P417 (450) | |
Text
{{#Wiki_filter:. _ . . _ . __. _ _ AP600 DOCUMENT COVER SHEET TDC: IDS: 1 S Form 58202G(S/94)ItAxxxx.wpf:1x1 AP600 CENTRAL FILE USE ONLY: 0058.FRM RFS#: RFS ITEM #: 1P600 DOCUMENT NO. REVISION NO. ASSIGNED TO Control Copy #
- GW-GL-021 5 Page 1 of___ y/@, c,77 ALTERNATE DOCUMENT NUMBER: N/A WORK BREAKDOWN #: 3.2.3 l DESIGN AGENT ORGANIZATION: Westinghouse TITLE: AP600 Standard Safety Analysis Report 1
I ATTACHMENTS: DCP #/REV. INCORPORATED IN THIS DOCUMENT REVISION: CALCULATION / ANALYSIS
REFERENCE:
ELECTRONIC FILENAME ELECTRONIC FILE FORMAT ELECTRONIC FILE DESCRIPTION (C) WESTINGHOUSE ELECTRIC CORPORATION 19_S6 g @ WESTINGHOUSE PROPRIETARY CLASS 2 ( This docun cnt contains information proprieutry to Westinghouse Electric Corpo>ation; it is submitted in confidence and is to be used solely for the ( - purpose for which it is furnished and returned upon request. This document and such information is not to be reproduced, transmitted, disclosed or used otherwise in whole or in part without prior wntten authorization of Westinghouse Electric Corporation, Energy Systems Business Unit, subject to the legends contained hereof. O WESTINGHOUSE PROPRIETARY CLASS 2C This document is the property of and contains Proprietary information owned by Westinghouse Electric Corporation and/or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document in strict accort;ance with the terms and condtions of the agreement under which it was provided to you. O WESTINGHOUSE CLASS 3 (NON PROPRIETARY) COMPLETE 1 IF WORK PERFORMED UNDER DESIGN CERTIFICATION OR ~ COMPLETE 2 IF WORK PERFORMED UNDER FOAKE 1 @ DOE DESIGN CERTIFICATION PROGRAM - GOVERNMENT LIMITED RIGHTS STATEMENT ISee page 21 Copyright statement A license is reserved to the U.S. Govemment under contract DE ACO3-90SF18495.
@ DOE CONTRACT DELIVERABLES (DELIVERED DATA)
Subject to specified exceptions, disclosure of this data is restricted until September 30,1995 or Design Certif'cation under DOF contract DE-AC03-90SF18495, whichever is later. EPRI CONFIDENTIAL: NOTICE: 10 2 30 4 s O CATEGORY: A3 B C D E F 2 O ARC FOAKE PROGRAM - ARC LIMITED RIGHTS STATEMENT ISee page 21 Copyright statement A license is reserved to the U.S. Govemment under contract DE-FC02-NE34267 and subcontract ARC-93-3-SC401. O ARC CONTRACT DELIVERABLES (CONTRACT DATA) Subject to specified excephons, disclosure of this data is restncted under AJig Sybcontract ABC 93-3 SC401. ORIGINATOR SIGNA RE/DATE / / Donald A. Lindgren n (( . yg/ g ~ pg AP600 RESPONSIBLE MANAGER SIG APPROVAL DATE Brian A. MCIntyre TURE[*[h' dd gggg g,= ' fes t document is complete, all Egdited revnws are complete, electfbne file is attached and document is PDR ADOCK 052000031 E pg g c
AP600 DOCUMENT COVER SHEET Pago 2 Form 58202G(5/9A) LIMITED RIGHTS STATEMENTS DOE GOVERNMENT LIMITED RIGHTS STATEMENT These data are submitted with limited nghts under govemment contract No. DE-AC03-90SF18495. The se data may be reproduced and (A) used by 8 e govemment with the express limitation that they wili not, without wntten permission of the contractor, be used for purposes of manufa':turer nor disclosed outside the govemment; except that the govemment may disclose these data outside the govemmerit for the folicwing purposes, if any, provided that the govemment makes such disclosure subject to prohibition against further use and disclosure: (1) TNs "Propnetary Data
- may be disclosed for evaluation purposes under the restnctions above.
(11) The 'Propnetary Data
- may be dsclosed to the Electnc Power Research inst;tute (EPRI), electric utility representatives and their direct consultants, exduding drect commercial competitors, and the DOE National Laboratones under the prohibitions and restnctons above.
TNs notice shall be marked on any reproduction of these data, in whole or in part. (B) ARC LIMfTED RIGHTS STATEMENT: TNs propnetary data, fumished under Subcontract Number ARC-93-3-SC-001 with ARC may be duplicated and used by the govemment and ARC, subject to the limstahons of Arbde H 17.F. of that subcontract, with the express limitations that the proprietary data may not be disdosed 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 ths Subcontractor, except that further disclosure or use may be made solely for the following purposes: This proprietary data may be disdosed to other than commercial competitors of Subcontractor for evaluation purposes of this subcontract under th3 restnction that the proprietary data be retained in confidence and not be further dsclosed, and subject to the terms of a non-disclosure agreement between the Subcontractor and that organization, exduding DOE and its contractors. DEFINITIONS CONTRACT / DELIVERED 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 / OBLIGATIONNOTICES NOTICE 1: The data in this document is subject to no conhdentiality obligations. NOTICE 2: The data in this documentis proprietary and confidential to Westinghouse Electric Corporation and/orits Contractors, itis forwarded to recipient under an obligation of Confidence andTrust for limited purposes only. Any use, dsclosure to unauthorized persons, or copying of this document or parts thereof is prohibited except as agreed to % advance by the Electric Power Research Insttute (EPRI) and Westinghouse Electric Corporation. Recipient of this data has a duty to inquire o' EPRI andor Westinghouse as to the uses of the informabon contained herein that are permitted. and con 6dential to Westinghouse Electric Co ration andorits Contractors. Itis forwarded NOTICE to recipient3:under The andata in this of obligation document Confidenceis and proprietar; Trust for use only in evaluation tasks specif fy authonzed by the Electric Power Research Insttute (EPRI). Any use, dtsdosure to unauthorized persons, or copying this document or parts thereof is prohibited except as agreed to in advance by EPRI and Westinghouse Electne Corporation. Recipient of his data has a duty to inquire of EPRI andor Weshnghase as to the uses of the information contained herein that are permitted. This docui vnt and any copies or excerpts thereof that may have een generated are to be retumed to Westnghouse, drectly or through EPRI, when rer4uested to do so. NOTICE 4: The data in this document is proprietary and confiderstial to Weshnghouse Electric Corporation andor its Contractors. It is being r:vealed in confidence and trust only to Employees of EPRI and to certain contractors of EPRI for limited evaluation tasks authorized by EPRI. Any use, declosure to unauthonzed persons, or copying of this document or parts thereof is proNbited. This Document and any copies or excerpts thereof that may have been generated are to be returned to Westinghouse, directly or through EPRI, when requested to do so. NOTICE 5: The data in this document is proprietary and confidential to Westinghouse Electric Corporation andor its Contractors. Access to this data is given in Confidence and Trust only at Westnghouse facilites for limited evaluation tasks assigned by EPRI. Any use, dis-losure to unauthonzod persons, or copying of this document or parts thereof is proNbited. Neither tNs document not any excerpts therefrom are to bs removed from West nghouse facilities. EPRI CONFIDENTIALITY / OBLIGATION CATEGORIES CATEGORY " A* -(See Delivered Data) Consists of CONTRACTOR Foreground Data that is contained in an issued reported. CATEGORY 'B' -(See Delivered Data) Consists of CONTRACTOR Foreground Data that is not contained in an issued report, except for computer programs. CATEGORY 'C"- Consists of CONTRACTOR Background Data except for computer programs. CATEGORY *D' - Consists of computer programs developed in the tx>urse of performing the Work. CATEGORY *E'- Consists of cornputer programs developed prior to the Effective Date or after the EffecCve Date but Ntside the scope of the Work. CATEGORY 'F'- Consists of administrative plans and administrative reports. O
i Instruction Sheet for SSAR Revision 5 : Enclosed you will find chapters to replace your current version of the SSAR: PACKAGE I Remove Present Pages: Replace with the Enclosed Pages: Volume 1 (Only) Master Table of Contents tab . ; List of Effective Pages pages 1 through 41 L.ist of Effective Pages tab I List of Effective pages 1 through 39 Change Roadmap tab Change Roadmap , 'ges R-1 through R-23 : Volumes 1 through 11 Master Table of Contents pages iii through xxxiv Master Table of Contents pages lii through xxx Chapters 1 through 18 Tables of Contents Chapters 1 through 18 Tables of Contents ' PACKAGE 2 (Information already sent to NRC in 1995 as a hand markup) i Volume 4 6.2-1 through 6.2-127 6.2-1 through 6.2-236 1 Volume 8 15.0-1 through 15.0-33 15.0-1 through 15.0-36 15.1-1 through 15.1-85 15.1-1 through 15.1-89 15.2-1 through 15.2-77 15.2-1 through 15.2-80 : 15.3-1 through 15.3-41 15.3-1 through 15.3-33 I 15.4-1 through 15.4-80 15.4-1 through 15.4-87 Volume 9 15.5-1 through 15.5-23 15.5-1 through 15.5-26 l 15.6-1 through 15.6-208 15.6-1 through 15.6-227 15.7-1 through 15.7-10 15.7-1 through 15.7-P ' 15.8-1 15.8-1 SSAR Rev 5 Change Sheet Instructions - Page 1 of 4
l l ( INSTRUCTION SHEET FOR SSAR REVISION 5 (CONT) REMOVE PRESENT PAGES: REPLACE %Tni UIE ENCLOSED PAGES: l, PACKAGE 3 (Revised Information) i Volume 1 i 1.1-1 through 1.1-6 1.1-1 through 1.1-10 ! l 1.3-1 through 1.3-7 1.3-1 through 1.3-7 2-1 through 2-15 2-1 through 2-18 l 2A-25 and 2A-26 2A-25 and 2A-26 2A-29 and 2A-30 2A-29 and 2A-30 2A-37 through 2A-91 2A-37 through 2A-90 28-1 through 2B-115 2B-1 through 2B-118 Volume 2
** Add new Appendix 2C tab and Section Appendix 2C - pages 2C-1 through 2C-16 **
3.5-1 through 3.5-15 3.5-1 through 3.5-16 Volume 3 3.9 -1 through 3.9-169 3.9-1 through 3.9-172 3,10-1 through 3.10-6 3.10-1 through 3.10-8 3C-1 through 3C-9. 3C-1 through 3C-5 3D-1 through 3D-110 3D-1 through 3D-126 4.4-1 and 4.4-2 4.4-1 and 4.4-2 4.4-31 and 4.4-32 4.4-31 and 4.4-32 4.5-1 through 4.5 4 4.51 through 4.5-4
- Volume 4 5.1-1 through 5.1-12 5.1-1 through 5.1-23 5.2-1 through 5.2-36 5.2-1 through 5.2-37 h
5.3-1 through 5.3-32 5.3-1 through 5.3-32 5.4-1 through 5.4-99 5.4-1 through 5.4-103 l SSAR Rev 5 Change Sheet Instructions - Page 2 of 4
Instruction Sheet for SSAR Revision 5 (Cont) Remove Present Pages: Replace with the Enclosed Pages: -. 6.0-1 and 6.0-2 6.01 and 6.0-2 6.3-1 through 6.3-75 6.3-1 through 6.3-79 6.4-1 through 6.4-12 6.4-1 through 6.4-15 6.6-1 through 6.6-4 6.6-1 through 6.6-4 7.1-1 through 7.1-60 7.1-1 through 7.1-61 f 7.2-1 through 7.2-115 7.2-1 through 7.2-65 7.3-1 through 7.3-34 7.3-1 through 7.3-34 7.4-1 through 7.4-15 7.4-1 through 7.4-15 7.6-1 through 7.6-9 7.6-1 through 7.6-5 7.7-1 through 7.7-2 7.7-1 through 7.7-26 Volume 5 9.1-1 through 9.1-65 9.1-1 through 9.1-59 Volume 7 10.1-1 through 10.1-9 10.1-1 through 10.1-7 10.2-1 through 10.2-19 10.2-1 through 10.2-27 Volume il 17.1-1 and 17.1-2 17.1-1 and 17.1-2 SSAR Rev 5 Change Sheet Instructions - Page 3 of 4 (' _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
J l INSTRUCTION SHEET FOR SSAR REVISION 5 (CONT) 4 l REMOVE PRESENT PAGES: REPLACE WTTH THE ENCLOSED PAGES: PROPRIETARY PACKAGE , I i Remove the Master Table of Contents lasert the Master Table of Contents in each in each of the three Proprietary Volumes - pages of the three Proprietary Volumes - pages lii through Pxii Piii through Pxi Remove all front matter from behind chapter tabs Volume 2 Remove 5.1 tab and pages P5.1-1 through P5.1-9 Remove page PS.4-3 Remove 6.3 tab and pages P6.3-1 through P6.3-7 Remove 6.4 tab and page P6.4-1 Remove Chapter 7 tab and 7.1, 7.2, and 7.7 tabs Remove Chapter 8 and 8.3 tabs Volume 3 i Remove page P9.1-1 Insert page P9.1-1 (Format Change) i L Remove Appendices 15A, ISB,15C, ISD and 15E text and tabs [ Replaced by WCAP-14601 (Proprietary)] l i l i ) s SSAR Rev 5 Change Sheet instructions - Page 4 of 4
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o Package 1 of 3 o O e
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N - Volume 1 Master Table of Contents C) List of Effective Pages Change Roadmap f (Insert infront of Volume 1) o O
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i l QJ MASTER TABLE OF CONTENTS Section Title P_aage VOLUME 1 CHAPIER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT . . ...... 1.1-1 1.1 Introduction . . . ........ ..... .... . ..................... 1.1-1 1.1.1 Plant Location . . . . . .............. ............ ..... 1.1-1 1.1.2 Containment Type . . . ........ ....... ......... ..... 1.1-1 1.1.3 Reactor Type . . . . . . . . . . . ..... ................... .. 1.1-1 1.1.4 Power Output . . . . . . . . . . . ............................ 1.1-1 1.1.5 Schedule . . . . . . . ................................... 1.1-1 1.1.6 Format and Content ..... ........................ .... 1.1-2 1.1.7 Combined License Information . . . . . . . . . . . . . . . . . .......... 1.1-3 1.2 General Plant Description . . . . .. .............................. 1.2-1 1.2.1 Design Criteria, Operating Characteristics, and Safety Considerations . 1.2-1 1.2.2 Site Description . . . . . ... ................... . . . . . . . 1.2-14 1.2.3 Plant Arrangement Description . . . . . . . . ............ . . . . . 1.2-16 1.2.4 Nuclear lsland . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . 1.2-16 1.2.5 Annex Building ........ ....................... ... 1.2-25 ,, 1.2.6 Diese1 Generator Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-25
; 1.2.7 Radwaste Building . . . . . . . . . . . ..................
V . . . . 1.2-26 1.2.8 Turbine Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1,2-27 1.2.9 Combined License Informa* ion . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-27 1.3 Comparisons with Similar Facility Designs . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-1 1.4 Identification of Agents and Contractors . . . . ...................... 1.4-1 1.4.1 Applicant - Program Manager . . . . . . . . . . . . . . . . ........... 1.4-1 1.4.2 Other Contractors and Participants ................... .... 1.4-1 1.4.3 Combined License Inforrnation . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.4-3 1.5 Requirements for Further Technical Information . . . . . . . . . . . . . . . . . . . . . . 1.5-1 1.5.1 AP600 Safety-Related Tests . . . . . . . . . . . . . . . . . . . .......... 1.5-1 1.5.2 AP600 Component Design Tests . . . . . . . . . . . . . . . . . . . . . . .. 1.5-7 1.5.3 Combined License Information . . . . . . . ....... ........... 1.5-8 1.5.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... 1.5-8 1.6 Materia 1 Referenced ..................................... ... 1.6-1 1.7 Drawings and Other Detailed Information ...... .............. ... 1.7-1 1.7.1 Electrical and Instrumentation and Control Drawings . .......... 1.7-1 1.7.2 Piping and Instrumentation Diagrams . . . . . . . . . . . . . . . . . . . . . . . 1.7-1 1.8 Interfaces for Standard Design . . . . . . . . . . . . . . . .................. 1.8-1 1.9 Compliance with Regulatory Criteria . . . . . . . . . . . ....... ....... 1.9-1 1.9.1 Regulatory Guides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-1 1.9.2 Compliance with Standard Review Plan (NUREG-0800) ..... ... 1.9-2 1.9.3 'Ihree Mile lsland Issues . . . . . . . . . . . . . . . . . . . . . . . . . ..... 1.9-2 m 1
/
Revision: 5 ' [ W85Dngh00S8 Febnrary 29,1996
samrm i ' i l O! 1 MASTER TABLE OF CONTENTS (Cont.) Section Title .P_ age 1.9.4 Unresolved Safety Issues and Generic Safety Issues . . . . . . . . . . . 1.9-19 1.9.5 Advanced Light Water Reactor Certillcation Issues . . . . . . . . . . . 1.9-44 IS.6 ' References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-6 8 APPENDIX 1 A CONFORMANCE WITH REGULATORY GUIDES . . . . . . . . . . . . . . . . . l A- 1 1 A.1 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 1A-86 APPENDIX IB SEVERE ACCIDENT MITIGATION DESIGN ALTERNATIVES . . . . , . 1B-1 IB.1 Introduction . . . . . . .............. ................ .... . . IB-1 IB.2 Summary . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . I B - 1 IB.3 Selection of S AMDAS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I B-2 1B.4 Methodology . . . . . . . . . . . . . . . . . . . . . . . . ........ ... .... . . . I B-2 IB.4.1 Risk Reduction ............ ... . . . . . . . . . . . . . . . . . . . . I B -2 IB.4.2 Capital Cost Estimates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1B-3 ! 1B.4.3 Cost Benefit Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . IB-3 IB.5 PRA Release Categories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 B-3 1B3.1 Release Category OK . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I B -4 IB3.2 Release Category OKP . . . . . . . . . . . . . . . . . . . . ....... . . . I B-4 IBS.3 Release Category CC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I B -5 IB5.4 Release Category CI . . . . . . . . . . . . . . . . . . . . ....... . . . . . I B-5 1B.6 Total Population Dose . . . . . . . . ...............................IB-5 IB.7 SAMDA Description and Benefit . . . . . . . . . . . . ...................IB-6 IB.7.1 Upgrade the CVCS for Small LOCAs . . . . . . . . . . . . . . . . . . . . . . IB-6 IB.7.2 Filtered Vent . . . ........................ .. ..... . IB-6 IB.7.3 Locate Normal RHR Inside Containment . . . . . . . . . . . . . . . . . . . I B -7 1B.7.4 Self-Actuating Containment Isolation Valves . . . . . . . . . . . . . . . . . IB-7 IB.7.5 Passive Containment Sprays . . . . . . . . . . . . . . . . . . . . . . . . . . . . IB-7 IB.7.6 Active High Pressure Safety Injection System . . . . . . . . . . . . . . 1 B -7 IB.7.7 Steam Generator Shell-Side Heat Removal System . . . . . . . . . . . . IB-8 1B.7.8 Direct Steam Generator Relief Flow to the IRWST . . . . . . . . . . 1 B-8 l 1B.7.9 Increased Steam Generator Pressure Capability . . . . . . . . . . . . . . I B -8 1B.7.10 Secondary Containment Filtered Ventilation ..... .. . . . . . . . I B-8 IB.7.11 Diversify the IRWST Discharge Valves .. ........ . . . . . 1 B-9 1B.7.12 Ex-Vessel Core Catcher . . . . ................. . . . . . . I B -9 l 1B.7.13 High Pressure Containment Design . . . . . . . . . . . . . . . . . . . . . . 1 B-9 1B.7.14 Increase Reliability of Diverse Actuation System . . . . . . . . . . . . 1 B-9 1B.8 Results . . . . . . . . . . .............................. .. . . . . . I B-9 1B.9 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 B-11 i O l Revision: 5 February 29,1996 ij T Westinghouse
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g U MASTER TABLE OF CONTENTS (Cont.) Section Title M CHAITER 2 SITE CHARACTERISTICS ..
........ ... ... ... . . ... 2-1 2.1 Geography and Demography . . ... . . . .... .... . . 2-1 2.1.1 Combined License Information for Geography and Demography . . . 2-1 2.2 Nearby Industrial, Transportation, and Military Facilities . . . . . . . . ... . 2-1 2.2.1 Combined License Information for Identification of Site-Specific Potential Hazards . . . . ....... . .. . .. . ........ . 2-2 2.3 Meteorology ...... ... ...... . . ....... ... .. ..
..... 2-2 2.3.1 Regional Climatology . . . .. . . . . . . . 2-3 2.3.2 Local Meteorology . . ....... .... ......... ... . 2-3 2.3.3 Onsite Meteorological Measurement Programs . ... . . ... 2-3 2.3.4 Short-Term Diffusion Estimates . . ........... . ... .. 2-3 2.3.5 Long-Term Diffusion Estimates . . .. . . . . ... ... 2-4 2.3.6 Combined Licensed Information . . . . . ..... . . ....... . 2-4 2.4 Hydrologic Engineering . . . . . .. .... .. .... ... . ......... 2-5 2.4.1 Combined License Information ...... .. ........ . . . . 2-5 2.5 Geology, Seismology, and Geotechnical Engineering . . . . . . .......... 2-6 2.5.1 Basic Geological and Seismic Combined License Information . . . 2-6 2.5.2 Vibratory Ground Motion ............. .. 2-7
[] v 2.5.3 2.5.4 Surface Faulting Combined License Information . . . . . . . .. ... 2-7 Stability of Subsurface Materials and Foundation . . ... .. .... 2-8 2.5.5 Combined License Information for Stability of Slopes . . . . . . . . 2-12 2.5.6 Combined License Information for Embankments and Dams . . . .. 2-12 2.5.7 References . . . . . . . . . . . . . . ................. ... . 2-12 APPENDIX 2A DESIGN SOIL PROFILES . . . ................. . . . . . . . . . . . . . 2 A- 1 2A.1 Survey of a Selection of Existing Nuclear Power Plant Sites in the U.S. . . . . 2A-1 2A.2 Generic Soil Profiles . .. ........ ............. . ....... 2A-2 2A.3 Applicability of Design Ground Motion to Generic Soil Proflies . . . . . . . . 2A-4 2A.4 Free-Field Site Response Analysis . . . . . . . . . . . . . . ...... ... 2A.5
. . 2A-5 Two-Dimensional Soil-Structure Interaction Analysis .. ........... . 2A-7 2A.6 AP600 Design Soil Profiles . . . . . . . . . .......... . .. ...... 2A-12 2A.7 References . . . ......................... ......... .... 2A-14 APPENDIX 2B PARAMETRIC STUDIES RELATED TO AP600 DESIGN SOIL PROFILES . . . . ............. ... .. ....... ..... ... . 2B-1 2B.1 2D SSI Analysis Parameters and Procedures . ..... . ..... .... . 2B-2 2B.1.1 Structural Models . . . . . . . . .. ... .... ......... . . 2 B-2 2B.I.2 SASSI Foundation Models ...... .. . .... . . . . . . . . 2B-2 2B.1.3 Soil Properties for SSI Analysis . .... .. ....... ... . 2B-2 2B.1.4 Input Motion . . . . . . . .... .... . ... .... .. . . 2B-3 2 B.1.5 Two-Dimensional Enveloped SSI Results . . . . ...... . . . . . 2B-3 A
V Revision: 5 J S8 WBSdnl$10' iij February 29,1996
m O MASTER TABLE OF CONTENTS (Cont.) Section Title a Page 2B.2 Design Soil Profiles . . . . . . . . . . . . . . . . . . . . . . . ............. . . 2B-3 2B.2.1 Lower Bound of Soft Soil Profile (V,=707 fps) .... ... . . . . 2B -4 2B.2.2 Firm Rock Profiles (V,=3500 fps) ............... . . . . 2B.4 2B.2.3 Parabolic Distribution of Soil Profile with Sandy Soil Material . . . 2B-5 2B.2.4 Variation of Soil Properties and 60 Percent Requirement of Free-Field Spectra at Foundation Level . . . . . . ....... .... 2B-5 2B.2.5 AP600 Design Soil Profiles . . . . . . . . . . . . . .. .... . 2B-6 2B.3 Soil Profile Parametric Studies: . ................. ....... .. 2B-6 2B.3.1 Depth-to-B ase Rock . . . . . . . . . . . . . . . . . . . ...... . . . 2B-7 2B.3.2 Soil Curves for Clay Material . . . . . . . . . . . . . .. . . .. . . 2B-7 2B.3.3 Poisson's Ratio for Soil above Water Table . . . . . . . . . . . . . . . . . 2B-8 2B.3.4 Fixed-Base vs. SSI Analysis for the Hard Rock Site . . . . . . . . . . . 2B-8 2B.3.5 Idriss 1990 Soil Curves . . . . . ........ . ..... .. ... 2B-9 2B.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2B-9 VOLUME 2 APPENDIX 2C SEISMIC LATERIAL EARTH PRESSURES . . . . . . . . . . .. .. .. . 2C-1 2C.1 2D SASSI Analysis for Seismic Lateral Earth Pressure . . . . . ...... . . 2C-1 2C.1.1 Analysis Method . . . . . . . . . . . . . . . . . .............. . . 2C-1 2C.I.2 Structural Models . . .......... ......... .. ... . . 2C-1 2C.I.3 SASSI Foundation Models . . . . . . .... ......... ..... . 2C-1 2C.1.4 Representation of Adjacent Structures . . ................ . 2C-2 2C.1.5 Input Motion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2C-2 2C.I.6 Soil Properties . . . . . . . . . . . . . . ............ ..... . . 2C-2 2C.1.7 2D SASSI Analyses Results . . . . .. ...... ........ . . . 2C-2 2C.2 Effect from Torsional Motion of the Nuclear Island . . . . . . . . . . . . . . . . 2C-2 2C.3 Local Distribution of Lateral Earth Pressure ....... ....... ... ... 2C-3 2C.4 References . . . . ............... .. . .. .... ......... . . 2C-4 CHAPIER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS . . . . . . . . . . . . . . . . . . . .. ...... ........ . 3.1-1 3.1 Conformance With Nuclear Regulatory Commission General Design Criteria . 3.1-1 3.1.1 Overall Requirements . . . . . . . . . ......... . ...... .. 3.1-1 3.1.2 Protection by Multiple Fission Product Barriers . . . . . . ..... 3.1-3 3.1.3 Protection and Reactivity Control Systems . . ....... . .. .. 3.1-9 3.1.4 Fluid Systems . . . . . . . . . . . . . . . . . . . . . ....... .. . . 3.1-13 3.1.5 Reactor Containment .. ....... ...... .. . . ... . 3.1-21 l 3.1.6 Fuel and Reactivity Control . .......... ... ... ..... . 3.1-25 l 3.1.7 References . . . . . . . . . . ..... .. . . .. ... .... . 3.1-28 1 1 Revision: 5 February 29,1996 ;y T Westinghouse 1-______.__-___ _ - __-- __ _-. - - _ - - _ . _ _ - _ _ _ . _ - _ . - -
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MASTER TABLE OF CONTENTS (Cont.) Section Title Page 3.2 Classification of Structures, Components, and Systems ........ . . 3.2-1 3.2.1 Seismic Classification . . . .... ....... ...... . 3.2-1 3.2.2 AP600 Classification System .... . .. ........ .... . 3.2-2 3.2.3 Inspection Requirements . . . . . . . . . . . . . . .. .. . .... 3.2-9 3.2.4 Application of AP600 Safety-Related Equipment and Seismic Classification System . . . .............. . ........ . 3.2-9 3.2.5 References . ..... ... .. .... .. . .. ... ... 3.2-12 3.3 Wind and Tornado Loadings . .. ......... ..... .... ... ... . 3.3-1 l 3.3.1 Wind Loadings . ... ..... .... 3.3-1 3.3.2 Tornado Loadings . . . . . ....... . ......... ... . . . 3.3-2 l 3.3.3 Combined License Information ... . ..... 3.3-4 l 3.3.4 References . . . . . ... ... .. ..... .... ... . . 3.3-4 ! 3.4 Water Level (Flood) Design .. ... ..... .. .. .. . . .. ..... 3.4-1 3.4.1 l Flood Protection .. ... ....... .. .. .... . . .. . 3.4-1 1 3.4.2 Analytical and Test Procedures . . . . . . . . ............... 3.4-23 3.4.3 Combined License Information .... . . . ....... . . . 3.4-23 3.4.4 References . . . . ............ ...................... 3.4-24 3.5 Missile Protection . . . . . . . . ....... . .... 3.5-1 O V 3.5.1 Missile Selection and Description . . . . . . . . . . . . ...... .. . . 3.5-4 3.5.2 Protection from Externally Generated Missiles . . . . . . .... . . 3.5-13 3.5.3 Barrier Design Procedures . . . . . . . . . . . . . . .. . ..... . 3.5-13 3.5.4 Combined License Information ... ............. . . . . . 3.5-16 3.5.5 References . . . . . . ......... ................... . . 3.5-16 3.6 Protection against the Dynamic Effects Associated with the Postulated Rupture of Piping . . . . . . . . . . . . . . . . . . .......... .... 3.6-1 3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside Containment ................ ... ....... ........ 3.6-2 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping . . . . . . ... . 3.6-12 3.6.3 Leak-before-Break Evaluation Procedures . . . . . . . . . .. . . . 3.6-28 3.6.4 Combined License Information . .. ........... . . . . . . 3.6-35 3.6.5 References . . . . . .. ...... . ... . .... . . .... 3.6-35 3.7 Seismic Design . .......... ......... ...... . . .... 3.7-1 3.7.1 Seismic Input . . . . .......... ... ... .. ... .. . 3.7-1 3.7.2 Seismic System Analysis . . ............ ... ...... .. 3.7-7 3.7.3 Seismic Subsystem Analysis . . ... .. ....... .... ... 3.7-19 3.7.4 Seismic Instrumentation . . . . . . . . . . . . . . . . .. .... .. . 3.7 36 3.7.5 Combined License Information .. . .......... . . . . . 3.7-39 3.7.6 References . . . . . . . ............. ...... . . . . 3.7-39 i m
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O MASTER TABLE OF CONTENTS (Cont.) Section Title Page 3.8 Design of Category I Stmetures .... .. .......... .... ....... 3.8-1 3.8.1 Concrete Containment . . . . . . ... ........ . .. ... .. 3.8-1 3.8.2 Steel Containment ........... ............... . 3.8-1 3.8.3 Concrete and Steel Internal Structures of Steel Containment . . 3.8-16 3.8.4 Other Category I Structures . . . . .. .......... . . . . . 3.8-24 3.8.5 Foundations . . . .. ........... .. .. . . .... . 3.8-37 3.8.6 Combined License Information .... ....... . . . . . . . 3.8-43 3.8.7 References . . . . . . ............. .... . .... . 3.8-43 VOLUME 3 3.9 Mechanical Systems and Components . . . . .. ..... ... .......... 3.9-1 3.9.1 Special Topics for Mechanical Components . . ...... . ..... 3.9-1 3.9.2 Dynamic Testing and Analysis . . . . . . . . . . . . . . . . . . . . . . . . 3.9-30 3.9.3 ASME Code Classes 1,2, and 3 Components, Component Supports, and Core Support Structures . . . . . . . . . . . . . . . . . 3.9-45 3.9.4 Control Rod Drive System (CRDS) . . ........ . . . . . . . 3.9-71 3.9.5 Reactor Pressure Vessel Internals . . . . . . ........... . . . 3.9-82 3.9.6 Inservice Testing of Pumps and Valves . . . . . . . . . . . . . . . . 3.9-89 3.9.7 Integrated Head Package . . . . . . . . . . . . . . ........ . . . . . 3.9-97 3.9.8 Combined License Infonnation .. ...... .. ........... 3.9-102 3.9.9 References . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . . . 3.9-103 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... 3.10-1 3.10.1 Seismic and Dynamic Qualification Criteria .... ....... . . 3.10-2 3.10.2 Methods and Procedures for Qualifying Electrical Equipment, Instrumentation, and Mechanical Components . . . . . . . . .. . . 3.10-3 110.3 Method and Procedures for Qualifying Supports of Electrical Equipment, Instrumentation, and Mechanical Components . . . . . . 3.10-7 3.10.4 Documentation . . . .. . . . ........ ..... .. ..... 3.10-7 3.10.5 Standard Review Plan Evaluation . . . . . . . . . . . ...... . 3.10-7 3.10.6 Combined License Information Items . . . . . . . . ...... . . 3.10-8 3.10.7 References . . . ....... .................... . . . . . 3.10-8 l 3.11 Environmental Qualification of Mechanical and Electrical Equipment . . . . . 3.11-1
- APPENDIX 3A CONTAINMENT STRUCTURAL MODULES (*) . . .... ... . 3A-1 3 A.1 Codes and Standards (*) .......... .. .... ..... .... .... . 3A-1 3A.2 Loads and Load Combinations (*) . . . . . ...... . . ... ... . . . 3A-1 3A.3 Design and Analysis Procedures (*) .... ............ . ... . . 3 A- 1 3A.3.1 Wall Modules (*) . . .. .... . .. . .... ...... .. 3A-1 3A.3.2 Floor Modules (*) . . . . ... .. .......... .. . . . . 3 A-1 1
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u--u I t N V MASTER TABLE OF CONTENTS (Cont.) l Section Title .P_ age 3A.4 Effective Width (*) ..... ... . . . . . .. ........ ... . . . . . . . 3A-1 3A.5 Design Details (*) . . . . . . . . . . . .... ...... ......... .. ..... 3A-1 3A.6 Site Fabrication (*) . . . . . . . . . . . . . . . . ..... ............... . . . 3A-1 3A.7 Materials (*) . . . . . . . . . ........... .......... .... . . . . . . . . . 3 A- 1 3A.8 References (*) . . . . . . . . .... ..... .. .............. . ... . 3A-1 APPENDIX 3B LEAK-BEFORE-BREAK EVALUATION OF THE REACTOR COOLANT LOOP PIPING OF THE AP600 . . . . . .. . ..... . . . 3B-1 l 3B.1 Leak-Before-Break Criteria for AP600 Primary Loop Piping . 3B-2 3B.2 l Potential Failure Mechanisms for AP600 Primary Loop Piping . . ... ... . 3B-2 3B.2.1 Erosion-Corrosion Induced Wall Rinning . . . ... . . . . . 3B-2 3B.2.2 Stress Corrosion Cracking ........ ... .......... . . . . 3B-2 3B.2.3 Water Hammer .... .. ...... ..... ...... .... . . 3B-3 3B.2.4 Fatigue . . . . . . . . . . .................................3B3 3B.2.5 hermel Aging .... ................ ..... ... . . . . . 3B-4 3B.2.6 Bermal Stratitication . . . . . . . . . . . ....... ............3B4 3B.3 Material and Fracture Toughness Properties . . . . ..... . .... . .. . . 3B-4 3B.3.1 Tensile Properties ............. ......... ............ 3B-4 (' ) 3B.3.2 Fracture Toughness Properties . . . . . . . . . . . . . . . ........ . . 3B-5 3B.4 Pipe Geometry, Loads and Critical Locations .............. .... . . 3B-5 3B.4.1 Calculation of Loads and Stresses . . . . . . . . ...... .... . . . . 3B-6 3B.4.2 Combination of Loads for Leak Rate . . . .... . . . . . . . . . . . . . 3 B -6 3B.4.3 Faulted Loads for Stability Evaluations . . . . . . . . . . ...... . . 3D-6 3B.4.4 Critical Location . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 B -7 3B.5 Leak Rate Evaluations . . . . . . . . . . . ..... .. . . . . . . . . . . . . . . . . . . . 3 B -7 3B.5.1 Phenomenological Considerations (*) ................. . . . . 3 B -7 3B.5.2 Calculational Method (*) ........ ......................3B-7 3B.S.3 Leakage Size Flaws . . . . . . . . . . . . . . ..... ... ... . . . . 3B-7 3B.6 Stability Evaluations for the Primary Loop Piping . . . . . . . . . . . . . . ... . 3B-7 3B.6.1 Stability Methodology (*) . ........ .... ... .. .. . . 3B-8 3B.6.2 Stability Evaluations . . . . . . .. ........ ...... ... . . . 3B-8 3B.7 Fatigue Crack Growth Methodology . . . . . . . . . . . ...... .. .. . . . 3B-8 3B.8 Discussion and Conclusions . . . . . . . . . . . . . . . . . . ...... . . . . . . . . . . 3 B -8 38.9 References . . . . . . . . ......................................3B-10 APPENDIX 3C REACTOR COOLANT LOOP ANALYSIS METHODS . . . . . . . ..... . 3C-1
' 3C.1 Reactor Coolant Loop Model Description . . . . ...... ........... . . 3C-1 3C.1.1 Steam Generator Model . . . .......... .. ..... . . . . . . . 3C-1 3C.1.2 Reactor Coolant Pump Model ... ........ .......... . . 3C-2 3C.I.3 Reactor Pressure Vessel Model .. .... .......... ... . 3C-2 m
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1 l l meu O' MASTER TABLE OF CONTENTS (Cont.) Section Title Page 3C.1.4 Containment Interior Duilding Structure Model . . . . . . . . . . . . . . . . 3C-3 3C.1.5 Reactor Coolant Loop Piping Model . . . . . . . . . . . . . . . . . . . . . . 3 C-3 3C.2 Design Requirements . . . . . . . . . . . . ............................3C-3 3C.3 Static Analyses . . . . . . . . . ...................................3C-4 3C.3.1 Deadweight Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-4 3C.3.2 Internal Pressure Analysis . . . . . . . . . . . . ........ ...... . . 3C-4 3.3.3 'Ihermal Expansion Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-4 3C.4 Seismic Analyses . . . . . . . . . . . . . . . . . . . . . ............... . . . . 3C-4 3C.5 Reactor Coolant Loop Piping Stresses . . . . . . .... . . ... . . . . . . . . . 3C-5 3C.6 Description of Computer Programs . . . . . . ..... . . . . . . . . . . . . . . 3C-5 APPENDIX 3D METHODOLOGY FOR QUALIFYING AP600 SAFE'IT-RELATED ElFCTRICAL AND MECHANICAL EQUIPMENT . . . . . . . . . . . . . . . . . 3D-1 3D.1 Purpose ........... ........... ..... . . . . . . . . . . . . . . . . . . . . 3 D-2 3D.2 Scope............................. . . . . . . . . . . . . . . . . . . . . . 3 D-2 3D.3 Introduction . . . . . . . . . . . . . . . . . . ............................3D-3 3D.4 Qualification Criteria . . . . . ...... ............................3D-3 3D.4.1 Qualification Guides . . . . . . . . . . . . . . . . .................3D-3 3D.4.2 Defi nitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-7 3D.4.3 Mild versus Harsh Environments . . . . . ........... . . . . . . . . 3 D-7 3D.4.4 Test Sequence . . . . . . . . . . . . . . . . . .....................3D-8 3D.4.5 Aging . . . . . . . ........................... . . . . . . . . . 3 D-9 3D.4.6 Operability Time . . . . . ................ . . . . . . . . . . . . . 3 D- 12 3D.4.7 Performance Criterion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-12 3D.4.8 Margin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......... . . 3D-12 3D.4.9 Treatment of Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-15 3D.4.10 Traceability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 D- 16 3D.5 Design Specifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 D- 17 3D.5.1 Normal Operating Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-18 3D.5.2 Abnormal Operating Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-19 3D.5.3 Seismic Events . . . . . . . . . . . . . . . . . . . . . . . . . . ........ . . 3D-20 3D.5.4 Containment Test Environment ......... . . . . . . . . . . . . . . . . 3 D -20 3D.5.5 Design Basis Event Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-20 3D.6 Qualification Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-24 3D.6.1 Type Test ................ ................ ...... . 3D-25 I 3D.6.2 Analysis . . . . . . . . . . . . . . . . . . . . . ..... . . . . . . . . . . . . . . 3 D-25 I 3D.6.3 Operating Experience . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 D -30 l 3D.6.4 On-Going Qualification . . . . . .... ....... l'
. . . . . . . . . . . . 3 D-30 3D.6.5 Combinations of Methods . .................... ... . . . 3D-30 1 <
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V MASTER TABLE OF CONTENTS (Cont.) Section _ Title Pm 3D.7 Documentation . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . . . . . . . . 3 D-31 3D.7.1 Equipment Qualification Data Package . . ...... .... . . . 3D-32 0 3D.7.2 Specifications . . . . . . . . . . ..... .... . . .... ... . 3D-32 3D.7.3 Qualification Program ... ... .. . .. .... ... . . . 3D-34 3D.7.4 Qualification by Test . . . . . . . . . .. .. .. . . ...... . . . 3D-34 3D.7.5 Qualification by Analysis .... ...... ...... ...... . . 3D-37 3D.7.6 Qualification by Experience .................. .. . . 3D-37 3D.7.7 Qualification Program Conclusions . . . . .. .. . . . . 3D-37 3D.7.8 Combined License Information . ....... ... . ...... . . 3D-38 3D.8 References . . . . . . . . . . . . . .... ... .. ... . .. .. . . 3D-38 Appendix 3D-Attachment A - Sample Equipment Qualification Data Package ... .. . . 3D-69 Appendix 3D-Attachment B - Aging Evaluation Program . . . . .. ... ... ..... .... 3D-87 B.1 Introduction . . . . . .. ....... .... .. ... .... ... . . 3D-87 B.2 Objectives . . . . ....... . . . .. ... ... . ... ...... . 3D-87 B.3 Basic Approach . . ............. . . .... ...... .......... 3D-87 B.4 Subprogram A . . . .
... . ................... .. ..... . 3D-88 '
;'} B.4.1 Scope ... .. ..... ...... ... ... .. . .. . . . . 3D-88 B.4.2 Aging Mechanisms . . . . . . . . . . . . . . ... .... . .... . . . 3D-88 B.4.3 Ti me . . . . . . . . . . . . ........... ............. . . . . . 3D-88 B.4.4 Operational Stresses . . . . . ..... ........ ........ . . . . 3D-89 B.4.5 External Stresses .. ... ..... .. ......... .. . . 3D-89 B.4.6 Synergism . . . . . . . . . . . . . ..... .... . ........... 3D-91 B.4.7 Design Basis Event Testing . ........ ... ...... .. ... 3D-91 B.4.8 Aging Sequence . . .. . . .... ..... ........ . . . 3D-91 B.4.9 Performance Criterion . . ..... .. ....... . ..... . . . 3D-91 B.4.10 Failure Treatment . ........... .... . .. ....... .. . 3D-91 B.5 Subprogram B .. ... . .. ...... ...... ... ... ..... 3D-92 B.5.1 Scope ............ . .. .. ...... ... . . . . . . . . . 3 D-93 B.S.2 Performance Criteria . . . . . . . . . . ...... .... ...... . 3D-93 B.5.3 Failure Treatment . . . . ........ .... .... ............ 3D-93 Appendix 3D-Attachment C - Effects of Gamma Radiation Doses Below 10' Rads on the Mechanical Properties of Materials . . ...... .. . 3D-96 C.1 Introduction . . . . . . . . . . . . . . . ... ... . . .... ...... . . . . 3D-96 C.2 Scope. ...
............ ........ ......... . . .... 3D-97 C.3 Discussion . . . . . . . . . . . . . .... ..... .... ... . . .. . 3D-97 C.4 Conclusions . . . ......... ........ ... . .. .... . 3D-98 C.5 References . . . . ..... ...... ......... . . . . . . 3D-98 p
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O MASTER TABLE OF CONTENTS (Cont.) Section Title Page Appendix 3D-Attachment D - Accelerr.ted Thermal Aging Parameters . . . . . . . . . . . . . . . 3D-103 .. D.1 Introduction . . . . . . . . . . . ................... ... . . . . . . . . 3 D- 103 l D.2 Arrhenius Mode 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 3D-103 i D.3 Activation Energy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-105 D.4 'Ihermal Aging (Normal / Abnormal Operating Conditions) . . ......... . 3D-106 D.4.1 Normal Operation Temperature . . . . . . . . . . . . . . . . . . . . . . . . 3D-106 D.4.1.1 External Ambient Temperature . . . . . . . . . . . . . . . . . 3 D- 106 D.4.1 ' Temperature Rise in Enclosure .. ........ .. ... 3D-107 D.4 Self-Heating Effects . . . . . . . . . . . . . . . . . . . . . . . . . 3D-107 D.4.2 Accca rated Aging Temperature . . . . . . . . . . . . . . . . . . . . . . . . 3 D- 107 D.4.3 Examples of Arrhenius Calculations . . . . . . . . . . . . . . . . . . . . . 3 D- 107 D.4.3.1 For a Normally Energized Component Aged Energized . 3D-107 D.4.3.2 For a Normally De-energized Component Aged Energized 3D-108 D.5 Post-Accident Thermal Aging . . . . . . . . . . . . . . . ......... ....... 3D-108 D.5.1 Post-Accident Operating Temperatures . . . . . . . . . . . . . . . . . . 3D-108 D.5.2 Accelerated Thermal Aging Parameters for Post-Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 D- 109 D.6 References . . . . . . . . . . . . .............................. . . 3D-109 Appendix 3D-Attachment E - Seismic Qualification Techniques . . . . . . . . . .......... 3D-115 E.1 Purpose . . . . . . . . . . . . . . . . . . . . . ......... ................ 3D-115 E.2 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 D- 1 15 E.2.1 1/2 Safe Shutdown Earthquake . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-115 E.2.2 Seismic Category I Equipment . . . . . . . . . . . . . . . . . . . . . . . . . 3 D- 1 15 E.2.3 Active Equipment . . . . . . . . . . . . . . . . . . . . . . . .... .. . . 3D-115 E.2.4 Passive Equipment . . . . . . . . . . . . . . . ...... . . . . . . . . . . . 3 D-115 E.3 Qualification Methods ............................... . . . . . 3D-116 E.3.1 Use of Qualification by Testing . ............ ......... 3D-116 E.3.2 Use of Qualification by Analysis . . . . . . . . . . . . . . . . . . . . . . 3 D- 1 16 E.4 Requirements . . . . . . . . . . . . . . . . . . ....... .............. . 3D-117 E.4.1 Dampi ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... .. . . 3D-117 E.4.1.) Testing . . . ... ........................ . . 3D-117 E.4.1.2 Analysis . . . . . . . . . ...... ...... . . . . . . . . . 3D- 117 E.4.2 Interface Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 D- 117 E.4.3 Mounting Simulation . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 3D- 118 1 E.4.4 1/2 Safe Shutdown Earthquake . . . . . .......... . .... . 3D-118 l E.4.5 Safe Shutdown Earthquake ...................... .. . . 3D-118 ) E.4.6 Other Dynamic Loads . . . . . . . . . . . . . . . . . . . . . .. .. . . 3D-118 O Revision: 5 February 29,19% x 3 Westifigh00Se
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MASTER TABLE OF CONTENTS (Cont.) 6ection Title h E.5 Qualification by Test . . . . . . . . . . . . . . . . . . . . . . . . . .. ... . . . . . . 3D-118 E.5.1 Qualification of Hard-Mounted Equipment .. ..... . . . . . . . 3D-119 E.5.2 Qualification of Line-Mounted Equipment . . . . . . . . . . . . ... . 3D-120 E.5.3 Operational Conditions . . . ...... .. .... . .. .. 3D-122 E.5.4 Resonant Search Testing . ...................... . . . . . 3D-123 E.6 Qualification by Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . 3D-123 E.6.1 Mode'i ng . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . . . . . . 3 D- 124 E.6.2 Qualification by Static Analysis . . .. ... . .. . . 3D-124 E.6.3 Qualification by Dynamic Analysis . . . . . . . ........ .. . . . 3D-124 E.7 Performance Criteria . . . . . . . . . . . . . . . . . . . .... ....... . . . . 3D-126 E.7.1 Equipment Qualification by Test ................... . . . . 3D-126 E.7.2 Equipment Qualification by Analysis ...... . . . . ..... 3D-126 - APPENDIX 3E HIGH-ENERGY PIPING IN THE NUCLEAR ISLAND . ..... .. . . 3E-1 APPENDIX 3F HYDRODYNAMIC ANALYSES OF THE IN-CONTAINMENT REFUELING WATER STORAGE TANK GRWST) (*) ....... . . . . . . 3F- 1 m ( \ APPEND 1X 3G
! CABLE TRAYS AND CABLE TRAY SUPPORTS . ...............3G-1 V 3G.1 Codes and Standards . . . . . . . . . .................... . . . . . . . . . . 3 G- 1 3G.2 Loads and Load Combinations . . . . . .... ...................... 3G-1 3G.2.1 Loads . . . . . ............................. . . . . . . . . . 3 G- 1 3G.2.2 Load Combinations . . . . . . . . . . . . . . . . . ... .. ........... 3G-2 3G.3 Analysis and Design . . . . . . ..................... ...... . . . . 3G-2 3G.3.1 Damping . . . . . . . . . . . . . . . . . ............ . . . . . . . . 3 G-2 3G.3.2 Seismic Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3G-3 3G.3.3 Allowable Stresses . . . ........ .. .... ..... ....... 3G-3 3G.3.4 Connections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 G-3 APPENDIX 3H HVAC DUCTS AND DUCT SUPPORTS . ... ..... .... ... . 3H-1 3H.1 Codes and Standards . . . . . . . . . ....... ...... .. ..... .. . . 3H-1 3H.2 Loads and Load Combinations . . . . . . . . . . . . . . . . . . . ... .. .... . . 3H-1 3H.2.1 Loads . . . . . . . . . . . . . . . . . .... ....... .. ... . . . . . . 3H-1 3H.2.2 Load Combinations . . . ...... .......... .... . . . . . . . . . 3 H-3 3H.3 Analysis and Design . . . . . . . . . . . . .... .......... .... ..... 3H-3 3H.3.1 Response Due to Seismic Loads . . . . . . ........... ... .... 3H-3 3H.3.2 Deflection Criteria . . . . .......... .............. . . . . 3H-3 3H.3.3 Relative Movement . . ........ . ........ . . . . . . . 3H-3 3H.3.4 Allowable Stresses . . . . . . ........................ . . . . 3 H-4 3H.3.5 Connections . . . . . . ............ .... ................ 3H-4 m
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O MASTER TABLE OF CONTENTS (Cont.) S.ection Title _ Page, CHAPTER 4 REACTO R . . . . . . . . . . . . . . . . . . . . . . . . . . ....... ............ 4.1-1 4.1 Sununary Description . . . . . . . . . . . . . . . . . . .............. . . . . 4.1-1 4.1.1 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1-4 4.1.2 References . . .. ............. ........ .............. 4.1 4 4.2 Fuel System Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2- 1 4.2.1 Design Basis . . . . . . ............................ . .. 4.2-2 4.2.2 Description and Design Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-9 4.2.3 Design Evaluation . . . ................................4.2-20 4.2.4 Testing and Inspection Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-32 4.2.5 Combined License Information . . . . .. . . . . . . . . . . . . . . . . . . . 4.2-3 7 4.2.6 References . . . . . . . . . . . . . . . . . . . . . . . . ........... . . 4.2-37 4.3 Nuclear Design . . . . . . . . . . . . . . . . . . . . . . . ..................... 4.3-1 4.3.1 Design Basis . . . . . . ........... ................ .. 4.3-1 4.3.2 Description .................. .. ....... ... ......... 4.3-7 4.3.3 Analytical Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.3-39 4.3.4 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . 4.3-42 4.3.5 References . . . . . ............... .. .......... .. .. 4.3 42 4.4 'Ihermal and Hydraulic Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.4-1 4.4.1 Design Basis . . . . . . . . . . . . . . . . .................. .. .. 4.4-1 4.4.2 Description of Thermal and Hydraulic Design of the Reactor Core . . . 4.4-4 4.4.3 Description of the 'Ihermal and Hydraulic Design of the Reactor Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.4-20 4.4.4 Evaluation . . . ............ ........... ...... .. . . 4.4-21 4.4.5 Testing and Verification . . . . . . . . . ................... . 4.4-30 4.4.6 Instrumentation Requirements ...... ... . . . . . . . . . . . . . . . 4.4-31 4.4.7 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . 4.4-35 4.4.8 References . . . . . ............ ..................... 4.4-35 4.5 Reactor Materials . . . . . . . . . . . . . . . . . . . . . . . . . . ......... .... 4.5-1 4.5.1 Control Rod and Drive System Structural Materials . . . . . . . . . . . . 4.5-1 4.5.2 Reactor Internal and Core Support Materials . .. . . . . . . . . .. . ... 4.5-3 4.5.3 Combined License Information ............... ... ....... 4.5-4 4.5.4 References . ..................... ........ . ... .. .. 4.5-4 4.6 Functional Design of Reactivity Control Systems . . . .... ............ 4.6-1 4.6.1 Information for Control Rod Drive System . . . . . . . . . . . ..... 4.6-1 4.6.2 Evaluations of the Control Rod Drive System ........ ...... 4.6-1 4.6.3 Testing and Verification of the Control Rod Drive System . . . . . . . . 4.6-2 4.6.4 Informa' ion for Combined Performance of Reactivity Systems . .. . 4.6-2 4.6.5 Evaluation of Combined Performance . . . . . . . . . . . . . . . . . . . . . 4.6-3 4.6.6 Combined License Information . . . . . . . . . . .... . ... . . 4.6-4 O Revision: 5 February 29,1996 xil 3 Westirighouse
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v l MASTER TABLE OF CONTENTS (Cont.) Section Title aage _P_ ; VOLUME 4 CHAI'IER 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS . . .. . 5.1-1 5.1 Summary Description . . . . . . . . . . . . ..... .... ..... . . . 5.1-1 5.1.1 Design Bases . . . . . . . . . . . . . . . . . ... . .. ... ... 5.1-1 5.1.2 Design Description . . . .. ........ ....... ........ 5.1-2 5.1.3 System Components . . . . . . . . . . . . . .... .. ....... ... 5.1-4 ; 5.1.4 System Performance Characteristics . . . ..... . .. . .. . 5.1-7 5.1.5 Combined License Information . . . . . . . . . . . ...... . . .... 5.1-9 5.2 Integrity of Reactor Coolant Pressure Boundary . . . . .... .. .. .. 5.2-1 5.2.1 Compliance with Codes and Code Cases . . . ....... . ...... 5.2-1 5.2.2 Overpressure Protection . . ........... .. .... .. . . 5.2-2 5.2.3 Reactor Coolant Pressure Boundary Materials . . . . ....... .... 5.2-7 5.2.4 Inservice Inspection auf Testing of Class 1 Components . . . . . . . . 5.2-17 5.2.5 Detection of Leakage 'Ihrough Reactor Coolant Pressure Boundary . 5.2-20 5.2.6 Combined License Information Items ........ . . . . . 5.2-29 5.2.7 References . . ....................... ... . . . . . . . . . . 5.2-29 5.3 Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . . ...... ........... . 5.3-1 l,,) 5.3.1 Reactor Vessel Design s / ............... ...... ... ..... 5.3-1 v 5.3.2 Reactor Vessel Materials . . . . ........................... 5.3-4 5.3.3 Pressure-Temperature Limits . . . . . . ................... . 5.3-13 5.3.4 Reactor Vessel Integrity . . . . . . . . ...... ....... . . . . 5.3-14 5.3.5 Combined License Information . . . . . . . . . . . . . ... . . 5.3-19 5.3.6 References . . . . . . . . . . . . . . . ............. ..... . . . 5.3-20 5.4 Component and Subsystem Design . . . . . . . . . ............ . . .. 5.4-1 5.4.1 Reactor Coolant Pump Assembly . . . . . . .................. . 5.4-1 5.4.2 Steam Generators . . . . . . . . . . . . . . . ....... ...... . . 5.4-10 5.4.3 Reactor Coolant System Piping . . . . . . . . ............ . . . . 5.4-23 5.4.4 Main Steam Line Flow Restriction . . . . . . ........... .. . . 5.4-28 5.4.5 Pressurizer . . . . . . . . . . . . . . . . . . . . . . . . ........... . . . 5.4-29 5.4.6 Automatic Depressurization System Valves . . ....... . . . 5.4-36 5.4.7 Normal Residual Heat Removal System . . . . . . . . . . . . . . . . . 5.4-38 5.4.8 Valves . . . . . . . . . ........ .......... . ...... .. . 5.4-51 5.4.9 Reactor Coolant System Pressure Relief Devices . . . . . . . .. . 5.4-57 5.4.10 Component Supports . . . . . . . . . . . .... ....... ... . 5.4-60 5.4.11 Pressurizer Relief Discharge .................. ... . . . . . 5.4-63 5.4.12 Reactor Coolant System High Point Vents . . . . . . . . . . . . . . . . 5.4-65 5.4.13 Core Makeup Tank . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . 5.4-68 5.4.14 Passive Residual Heat Removal Heat Exchanger . . . . . . 5.4-71 5.4.15 Combined License Information . . . . . . . . . . . . . . .. ... .. 5.4-74 5.4.16 References ............................... ... . . . 5.4-74 g) i QJ Revision: 5 f 3 Westingh00Se xii February 29,1996
O MASTER TABLE OF CONTENTS (Cont.) Section Title .P_a,.ga CHAPTER 6 ENGINEERED SAFETY FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1-1 6.0 Engineered Safety Features . . . . . . . . . . . . . . . . . . . . . . . ,,...... .... .... .. . . 6.0-1 6.1 Engineered Safety Features Materials . . . . . . . . . . . . . . . . . . . . . . . . . .. . 6.1-1 6.1.1 Metallic Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1 -1 6.1.2 Organic Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1 -5 6.1.3 Combined License Information Items . . . . ............. . . . 6.1-10 6.1.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1 - 10 6.2 Containment Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.2-1 6.2.1 Containment Functional Design . . . . . . . . . . . . . . . . . . . . . . . . . . 6.2- 1 6.2.2 Passive Containment Cooling System . . . . . . . . . . . . . . . . . . . . . . 6.2-20 6.2.3 Containment Isolation System . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.2-28 6.2.4 Containment Hydrogen Control System . . . . . . . . . . . . . . . . . . . . . 6.2-37 6.2.5 Containment Leak Rate Test System . . . . . . . . . . . . . . . . . . . . . . 6.2-48 6.2.6 References . . . . . . . . . . . . . ........................... 6.2-54 6.3 Passive Core Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3 - 1 6.3.1 Design B asis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3-2 6.3.2 System Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3 -6 6.3.3 Performance Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3-2 8 6.3.4 Post-72 Hour Actions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3-41 6.3.5 Limits on System Parameters . . . . . . . . . .. . . . . . . . . . . . . . . . 6.3-41 6.3.6 Inspection and Testing Requirements . . . . . . . . . . . . . . . . . . . . . 6.3 -42 6.3.7 Instrumentation Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3-43 6.3.8 Combined Liceme Information . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3-48 6.3.9 References . . ....................................6.3-48 6.4 Habitability Systems . . ............................. . . . . . . 6.4 - 1 6.4.1 Safety Design Bals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 6.4-1 6.4.2 System Description . ...,.. . . . .. . . .. ... .. .... . . .. . . . .. 6.4-2 6.4.3 System Operation ....... ... . .. . . . .... .. . . ... . . .. . . . . . 6.4-6 6.4.4 System Safety Evaluation . . . . . . . . . . . . . . .. . . .. .. .. . . .. . . 6.4-8 6.4.5 Inservice Inspection / Inservice Testing . . . . . . . . . . . . . . . . . . . . . . 6.4-10 6.4.6 Instrumentation Requirements ................. . . . . . . . . . 6.4- 10 6.4.7 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . 6.4- 1 1 6.4.8 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.4-1 1 6.5 Fission Product Removal and Control Systems . . . . . . . . . . . . . . . . . . . . . . 6.5-1 6.5.1 Engineered Safety Feature (ESF) Filter Systems . . . . . . . . . . . .. . 6.5-1 6.5.2 Containment Spray System . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.5-1 6.5.3 Fission Product Control Systems . . . . . . . . . . . . . . . . . . . . . . . 6.5-1 6.6 Inservice Inspection of Class 2 and 3 Components . . . . . . . . . . . . . . . .. 6.6-1 6.6.1 Components Subject to Examination . . . . . . . . . . . . . . . . . ... .. 6.6-1 6.6.2 Accessibility . . . . . . . . . . . . . . . . . . . . . . . ............ .. 6.6-1 O Revision: 5 February 29,1996 xiv 3 WeStingh00Se 1
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i / v MASTER TABLE OF CONTENTS (Cont.) Section Title P_y,e 6.6.3 Examination Techniques and Procedures . . . . ... .... ... . 6.6-2 6.6.4 Inspection Intervals . . . . . . . . . . . . .. .... . ... .. . 6.6-3 6.6.5 Examination Categories and Requirements . .
. .. . . . ... 6.6-3 6.6.6 Evaluation of Examination Results ....... ... .. .. . 6.6-3 6.6.7 System Pressure Tests ... . ........ . .... . .... 6.6-3 6.6.8 Augmented Inservice Inspection to Protect against Postulated Piping Failures . . . . . . . . . ........ . ... .. ....... . 6.6-3 6.6.9 Combined License Information Items . ...... . . .. . . 6.6-4 CHAPER7 INSTRUMENTATION AND CONTROLS ..... . . . .... .... . 7.1-1 7.1 Introduction . . ... .... ... . .......... .. . ..... 7.1-1 7.1.1 The AP600 Instrumentation and Control Architecture . . .. .. . 7.1-2 7.1.2 General Protection Subsystem Configuration .... .. . . 7.1-6 7.1.3 Plant Control System . . . . . . . . . .. ..... . .... . .. 7.1-27 7.1.4 Identification of Safety Criteria . . ... ..... .. .. . 7.1-32 7.1.5 AP600 Protective Functions . . . . . . ... .. .. .. . . 7.1-48 7.1.6 Combined License Information . . . . . ... ... .. . .. ... 7.148
, 7.1.7 References . . . . . . .... ................... .
Iv ,) 7.2 Reactor Trip . . . . . . . . . . . ..... ... ... .. .... ... . .. 7.1-49 7.2-1 7.2.1 Description ...... ......... .... .... ... . . .. 7.2-1 7.2.2 Analyses ............... .... . ........ ... 7.2-14 7.2.3 Combined License Information . ... . . .... . . . .. .. 7.2-17 7.2.4 References ..... ............ . . ...... ... 7.2-17 7.3 Engineered Safety Features . . . . . . .. .. ...... ... . .. .. . 7.3-1 7.3.1 Description . . . . . . . . .......... .... .. .. .. . . .. 7.3-1 7.3.2 Analysis for Engineered Safety Features Actuation . . . . . . . . . . . 7.3-19 7.3.3 Combined License Information . . . . . . .. .. . .. .. . . . 7.3-21 7.4 Systems Required for Safe Shutdown . . . . . . . . ........... ...... 7.4-1 7.4.1 Safe Shutdown . . . . . . . . . . . . .. .......... ........ . 7.4-3 7.4.2 Safe Shutdown Systems . .
.. .... . ....... . .. . . 7.4-10 7.4.3 Safe Shutdown from Outside the Main Control Room . . . . . . . . 7.4-11 7.4.4 Combined License Information . .. .. ........ ... . . . 7.4-14 7.4.5 References . ........ .............. .... ... . . . 7.4-14 7.5 Safety-Related Display Information . . . . . . .. .. .. . ...... 7.5-1 7.5.1 Introduction . . . . . . . . .......... ... ........ . . . . . . 7.5- 1 7.5.2 Variable Classifications and Requirements . . . ..... .... .... 7.5-1 7.5.3 Description of Variables . . . . . . . . . ........ ... .. ... 7.5-7 7.6 Interlock Systems important to Safety . . . ........... ...... . ... 7.6 1 7.6.1 Prevention of Overpressurization of Low-Pressure Systems . . ... 7.6 1 7.6.2 Availability of Engineered Safety Features ..... .. . .. ... 7.6-2 7.6.3 Combined License Information . . . .. .... . .. ...... .. . 7.6-5 t
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2 l i O' MASTER TABLE OF CONTENTS (Cont.) Section Title Page 7.7 Control and Instrumentation Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.7-1 7.7.1 De scription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 7.7-2 7.7.2 Analysi s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.7-20 7.7.3 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.7-21 VOLUME 5 CHAPIER 8 ELECTRIC POWER . . . . .......... ............. .. . . ... .. 8.1-1 8.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.1 - 1 8.1.1 Utility Grid Description . . . . . . . .... . . . . . . . . . . . . . . . . . . . . 8.1 - 1 8.1.2 Onsite Power Sy stem Description . . . . . . . . . . . . . . . . . . . . .. . . . . . 8.1-1 8.1.3 Safety-Related Loads ............... ........... ... .. 8.1-2 8.1.4 Design B asis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.1 -2 8.1.5 Combined License Information . . . . . . . . . . . . . . . . . . . .... . . 8.1-5 8.2 Offsite Power System . . . . . . .. .. . .. .. .. . .. . ... . .. . . ... ... . .. . 8.2-1 8.2.1 System Description . . . . . . . . . .. .. ...... .. .. .......... . 8.2-1 8.2.2 Conformance to Criteria . . . . . . . . . . . . . . . . . .... .. ... . . .. 8.2-2 8.2.3 Standards and Guides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.2-2 8.2.4 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.2-2 8.2.5 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.2-3 8.3 Onsite Power Systems . . . . . . . . . . . . . . . . . . . . . . ......... . . . . 8.3-1 8.3.1 AC Power Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.3-1 8.3.2 DC Power Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.3-11 8.3.3 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . 8.3-20 8.3.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.3-20 CHAPIER 9 AUXILIARY SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.1 -1 9.1 Fuel Storage and Handling . . . . . . . .............. ... .. .... ..... 9.1-1 9.1.1 New Fue1 Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. 9.1-1 9.1.2 Spent Fue1 Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ... . 9.1-5 9.1.3 Spent Fuel Pool Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . 9.1 - 10 9.1.4 Light Load Handling System (Related to Refueling) . . . . . . . . . . 9.1-20 9.1.5 Overhead Heavy Load Handling Systems . . . . . . . . . . . . . . . . . . . 9.1-34 9.1.6 Combined License Information for Fuel Storage Handling . . . . . . . 9.1-40 9.1.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ..... 9.140 9.2 Water Systems . . . . . . . . . . . . . . . ....................... .. . 9.2-1 9.2.1 Service Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.2-1 9.2.2 Component Cooling Water System . . . . . . . . .............. . 9.2-8 9.2.3 Demineralized Water Treatment System . . . . . . . ... ..... . . 9.2-17 9.2.4 Demineralized Water Transfer and Storage System . . . . . . ... . . 9.2-22 O Revision: 5 February 29,1996 xvj [ Westingh00Se
H --!!! m _- i \ C/ MASTER TABLE OF CONTENTS (Cont.) Section Title Page 9.2.5 Potable Water System . . . . . . . . . ..... ......... ... . . 9.2-26 9.2.6 Sanitary Drainage System . . . . . . . . . . ..... ... . . . . . . 9.2-30 9.2.7 Central Chilled Water System . . . ........... ... . . . . . . 9.2-31 9.2.8 Turbine Building Closed Cc,oling Water System . . . . . . . . . . . . . . . 9.2-37 9.2.9 Waste Water System ............. ............. . . . . . 9.2-42 9.2.10 Hot Wate; Heating System ...... ............... . . . . 9.2-46 9.2.11 References . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . 9.2 -50 9.3 Process Auxiliaries . . . . . . . . ....... .. . 9.3-1 9.3.1 Compressed and Instrument Air System ..... ......
........ . ... .. 9.3-1 9.3.2 Plant Gas System . . . . . . . . . . . . . .. . .. .. .... ..... . . 9.3-7 9.3.3 Primary Sampling System . . . . . . . . . . . .......... ..... . 9.3-12 9.3.4 Secondary Sampling System . . ,.. ....... .. . . . . . . . 9.3-25 9.3.5 Equipment and Floor Drainage Systems . . . . . . ....... . . . . . 9.3-27 9.3.6 Chemical and Volume Control System . . . . ....... ... . . 9.3-33 9.3.7 Combined License Information . . . . . . . . ..... ..... . . . . . . 9.3-55 9.3.8 References . . . . . . . . . . . . . . . . . . . . . ........... . . 9.3-56 '73 \ VOLUME 6 \'" /
9.4 Air-Conditioning, Heating, Cooling, and Ventilation System . . . . . . . . . . . . . 9.4-1 9.4.1 Nuclear Island Nonradioactive Ventilation System . . . . . . . . . . . . . . 9.4-1 9.4.2 Annex / Auxiliary Buildings Non-Radioactive HVAC System . . . . . . 9.4-10 9.4.3 Radiologically Controlled Area Ventilation System ...... . . . . 9.4-14 9.4.4 Balance-of. Plant Interface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9.4-18 9.4.5 Engineered Safety Features Ventilation System . . . . . . . . . . . . . . 9.4-18 9.4.6 Containment Recirculation Cooling System . . . . . . . . . . . . . . . . . 9.4-18 9.4.7 Containment Air Filtration System . . . . . . . . . . . . . . . . . . . . . . . 9.4-21 9.4.8 Radwaste Building HVAC System . . . . . . . . . . . . . . . . . . . . . . . 9.4-25 9.4.9 Turbine Building Ventilation System . . . . . . . . . . . . . . . . . . . . . . 9.4-2 8 9.4.10 Diesel Generator Building Heating and Ventilation System . . . . 9.4-31 9.4.11 Health Physics and Hot Machine Shop HVAC System .. . . . . . 9.4-33 9.4.12 References . . . . . . . . . . . . . . . . . . . . . . . . . . ..... ... .. 9.4-36 9.5 Other Auxiliary Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 9.5-1 9.5.1 Fire Protection System . . . . . . . . . ........... . .. .... .. .. 9.5-1 9.5.2 Communication System . . . . . . . . . . . . . . .......... . . . . . . 9.5- 15 9.5.3 Plant Lighting System . . . . . . . . . . . . . . . . . . ....... . . . . . . 9.5-18 9.5.4 Standby Diesel and Auxiliary Boiler Fuel Oil S ...... . . . 9.5-21 9.5.5 References . . . . . . . . . . . . . . . . . . . . . . . .............. . .ystem . 9.5-28 7 N h %d Revision: 5 UN xvii February 29,1996
O MASTER TABLE OF CONTENTS (Cont.) Section Title ,P_ag,,e APPENDIX 9A FIRE PROTECTION ANALYSIS .......................... . . 9A-1 9A.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . .... ..... ... . . . . 9A-1 9A.2 Fire Protection Analysis Methodology . . . . . . . . . . . . . . . . . . . ....... . 9A-1 9A.2.1 Fire Area Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9A- 1 9A.2.2 Combustible Material Survey . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 A- 1 9A.2.3 Fire Severity Categorization . . . . . . . . . ...... .... .... . 9A-1 9A.2.4 Combustible Loading and Equivalent Fire Duration Calculations . . . . . . ............................. . . . 9A-2 9A.2.5 Fire Protection Adequacy ...............................9A-2 9A.2.6 Fire Protection System Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . 9A-3 9A.2.7 Safe Shutdown Evaluation . . . . . . . . ... .................9A-3 9A.3 Fire ProtecJon Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . ....... . 9A-6 9A.3.1 Nuclear lsland . . ................. . ..... ..... . . . 9A-6 9A.3.2 Turbine Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9A-57 9A.3.3 Yard Area And Outlying Buildings . . . . . . . . . . . . . . . . . . . . . . . . 9A-63 9A.3.4 Annex I and II Buildings ............ . . . . . . . . . . . . . . . . . 9A-63 9A.3.5 Radwaste Building . . . . . . . . . . . . . ............... .... 9A-76 9A.3.6 Diesel Generator Building . . . . . . . . . . . . . . . . . . . . . ...... . 9A-85 9A.3.7 Special Topics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9A-87 9A.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 A-90 VOLUME 7 CIIAI'rER 10 STEAM AND POWER CONVERSION SYSTEM . . . . . ..... . . . . . . . 10.1 -1 10.1 Summary Description ....................................... 10.1-1 10.1.1 General Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.1- 1 10.1.2 Protective Features . . . . . . . . . . .............. . . . . . . . . . 10.1 -2 10.1.3 Combined License Information on Erosion-Corrosion Monitoring . . . 10.1-4 10.2 Turbine-Generator . . . . ....... .................. ......... . 10.2-1 10.2.1 Design B asis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10.2- 1 10.2.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 10.2-1 10.2.3 Ttirbine Rotor Integrity . . . . . . . . ............ ..... . 10.2-11 10.2.4 Evaluation . . . . . . . . . .......................... .. 10.2-17 10.2.5 Instnamentation Applications . . . . . . . . . . . . . . . . . . . . . . . . . . 10.2- 17 10.2.6 Combined License Information on Turbine Maintenance
- and Inspection . . . . . . . . . . . . . . . . . . . .......... . . . . . . 10.2-19 l 10.2.7 References . . . . . . . . . . . . . . . . . . . . ...... . . . . . . . . . . . . 10.2- 19 1
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MASTER TABLE OF CONTENTS (Cont.) Section Title ae fage, 10.3 Main Steam Supply System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10. 3- 1 10.3.1 Design Basis . . . . . .... .. ..... .... . . . . . . . . . . . . . . 10. 3- 1 10.3.2 System Description . . . . . . . . . . . . .......... ..... .... 10.3-4 10.3.3 Safety Evaluation . . . . . . . . . . . . . . . . . . . . . . . ...... .. .. 10.3-10 10.3.4 Inspection and Testing Requirements ......... .... ....... 10.3-12 10.3.5 Water Chemistry . . . . .............. .. ........... . 10.3-13 10.3.6 Steam and Feedwater System Materials . . . .... ........ 10.3-18 ' 10.3.7 Combined License Information . . ................... . . 10.3-19 10.3.8 References . . .................... . ........ . . 10.3-19 1 10.4 Other Features of Steam and Power Conversion System .......... . . . . 10.4- 1 10.4.1 Main Condensers . . ............ ...... ... . . . . . . 10.4-1 10.4.2 Main Condenser Evacuation System . . . . . ... . . . . . . . . . . 10.4-4 10.4.3 Gland Seal System . . . . ....... ... .......... . .. ... 10.4-6 10.4.4 Turbine Bypass System . . . . . . . ......... .. .... . . ... 10.4-9 10.4.5 Circulating Water System . . .... .. .. ....... ....... 10.4-12 10.4.6 Condensate Polishing System . ... ..... ..... .. .... . 10.4-18 10.4.7 Condensate and Feedwater Systems .......... ... ..... 10.4-21 10.4.8 Steam Generator Blowdown System . . . . .... .. . ........ 10.4-36 [mi 10.4.9 Startup Feedwater System . . . . . . . .. . .. .... .... 10.4-45 10.4.10 Auxiliary Steam System . . . . . . . . ..... . .. .... . .. . 10.4-54 10.4.11 Turbine Island Chemical Feed . . . . . . ..... .... . . .... 10.4-57 10.4.12 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . 10.4-59 10.4.13 References . . . . . . .. .............. ......... ..... 10.4-60 CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT . . . . . . . . . . . ... . . . . . 11.1-1 11.1 Source Terms . . . . . . . . . ... ........ .. ........... . . . . . 11.1-1 11.1.1 Design Basis Reactor Coolant Activity . . .......... . . . . . 11.1 1 11.1.2 Design Basis Secondary Coolant Activity . ... . . . . . . . . . . . . 1 1.1 -3 11.1.3 Realistic Reactor Coolant and Secondary Coolant Activity . . . . . . 11.1 -4 11.1.4 Core Source Term . . . . . . . . . . . . . . . . . . . . ...... . .. .. 11.1-4 11.1.5 Process Leakage Sources . . . . . . . . ......... .... .. . . 11.1-4 11.1.6 Combined Ucense Information . . . . . ........ ... . . . . . . . . 11.1 -4 11.L7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1 -5 11.2 Liquid Waste Management Systems .. .. .......... ..... . . ... 11.2-1 11.2.1 Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.2- 1 11.2.2 System Description . . . . . . . . . . . . .. .... ........ . . . . . 11.2-4 11.2.3 Radioactive Releases . . . . . . . . . . . . .......... . .... . 11.2-13 11.2.4 References . . . . .......... ..... ... ...... .... . 11.2-15
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O MASTER TABLE OF CONTENTS (Cont.) Section Title Pane 11.3 Gaseous ' Waste Management System .................. .... . . . . 11.3-1 11.3.1 Design B asis . . . . . . . . . ............................. 11.3- 1 11.3.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.3-3 11.3.3 Radioactive Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.3-7 11.3.4 References . . . . . . . . ...................... ...... . . 11.3-9 11.4 Solid Waste Management ...................... ..... . . . . . . . 1 1.4- 1 11.4.1 Design Basis . . . . . . . . . . . . . . ... .... .... . . . . . . . 11.4-1 11.4.2 System Description . . . . . . . . . . . . . . . ....... ....... . . . 11.4-4 11.4.3 System Safety Evaluation . . . . ....... ... ........ .... 11.4-13 11.4.4 Tests and Inspections . . . . . . . . . . . . . . . . ............. .. 11.4-13 11.4.5 Comt sed License Information . . . . . . . . . . . . . . . . . . . . . . . . 11.4- 14 11.4.6 References . . . . . . . . . . . . . . ....... . . . . . . . . . . . . . . . . . 1 1.4- 15 11.5 Radiation Monitoring . . . . . . . . . . . . . . . . . . . . . . . .... . . . . . . . . . . 1 1.5- 1 11.5.1 Design B asis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.5- 1 11.5.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 11.5-2 11.5.3 Effluent Monitoring and Sampling . . . . . . . . . . . . . . ...... .. 11.5-14 11.5.4 Process Monitoring and Sampling . . . . .......... .... .. . 11.5-14 11.5.5 Post-Accident Radiation Monitoring . . . . . . . . . . . ....... .. 11.5-15 11.5.6 Area Radiation Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . 11.5-15 11.5.7 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . 11.5-17 CHAIrIER 12 RADIATION PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 12.1-1 12.1 Assuring that Occupational Radiation Exposures are As-Low-As-Reasonably Achievable (ALARA) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1- 1 12.1.1 Policy Considerations .............. ................. 12.1-1 12.1.2 Design Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1 -2 12.1.3 Combined License Information . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.1-6 12.2 Radiation Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ..... . 12.2-1 12.2.1 Contained Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2-1 12.2.2 Airborne Radioactive Material Sources . . . . . . . . . . . . . . . . . . . . . 12.2-6 12.2.3 Combined License Information . . . . . . . . . . . . ....... . . . . . . 12.2- 8 12.2.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.2- 8 12.3 Radiation Protection Design Features . . . . . . ............. . . . . . . . . 12.3- 1 12.3.1 Facility Design Features . .............................12.3-1 12.3.2 Fhielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12.3-9 12.3.3 Ventilation . . . . . . . . . . . ................... .... .. 12.3-16 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... 12.3-20 12.3.5 Combined License Information . . . . . . . . . . . . . . . . . . . ....... 12.3-20 l 12.3.6 Refe.rences . ........................ .... .... . 12.3-20 O Revisten: 5 February 29,1996 u Westifigh0USB l
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i V: MASTER TABLE OF CONTENTS (Cont.) Section Title Page 12.4 Dose Assessment ................. .............. . . . . . . . . 12.4-1 12.4.1 Occupational Radiation Exposure . . . . . . . . . . . . ... ........ 12.4-1 12.4.2 Radiation Exposure at the Site Boundary . . . . . . ..... .... . 12.4-4 12.4.3 Combined License Information . . . . . . ....... . . . . . . . . . . . . 12.4-4 12.5 Health Physics Facilities Design . . ... .. .............. . . . . . 12.5-1 12.5.1 Objectives . . . . . . . . . . . . ............... .. ..... . 12.5-1 12.5.2 Equipment, Instmmentation, and Facilities . . . . . . . . . . . . . . . . . . 12.5- 1 12.5.3 Other Design Features . . . . . ...... ....... ........ . . 12.5-3 12.5.4 Controlling Access and Stay Time . .............. ...... 12.5-4 12.5.5 Combined License Information . . . . . . . . . .. .... . ... . . 12.5-5 VOLUME 8 CHAI'IER 13 CONDUCT OF OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 13-1 13.1 Organizational Structure of Applicant . . . . . . . . . . . . . . . . . . . . . . . . ..... 13-1 13.1.1 Combined License Information Item . . . . . . . . ..... ........ 13-1 13.2 Training . . . . . . . . . . . . . . .................................... 13-1 n 13.2.1 Combined License Information Item . . . . . . . . . . . . . . . . . . . . . . . 13-1 ( ) 13.3 Emergency Planning . . . . . . . ................. ............ ... 13-1
13.3.1 Combined License Information Item . . . . . . . . ........ . ..... 13-2 13.4 Operationa1 Re view . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... 13-2 13.4.1 Combined License Information Item . . . . . . . . . . . . . . . . . . . . . . . . 13-2 13.5 Plant Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-2 13.5.1 Administrative Procedures . . . . . . . . . . ........ ............ 13-2 13.5.2 Operating and Maintenance Procedures . . . . . . . . . . . . . . . . . . . . . . . 13-2 13.5.3 Combined License Information Item ........................ 13-2 13.6 Industria1 Security . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-2 13.6.1 Preliminary PI'anning . . . . . . . . . . . ........................ 13-2 13.6.2 Security Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . 13-3 13.6.3 Plant Protection System . . . . ................ ..... ..... 13-3 13.6.4 Physical Security Organization . . . . . . . . . . . . . . . . . . . . . . . ..... 13-5 13.6.5 Physical B arriers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-5 13.6.6 Access Requirements .... ..... .............. . ..... 13-6 13.6.7 Detection Aids . . . . . . . . . . . . . . . . . . .... ... .......... 13-7 13.6.8 Security Lighting . . . . . . . . . . . . . . . . . ............... ... 13-7 13.6.9 Security Power Supply System . . . . . . ................. .. 13-7 13.6.10 Communications . . . . . . . . . . . . . . . ... .... ... .. ....... 13-7 13.6.11 Testing and Maintenance ............ ............ ...... 13-8 13.6.12 Response Requirements . . . . . . . . . . . . . . . . . . . . . . ... . ..... 13-8 l 13.6.13 Combined License Information Item . . . . . . . . . . . . . . . . . ...... 13 8 l 13.7 References . . . . . . . . . . . . . . . ................... ........ .. 13-8 n
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MASTER TABLE OF CONTENTS (Cora.\ . I l Section Title Page CHAPIER 14 INITIAL TEST PROGRAM . . . . . . . . . . . . . . . . . . . . . . . ............ 14.2-1 14.1 Specific Information to be Included in Preliminary / Final Safety Analysis Reports . . . . . . . . . . . . . . . .................. .... ..... . 14.2-1 14.2 Specific Information to be Included in Standard Safety Analysis Reports . . . . 14.2-1 14.2.1 Sununary of Test Program and Objectives . . . . . . . . .. . . . 14.2-1 14.2.2 Test Procedures ... ...... . . ... ...... .... . 14.2-4 14.2.3 Compliance of Test Program with Regulatory Guides . . . . . . .... 14.2-5 14.2.4 Utilization of Reactor Operating and Testing Experience in the Development of Test Program . . . . . . . . . . ... . .... . 14.2-6 14.2.5 Use of Plant Operating and Emergency Procedures . . .. . ..... 14.2-6 14.2.6 Initial Fuel Loading and Initial Criticality . . .... ..... .. . 14.2-6 14.2.7 Test Program Schedule . . . . . .... .... . ........ . . . . 14.2-9 14.2.8 Individual Test Descriptions . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-10 14.2.9 Interfaces . . . . . . . . . . ............ .. . ......... 14.2-111 CHAPIER 15 ACCIDENT ANALYSES ............................ ... . 15.0-1 15.0.1 Classification of Plant Conditions . . . . ............. .. . .. . 15.0-1 15.0.1.1 Condition I: Normal Operation and Operational Transients .. .. 15.0-1 15.0.1.2 Condition II: Faults of Moderate Frequency . . . . . . . . . .. . . 15.0-2 15.0.1.3 Condition III: Infrequent Faults . . . . . . . . . . . . ......... . . 15.0-4 15.0.1.4 Condition IV: Limiting Faults . . . . . . . .. . .. ........ . . 15.0-4 15'.0.2 Optimization of Control Systems . . . . . . . . . . . . . . . . . . . . ...... . . . 15.0-5 15.0.3 Plant Characteristics and Initial Conditions Assumed in the Accident Analyses . . . . ....... ....... .... .. . . . . . . . 15.0-5 15.0.3.1 Design Plant Conditions . . . . . . . . . .............. . .. . 15.0-5 15.0.3.2 Initial Conditions . . . . . . . . .................. . . . . . . . 15.0-6 15.0.3.3 Power Distribution . . ..............,......... . . . . 15.0-6 15.0.4 Reactivity Coefficients Assumed in the Accident Analysis . ............ 15.0-7 15.0.5 Rod Cluster Control Assembly Insertion Characteristics . . . . .. .. . 15.0-7 15.0.6 Trip Points and Time Delays to Trip Assumed in Accident Analyses . . . . . . 15.0-8 15.0.7 Instrumentation Drift and Calorimetric Errors, Power Range Neutron Flux . . 15.0-9 15.0.8 Plant Systems and Components Available for Mitigation of Accident Effects . 15.0-9 15.0.9 Fission Product Inventories . . . . . . . . ... ...... . .. .. .... 15.0-10 15.0.10 Residual Decay Heat . . . . . . . . . . . . . ...... ....... .... . . . . 15.0-10 15.0.10.1 Total Residual Heat . . . . . . ...., .... ....... ..... 15.0-10 15.0.10.2 Distribution of Decay Heat Following Loss-of-Coolant Accident . 15.0-10 15.0.11 Computer Codes Utilized . . . . . . . . .... .... ..... .... 15.011 15.0.11.1 FACTRAN Computer Code ...... . .. ....... . .. 15.0-11 15.0.11.2 LOFTRAN Computer Code . . . . . . . . . . . .. .... ..... 15.0-11 15.0.11.3 TWINKLE Computer Code . . . . . . . . ... . . ... 15.0-12 O Revision: 5 February 29,1996 xxii [ W85tingt10USS
i w/ MASTER TABLE OF CONTENTS (Cont.) Section Title Pm 15.0.11.4 THINC Computer Code . . . . . .................... .... 15.0-12 15.0.11.5 WESTAR Computer Code ..... ... ........ ......... 15.0-13 15.0.12 Component Failures . . . . . . . . . . . . .. ..... ..... ..... . . . 15.0-13 15.0.12.1 Active Failures . . . . . . . . . . . . ... ..... ..... .... .. 15.0-13 15.0.12.2 Passive Failures . . . . . . . . . . . . . . . . .. . ....... . . 15.013 15.0.12.3 Limiting Single Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.0- 14 15.0.13 Operator Actions . . ............. . . ... .. ........ ... 15.014 15.0.14 Combinmi License Information . . . . . . . .......... ... .... . . . 15.0-14 15.0.15 References ........ ........ ....... ........ ........ . 15.0-14 15.1 Increase in Heat Removal from the Primary System . . . . . . . . . . . . . . . . 15.1-1 15.1.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature . . . . . . . . . . . . . . . . . .... ...... . 15.1-1 15.1.2 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow . . . . . . . . . . . . . . . . . . . ............... . 15.1-3 15.1.3 Excessive Increase in Secondary Steam Flow . . . . .... ..... . 15.1-5 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve . . 15.1-8 15.1.5 Steam System Piping Failure . . ....... ................. 15.1-11 15.1.6 Inadvertent Operation of the Passive Residual Heat Removal l(,,) l Heat Exchanger . . . . . . . . ... ........... . . . . . . . . . . . 15.1-20 15.1.7 Combined License Information . . . . . . . . . . ................ 15.1-24 15.1.8 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.1-24 15.2 Decrease in Heat Removal by the Secondary System . . . . . . . . . . . . . . . . . . 15.2-1 15.2.1 Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steamflow . . . . ...................... ..... 15,2-1 15.2.2 Loss of External Electrical Load . ......... ......... . . 15.2-1 15.2.3 _ '1brbine Trip . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15.2-3 15.2.4 Inadvertent Closure of Main Steam Isolation Valves . ......... . 15.2-7 15.2.5 Loss of Condenser Vacuum and Other Events Resulting in Turbine Trip . . . . . . . . . . . . . . . . ....... ......... ... 15.2-7 15.2.6 Less of ac Power to the Plant Auxiliaries . . . . ........ . . . 15.2-8 15.2.7 Loss of Normal Feedwater Flow . ............... .... ... 15.2-12 15.2.8 Feedwater System Pipe Bre?1 . ........... ............ 15.2-15 i 15.2.9 Combined License Information . .............. ........, 15.2-20 15.2.10 References . . . . . . . . . . . . . . . . . . . . . . ........ . ....... 15.2-20 15.3 Decrease in Reactor Coolant System Flow Rate . . . . . . . . . . . . . . . . . . . 15.3- 1 15.3.1 Partial Loss of Forced Reactor Coolant Flow . . . . . . . . . . . . . . . . 15. 3- 1 I 15.3.2 Complete Loss of Forced Reactor Coolant Flow . . . . . . . . . . . . . . . 15.3-3 15.3.3 Reactor Coolant Pump Shaft Seizure (Locked Rotor) . . . . . . . . . . 15.3-5 15.3.4 Reactor Coolant Pump Shaft Break . . . . . . . . . . . . . . . . . .... . 15.3-10 g t j
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m-O MASTER TABLE OF CONTENTS (Cont.) Section Title Page 15.3.5 Combined License Information . . .. .. ...... .. ....... 15.3-10 15.3.6 References . . ...... ....... . . .... .. .. ........ 15.3-11 15.4 Reactivity and Power Distribution Anomalies .... ...... ..... .. 15.4-1 15.4.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Suberitical or Low-Power Startup Condition . . . . . . . . . . . . . . 15.4-2 15.4.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power . . . . . . . . . . ........... ..... ..... . . . 15.4-5 15.4.3 Rod Cluster Control Assembly Misalignment (System Malfunction or Operator Error) ..... ...... ......... ... 15.4-11 15.4.4 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature . . . . . . . . . . . . .... .. ........ . .. . 15.4-17 15.4.5 A Malfunction or Failure of the Flow Controller in a Boiling Water Reactor Loop That Results in an Increased Reactor Coolant Flow Rate . . . . . . . . .... .. ........ .... 15.4-19 15.4.6 Chemical and Volume Control System Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant . . . . . . . . . . ............... .. .. 15.4-19 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position . . . ............. .. .... . . .... 15.4-25 15.4.8 Spectrum of Rod Cluster Control Assembly Ejection Accidents . . 15.4-28 15.4.9 . Combined License Information ... ..... .. . . ........ 15.4-39 15.4.10 References . . . . ...... ....... . . ..... . ... 15.4-39 VOLUME 9 15.5 Increase in Reactor Coolant Inventory ... ... ... ........ . . 15.5-1 15.5.1 Inadvertent Operation of the Core Makeup Tanks (CMT) During Power Operation . . . . . . ..... . ... .. ... .... . 15.5-1 15.5.2 Chemical and Volume Control System Malfunction That Increases Reactor Coolant inventory . . ........ .... . . . 15.5-4 15.5.3 Boiling Water Reactor Transients . . . .... . . .. . .. . . 15.5-8 15.5.4 Combined License Information . . . . . . . .. . ............ . 15.5-8 15.5.5 References . . . . . . . . . . . ..... ....... . ....... . . . . 15.5-8 15.6 Decrease in Reactor Coolant Inventory . . . ....... . .. .. . .... 15.6-1 15.6.1 Inadvertent Opening of a Pressurizer Safety Valve or Inadvertent Operation of the ADS ........... . ... . 15.6-1 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment . . . . . . . . . . . . . . . . . . . . .......... . . . . . 15.6-4 15.6.3 Steam Generator Tube Rupture ... ....... ... .. . . . 15.6-5 O Revision: 5 February 29,1996 xxiy 3 Westinghouse
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l t n t LJ MASTER TABLE OF CONTENTS (Cout.) l Section Title ,P_ a_gg 15.6.4 Spectrum of BWR Steam System Pipng Failures Outside of Containment . . . . ... . .... . .... . . 15.6-15 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary . . . . . . . . . . . ........... ...... .. 15.6-15 15.6.6 References . . . . . . . . . . . . . . .......... .. ..... ...... 15.6-64 15.7 Radioactive Release From a Subsystem or Component ... .. . . . . . . . 15.7- 1 15.7.1 Gas Waste Management System Leal; or Failure . . . .. . ... 15.7-1 15.7.2 Liquid Waste Management System Leak or Failure (Atmospheric Release) ... ........... ............... 15.7-1 15.7.3 Release of Radioactivity to the Environment Due to a Liquid Tank Failure ..... .. . .. ........... .... ....... . 15.7-1 15.7.4 Fuel Handling Accident . . ... .. ..... .. ...... ... . 15.7-1 15.7.5 Spent Fuel Cask Drop Accident . . . ..... . ..... . . . . . . 15.7-6 15.7.6 References . . . . . . . . ..... ... ................. .. 15.7-6 ; 15.8 Anticipated Transients without Scram . .... .... . . . . . . . . . . . . 15. 8- 1 15.8.1 General Background .. ...... ........ ... ....... . 15.8-1 15.8.2 Anticipated Transients without SCRAM in the AP600 . . . . . . . . . 15.8- 1 15.8.3 Conclusion . . . . . . . . . . .... ......... ('v') 15.8.4 References . . . . . . . . . . . . . . . . ..... . .... ..... ....
...... . 15.8-1 15.8-1 VOLUME 10 CHAPTER 16 TECHNICAL SPECIFICATIONS ...... .. ... .. .... . . 16.1-1 16.1.1 Introduction to Technical Specifications ..... .. ........ . . . 16.1-1 1.0 Use and Application . . . . ..... ... . . ...... ..... . 16.1-4 1.1 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16.1 -4 1.2 Logical Connectors . . . . ... . .... ... ... . 16.1-11 1.3 Completion Times ......... . .... ... .. . . 16.1-14 1.4 Frequency . . . . . . . . ..... .. .. ..... .... 16.1-26 2.0 Safety Limits . . . . . . . . . . . . . . . . . . .. . .......... . 16.1-30 2.1 Safety Limits . ............... .. ... ... .. 16.1-30 1 2.2 Safety Limit Violations . . . . . . . . . . . ..... . ... ... 16.1-30 l
3.0 Applicability . .......... ...... .... . . . ..... 16.1-41 { 3.0 Limiting Conditions for Operation (LCO) Applicability . .. 16.1-41 1 3.0 Surveillance Requirement (SR) Applicability . . . . . . . . . . . . 16.1-44 3.0 Limiting Conditions for Operation (LCO) Applicability . . . . 16.1-46 3.0 Surveillance Requirement (SR) Applicability . . . . ..... 16.1-55 3.1 Reactivity Control Systems . . . . . . . . . . ...... ... 16.1-60 3.1.1 Shutdown Margin (SDM) - T,,, > 200'F . . ... . 16.1-60 { 3.1.2 Shutdown Margin (SDM) - T,,, s 200*F . . ... . 16.1-66 n
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M e MASTER TABLE OF CONTENTS (Cont.) ! Section Title Page l 3.1.3 Core Reactivity . . . . . . . . . . . . . . . . ..... ... 16.1-71 l 3.1.4 Moderator Temperature Coefficient (MTC) . . . .. 16.1-79 3.1.5 Rod Group Alignment Limits . . . . . . . . . . . . . . . 16.1 - 8 8 3.1.6 Shutdown Bank Insertion Limits . . . . . . ... .. 16.1-103 3.1.7 Control Bank Insertion Limits . . . . . ..... .. 16.1-110 3.1.8 Rod Position Indication . . . . . . . . . ...... .. 16.1-119 3.1.9 Mode 2 Physics Test Exceptions . . . . . . . . . . . . . 16.1-127 3.1.10 Chemical and Volume Control System (CVS) Demineralized Water Isolation Valves . . . . . . . . . 16.1-138 3.2 Power Distribution Limits . . . . . . . ........ .... . 16.1-143 3.2.1 Heat Flux Hot Channel Factor qF (Z) . . . . . . .. 16.1-143 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F5) . . 16.1-157 3.2.3 Axial Flux Difference (AFD) . . .. ........ 16.1-167 3.2.4 Quadrant Power Tilt Ratio (QI'rR) ......... . 16.1-175 3.2.5 OPDMS-Monitored Power Distribution Parameters 16.1-185 3.3 Instrumentation ............ ...... ... .... .. 16.1-192 3.3.1 Reactor Trip System Instmmentation . . . . . . . . . . 16.1-192 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation . . . . . ................ .. 16.1-258 3.3.3 Post-Accident Monitoring Instrumentation . . . . . . 16.1-309 3.3.4 Remote Shutdown Workstation . . . . . . . . . . . .. 16.1-323 3.3.5 Main Control Room Habitability System (VES) Actuation Instmmentation . . . . ........ .. 16.1-331 ! 3.4 Reactor Coolant System (RCS) . . . . . . . . . . . . . . . . . . . 16.1-340 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits . . . . . . . . . . . . . 16.1-340 3.4.2 RCS Minimum Temperature for Criticality ..... 16.1-350 3.4.3 RCS Pressure and Temperature (P/T) Limits . . . . . 16.1-356 I 3.4.4 RCS Loops - Modes 1 and 2 . . . . ........... 16.1-369 3.4.5 Pressurizer . ..................... .... 16.1-374 3.4.6 Pressurizer Safety Valus . .. ..... .. 16.1-378 l 3.4.7 RCS Operational Leakage . . . .. ..... . .. 16.1-384 3.4.8 RCS Pressure Isolation Valve (PIV) Leakage ... 16.1-392 3.4.9 RCS Leakage Detection Instrumentation . . . . . . 16.1-401 3.4.10 RCS Specific Activity . .. ......... .... . 16.1-411 3.4.11 RCS Loops . . . . ........... ........ .. 16.1-417 3.4.12 Automatic Depressurization System . . . ...... 16.1-421 3.4.13 Low Temperature Overpressure Protection (LTOP) System ... ... .. ....... . .. .. 16.1-427 3.4.14 Minimum RCS Flow . ..... ... . ... . 16.1-437 O Revision: S February 29,1996 xxvi [ W85tingl10USS
l s-s En - MASTER TABLE OF CONTENTS (Cont.) Section Title g 3.5 Passive Core Cooling System (PXS) . . . . ............ 16.1-441 3.5.1 Accumulators . . . . . . . . . . . . . . ............ 16.1-441 3.5.2 Core Makeup Tanks . . . . . . . . . ....... ... 16.1-449 3.5.3 Passive Residual Heat Removal System (PRHR) . . 16.1-459 3.5.4 In-containment Refueling Water Storage Tank (IRWST) . . . . ........... ......... ... 16.1-466 3.6 Containment Systems .. ............. ...... 16.1-473 3.6.1 Containment ..... .. . .... ..... . 16.1-473 3.6.2 Containment Air Locks . . . . . . .. . . ... 16.1-479 3.6.3 Contamment Isolation Valves . . . ........... 16.1-491 3.6.4 Contamment Pressure . . . . . . . ............ 16.1-506 3.6.5 Containment Air Temperature . . . . . . .... . 16.1-510 3.6.6 Passive Containment Cooling System . ...... 16.1-514 3.6.7 Hydrogen Recombiners . .. .............. 16.1-522 3.6.8 pH Adjustment Tank . .... .. .. . ... . 16.1-529 3.7 Plant Systems . . . . . . . . . . . .. ....... .......... 16.1-534 3.7.1 Main Steam Safety Valves (MSSVs) .... .. 16.1-534 3.7.2 Main Steam Isolation Valves (MSIVs) . . . . . . . . . 16.1-544 l 3.7.3 Main Feedwater Isolation and Control Valves . . 16.1-552 3.7.4 Secondary Specific Activity . ....... .... 16.1-559 3.7.5 Spent Fuel Pit Water Level . . . . . ........... 16.1-563 3.7.6 Main Control Room (MCR) Room Habitability S ystem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16.1-567 3.7.7 Startup Feedwater Isolation and Control Valves 16.1-577 3.8 Electrical Power Systems . . . . . . . . . . . .. .......... 16.1-581 3.8.1 Class IE DC Sources - Operating . ......... 16.1-581 3.8.2 Class 1E DC Sources - Shutdown . . . . . . . . . .. 16.1-595 3.8.3 Class 1E Uninterruptible Power Supply (UPS) Inverters - Operating ... ...... .. ... .. 16.1-601 3.8.4 Class 1E Uninterruptible Power Supply (UPS) Inverters - Shutdown . . . . . . . . . .. .... 16.1-608 3.8.5 Class IE Didribution Systems - Operating . . .. 16.1-614 l 3.8.6 Class IE Distribution Systems - Shutdown . . . . . 16.1-623 3.8.7 Class 1E Battery Cell Parameters ... .. . . 16.1-629 3.9 Refueling Operations . . . . . ... ......... .. ... . 16.1-640 3.9.1 Boron Conceatration . . . . . . . . . . . . . . . ...... 16.1-640 3.9.2 Unborated Water Source Isolation Valves . . . 16.1-645 3.9.3 Nuclear Instrumentation . . . . . . . . . . . . ... 16.1-650 3.9.4 Refueling Cavity Water Level . . . . . .. . ... 16.1-655 p V Revision: 5 W8 @ 00S8 xxvii February 29,1996 1 I
l 1 l l MASTER TABLE OF CONTENTS (Cont.) Section Title hage 4.0 Design Features . . ...... ....... ...... .... . 16.1-659 4.1 Site . . . . . . . . . . . . . . . . ....................... 16.1-659 4.1.1 Site and Exclusion Boundaries . .. .......... 16.1-659 4.1.2 Low Population Zone (LPZ) . .............. 16.1-659 4.2 Reactor Core . . . . . . . . . . . . . . . . . . ... ...... ... 16.1-659 4.2.1 Fuel Assemblies . . . . . . . . . . . . . . . . . . . ..... 16.1-659 4.2.2 Control rod and Gray Rod Assemblies . . . . .... 16.1-659 4.3 Fuel Storage . . . . . . . . . . . . . . . . . . ... ... ... .. 16.1-660 4.3.1 Criticality . . . . . . . . ..... ... ....... .. 16.1-660 4.3.2 Drainage . . . . . . . . ..... .. .. ... ... 16.1-660 4.3.3 Capacity . . . . . . . . ................... 16.1-660 5.0 Administrative Controls . . . . . . . . . . . . . . . . . .. ....... .. 16.1-661 VOLUME 11 16.2 Reliability Assurance Program . . . . . . ..... ................ . . . 16.2-1 16.2.1 Introduction . . . . ........ ................. ... . 16.2-1 16.2.2 Objective . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .... .. . 16.2-1 16.2.3 Design Reliability Assurance Program (D-RAP) . . . . . . . . . . . . . . 16.2-2 16.2.4 Combined License Applicant RAP (0-RAP) . . . . . . . . . . . . . . . . . 16.2-3 16.2.5 Glossary of Terms ... ............. . .. . ...... 16.2-4 16.2.6 References . . . . . . . . . . . . ... .......... . . . . . . . . . . . . 16.2-5 CHAP'ER 17 QUALITY ASSURANCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17.1-1 17.1 Quality Assurance During the Design and Construction Phases . . . . . . . . . . 17.1 - 1 17.2 Quality Assurance During the Operations Phase . . . . . . . . . . . . . . . . . . 17.1 - 1 17.3 Quality Assurance During Design, Procurement, Fabrication, Inspection and/or Testing of Nuclear Power Plant Items and Services . ... . . 17.1-1 17.4 Combined License Information Items ............. . ...... .... 17.1-2 l 17.5 References ........ ......... . ...... .... . . .. . . 17.1-2 1 l CHAPTER 18 HUMAN FACTORS ENGINEERING . ........ ...... ..... . 18.1-1 L 18.1 Overview . . . . . . . . ................................ . .. . 18.1-1 l 18.1.1 General . . . . ..................... . . . . . . . . . . . . . . . 1 8.1 - 1 l 18.1.2 References . . . . . . . . . . . . .. ..... .. . ... .. .... . 18.1-3 l 18.2 Introduction . . . . . . . . . . . . .................... ......... . 18.2-1 l 18.3 Past Experience and Lessons Learned .......... .......... .. . 18.3-1 18.3.1 Nuclear Power Plant Operating Experience . . . . . . . . . . . . . . . 18.3- 1 18.3.2 Lessons Learned from Other Process Control Industries ...... .. I8.3-1 O
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m (% )) MASTER TABLE OF CONTENTS (Cont.) Section Title Page 18.4 M-MIS Design Team . ................................ ..... 18.4-1 18.4.1 Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.4- 1 18.4.2 Qualifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.4- 1 18.4.3 M-MIS Design Team Role . . . . . . . . . . . . . . . . . . . . .. . .... 18.4-2 18.5 Human Engineering Evaluation Cdteria - Global View . . . . . . . . .. . . . . 18.5-1 18.5.1 Basis for Evaluation Criteria . . . . . . . . . . ..... ........ . . 18.5-1 18.5.2 Design Evaluation Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 8.5- 1 18.5.3 References . . . . . ....................... .. . . . . . . 18.5-3 I 18.6 User Behavior / Decision Making Model . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.6-1 18.6.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.6- 1 l 18.6.2 Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.6-1 18.6.3 Approaches to Improve Human Reliability . . . . ... .. . . . . . . 18.6- 1 i 18.6.4 The Decision-Sets Organization Model (*) .... . . . . . . . . . . . . 18. 6-2 18.6.5 'Ihe User Behavior / Decision Making Model (*) . ............. 18.6-2 l 18.6.6 Role of the Operators in the AP600 Main Control Room (*) . . . . . . 18.6-2 , 18.6.7 Summary (*) . . . . .............. . . . . . . . . . . . . . . . . . . . 1 8. 6-2 i 18.6.8 References . . . . . ... ............................. . 18.6-3 18.7 Allocation and Determination of Staffing . . . . . . . . . . . . . . . . . . . . . . . . . . 18.7-1 I 18.7.1 Basis for Staffing Requirements . . . . . . . . . . . . . . . . . . . . . . . . . 18.7-1 (J) k 18.7.2 Starfing Requirements . . . . . . . . . . . ...... ............ . 18.7-1 18.8 Human Engineering Design and Implementation Process . . . . . . . . . . . . . . . 18.8-1 l 18.8.1 Scope and Mission of the Human Engineering Design and I Implementation Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18. 8- 1 18.8.2 Detailed Explanation of the Human Engineering Design Process . 18.8-2 18.8.3 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.8-22 ! 18.9 Design Process Results for the Main Control Area . . . . . . . . . . ... . . . . 18.9-1 18.9.1 Main Control Area Functional Design . . . . . . . ... . . . . . . . . . . 18.9- 1 18.9.2 'Ihe Alarm System . . . . . . . . . . . . .................... 18.9-4 18.9.3 Plant Information System ..............................18.9-7 { 18.9.4 Wall Panel Information Station .. ....... . . . . . . . . . . . . . . . 18.9- 8 18.9.5 The Qualified Data Processing System . . . . .......... ..... 18.9-8 18.9.6 Cdsis Management Information System ............ . . . 18.9-10 18.9.7 The Design for Controls . . . . . . . . ..................... 18.9-10 18.9.8 Design of Plant Procedure . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.9- 12 l 18.9.9 The AP600 Training Program . . ..... . . . . . . . . . . . . . . . . . 18.9- 16 18.9.10 Workstation Design .... . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.9- 19 18.9.11 Main Control Area Design . . . . . . . . . . . . . . . . . . . . . . . . .. . 18.9-20 18.9.12 Recommended Operating Staff Qualifications . . . . . . . . . . . . . . . 18.9-21 18.9.13 Recommended Staffing and Operating Philosophy for Each Plant Mode . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... ..... . 18.9-21 l n 1 J V Revision: 5 l [ W95tingh00$8 xxix February 29,1996
O MASTER TABLE OF CONTENTS (Cont.) Section Title Page 18.9.14 Main Control Area Validation . . . . . . . . . . . . . . . . . . . . . . . . . 18.9-21 18.9.15 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.9-22 18.10 Design Process Results for Other Areas Within the Main Control Room Envelope ..................... . . . . . . . . . . . . . . . . . . . . 18.10- 1 18.10.1 Switching and Tagging Area . . . . . . . . . . . . . . . . . . . . . . . . . . 18.10-1 18.10.2 Shift Supervisor's Office Design ......................... 18.10-1 18.10.3 Shift Supervisor's Clerk's Office Design . . . . . . . . . . . . . . . . . . . 18.10-2 18.10.4 Operations Staff Area Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.10-2 18.11 Design Process Results for Other Operations and Control Centers . . . . . 18.11-1 18.11.1 Remote Shutdown Room . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.11-1 18.11.2 Technical Support Center . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.1 1 -2 18.12 M-MIS Integration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.12-1 18.12.1 Areas Developed by the Combined License Applicant . . . . . . . . . 18.12- 1 18.12.2 Areas Included in AP600 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.12-1 18.13 Human Factors Design for the Non Man-Machine Interface Systems Portion of the Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18.13-1 18.13.1 General Plant Layout and Design . . . . . . . . . . . . . ........... 18.13-1 O (*) - Westinghouse Proprietary, provided under separate cover. O Revision: 5 February 29,1996 xxx T Westinghouse
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List of Effective Pcges AP600 O List of Effective Pages Volume 1 Page Revision Page Revision Page Revision Page Revision 1.9-23 1 1.9-67 1 1.9-111 1 1A-12 1 1.9-24 1 1.9-68 1 1.9-112 1 1 A-13 1 1.9-25 1 1.9-69 1 1.9-113 1 1A-14 1 1.9-26 1 1.9-70 1 1.9-114 1 !A-15 1 1.9-27 1 1.9-71 1 1.9-115 1 1 A-16 1 1.9-28 1 1.9-72 1 1.9-116 1 l A-17 1 1.9-29 1 1.9-73 1 1.9-117 1 1 A-18 1 1.9-30 1 1.9-74 1 1.9-118 1 1 A-19 1 1.9-31 1 1.9-75 1 1.9-119 1 1A-20 1 1.9-32 1 1.9-76 1 1.9-120 1 1A-21 1 1.9-33 1 1.9-77 1 1.9-121 1 1A-22 1 1.9-34 1 1.9-78 1 1.9-122 1 1A-23 1 1.9-35 1 1.9-79 1 1.9-123 1 1A-24 1 1.9-36 1 1.9-80 1 1.9-124 1 1A-25 1 1.9-37 1 1.9-81 1 1.9-125 1 1A-26 1 1.9-38 1 1.9-82 1 1.9-126 1 1A-27 1 1.9-39 1 1.9-83 1 1.9-127 1 1A-28 1 1.9-40 1 1.9-84 1 1.9-128 1 1A-29 1 1.9-41 1 1.9-85 1 1.9-129 1 1A-30 1 1.9-42 1 1.9-86 1 1.9-130 1 1 A-31 1 1.9-43 1 1.9-87 1 1.9-131 1 1A-32 1 1.9-44 1 1.9-88 1 1.9-132 1 1A-33 1 1.9-45 1 1.9-89 1 1.9-133 1 1A-34 1 1.9-46 1 1.9-90 1 1.9-134 1 1A-35 1 1.9-47 1 1.9-91 1 1.9-135 1 1A-36 1 1.9-48 1 1.9-92 1 1.9-136 1 1A-37 1 1.9-49 1 1.9-93 1 1.9-137 1 1A-38 1 1.9-50 1 1.9-94 1 1.9-138 1 1A-39 1 1.9-51 1 1.9-95 1 1.9-139 1 1A-40 1 1.9-52 1 1.9-96 1 1.9-140 1 1A-41 1 1.9-53 1 1.9-97 1 1.9-141 1 1A-42 1 1.9-54 1 1.9-98 1 1.9-142 1 1A-43 1 1.9-55 1 1.9-99 1 1 A-44 1 1.9-56 1 1.9-100 1 1 A-1 0 1A-45 1 1.9-57 1 1.9-101 1 1A-2 0 1A-46 1 1.9-58 1 1.9-102 1 1A-3 0 1A-47 1 1.9-59 1 1.9-103 1 1A-4 0 1A-48 1 1.9-60 1 1.9-104 1 1A-5 0 1A-49 1 1.9-61 1 1.9-105 1 1A-6 0 1A-50 1 1.9-62 1 1.9 106 1 1A-7 1 1A-51 1 1.9-63 1 1.9-107 1 1A-8 1 1A-52 1 1.9-64 1 1.9-108 1 1A-9 1 1A-53 1 1.9-65 1 1.9-109 1 1A-10 1 1A-54 1 1.9-66 1 1.9-110 1 1A-11 1 1A-55 1 Revision: 5 February 29,1996 2
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- 2C-4 5 3.1-9 0 3.2-24 0 3.2-68 0 2C-5 5 3.1-10 0 3.2-25 0 3.2-69 0 2C-6 5 3.1-11 0 3.2-26 0 3.2-70 0 2C-7 5 3.1-12 0 3.2-27 0 3.2-71 0 2C-8 5 3.1-13 0 3.2-28 0 3.2-72 0 2C-9 5 3.1-14 0 3.2-29 0 3.2-73 0 l 2C-10 5 3.1-15 0 3.2-30 0 3.2-74 0 2C-11 5 3.1-16 0 3.2-31 0 3.2-75 0 2C-12 5 3.1-17 0 3.2-32 0 3.2-76 0 L 2C-13 5 3.1-18 0 3.2-33 0 3.2-77 0 2C-14 5 3.1-19 0 3.2-34 0 3.2-78 0 2C-15 5 3.1-20 0 3.2-35 0 3.2-79 0 2C-16 5 3.1-21 0 3.2-36 0 3.2-80 0 3.1-22 0 3.2-37 0 3.2-81 'O 3-i 5 3.1-23 0 3.2-38 0 3.2-82 0 i
3-ii 5 3.1-24 0 3.2-39 0 3.2-83 0 [ 3-iii 5 3.1 25 0 3.2-40 0 3.2-84 0 i 3-iv 5 3.1-26 0 3.2-41 0 3.2-85 0 3-v 5 3.1-27 0 3.2-42 0 3.2-86 0 3-vi 5 3.1-28 0 3.2-43 0 3.2-87 0 3-vii 5 3.2-44 0 3.2-88 0 3-viii 5 3.2-1 1 3.2-45 0 3.2-89 0 3-ix 5 3.2-2 1 3.2-46 0 3.2-90 0 3-x -5 3.2-3 1 3.2-47 0 3.2-91 0 3-xi 5 3.2-4 1 3.2-48 0 3.2-92 0 3-xii 5 3.2-5 1 3.2-49 0 3.2-93 0 3-xiii 5 3.2-6 1 3.2-50 0 3.2-94 0 3-xiv- 5 3.2-7 1 3.2-51 0 3.2-95 0 3-xv 5 3.2-8 1 3.2-52 0 3.2-95 0 3-xvi 5 3.2-9 1 3.2-53 0 3.2-97 0 3-xvii 5 3.2-10 1 3.2-54 0 3.2-98 0 3-xviii 5 3.2-11 1 3.2-55 0 3.2-99 0 3-xix 5 3.2-12 1 3.2-56 0 3.2-100 0 3-xx 5 3.2-13 1 3.2-57 0 3.2-101 0 3-xxi 5 3.2-14 1 3.2-58 0 3.2-102 0 3.2 1 3.2-59 0 3.2-103 0 3.1-1 0 3.2-16 0 3.2-60 0 3.2-104 0 3.1-2 0 3.2-17 0 3.2-61 0 3.2-105 0 3.1-3 0 3.2-18 0 3.2-62 0 3.2-106 0 3.1-4 0 3.2-19 0 3.2-63 0 3.2-107 0 3.1 0 3.2-20 0 3.2-64 0 3.2-108 0 '/^\ Q Revision: 5 5 February 29,1996
List tf Effectiva Pages AP600 List of Effective Pages Volume 2 Page Revision Page Revision Page Revision Page Revision I- 3.2-109 0 3.4-23 4 3.6-25 4 3.7-21 2 j 3.2-110 0 3.4-24 4 3.6-26 4 3.7-22 2 3.2-111 0 3.6-27 4 3.7-23 2 3.2-112 0 3.5-1 5 3.6-28 4 3.7-24 2 l 3.2-113 0 3.5-2 5 3.6-29 4 3.7-25 2 3.2 114 0 3.5-3 5 3.6-30 4 3.7-26 2 3.2-115 0 3.5-4 5 3.6-31 4 3.7-27 2 3.2-116 0 3.5-5 5 3.6-32 4 3.7-28 2 3.2-117 0 3.5-6 5 3.6-33 4 3.7-29 2 3.2-118 0 3.5-7 5 3.6-34 4 3.7-30 2 3.2-119 0 3.5-8 5 3.6-35 4 3.7-31 2 3.2-120 0 3.5-9 5 3.6-36 4 3.7-32 2 3.2-121 0 3.5-10 5 3.6-37 4 3.7-33 2 l 3.2-122 0 3.5-11 5 3.6-38 4 3.7-34 2 3.5-12 5 3.6-39 4 3.7-35 2 3.3-1 2 3.5-13 5 3.6-40 4 3.7-36 2 3.3-2 2 3.5-14 5 3.6-41 4 3.7-37 2 3.3-3 2 3.5-15 5 3.6-42 4 3.7-38 2 3.3-4 2 3.5-16 5 3.6-43 4 3.7-39 2 3.3-5 2 3.6-44 4 3.7-40 2 i 3.3-7 2 3.6-1 4 3.6-45 4 3.7-41 2 l 3.6 2 4 3.6-46 4 3.7-42 2 3.4-1 4 3.6-3 4 3.6-47 4 3.7-43 2 l 3.4-2 4 3.6-4 4 3.7-44 2 3.4-3 4 3.6-5 4 3.7-1 2 3.7-45 2 3.4-4 4 3.6-6 4 3.7-2 2 3.7-46 2 , 3.4-5 4 3.6-7 4 3.7-3 2 3.7-47 2 3.4-6 4 3.6-8 4 3.7-4 2 3.7-48 2 3.4-7 4 3.6-9 4 3.7-5 2 3.7-49 2 3.4-8 4 3.6-10 4 3.7-6 2 3.7-50 2 3.4-9 4 3.6-11 4 3.7-7 2 3.7-51 2 3.4-10 4 3.6-12 4 3.7-8 2 3.7-52 2 3.4-11 4 3.6-13 4 3.7-9 2 3.7-53 2 3.4-12 4 3.6-14 4 3.7-10 2 3.7-54 2 3.4-13 4 3.6-15 4 3.7-11 2 3.7-55 2 3.4-14 4 3.6-16 4 3.7-12 2 3.7-56 2 3.4-15 4 .' .c.,- 17 4 3.7-13 2 3.7-57 2 3.4-16 4 3.6-18 4 3.7-14 2 3.7-58 2 3.4-17 4 3.6-19 4 3.7-15 2 3.7-59 2 3.4-18 4 3.6-20 4 3.7 16 2 3.7-60 2 3.4-19 4 3.6-21 4 3.7-17 2 3.7-61 2 3.4-20 4 3.6-22 4 3.7-18 2 3.7-62 2 3.4-21 4 3.6-23 4 3.7-19 2 3.7-63 2 3.4-22 4 3.6-24 4 3.7-20 2 3.7-64 2 O Revision: 5 February 29,1996 6
List of Fifective Pages AP600 List of Effective Pages Volume 2 Page Revision Page Revision Page Revision Paic Revision 3.7-65 2 3.7-135 2 3.7-223 2 3.7-311 2 3.7-66 2 3.7-137 2 3.7-225 2 3.7-313 2 3.7-67 2 3.7-139 2 3.7-227 2 3.7-315 2 3.7-68 2 3.7-141 2 3.7-229 2 3.7-317 2 3.7-69 2 3.7-143 2 3.7-231 2 3.7-319 2 3.7-70 2 3.7-145 2 3.7-233 2 3.7-321 2 3.7-71 2 3.7-147 2- 3.7-235 2 3.7-323 2 3.7-72 2 3.7-149 2 3.7-237 2 3.7-325 2 3.7-73 2 3.7-151 2 3.7-239 2 3.7-327 2 3.7-74 2 3.7-153 2 3.7-241 2 3.7-329 2 3.7-75 2 3.7-155 2 3.7-243 2 3.7-331 2 3.7 76 2 3.7-157 2 3.7-245 2 3.7-333 2 3.7-77 2 3.7-159 2 3.7-247 2 3.7-78 2 3.7-161 2 3.7-249 2 3.8-1 3
.3.7-79 2 3.7-163 2 3.7-251 2 3.8-2 3 3.7-80 2 3.7-165 2 3.7-253 2 3.8-3 3 3.7-81 2 3.7-167 2 3.7-255 2 3.8-4 3 3.7-82 2 3.7-169 2 3.7-257 2 3.8-5 3 ,
3.7-83 2 3.7-171 2 3.7-259 2 3.8-6 3 3.7-85 D) C 3.7-87 2 2 3.7-173 3.7-175 2 2 3.7-261 3.7-263 2 2 3.8-7 3.8-8 3 3 3.7-89 2 3.7-177 2 3.7-265 2 3.8-9 3 3.7-91 2 3.7-179 2 3.7-267 2 3.8-10 3 3.7-93 2 3.7-181 2 3.7-269 2 3.8-11 3 3.7-95 2 3.7-183 2 3.7-271 2 3.8-12 3 3.7-97 2 3.7-185 2 3.7-273 2 3.8-13 3 3.7-99 2 3.7-187 2 3.7-275 2 3.8-14 3 3.7-101 2 3.7-189 2 3.7-277 2 3.8-15 3 3.7-103 2 3.7-191 2 3.7-279 2 3.8-16 3 3.7 105 2 3.7-193 2 3.7-281 2 3.8-17 3 3.7 107 2 3.7-195 2 3.7-283 2 3.8-18 3 3.7-109 2 3.7-197 2 3.7-285 2 3.8-19 3 3.7-111 2 3.7-199 2 3.7-287 2 3.8-20 3 3.7-113 2 3.7-201 2 3.7-289 2 3.8-21 3 3.7-115 2 3.7-203 2 3.7-291 2 3.8-22 3 3.7-117 2 3.7-205 2 3.7-293 2 3.8-23 3 3.7-119 2 3.7-207 2 3.7-295 2 3.8-24 3 3.7-121 2 3.7-209 2 3.7-297 2 3.8-25 3 3.7-123 2 3.7-211 2 3.7-299 2 3.8-26 3 3.7-125 2 3.7-213 2 3.7-301 2 3.8-27 3 3.7-127 2 3.7-215 2 3.7-303 2 3.8-28 3 3.7-129 2 3.7-217 2 3.7-305 2 3.8-29 3 3.7-131 2 3.7-219 2 3.7-307 2 3.8-30 3 3.7-133 2 3.7-221 2 3.7-309 2 3.8-31 O 3 O Revision: 5 7 February 29,1996 y .. .
List of Effectiv2 Pages AM List of Effective Pages Volume 1 Page Revision Page Revision Page Revision Page Revision 3.8-32 3 3.8-76 3 3.8-33 3 3.8-77 3 3.8-34 3 3.8-78 3 3.8-35 3 3.8-79 3 3.8-36 3 3.8-80 3 3.8-? 3 3.8-81 3 3.8-38 3 3.8-82 3 3.8-39 3 3.8-83 3 3.8-40 3 3.8-84 3 3.8-41 3 3.8-85 3 3.8-42 3 3.8-86 3 3.8-43 3 3.8-87 3 3.8-44 3 3.8-89 3 3.8-45 3 3.8-91 3 3.8-46 3 3.S-92 3 3.8-47 3 3.8-93 3 1 3.8-48 3 3.8-94 3 3.8-49 3 3.8-95 3 3.8-50 3 3.8-51 3 3.8-52 3 3.8-53 3 3.8-54 3 3.8-55 3 3.8-56 3 3.8-57 3 3.8-58 3 3.8-59 3 3.8-60 3 3.8-61 3 3.8-62 3 3.8-63 3 3.8-64 3 3.8-65 3 3.8-66 3 3.8-67 3 3.8-68 3 3.8-69 3 3.8-70 3 3.8-71 3 3.8-72 3 3.8-73 3 3.8-74 3 l 3.8-75 3 Revision: 5 February 29,1996 8
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List of Effective Pages AP600 i' List of Effective Pages Volume 3 Page Revision Page Revision Page Revision Page Revision 3.9-1 5 3.9-45 5 3.9-89 5 3.9-135 4 i t 3.9-2 5 3.9-46 5 3.9-90 a 3.9-137 4 3.9-3 5 3.9-47 5 3.9-91 5 3.9-139 4-3.9-4 5 3.9-48 5 3.9-92 5 3.9-141 4 3.9-5 5 3.9-49 5 3.9-93 5 3.9-143 4 3.9-6 5 3.9-50 5 3.9-94 5 3.9-145 4 3.9-7 5 3.9-51 5 3.9-95 5 3.9-147 4 3.9-8 5 3.9-52 5 3.9-96 5 3.9-149 4 l 3.9-9 5 3.9-53 5 3.9-97 5 3.9-151 4 3.9-10 5 3.9-54 5 3.9-98 5 3.9-153 4 3.9-11 5 3.9-55 5 3.9-99 5 3.9-155 4 i 3.9-12 5 3.9-56 5 3.9-100 5 3.9-157 4 3.9-13 5 3.9-57 5 3.9-101 5 3.9-159 4 3.9-14 5 3.9-58 5 3.9-102 5 3.9-161 4 3.9-15 5 3.9-59 5 3.9-103 5 l 3.9-163 4 3.9-16 5 3.9-60 5 3.9-104 5 3.9-165 4 3.9-17 5 3.9-61 5 3.9-105 5 3.9-166 4 3.9 18 5 3.9-62 5 3.9-106 5 3.9-167 4 3.9-19 5 3.9-63 5 3.9-107 5 l O 3.9-20 5 3.9-64 5 3.9-108 3.9-168 3.9.169 4 l V ' 3.9-21 '5 3.9-65 5 3.9-109 5 5 3.9-170 5 5 4 3.9-22 5 3.9-66 5 3.9-110 5 3.9-171 4 3.9-23 5 3.9-67 5 3.9-111 5 3.9-172 5 3.9-24 5 3.9-68 5 3.9-112 5 3.9-25 5 3.9-69 5 3.9-113 5 3.10-1 5 3.9-26 5 3.9-70 5 3.9-114 5 3.10-2 5 l 3.9-27 5 3.9-71 5 3.9-115 5 3.10-3 5 3.9-28 5 3.9-72 5 3.9-116 5 3.10-4 5 3.9-29 5 3.9-73 5 3.9-117 5 3.10-5 5 3.9-30 5 3.9-74 5 3.9-118 5 3.10-6 5 3.9-31 5 3.9-75 5 3.9-119 5 3.10-7 5 3.9-32 5 3.9-76 5 3.9-120 5 3.10-8 5 3.9-33 5 3.9-77 5 3.9-121 4 3.9-34 5 3.9-78 5 3.9-122 4 3.11-1 1 l 3.9-35 5 3.9-79 5 3.9-123 4 3.11-2 1 L 3.9-36 5 3.9-80 5 3.9-124 4 3.11-3 1 3.9-37 5 3.9-81 5 3.9-125 4 3.11-4 1 3.9-38 5 3.9-82 5 3.9-126 4 3.11-5 1 3.9-39 5 3.9-83 5 3.9-127 5 3.11-6 1 l 3.9-40 5 3.9-84 5 3.9-128 5 3.11-7 1 3.9-41 5 3.9-85 5 7.9-129 5 3.11-8 1 3.9-42 5 3.9-86 5 3.9-130 5 3.11-9 1 3.9-43. 5 3.9-87 5 3 9-131 4 3.11-10 1 3.9-44 5 3.9-88 5 3.9-133 4 3.11-11 1 l: Revision: 5 9 February 29,1996
1 List of Effectiv2 Pcges APW List of Effective Pages Volume 3 Page Revision Page Revision Page Revision Page Revision 3.11-12 1 3B-19 0 3D-32 5 3D-76 5 3.11-13 1 3B-20 0 3D-33 5 3D-77 5 3.11-14 1 3B-21 0 3D-34 5 3D-78 5 3.11-15 1 3B-22 0 3D-35 5 3D-79 5 3.11-16 1 3B-23 0 3D-36 5 3D-80 5 3.11-17 1 3B-24 0 3D-37 5 3D-81 5 3.11-18 1 3D-38 5 3D-82 5 3.11-19 1 3C-1 5 3D-39 5 3D-83 5 3.11-20 1 3C-2 5 3D-40 5 3D-84 5 3.11-21 1 3C-3 5 3D-41 5 3D-85 5 3.11-22 1 3C-4 5 3D-42 5 3D-86 5 3.11-23 1 3C-5 5 3D-43 5 3D-87 5 3.11-24 1 3D-44 5 3D-88 5 3.11-25 1 3D-1 5 3D-45 5 3D-89 5 3.11-26 1 3D-2 5 3D-46 5 3D-90 5 3.11-27 1 3D-3 5 3D-47 5 3D-91 5 3.11-28 1 3D-4 5 3D-48 5 3D-92 5 3.11-29 1 3D-5 5 3D-49 5 3D-93 5 3.11-30 1 3D-6 5 3D-50 5 3D-94 5 3.11-31 1 3D-7 5 3D-51 5 3D-95 5 3.11-32 1 3D-8 5 3D-52 5 3D-96 5 3.11-33 0 3D-9 5 3D-53 5 3D-97 5 3D-10 5 3D-54 5 3D-98 5 3A-1 0 3D-11 5 3D-55 5 3D-99 5 3A-2 0 3D-12 5 3D-56 5 3D-100 5 3D-13 5 3D-57 5 3D-101 5 3B-1 0 3D-14 5 3D-58 5 3D-102 0 l 3B-2 0 3D-15 5 3D-59 5 3D-103 5 i 3B-3 1 3D-16 5 3D-60 5 3D-104 5 3B-4 l 1 3D-17 5 3D-61 5 3D-105 5 : 3B-5 1 3D-18 5 3D-62 0 3D-106 5 l 3B-6 1 3D-19 5 3D-63 0 3D-107 5 l 3B-7 0 3D-20 5 3D-64 0 3D-108 5 3B-8 0 3D-21 5 3D-65 0 3D-109 5 3B-9 0 3D-22 5 3D-66 0 3D-110 5 3B-10 0 3D-23 5 3D-67 0 3D-111 5 3B-Il 1 3D-24 5 3D-68 0 3D-112 0 3B-12 1 3D-25 5 3D-69 5 3D-113 0 3B-13 1 3D-26 5 3D-70 5 3D-114 0 3B-14 1 3D-27 5 3D-71 5 3D-Il5 5 3B-15 0 3D-28 5 3D-72 5 3D-116 5 3B-16 0 3D-29 5 3 0-73 5 3D-117 5 3B-17 0 3D-30 5 3D-74 5 3D-Il8 5 i 3B-18 0 3D-31 5 3P-75 5 3D-119 5 ) I Revision: 5 February 29,1996 10
I h List of Effective Pages - C AP600 ' List of Effective Pages Volume 3 Page Revision Page Revision Page Revision Page Revision 3D-120 5 4.1-6 3 4.2-39 3 i 4 3-28 3 ! 3D-121 - 5 4.1-7 3 4.2-40 3 4 3-29 3 3D-122 5 4.1-8 3 4.2-41 3 4 3-30 3 3D-123 5 4.1-9 3 4.2-42 3 4.3-31 3 3D-124 5 4.1-10 3 4.2-43 3 4 3-32 3 3D-125 5' 4.2-44 3 4 3-33 3 3D-126 5 4.2-1 3 4.2-45 3 4.3-34 3 i 4.2-2 3 4.2-46 3 4 3-35 3 1 3E-1 4 4.2-3 3 4.2-47 3 4 3-36 3 3E-3 4 4.2-4 3 4.2-48 3 4 3-37 3 3E-5 4 4.2-5 3 4.2-49 3 4.3-38 3 3E-7 4 4.2-6 3 4.2-50 3 43-39 3 3E-9 4- 4.2-7 3 4.2-51 3 43 40 3 3E-11 4 4.2 3 4.2-52 3 4 3 3 3E-13 4 4.2-9 3 4.2-53 3 43-42 3 3E-15 4 4.2-10 3 4.2-54 3 4 3-43 3 3E-17 4 4.2-11 3 4 3-44 3 3E-19 4 4.2-12 3 43-1 3 4.3-45 3 4.2-13 3 43-2 O 3F-1 ~1 4.2-14 3 43-3 3 4 3-46 3 O 3F-2 1 4.2-15 3 43-4 3 3 4 3-47 43-48 3 3 ' 4.2-16 3 43-5 3 43-49 3 3G-1 3 4.2-17 3 4.3-6 3 4 3-50 3 3G-2 3 4.2-18 3 43-7 3 4 3-51 3 30-3 3 4.2-19 3 43-8 3 43-52 3 i 4.2-20 3 43-9 3 43-53 3 3H-1 3 4.2-21 3 43-10 3 4 3-54 3
' 3H-2 3 4.2-22 3 43-11 3 4.3-55 3 3H-3 3 4.2-23 3 43-12 3 4 3-56 3 3H-4 3 4.2-24 3 4 3-13 3 4 3-57 3 4.2-25 3 4 3-14 ' 3 43-58 3 4-i 5 4.2-26 3 4.3-15 3 4 3-59 3 i
4-il 5 4.2-27 3 43-16 3 4 3-60 3 i 4-iii 5 4.2-28 3 4 3-17 3 4 3-61 3 4-iv 5 4.2-29 3 4 3-18 3 4 3-62 3 l 4-v 5 4.2-30 3 4 3-19 3 4 3-63 3 l 4-vi 5 4.2-31 3 4 3-20 3 4 3-64 3 ! 4-vii 5 4.2-32 3 4 3-21 3 4 3-65 3 4.2-33 3 4.3 22 3 4 3-66 3 4.1-l' 3 4.2-34 3 4 3-23 3 4.3-67 3 4.1-2 3 4.2-35 3 4 3-24 3 4 3-68 3 4.1-3 3 4.2-36 3 4 3-25 3- 4 3-69 3 4.1-4 3 4.2-37 3 43-26 3 43-70 3 4.1-5 3 4.2-38 3 4 3-27 3 4 3-71 3 f~h l d' Revision: 5 11 February 29,1996 i
List of Effective Pcges AP600 O List of Effective Pages Volume 3 Page Revision Page Revision Page Revision Page Revision 4.3-72 3 4.4-20 3 4 3-73 3 4.4-21 3 4 3-74 3 4.4-22 3 l 4 3-75 3 4.4-23 3 l 43-76 3 4.4-24 3 4.3-77 3 4.4-25 3 4.3-78 3 4.4-26 3 4 3-79 3 4.4-27 3 4 3-80 3 4.4-28 3 4 3-81 3 4.4-29 3 4 3-82 3 4.4-30 3 4 3-83 3 4.4-31 5 4 3-84 3 4.4-32 3 4 3-85 3 4.4-33 3 4 3-86 3 4.4-34 3 4 3-97 3 4.4-35 3 4.3-r , 3 4.4-36 3 4 3-89 3 4.4-37 3 4.3 3 4.4-38 3 4 3-91 3 4.4-39 3 4.3-92 3 4.4-40 3 4.3-93 3 4.4-41 3 4 3-94 3 4.4-42 3 4 3-95 3 4.4-43 3 4.4-44 3 4.4-1 3 4.4-45 3 4.4-2 5 4.4-46 3 4.4-3 3 4.4-4 3 4.5-1 5 4.4-5 3 4.5-2 5 4.4-6 3 4.5-3 5 4.4-7 3 4.5-4 5 4.4-8 3 4.4-9 3 4.6-1 3 4.4-10 3 4.6-2 3 4.4-11 3 4.6-3 3 4.4-12 3 4.6-4 3 l 4.4-13 3 4.4-14 3 4.4-15 3 4.4-16 3 !- 4 t-17 3 4.4-18 3 4.4-19 3 Revision: 5 February 29,1996 12
List of Effective Pages V AP600 List of Effective Pages Volume 4 Page Revision Page Revision Page Revision Page ' Revision 5-i 5 5.2-16 5 5.3-22 5 5.4-33 5 5-ii '5 5.2-17 5 5.3-23 5 5.4-34 5 5-iii 5 5.2-18 5 5.3-24 5 5.4-35 5 5-iv 5 5.2-19 5 5.3-25 5 5.4-36 5 5-v. 5 5.2-20 5 5.3-26 5 5.4-37 5 ; 5-vi 5 5.2-21 5 5.3-27 3 5.4-38 5 5-vii 5 5.2-22 5 5.3-28 5 5.4-39 5 5.2 23 5 5.3-29 5 5.4-40 5 5.1-1 5 5.2-24 5 5.3-30 3 5.4-41 5 5.1-2 5 5.2-25 5 5.3-31 3 5.4-42 5 5.1-3 5 5.2-26 5 5.3-32 3 5.4-43 5 5.1-4 5 5.2-27 5 5.4-44 5 5.1-5 5- 5.2-28 5 5.4-1 5 5.4-45 5 5.1-6 5 5.2-29 5 5.4-2 5 5.4-46 5 5.1-7 5 5.2-30 5 5.4-3 5 5.4-47 5 5.1-8 5 5.2-31 5 5.4-4 5 5.4-48 5 5.1-9 5 5.2-32 5 5.4-5 5 5.4-49 5 5.1-10 5 5.2-33 5 5.4-6 5 5.4-50 5 5.1 5 5.2-34 5 5.4-7 5 5.4-51 5 5.1-12 5 5.2-35 5 5.4-8 5 5.4-52 5 LL 5.1-13 4 '5.2-36 5 5.4-9 5 5.4-53 5 5.1-14 5 5.2-37 5 5.4-10 5 5.4-54 5 5.1-15 5 5.4-11 5 5.4'55 5 5.1-17 5 5.3-1 5 5.4-12 5 5.4-56 5 5.1-10 5 5.3 2 5 5.4-13 5 5.4-57 5 5.1-19 5 5.3-3 5 5.4-14 5 5.4-57 5 5.1-21 ' 5 5.3-4 5 5.4-15 5 5.4-5) 5 5.1-23 5 5.3-5 5 5.4-16 5 5.4-60 5 5.3-6 5 5.4-17 5 5.4-61 5 5.2-1 5 5.3-7 5 5.4-18 5 5.4-62 5 i 5.2-2 5 5.3-8 5 5.4-19 5 5.4-63 5 l 5.2-3 5 5.3-9 5 5.4-20 5 5.4-64 5 ; 5.2-4 5 5.3-10. 5 5.4-21 5 5.4-65 5 I 5.2-5 5 5.3-11 5 5.4-22 5 5.4-66 5 l 5.2-6 5 5.3-12 5 5.4-23 5 5.4-67 5 l 5.2 5- 5 3-13 5 5.4-24 5 5.4-68 5 i 5.2-8 5 5.3-14 5 5.4-25 5 5.4-69 5 5.2-9 5 5.3-15 5 5.4-26 5 5.4-70 5 ) 5.2-10 5 5.3-16 5 5.4-27 5 5.4-71 5 5.2-11 5 5.3-17 5 5.4-28 5 5.4-72 5 5.2-12 5 5.3-18 5 5.4-29 5 5.4-73 5 ! 5.2-13 5 5.3-19 5 5.4-30 5 5.4-74 5 5.2-14 5 5.3-20 5 5.4-31 5 5.4-75 5 i 5.2-15 5 5.3-21 5 5.4-32 5 5.4-76 5 r9 G. Revision: 5 13 February 29,1996 i
l I i List of Effectiva Pages l AP600 List of Effective Pages Volume 4 Page Revision Page Revision Page Revision Page Revision ' 5.4-77 5 6.1-6 3 6.2-38 5 6.2-82 5 5.4-78 5 6.1-7 3 6 2-39 5 6.2-83 5 5.4-79 5 6.1-8 3 6.2-40 5 6.2-84 5 5.4-80 5 6.1-9 3 6.2-41 5 6.2-85 5 5.4-81 5 6.1-10 3 6.2-42 5 6.2-86 5 5.4-82 5 6.1-11 3 6.2-43 5 6.2-87 5 5.4-82 5 6.2-44 5 6.2-88 5 5.4-83 5 6.2-1 5 6.2-45 5 6.2-89 5 5.4-84 5 6.2-2 5 6.2-46 5 6.2-90 5 5.4-85 5 6.2-3 5 6.2-47 5 6.2-91 5 5.4-86 5 6.2-4 5 6.2-48 5 6.2-92 5 5.4-87 5 6.2-5 5 6.2-49 5 6.2-93 5 5.4-88 5 6.2-6 5 6.2-50 5 6.2-94 5 5.4-89 5 6.2-7 5 6.2-51 5 6.2-95 5 5.4-90 5 6.2-8 5 6.2-52 5 6.2-96 5 5.4-91 5 6.2-9 5 6.2-53 5 6.2-97 5 5.4-92 5 6.2-10 5 6.2-54 5 6.2-98 5 5.4-93 5 6.2-11 5 6.2-55 5 6.2-99 5 5.4-94 4 6.2-12 5 6.2-56 5 6.2-100 5 5.4-95 4 6.2-13 5 6.2-57 5 6.2-101 5 5.4-% 4 6.2-14 5 6.2-58 5 6.2-102 5 5.4-97 4 6.2-15 5 6.2-59 5 6.2-103 5 5.4-98 5 6.2-16 5 6.2-60 5 6.2-104 5 5.4-99 5 6.2-17 5 6.2-61 5 6.2-105 5 5.4-100 5 6.2-18 5 6.2-62 5 6.2-106 5 5.4-101 5 6.2-19 5 6.2-63 5 6.2-107 5 5.4-103 4 6.2-20 5 6.2-64 5 6.2-108 5 6.2-21 5 6.2-65 5 6.2 109 5 6-i 5 6.2-22 5 6.2-66 5 6.2-110 5 6-ii 5 6.2-23 5 6.2-67 5 6.2-111 5 6-iii 5 6.2-24 5 6.2-68 5 6.2-112 5 6-iv 5 6.2-25 5 6.2-69 5 6.2-113 5 6-v 5 6.2-26 5 6.2-70 5 6.2-114 5 6-vi 5 6.2-27 5 6.2-71 5 6.2-115 5 6-vii 5 6.2-28 5 6.2-72 5 6.2-116 5 6.2-29 5 6.2-73 5 6.2-117 5 6.0-1 5 6.2-30 5 6.2-74 5 6.2-118 5 6.0-2 5 6.2-31 5 6.2-75 5 6.2-119 5 6.2-32 5 6.2-76 5 6.2-120 5 6.1-1 3 6.2-33 5 6.2-77 5 6.2-121 5 6.1-2 3 6.2-34 5 6.2-78 5 6.2-122 5 6.1-3 3 6.2-35 5 6.2-79 5 6.2-123 5 6.1-4 3 6.2-36 5 6.2-80 5 6.2-124 5
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6.1-5 3 6.2-37 6.2-81 5 6.2-125 5 O Revision: 5 February 29,1996 14 l l l
List of ENective Pages
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AP600 List of Effective Pages Volume 4 Page Revision Page Revision Page Revision Page Revision 6.2-126 5 6.2-170 5 6.2-219 5 6.3-26 5 6.2-127 5 6.2-171 5 6.2-220 5 6.3-27 5 6.2-128 5 6.2-172 5 6.2-221 5 6.3-28 5 6.2-129 5 6.2-173 5 6.2-222 5 6.3-29 5 6.2-130 5 6.2 174 5 6.2-223 5 6.3-30 5 6.2-131 5 6.2-175 5 6.2-224 5 6.3-31 5 6.2-132 5 6.2-176 -5 6.2-225 5 6.3-32 5 6.2-133 5 6.2-177 5 6.2-226 5 6.3-33 5 6.2-134 5 6.2-178 5 6.2-227 5 6.3-34 5 6.2-135 5 6.2-179 5 6.2-228 5 6.3-35 5 6.2 136 5 6.2-180 5 6.2-229 5 6.3-36 5 6.2-137 5 6.2-181 5 6.2-230 5 6.3-37 5 6.2-138 5 6.2-182 5 6.2-231 5 6.3-38 5 6.2-139 5 6.2-183 5 6.2-232 5 6.3-39 5 6.2-140 5 6.2-184 5 6.2-233 5 6.3-40 5 6.2-141 5 6.2-185 5 6.2-234 5 6.3-41 5 6.2-142 5 6.2-187 5 6.2-235 5 6.3-42 5 6.2-143 5 6.2-189 5 6.2-236 5 6.3-43 5 6.2-144 5 6.2-191 5 6.3-44 A 6.2-145 5 6.2-193 5 6.3-1 5 C 6.2-146 6.2-147 5 6.2-195 5 6.3-2 5 5 6.3-45 6.3-46 5 5 5 6.2-l% 5 6.3-3 5 6.3-47 5 6.2-148 5 6.2-197 5 6.3-4 5 6.3-48 5 6.2-149 5 6.2-198 5 6.3-5 5 6.3-49 5 6.2-150 5 6.2-199 5 6.3-6 5 6.3-50 5 6.2 151 5 6.2-200 5 6.3-7 5 6.3-51 5 6.2-152 5 6.2-201 5 6.3-8 5 6.3 52 5 6.2-153 5 6.2-202 5 6.3-9 5 6.3-53 5 6.2-154 5 6.2-203 5 6.3-10 5 6.3-54 5 6.2-155 5 6.2-2M 5 6.3-11 5 6.3-55 5 6.2-156 5 6.2-205 5 6.3-12 5 6.3-56 5 6.2-157 5 6.2 206 5 6.3-13 5 6.3-57 5 6.2-158 5 6.2-207 5 6.3-14 5 6.3-58 5 6.2-159 5 6.2-208 5 6.3-15 5 6.3-59 5 6.2-160 5 6.2-209 5 6.3-16 5 6.3-60 5 6.2-161 5 6.2-210 5 6.3-17 5 6.3-61 5 6.2-162 5 6.2-211 5 6.3-18 5 6.3-62 5 6.2-163 5 6.2-212 5 6.3-19 5 6.3-63 5 6.2-164 5 6.2-213 5 6.3-20 5 6.3-64 5 6.2-165 5 6.2-214 5 6.3-21 5 6.3-65 5 6.2-166 5 6.2-215 5 6.3-22 5 6.3-66 5 6.2-167 5 6.2-216 5 6.3-23 5 6.3-67 5 6.2-168 5 6.2-217 5 6.3-24 5 6.3-68 5 6.2-169 5 6.2-218 5 6.3-25
/-m 5 6.3-69 5 Revision: 5 15 February 29,1996 l.
List of Effective Pages AP600 List of Effective Pages Volume 4 Page Revision Page Revision Page Revision Page Revision 6.3-70 5 7.1.6 5 7.1-50 5 7.2-39 3 6.3-71 5 7.1-7 5 7.1-51 5 7.2-41 5 6.3-72 5 7.1-8 5 7.1-53 5 7.2-43 5 6.3-73 5 7.1-9 5 7.1-54 5 7.2-45 5 6.3-75 5 7.1-10 5 7.1-55 5 7.2-47 5 6.3-77 5 7.1-11 5 7.1-56 5 7.2-49 5 6.3-78 5 7 1-12 5 7.1-57 5 7.2-51 5 6.3-79 5 7.1-13 5 7.1-58 5 7.2-53 3 7.1-14 5 7 1-59 5 7.2-55 5 6.4-1 5 7.1-15 5 7.1-60 5 7.2-57 5 6.4-2 5 7.1-16 5 7.1-61 3 7.2-59 5 6.4-3 5 7.1-17 5 7.2-61 5 6.4-4 5 7.1-18 5 7.2-1 5 7.2-63 3 6.4-5 5 7.1-19 5 7.2-2 5 7.2-65 5 6.4-6 5 7.1-20 5 7.2-3 5 6.4-7 5 7.1-21 5 7.2-4 5 7.3-1 5 6.4-8 5 7.1-22 5 7.2-5 5 7.3-2 5 6.4-9 5 7.1-23 5 7.2-6 5 7.3-3 5 6.4-10 5 7.1-24 5 7.2-7 5 7.3-4 5 6.4-11 5 7.1-25 5 7.2-8 5 7.3-5 5 6.4-12 .5 7.1-26 5 7.2-9 5 7.3-6 5 l 6.4-13 5 7.1-27 5 7.2-10 5 7.3-7 5 6.4-14 5 7.1-28 5 7.2-11 5 7.3-8 5 6.4-15 5 7.1-29 5. 7.2-12 5 7.3-9 5 7.1-30 5 7.2-13 5 7.3-10 5 6.5-1 1 7.1-31 5 7.2-h 5 7.3-11 5 6.5-2 1 7.1-32 5 7.2-15 5 7.3-12 5 7.1-33 5 7.2-16 5 7.3-13 5 6.6-1 5 7.1-34 5 7.2-17 5 7.3-14 5 6.6-2 5 7.1-35 5 7.2-18 5 7.3-15 5 l 6.6-3 5 7.1-36 5 7.2-19 5 7.3-16 5 l 6.6-4 5 7.1-37 5 7.2-20 5 7.3-17 5 7.1-38 5 7.2-21 5 7.3-18 5 7-i 5 7.1-39 5 7.2-22 5 7.3-19 5
' 7-ii 5 7.1-40 5 7.2-23 5 7.3-20 5 i
7-iii 5 7.1-41 5 7.2-24 5 7.3-21 5 l 7-iv 5 7.1-42 5 7.2-25 5 7.3-22 5 i 7-v 5 7.1-43 5 7.2-26 5 7.3-23 5 7.1-44 5 7.2-27 5 7.3-24 5 7.1-1 5 7.1-45 5 7.2-29 5 7.3-25 5 7.1-2 5 7.1-46 5 7.2-31 5 7.3-26 5 7.1-3 5 7.1-47 5 7.2-33 5 7.3-27 5 7.1-4 5 7.1-48 5 7.2-35 5 7.3-28 5 7.1-5 5 7.1-49 5 7.2-37 5 7.3-29 5 Revision: 5 February 29,1996 16
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List of Effective Pages j s
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AP600 List of Effective Pages Volume 4
'Page - Revision - Page Revision Pt.ge Revision Page Revision 7.3 5 7.5-23 0 7.7-24 5 7.3-31 5 7.5-24 0 7.7-25 5 7.3-32 5 7.5-25 0 7.7-26 5 7.3-33 5 7.5 0 7.3-34 .5 7.5-27 0 7.5-23 0 7.4-1 5 7.5-29 0 7.4-2 5 7.5-30 0 7.4-3 5 7.5-31 0 7.4-4 5 7.5-32 0 7.4-5 5 7.5-33 1
'7.4-6 5 7.5-34 1 7.4-7 5 7.5-35 1 7.4-8 5 7.5-36 1 7.4-9 5-7.4-10 3 7.6-1 5 7.4-11 5 7.6-2 5 7.4-12 5 7.6-3 5 7.4-13 5 7.6-4 5
[3 7.4-14 5 7.6-5 5 V 7.4-15 5 7.7-1 5 7.5-1 0 7.7-2 5
-7.5-2 0 7.7-3 5 7.5-3 0 7.7-4 5 7.5-4 0 7.7-5 -5
- 7.5-5 0 7.7-6 5 7.5-6 0 7.7-7 5 7.5-7 0 7.7-8 5 7.5-8 0 7.7-9 5 7.5-9 1 7.7-10 5 7.5-10 1 7.7-11 5 7.5-11 0- 7.7-12 5 7.5-12 0 7.7-13 5 7.5-13 0 7.7-14 5
- 7.5-14 0 7.7-15 5 7.5-15 1 7.7-16 5
. 7.5-16 1 7.7-17 5 7.5-17 0 7.7-18 5 l 7.5-18 0 7.7-19 5 7.5 1 - 7.7-20 5 7.5-20 1 7.7-21 ,5 i-7.5-21 0 7.7-22 5
. 7.5-22 0 7.7-23 5 f Revision: 5 17 February 29,1996 l
List of Effective Pages , l AF600 List of Effective Pages Volume 5 Page Revision Page Revision Page Ruision I-age Revision 8-i 5 8.3-26 3 9.1-1 5 9.1-45 5 8-ii 5 8.3-27 3 9.1-2 5 9.1-46 5 8-iii 5 8.3-28 3 9.1-3 5 9.1-47 5 8.3-29 3 9.1-4 5 9.1-48 5 8.1-1 3 8.3-30 3 9.1-5 5 9.1-49 5 8.1-2 3 8.3-31 3 9.1-6 5 9.1-50 5 8.1 3 3 8.3-32 3 9.1-7 5 9.1-51 5 8.1-4 3 8.3-33 3 9.1-8 5 9.1-52 5 8.1-5 3 8.3-34 3 9.1-9 5 9.1-53 5 8.1-6 3 8.3-35 3 9.1-10 5 9.1-54 5 8.1-7 3 8.3-36 3 9.1-11 5 9.1-55 5 8.1-8 3 8.3-37 3 9.1-12 5 9.1-56 5 8.1-9 3 8.3-38 3 9.1-13 5 9.1-57 5 8.1-10 3 8.3-39 3 9.1-14 5 9.1-58 5 8.3-40 3 9.1-15 5 9.1-59 5 8.2-1 3 8.3-41 3 9.1-16 5 8.2-2 3 8.3-43 3 9.1-17 5 9.2-1 3 8.2 3 3 8.3-45 3 9.1-18 5 9.2-2 3 8.3-47 3 9.1-19 5 9.2-3 3 8.3-1 3 8.3-49 3 9.1-20 5 9.2-4 3 8.3-2 3 8.3-51 3 9.1-21 5 9.2-5 3 8.3-3 3 8.3-53 3 9.1-22 5 9.2-6 3 8.3-4 3 8.3-55 3 9.1-23 5 9.2-7 3 8.3-5 3 8.3-57 3 9.1-24 5 9.2-8 3 8.3-6 3 8.3-59 3 9.1-25 5 9.2-9 3 8.3-7 3 8.3-61 3 9.1-26 5 9.2-10 3 8.3-8 3 8.3-63 3 9.1-27 5 9.2-11 3 8.3-9 3 8.3-65 3 9.1-28 5 9.2-12 3 8.3-10 3 8.3-67 3 9.1-29 5 9.2-13 3 8.3-11 3 9.1-30 5 9.2-14 3 8.3-12 3 9-i 5 9.1-31 5 9.2-15 3 8.3-13 3 9-ii 5 9.1-32 5 9.2-16 3 8.3-14 3 9-iii 5 9.1-33 5 9.2-17 3 8.3-15 3 9-iv 5 9.1-34 5 9.2-18 3 8.3-16 3 9-v 5 9.1-35 5 9.2-19 3 8.3-17 3 9-vi 5 9.1-36 5 9.2-20 3 8.3-18 3 9-vii 5 9.1-37 5 9.2-21 3 8.3-19 3 9-viii 5 9.1-38 5 9.2-22 3 8.3-20 3 9-ix 5 9.1-39 5 9.2-23 3 8.3-21 3 9-x 5 9.1-40 5 9.2-24 3 8.3-22 3 9-xi 5 9.1-41 5 9.2-25 3 8.3-23 3 9-xii 5 9.1-42 5 9.2-26 3 8.3-24 3 9-xiii 5 9.143 5 9.2-27 3 8.3-25 3 9-xiv 5 9.1-44 5 9.2-28 3 O l Revision: 5 l February 29,1996 18 i ______________ _
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List of Effective Pages (-m _ AP600 List of Effective Pages j Volume 5 l Page Revision Page Revision Page Revision Page Revision ! 9.2-29 3 9.2-75 '3 9.3-16 4 I 9.3-60 4 9.2-30 3 9.2-77 3- 9.3-17 4 9.3-61 4 9.2-31' 3 9.2-79 3 9.3-18 4 9.3-62 4 9.2-32 3- 9.2-81 3 9.3-19 4 9.3-63 4 9.2-33 3 9.2-83 3 9.3-20 4 9.3-64 4 9.2-34 3 9.2-85 3 9.3-21 4 9.3 4 9.2-35 3 9.2-87 3 9.3-22 4 9.3-66 4 9.2-36 3 9.2-89 3 9.3-23 4 9.3-67 4 9.2-37 3 9.2-91 3 9.3-24 4 9.3-68 4 9.2-38 3 9.2-93 3 9.3-25 4 9.3-69 4 ! 9.2 3 9.2 3 9.3-26 4 9.3-70 4 l
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- 9.2-49 3' ~ 9.2-115 3 9.3-36 4 9.3-80 4 9.2-50 3 9.2-117 3 9.3-37 4 9.3-81 4
' 9.2 3 9.2-119 3 9.3-38 4 9.3-82 4 9.2 3 9.2-121 3 9.3-39 4 9.3-83 4 9.2-53 3 9.2-123 3 9.3-40 4 9.3-85 4 i 9.2-54 3 9.2-125 3 9.3-41 4 9.3-87 4 9.2-55 3 9.2-127 3 9.3-42 4 9.3-89 4 9.2-56 3 .9.2-129 3 9.3-43 4 9.3-91 4 9.2-57 3' 9.3-44 4 9.3-92 4
-9.2-58 '3 9.3-1 4 9.3-45 4 9.3-93 4 9.2-59 3 -9.3-2 4 9.3-46 4 9.3-95 4 9.2-60 3 9.3-3 4 9.3-47 4 9.3-97 4
, 9.2-61 3 9.3-4 4 9.3-48 4 9.3-99 4
.9 2-62 3 9.3-5 4 9.3-49 4 9.3-101 4
,9.2-63 3 9.3-6 4 9.3-50 4 9.3-103 4 9.2-64 3 9.3 4 9.3-51 4 9.3-104 4 9.2-65 3 9.3-8 4 9.3-52 4 9.3-105 4 9.2-66 3 9.3-9 4 9.3-53 4 9.3-107 4 9.2-67 3 93-10 4 9.3-54 4 9.2-68 3 9.3-11 4 9.3-55 4 L 9.2-69' 3 9.3-12 4 9.3-56 4 i
' 9.2-70 3: 9.3-13 4' 9.3-57 4 9.2-71 3 9.3-14 4 9.3-58 4 9.2-73 3 9.3-15 4 9.3-59 4 ift/ \
4 Revision: 5 ! February 29,1996 ! 19 l'
List of Effective Pages AM List of Effective Pages Volume 6 Page Revision Page Revision Page Revision Page Revision 9.4-1 1 9.4-45 1 9.4-89 1 9.5-40 3 9.4-2 1 9.4-46 1 9.4-90 1 9.5-41 3 9.4-3 1 9.4-47 1 9.4-91 1 9.5-42 3 9.4-4 1 9.4-48 1 9.4-92 1 9.5-43 3 9.4-5 1 9.4-49 1 9.5-44 3 9.4-6 1 9.4-50 1 9.5-1 3 9.5-45 3 9.4-7 1 9.4-51 1 9.5-2 3 9.5-46 3 9.4-8 1 9.4-52 1 9.5-3 3 9.5-47 3 9.4-9 1 9.4-53 1 9.5-4 3 9.5-48 3 9.4-10 1 9.4-54 1 9.5-5 3 9.5-49 3 9.4-11 1 9.4-55 1 9.5-6 3 9.5-50 3 9.4-12 1 9.4-56 1 9.5-7 3 9.5-51 3 9.4-13 1 9.4-57 1 9.5-8 3 9.5-52 3 9.4-14 1 9.4-58 1 9.5-9 3 9.5-53 3 9.4-15 1 9.4-59 1 9.5-10 3 9.5.54 3 9.4-16 1 9.4-60 1 9.5-11 3 9.5-55 3 9.4-17 1 9.4-61 1 9.5-12 3 9.5-56 3 9.4-18 1 9.4-62 1 9.5-13 3 9.5-57 3 9.4-19 1 9.4-63 1 9.5-14 3 9.5-58 3 9.4-20 1 9.4-64 1 9.5-15 3 9.5-59 3 9.4-21 1 9.4-65 1 9.5-16 3 9.5-60 3 9.4-22 1 9.4-66 1 9.5-17 3 9.5-61 3 9.4-23 1 9.4-67 1 9.5-18 3 9.5-62 3 9.4-24 1 9.4-68 1 9.5-19 3 9.5-63 3 9.4-25 1 9.4-69 1 9.5-20 3 9.5-64 3 9.4-26 1 9.4-70 1 9.5-21 3 9.5-65 3 9.4-27 1 9.4-71 1 9.5-22 3 9.5-66 3 9.4-28 1 9.4-72 1 9.5-23 3 9.5-67 3 9.4-29 1 9.4-73 1 9.5-24 3 9.5-68 3 9.4-30 1 9.4-74 1 9.5-25 3 9.5-69 3 9.4-31 1 9.4-75 1 9.5-26 3 9.5-71 3 9.4-32 1 9.4-76 1 9.5-27 3 9.5-73 3 9.4-33 1 9.4-77 1 9.5-28 3 9.5-75 3 9.4-34 1 9.4-78 1 9.5-29 3 9.5-77 3 9.4-35 1 9.4-79 1 9.5-30 3 9.5-79 3 9.4-36 1 9.4-80 1 9.5-31 3 9.5-81 3 9.4-37 1 9.4-81 1 9.5-32 3 9.5-83 3 9.4-38 1 9.4-82 1 9.5-33 3 9.5-85 3 9.4-39 1 9.4-83 1 9.5-34 3 9.5-87 3 9.4-40 1 9.4-84 1 9.5-35 3 9.4-41 1 9.4-85 1 9.5-36 3 9A-1 1 9.4-42 1 9.4-86 1 9.5-37 3 9A-2 1 9.4-43 1 9.4-87 1 9.5-38 3 9A-3 1 9.4-44 1 9.4-88 1 9.5-39 3 9A-4 1 Revision: 5 February 29,1996 20
I List of Effectiv3 Pages
.b U AP6 o List of Effective Pages Volume 6 Page Revision Page Revision Page Revision Page Revision 9A-5 1 9A-49 1 9A-93 1 9A-137 1 I 9A-6 1 9A-50 1 9A-94 1 9A-138 1 9A-7 1 9A-51 1 9A-95 1 9A-139 1
. 9A-8 ' I 9A-52 1 9A-96 1 9A-140 1 9A-9 1 9A-53 1 9A-97 1 9A-141 1 9A-10 1- 9A-54 1 9A-98 1 9A-142 1 9A-Il 1 9A-55 1 9A-99 1 9A-143 1 9A-12 1 SA-56 1 9A-100 1 9A-144 1 9A 13 1 9A-57 I 9A-101 1 9A-145 1 9A-14 'I 9A-58 1 9A-102 1 9A-146 1 9A-15 1 9A-59 1 9A-103 1 9A-147 1 9A 16 1 9A-60 1 9A-104 1 9A-148 1 9A-17 1 9A-61 1 9A-105 1 9A-149 0 9A-18 1 9A-62 1 9A-106 1 9A-150 1 9A-19 1 9A-63 1 9A-107 1 9A-151 1 9A-20 1 9A-64 1 9A-108 1 9A-152 1 9A-21 1 9A-65 1 9A-109 1 9A 153 l 1
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- 9A-36 1 9A-80 1 9A-124 1 9A-37 1 9A-81 1 9A-125 1 9A-38 1 9A-82 1 9A-126 1 9A-39 1 9A-83 1 9A-127 1 9A-40 1- 9A-84 1 9A-128 1 9A-41 1 9A-85 1 9A-129 1 9A-42 1 9A-86 1 9A-130 1 9A-43 1 9A-87 1 9A-131 1 9A-44 1 9A-88 1 9A-132 1 9A-45 1 9A-89 1 9A-133 1 I
9A-46 1 9A-90 1 9A-134 9A-47 1 9A-91 1 9A-135 i 9A-48 1 9A-92 1 9A-136 1
- /7 V-Revision: 5 21 February 29,1996
i List of Effectiv3 Pcges l AP600 List of Effective Pages Volume 7 Page Revision Page Revision Page Revision Page Revision , 10-i 5 10.3-3 3 10.4-4 4 10.4-48 4 10-ii 5 10.3-4 3 10.4-5 4 10.4-49 4 10-lii 5 10.3-5 3 10.4-6 4 10.4-50 4 10-iv 5 10.3-6 3 10.4-7 4 10.4-51 4 10-v 5 10.3-7 3 10.4-8 4 10.4-52 4 10.vi 5 10.3-8 3 10.4-9 4 10.4-53 4 10.3-9 3 10.4-10 4 10.4-54 4 10.1-1 5 10.3-10 3 10.4-11 4 10.4-55 4 10.1-2 5 10.3-11 3 10.4-12 4 10.4-56 4 10.1-3 5 10.3-12 3 10.4-13 4 10.4-57 4 10.1-4 5 10.3-13 3 10.4-14 4 10.4-58 4 10.1-5 5 10.3-14 3 10.4-15 4 10.4-59 4 10.1-6 5 10.3-15 3 10.4-16 4 10.4-60 4 10.1-7 5 10.3-16 3 10.4-17 4 10.4-61 4 10.3-17 3 10.4-18 4 10.4-62 4 10.2-1 5 10.3-18 3 10.4-19 4 10.4-63 4 10.2-2 5 10.3-19 3 10.4-20 4 10.4-64 4 10.2-3 5 10.3-20 3 10.4-21 4 10.4-65 4 10.2-4 5 10.3-21 3 10.4-22 4 10.4-66 4 10.2-5 5 10.3-22 3 10.4-23 4 10.4-67 4 ! 10.2-6 5 10.3-23 3 10.4-24 4 10.4-68 4 10.2-7 5 10.3-24 3 10.4-25 4 10.4-69 4 10.2-8 5 10.3-25 3 10.4-26 4 10.4-70 4 10.2-9 5 10.3-26 3 10.4-27 4 10.4-71 4 10.2-10 5 10.3-27 3 10.4-28 4 10.4-72 4 l 10.2-11 5 10.3-28 3 10.4-29 4 10.4-73 4 10.2-12 5 10.3-29 3 10.4-30 4 10.4-74 4
- 10.2-13 5 10.3-30 3 10.4-31 4 10.4-75 4 10.2-14 5 10.3-31 3 10.4-32 4 10.4-76 4 10.2-15 5 10.3-32 3 10.4-33 4 10.4-77 4 10.2-16 5 10.3-33 3 10.4-34 4 10.4-78 4 10.2-17 5 10.3-34 3 10.4-35 4 10.4-79 4 10.2-18 5 10.3-35 3 10.4-36 4 10.4-80 4 10.2-19 5 10.3-36 3 10.4-37 4 10.4-81 4 10.2-20 5 10.3-37 3 10.4-38 4 10-4-83 4 10.2-21 5 10.3-38 3 10.4-39 4 10.4-85 4 10.2-22 5 10.3-39 3 10.4-40 4 10.4-85 4 10.2-?3 5 10.3-41 3 10.4-41 4 10.4-87 4 10.2-24 5 10.3-43 3 10.4-42 4 10.4-89 4 10.2-25 5 10.3-45 3 10.4-43 4 10.4-91 4 10.2-27 5 10.4-44 4 10.4-93 4 10.4-1 4 10.4-45 4 10.4-95 4 10.3-1 3 10.4-2 4 10.4-46 4 10.4-97 4 10.3-2 3 10.4-3 4 10.4-47 4 10.4-99 4 O
Revision: 5 February 29,1996 22
1 List of Effective Pages , ' List of Effective Pages Volume 7 i Page ' Revision Page Revision Page Revision Page Revision 10.4-101 4 11.2-16 1 11.4 4 11.5-1 3 ;
-10.4-103 4 11.2-17 I 11.4-3 4 11.5-2 3 ,
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List of Effective Pages AP600 List of Effective Pages Volume 7 Page Revision Page Revision Page Revision Page Revision 12.1-3 3 12.2-40 3 12.3-41 3 12.4-3 3 12.1-4 3 12.2-41 3 12.3-43 3 12.4-4 3 12.1-5 3 12.2-42 3 12.3-45 3 12.4-5 3 12.1-6 3 12.2-43 3 12.3-47 3 12.4-6 3 12.2-44 3 12.3-49 3 12.4-7 3 12.2-1 3 12.2-45 3 12.3-51 3 12.4-8 3 12.2-2 3 12.2-46 3 12.3-53 3 12.4-9 3 12.2-3 3 12.2-47 3 12.3-55 3 12.4-10 3 12.2-4 3 12.2-48 3 12.3-57 3 12.4-11 3 12.2-5 3 12.2-49 3 12.3-59 3 12.4-12 3 12.2-6 3 12.2-50 3 12.3-61 3 12.4-13 3 12.2-7 3 12.2-51 3 12.3-63 3 12.4-14 3 12.2-8 3 12.3-65 3 12.4-15 3 12.2-9 4 12.3-1 3 12.3-67 3 12.4-16 3 12.2-10 3 12.3-2 3 12.3-69 3 12.4-17 3 12.2-11 3 12.3-3 3 12.3-71 3 12.4-18 3 12.2-12 3 12.3-4 3 12.3-73 3 12.4-19 3 12.2-13 3 12.3-5 3 12.3-75 3 12.2-14 3 12.3-6 3 12.3-77 3 12.5-1 4 12.2-15 3 12.3-7 3 12.3-79 3 12.5-2 4 12.2-16 3 12.3-8 3 12.3-81 3 12.5-3 4 12.2-17 3 12.3-9 3 12.3-83 3 12.5-4 4 12.2-18 3 12.3-10 3 12.3-85 3 12.5-5 4 12.2-19 3 12.3-11 3 12.3-87 3 12.2-20 3 12.3-12 3 12.3-89 3 12.2-21 3 12.3-13 3 12.3-91 3 12.2-22 3 12.3-14 3 12.3-93 3 12.2-23 3 12.3-15 3 12.3-95 3 12.2-24 3 12.3-16 3 12.3-97 3 12.2-25 3 12.3-17 3 12.3-99 3 12.2-26 3 12.3-18 3 12.3-101 3 12.2-27 3 12.3-19 3 12.3-103 3 12.2-28 3 12.3-20 3 12.3-105 3 12.2-29 3 12.3-21 3 12.3-107 3 12.2-30 3 12.3-22 3 12.3-109 3 12.2-31 3 12.3-23 3 12.3-111 3 12.2-32 3 12.3-25 3 12.3-113 3 12.2-33 3 12.3-27 3 12.3-115 3 12.2-34 3 12.3-29 3 12.3-117 3 12.2-35 3 12.3-31 3 12.3-119 3 12.2-36 3 12.3-33 3 12.3-121 3 l 12.2-37 3 12.3-35 3 ' 12.2-38 3 12.3-37 3 12.4-1 3 12.2-39 3 12.3-39 3 12.4-2 3 O Revision: 5 February 29,1996 24
List of ENective Pages AP600 List of Effective Pages Volumme 8 Page Revision Page Revision Page Revision Page Revision 13-i 5 14.2-31 1 14.2-75 1 14.2-119 1 14.2-32 1 14.2-76 1 13-1 3 14.2-33 1 14.2-77 15-i 1 5 13-2 3 14.2 34 1 14.2-78 1 15-ii 5 13-3 3 14.2-35 14.2-79 1 1 15 iii 5 13-4 3 14.2 1 14.2-80 1 15-iv 5 13-5 3 14.2-37 1 14.2-81 15-v 1 5 13-6 3 14.2-38 1 14.2-82 1 15-vi 5 13-7 3 14.2-39 14.2-83 1 1 15-vii 5 13-8 3 14.2-40 14.2-84 1 1 15-viii 5
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LJ Revision: 5 25 February 29,1996
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t l l l List of Effective Pages l l l V AP600 List of Effective Pages Volume 8 Page Revision Page Revision Page ' Revision Page Revision 15.2-68 5 15 3-31 5 15.4-41 5 15.4-85 5 15.2-69 5 153-32 5 15.4-42 5 15.4-86 5 15.2-70 5 15 3-33 5 15.4-43 5 15.4-87 5 15.2-71 5 15.4-44 5 15.2-72 5 15.4-1 5 15.4-45 5 15.2-73 5 15.4-2 5 15.4-46 5 15.2-74 5 15.4-3 5 15.4-47 5 15.2-75 5 15.4-4 5 15.4-48 5 15.2-76 5 15.4-5 5 15.4-49 5 15.2-77 5 15.4-6 5 15.4-50 5 15.2 78 5 15.4-7 5 15.4-51 5 15.2-79 5 15.4-8 5 15.4-52 5 15.2-80 5 15.4-9 5 15.4-53 5 15.4-10 5 15.4-54 5 15 3-1 5 15.4-11 5 15.4-55 5 15 3-2 5 15.4-12 5 15.4-56 5 15 3-3 5 15.4-13 5 15.4-57 5 15 3-4 5 15.4-14 5 15.4-58 5 15 3-5 5 15.4-15 5 15.4-59 5
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Change Roadmap O V OIIAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF THE PLANT 1.1 Introduction
- 1. Schedule information was changed to COL information.
- 2. COL item was added on schedule.
- 3. The List of Acronyms was revised.
- 4. The list of system designators was added.
- 5. The discussion of tense used in SSAR was deleted.
1.3 Comparisons with Similar Facility Designs
- 1. Updated parameters to current design.
- 2. Added information on separate steam generator startup/ auxiliary feedwater nozzle.
(O) 3. Changed heat sink for service water. l l O V. i Revision: 5 [ W @ 00S8 R-1 February 29,1996
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Change Roadmap en-O CHAPTER 2 SITE CHARACTERISTICS
- 1. Added the basis for long-term diffusion estimates.
- 2. Added a discussion of bearing capacity.
- 3. Added a discussion of settlement.
- 4. Added a discussion of liquefaction.
- 5. Renumbered the Combined License information.
- 6. Added references for bearing capacity, settlement, and liquefaction.
- 7. Revised precipitations values in Table 2-1.
- 8. Added Table 2-2.
- 9. Added Table 2-3.
Appendix 2A Design Characteristics
- 1. Tables 2A-7 and 2A-li were revised to match analysis results.
- 2. Figure 2A-1 was deleted.
Appendix 2B Parametric Studies Related to AP600 Design Soil Profiles
- 1. Added Tables 2B-5 and 2B-8. Revised other table numbers.
- 2. Added more discussion of differences between the models of Appendices 2A,2B, and 3.72.
- 3. Changed fonaat of figure layouts (no change to plots).
Appendix 2C Seismic Lateral Earth Pressures
- 1. This is a new section.
Revision: 5 O February 29,1996 W Westirighouse R-2
ww Change Roadm:p g V CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.5 Missile Protection l
- 1. Revised criteria for high-pressure gas cylinders.
- 2. Added ASME Code, Section VIII gas storage tanks as vessel ruptures not considered i credible.
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- 3. Added a restriction on location of hydrogen supply. l l
- 4. Revised the discussion of turbine missiles and rotor material.
- 5. Added examples of rotating equipment excluded by the 2 percent rule.
- 6. Added information on steel ductility.
- 7. Reformatted missile protection interface as COL information.
p 8. Deleted two references. ( 3.9 Mechanical Systems and Components
- 1. Deleted reference to tests on reactor internals during seismic loads.
- 2. Revised reference to material in Appendix 3C.
- 3. Revised reference to 10 CFR 50.55a.
- 4. Clarified that inservice inspection program includes system pressure tests.
- 5. Specified baseline ASME OM Code.
- 6. Deleted reference to ASME Section XI.
- 7. Clarified the testing of RTNSS pumps.
- 8. Deleted reference to ASME Code for valves to be inservice tested. All safety-related valves are considered for inservice testing.
- 9. Added statement that AP600 does not have to be in cold shutdown to be in a safe shutdown.
(3 V Revision: 5 3 W8Stingh0088 R-3 February 29,1996
Change Roadm p l
- 10. Added provisions to check capability of RTNSS valves during operatior..
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- 11. Corrected reference to inservice testing table to 3.9-16.
- 12. Added nonreclosing pressure relief devices to inservice testing categories.
- 13. Added provision to normal plant operation in lieu of exercise test.
- 14. Changed the firing and replacing to a frequency of 20 percent every 2 years.
- 15. Added requirement for nonreclosing pressure relief devices.
- 16. Revised the description of deflectior ~.:imit for CRDM rod travel housings and added basis.
- 17. Specified that design specifications and design reports are made available by the COL to the NRC for audit.
- 18. Revised the reference for th ASME OM Code.
- 19. Deleted reference to Japanese test of reactor internals.
- 20. Revised load definitions in Table 3.9-3.
- 21. Revised the load combinations and added footnotes in Table 3.9-5.
- 22. Revised the load combinations and added footnotes in Table 3.9-8.
- 23. Added equations and footnotes in Table 3.9-11.
3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment
- 1. The number of 1/2 SSE earthquakes assumed for qualification purposes was specified as five.
- 2. The requirements for the use of experience data were added.
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- 3. Additional information was added to valve qualification discussion related to leakage.
- 4. Qualification requirements for valve disks were added.
- 5. Maintenance of the equipment qualification file by the Combined License applicant was added. <
l l 6. Two Combined License applicant information items were added. l l ' Revision: 5 e February 29,1996 [ W85tingh00S8 R-4
- Change Roadm
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('~h l iU Appendix 3C Reactor Coolant Loop Analysis Methods
- 1. Deleted test, table, and figures that provided unneeded detail /results.
- 2. Added requirement to consider the mass and stiffness effects of branch line piping on the loop piping when significant. Added reference to SSAR subsection 3.7.3.8.
- 3. Changed SSE damping values for internal concrete building from 7 to 5 percent.
Appendix 3D Methodology for Qualifying AP600 Safety-Related Electrical and Mechanical Equipment
- 1. References to IEEE 323-1983 were deleted throughout section; replaced with references to IEEE 323-1974.
- 2. Location of damping values was corrected to Table 3.7.1-1.
- 3. Information was added for radiation-harsh environment.
- 4. The definition of abnormal conditions was modified.
g3 5. Discussion of core inventory released to inventory was revised. U 6. Definition of similarity was modified.
- 7. The discussion number of component test was modified.
- 8. References were added for source terms.
- 9. Note 3 was added to Table 3D.5-4.
- 10. Sheets were added to Typical Abnormal Environmental Test Profile figure.
- 11. Discussion of limits of qualification based on analysis was modified.
- 12. Ambient temperature for equipment in air conditioned and ventilated areas was revised.
- 13. The number of small earthquakes for qualification was revised to five.
- 14. The discussion of the seismic and dynamic input motions used for qualification by testing was expanded.
,v Revision: 5 3 Westinghouse R-5 February 29,1996
n-m , Change Roadmip l CHAPTER 4 e REACTOR 4.4 . Thermal and Hydraulic Design
- 1. Page 4.4-31 revised to clarify incore instrumentation requirements (loose page amendment).
- 2. Spelling corrections in 4.4.2.
4.5 Reactor Materials
- 1. Changed the material specification for springs in the CRDM.
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Revision: 5 e February 29,1996 R-6 ggg
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# Reg Change Roadmap i
m CHAPTER 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS l 5.1 Summary Description
- 1. Revised discussion of ADS valve opening sequence.
- 2. Deleted size of pn:ssurizer nozzles.
- 3. Deleted safety valve set pressure.
- 4. Deleted system volumes.
- 5. Deleted thermal-hydraulic parameters not needed for design certification.
- 6. Revised mechanical design flow.
- 7. Moved Figures 5.1-3 through 5.1-5 to nonproprietary volume.
5.2 Integrity of the Reactor Coolant Pressure Isoundary
,ry
( ) 1. The section on leak detection was extensively revised to reflect NRC commitments and information developed as part of the detailed design. System designation was added for components used for leak detection.
- 2. Figure 5.2-1 was added to provide a schematic of the leak detection approach.
- 3. Information was added on the monitoring of identified leakage.
- 4. Valve stem leakoffs were deleted.
- 5. Reference to flow monitors in the leakoff lines was deleted. Reference to temperature indication was added.
- 6. Reference to other sections was added to intersystem leakage detection.
- 7. Radiation monitor for condenser air removal was changed to turbine island vent.
- 8. Pressure and temperature were added as parameters monitored in containment atmosphere.
, 9. Seismic requirements were expanded for containment atmosphere monitor.
- 10. The system used to calculate leak rate was corrected to the DDS.
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' \ ~V Revision: 5
[ W8Sfingh00S8 R-7 February 29,1996 l
q=a Change Roadarp O
- 11. Gas measurement was added to the containment atmosphere monitor description.
Explanations of the seismic requirements and Class IE power requirement were added.
-12. The description for the condensate monitor for the fan coolers was deleted.
- 13. Instruments were added to the list of instruments.
- 14. Reference to tech spec requirements was added.
- 15. The reference to ASME Code Case N-284 was corrected and changed to Rev.1.
5.3 Reactor Vessel
- 1. The pressure-temperature limit curves (Figures 5.3-2 and 5.3-3) were revised.
- 2. 'Iable 5.3-3, End-of-Life RTm and Upper Shelf Energy Projections was revised.
- 3. Additional information was provided in 5.3.3 on the use of the limit curves.
- 4. Paragraph 5.3.5.1 was revised.
5.4 Component and Subsystem Design
- 1. Added information on steam generator corrosion allowance.
- 2. Revised discussion of natural circulation with loss of offsite power.
- 3. Added information on bypass test line valves.
- 4. Added RNS design pressure.
- 5. Revised stem packing for RNS valves.
- 6. Revised dir,cussion of throttling of cooling water to RNS heat exchangers.
- 7. Added statements that venting of noncondensables by ADS valves from the pressurizer l steam space is not a safety-related function.
- 8. Deleted reference to valve leakoffs. Added information on valve misposition.
- 9. Revised the discussion of overflow from the IRWST. I J
- 10. Revised the figure reference for the PRHR heat exchanger. ;
l 1 Revisiom 5 February 29,1996 R-8 yg l l
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Change Roadm:p if 1 '. g i n , \.J
- 11. Removed suction temperature and motor power in pump parameters.
- 12. Removed heatup rate using only heaters in pressurizer parameters.
- 13. Changed design delta P for RNS isolation valves.
- 14. Revised specification of relieving capacity for pressurizer safety valve and RNS relief valve.
- 15. Moved Figure 5.4-7 to nonproprietary volume.
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i l !A > a V l Revision: 5 l W Westinghouse R-9 February 29,1996
Change Roadmap O CHAPTER 6 ENGINEERED SAFETY FEATURES
- 1. The description of the containment hydrogen control system was revised.
6.2 Containment Systems
- 1. Incorporated changes to Sections 6.2.1 and 6.2.4 previously submitted to the NRC as a hand markup.
6.3 Passive Core Cooling System
- 1. Incorporated design changes, including:
- Changed from NaOH to TSP for post-accident pH adjustment.
- Increased IRWST vent / overflow capacity.
6.4 Habitability Systems Deleted references to fire-smoke detection and alarm system (FDS) and fire protection 4 system (FPS). Changed all references from MCR " envelope" to MCR " pressure boundary." Increased maximum MCR occupancy during VES operation from five persons so eleven persons. Added design basis criteria for limiting CO2 concentration for both single and dual train operation. Modified general description of system to reflect a total of 24 air storage tanks, based on two redundant trains containing 12 tanks each. Clarified role of portable coolers, in that they are only used for cooling the MCR, not the electrical equipment rooms. s
- Identified that the emergency air storage tanks conform to Section VIII and Appendix 22 of the ASME code. (Justification of this was previouslyprovided in letter DCP/NRC0419 of 10/9/95 to the NRC.)
Changed reference from " pressure regulating valve" to " pressure control valve" Likewise, changed reference from " flow control orifice" to " flow metering orifice." Revision: 5 O February 29,1996 {gg R-10
4-w Changi Roadmap b Changed the location of the solenoid operated isolation valves from inside the MCR pressure boundary to upstream of the pressure regulating valve. Expanded the discussion on MCR leaktightness in response to an NRC meeting open item. Added discussions on isolation valve and pressure relief damper to component section. Changed terminology for system initiation from "High-2" MCR rc.diation to "Hi-Hi" MCR radiation. Added short discussion on CO2 toxicity, with appropriate references cited, per NRC request. Changed system flowrate from 20 scfm to 25 scfm, and identified it as a nominal value. Similarly, with both trains in operation, flowrate changes from 40 scfm to 50 scfm. t Added section on Combined License applicant responsibilities. Updated Table 6.4-2 to reflect the revised VES indications and alarms per current P&ID revision. c ) V
- Replaced reference to " Water Service Building" with " Turbine Building" in Table 6.4-2.
Updated Figures 6.4-1 and 6.4-2 to reflect curren MCR layout and P&ID drawing. Deleted references to respiratory, eye, and skin protection being provided for emergency use within the MCR pressure boundary. Added disclaimer on configuration of operator workstation and wall panel information system. Modified references to specific operator types, tank configurations, and MCR access doors in Component Description subsection 6.4.2.3. Deleted Component Data of Table 6.4-1 and renumbered remaining tables. Deleted nominal quantities for onsite chemicals table and referred to functional names rather than brand names. Included the system P&ID as Figure 6.4-2, thus making it non-proprietary. p L] Revision: 5 T WestingNUse R-11 February 29,1996
Change Roadmip 6.6 Inservice Inspection of Class 2 and 3 Components e
- 1. The description of the responsibility of the combined license applicain for the insenrice inspection programs was revised.
- 2. A reference to erosion-corrosion monitoring of nonsafety-relat d piping was added.
The discussion of this monitoring is found in Section 10.1. e' Revision: 5 O
. February 29,1996 g glg R-12
m _ 4 - i l 27 .:: l Change Roadmap * ! l
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(~) l CHAPTER 7 l INSTRUMENTATION AND CONTROLS 7.1 Introduction
- 1. Revised to correct minor typographical errors and inconsistent terminology usage.
- 2. Added definitions of " system-level" and " component-level" actuation.
- 3. Deleted refercnce to the waste processing control room as workscope of the Operations and Controls Centers System.
- 4. Added additional explanation of the functions performed by the trip enable subsystem.
- 5. Revised functional requirements of the soft control implementation to delete specific
! requirement for independent operator confirmation before control action is l implemented.
- 6. Removed implication that operation with two bypassed channels could be continued indefinitely.
i (m) 7.2 Reactor Trip l
- 1. Deleted temperature and pressure compensation of the steam generator level measurements.
- 2. Revised the range, accuracy, and response time of some reactor trip variables.
Clarified that accuracy and response times are typical values.
- 3. Deleted the detection of dropped rod (P-18 interlock) function.
- 4. Revised to correct minor typographical errors and inconsistent terminology usage.
- 5. Revised section numbers of other SSAR sections that are referred to in this section.
7.3 Engineered Safety Features
- 1. This revision incorporates numerous plant design changes that affect the details of engineered safety features actuation. These changes are summarized as follows:
Added reactor trip coincident with Low-2 reactor coolant system average i temperature as an actuation signal for main feedwater isolation and steam dump f block.
-O a j v
! Revision: 5 T W85tingh0088 R-13 February 29,1996 i
i h:: [ Change Roadmrp l 1 0' Deleted the IE confirmatory open signal to the accumulator and in-containment refueling water storage tank isolation valves. Added spent fuel pool isolation on low spent fuel pool water level. Deleted reference leg compensation for steam generator water level measurements. Modified the measurement of loss of ac power from the 4160 buses to the 480 buses. Revised the actuation logic for in-containment refueling water storage tank. i Added manual actuation for core makeup tank injection. l 1 Revised the actuation logic for the automatic depressurization system. Revised the actuation logic for the passive residual heat exchanger.
- Revised the actuation logic for steam generator blowdown isolation.
Revised the actuation logic for steam dump block.
- Deleted containment sump pH control. I
- 2. Revised the range, accuracy, and response time of some engineered safety features actuation variables. Clarified that accuracy and response times are typical values. '
- 3. Revised to cormet minor typographical errors and inconsistent terminology r se.
- 4. Revised section numbers of other SSAR sections that are referred to in this section.
7.4 Systems Required for Safe Shutdown
- 1. Revised to eliminate discussion of the condensate return to the IRWST in the discussion of safe shutdown using safety-related systems. The closure of the gutter drain to containment sump is a nonsafety-related function.
1
- 2. Added new discussion of safe shutdown using a combination of safety-related and ;
nonsafety-related systems in addition to the discussions of safe shutdown using only 1 safety-related or only nonsafety-related systems. 1 i l Revisiom 5 9 February 29,1996 T Westinghouse R-14 . 1 I l l
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Chang Roadm p n b
- 3. Revised the discussion of safe shutdown using nonsafety-related systems based on the design change, which provides a crossfeed of the startup feedwater line from the main feedwater pumps.
- 4. Revised to correct minor typographical errors and inconsistent terminology usage.
7.6 Interlock Systems Important to Safety 1. Revised to delete the discussion of the interlocks for the accumulator isolation valves and the IRWST injection isolation valves. These vaives are normally open, motor-operated with power locked during normal operation. For this reason, these valves have been designated as nonsafety-related and a discussion of the interlocks associated with these valves is not required to be included in this SSAR section. 2. Revised the section numbers of other SSA'Rsections referenced in the discussion.
- 3. Revised to delete specific reference to valve numbers. Valves are referred to by function.
- 4. Described the method by which diversity of the wide range pressure transmitters is achieved.
/ 5.
b Revised to delete the auto-closure function of the normal residual heat removal isolation valves.
- 6. Revised the discussion of the passive residual heat removal exchanger inlet isolation valve interlock to clarify that the control circuit power is not locked out and that this valve can be closed only for short periods of time during certain plant operating modes.
- 7. Revised the discussion of the ccre makeup tank balance line isolation valve interlock to clarify that the control circuit power is not locked out.
7.7 Control and Instrumentation Systems
- 1. Revised the description of the steam generator feedwater control to reflect the current plant design.
- 2. Revised the terminology used in the description of the rod control by changing references to the G control rod bank to the M0 control rod bank.
- 3. Deleted the functional diagrams for the nonsafety-related plant control system.
- 4. Revised the diverse actuation system manual actuations to add a manual containment recirculation function and a manual dump of in-containment refueling water storage tank to the containment sump function.
O. Revisiom 5 l [ W8Stingh0088 R-15 February 29,1996
Change Roadmcp l
- 5. Revised the diverse actuation system indications to add indication of core exit e
temperature. l l 6. Deleted the detection of dropped rod (P-18 interlock) function. l
- 7. Revised to correct minor typographical errors and inconsistent terminology usage.
l l I I O l r I a Revision: 5 February 29,19% R-16 M __ _ _ - - _ - _ - __ ____-_________________________a
Changs Roadmip - 7-
- 3 Ng 1
CHAPTER 9 AUXILIARY SYSTEMS 9.1 Fuel Storage and Handling
- 1. This revision resolves NRC Requests for Additional Information (RAls) 410.265, 410.266, 410.267, 410.268, 410.269, 410.271, 410.272, 410.273, 410.274, and 410.275. These RAIs inquire into the structural analysis methodology used in the analysis of the AP600 new, spent, and containment fuel racks. Since the AP600 fuel racks will be procured items, it will be the responsibility of the COL applicant to justify the analysis methodology used to qualify the fuel racks. Therefore, the methodology used in fuel rack analysis has been removed from the SSAR and replaced by a COL applicant requirement to perform the appropriate stress analysis and qualification.
- 2. A discussion about communication was added to subsection 9.1.4.2.2 rs rm i 4 Q)
Revision: 5 3 W8Stingh0088 R-17 February 29,1996
w. Change Roadntp
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CHAPTER 10 9 STEAM AND POWER CONVERSION SYSTEM 10.1 Smnmary Description
- 1. The subsection discussing Combined License infornntion (10.1.3) was modified to address COL responsibihty for an erosion / corrosion monitoring program.
- 2. Additional revisions were made to the system as follows:
- NRC requested that a system configuration heat balance be added to the nonproprietary section of the SSAR in DSER 1821. Instead, Figure 10.1-1 was made nonproprietary.
Table 10.1-1 (sheet 4 of 4) was modified to reflect a separation of the common driver motor from the feedwater pump information. Section 10.1.2 was modified to add a discussion of erosion / corrosion protection for steam and power conversion systems.
- Section 10.1.2 (Turbine Overspeed Protection) was revised to indicate that an electronic trip system was used for AP600 and that this trip system caused the steam supply valves to close if turbine speed exceeds 110 percent of rated speed.
10.2 Turbine-Generator
- 1. Combined License information was added as subsection 10.2.6 to address the DSER request for COL action items.
- 2. DSER numbers 359, 362,1133,1135,1136,1139,1822, and 1926 are included in the SSAR revision as follows:
DSER 359: NRC questioned the adequacy of the existing extraction nonreturn valve closure time to prevent turbine overspeed. SCS changed the response time of the extraction nonreturn valves with WTD input (reference SEI/FOK-0363 dated January 16, 1995). DSER 362 and 1133: NRC noted that the nominal T/G ratings were inconsistent in several SSAR locations. Subsection 10.2.2.1 and Table 10.2-1 were modified to incorporate the current nominal T/G rating. Revision: 5 O Februtry 29,1996 [ W85tlfigh00S8 R-18
it Chang) Roadm:p
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,f3 L} DSER 1135: NRC asked for additional information on the turbine rotor manufacturing process. A paragraph has been added to the Materials Selection subsection (10.2.3.1) to describe the basic turbine rotor steel manufacturing process. DSER 1136: NRC requested that the SSAR be modified to include information on basic chemical and material test requirements for the turbine rotor materials. The Materials Selection subsection (10.2.3.1) was modified to provide references to the appropriate ASTM standard for chemical and material properties. DSER 1139: NRC requested justification for not providing a surface examination of drilled and tapped areas on the T/G rotor. A statement was added to the Preservice Tests and Inspections subsection (10.2.3.5) which detailed the surface examination procedure. DSER 1822: NRC asked for a T/G outline drawing to include in the non-proprietary section of the SSAR. This drawing has been supplied as Figure 10.2-2. DSER 1926: NRC requested additional information on the turbine rotor materials with respect to fracture data on the turbine rotor. The Fracture
~
Toughness subsection (10.2.3.2) was supplemented to provide information on the turbine rotor steel tensile requirements, impact energy, and transition temperature. Additional revisions were made to the system as follows: Modified the descriptive language throughout Section 10.2 to provide a consistent spelling of moisture separator / reheater. Revised the description of the orientation of the turbine with respect to high-energy missiles (10.2.2). Power Generation Design Basis (10.2.1.2) text was modified to clarify the function of the MTS. Figure 10.2-1 was made nonproprietary. Modified the Turbine-Generator section (10.2.2.1) to include current nominal sizing and the change from two MSRs to a single MSR. Added an Exciter Description to subsection 10.2.2.3 in response to an RAI. A L) Revision: 5 W Westinghouse R-19 February 29,1996
r i l [ Change Roadatp 9' Changed " pressure switches" to " pressure sensors" in subsection 10.2.2.5.1 to provide consistent language throughout Section 10.2. Modified subsections 10.2.2.4,10.2.2.5.2, and 10.2.5 to reflect a language change from low condenser vacuum to high condenser backpressure. Modified information in Table 10.2-1 to provide new T/G ratings based on current heat balance and decreased the number of MSRs from two to one. DSER item numbers 358, 360, 361, 363,1134,1137,1138,1140,1141,1142, and 1925 were not addressed by SCS. These have been resolved and incorporated by Westinghouse. DSER 358: No mechanical overspeed trip provided for turbine-generator as specified in SRP 10.2; addressed in subsection 10.2.2.5.3. DSER 360: Turbine valve maintenance interval is different from that specified in SRP; addressed in subsection 10.2.3.6. DSER 361: NRC noted that the turbine valve test interval deviated from the interval specified in SRP 10.2; addressed in subsection 10.2.3.6.
- DSER 363: Responses to RAI numbers 410.139,410.143, and 410.144 were received after the DSER was prepared and are currently under review by NRC. Same as DSER items 358, 359, 360 and 361.
DSER 1134: NRC requests Westinghouse to address concern of diversity and conunon mode failure pertaining to a mechanical overspeed trip device; addressed in subsection 10.2.2.5.3. DSER 1137: Add information on Charpy V-notch properties and fracture toughness for the T/G rotor to the SSAR; addressed in subsection 10.2.3.1. DSER 1138: Westinghouse to provide a technical justification for not boring the HP T/G rotor centers; addressed in subsection 10.2.3.4. DSER 1140: Turbine maintenance program and missile probability calculation added to SSAR as a COL responsibility; addressed in subsection 10.2.t1
- DSER 1141: COL to submit test data and calculated toughness curves for the turbine rotor; addressed in subsection 10.2.6.
Revisiom 5 O February 29,1996 { Westinghouse R-20
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- DSER 1142: NRC has not determined acceptability of the T/G design due to questions on turbine overspeed trip, extraction nonreturn valve closure time, and valve testing intervals. Same as DSER items 359,360,361 and 363.
DSER 1925: NRC requests a submission from the COL of a turbine maintenance program and a turbine missile generation calculation or a volumetric rotor inspection program; addressed in subsection 10.2,6. (v\ l O Revision: 5 MDUSS R-21 February 29,1996
,y ~..
Change Roadmip CHAPTER 15 e ACCIDENT ANALYSES
- 1. Incorporated changes to Chapter 15 previously submitted to the NRC as a hand markup.
- 2. Removed appendices 15A, ISB,15C, ISD, ISE into new WCAP-14601 and changed the references to these appendices to WCAP-14601 in the body of Chapter 15.
O Revision: 5 February 29,1996 e l T Westinghouse ! R-22
_ _~ 1 1 ' Mnemate
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O CHAPTER 17 QUALITY ASSTIRANCE
- 1. The changes in Chapter 17 include:
Reference to the Quality Management System document as the applicable quality plan. Restructuring of the discussion of prior quality plans. l O O Revision: 5 T Westinghouse R-23 February 29,1996
1 i l i 4 Volumes 2 through 11 , 10 sets of Master Table of Contents (Insert infront of Volumes 2 through 11)
I o o . Package 2 of 3 P
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- 6. Engineered Safety Features
\ ! u 6.2 Contalmnent Systems 6.2.1 Containment Functional Design 6.2.1.1 Containment Structure 6.2.1.1.1 Design Basis he containment system is designed such that for all break sizes, up to and including the double-ended severance of a reactor coolant pipe or secondary side pipe, the containment peak pressure is below the design pressure. A summary of the results is presented in 1 I Table 6.2.1.1-1. 1 This capability is maintained by the containment system assuming the worst single failure affecting the operation of the passive containment cooling system (PCS). For primary system breaks, loss of offsite power (LOOP) is assumed. For secondary system breaks, offsite power ; is assumed to be available where it maximizes the mass and energy released from the break. l Additional discussion of the assumptions made for secondary side pipe breaks may be found in Subsection 6.2.1.4. he single failure postulated for the containment pressure / temperature calculations is the failure of one of the valves controlling the cooling water flow for the PCS. Failure of one (] V of these valves would lead to cooling water flow being delivered to the containment vessel through one of two delivery headers. His results in reduced cooling flow for PCS operation. No other single failures are postulated in the containment analysis. I The containment integrity analyses for the AP600 employed two types of models of the l containment. A multivolume lumped parameter model is used to study the long term I containment response to a postulated Loss of Coolant Accident (LOCA) and the Main Steam l Line Break (MSLB) accidents. The second modelis a detailed distributed or finite volume I model of the AP600 containment that is used to analyze the short-term response (i.e., the peak l pressure transient) of the containment. The analyses presented in this section are based on assumptions that are conservative with respect to the containment and its heat removal systems, such as minimum heat removal, and maximum initial containment pressure. The contamment design for the Safe Shutdown Earthquake (SSE) is discussed in Subsection 3.8.2. He minimum containment backpressure used in the Passive Core Cooling System (PXS) analysis is discussed in Subsection 6.2.1.5. lm v) Revision: 5 3 Westingh00S8 6.2-1 February 29,1996
- 6. Engineered S:fity Feat res m- l 6.2.1.1.2 Design Features 0'i 1
He operation of the PCS is discussed in Subsection 6.2.2. He arrangement of the l containment and internal structures is described in Section 1.2. I The reactor coolant loop is surrounded by stmetural walls of the containment internal l structures. These structural walls are a minimum of two feet - six inches thick and enclose the reactor vessel, steam generators, reactor coolant pumps, and the pressurizer. The containment vessel is designed and constructed in accordance with the ASME Code, Section III, Subsection NE, Metal Containment, including Addenda through 1989, as described in Subsection 3.8.2. Structural steel non-pressure retaining parts such as ladders, walkways, and handrails are designed to the requirements for steel structures defined in Subsection 3.8.4. The design features provide adequate containment sump levels following a design basis event as described in Subsection 3.4. Containment and subcompartment atmospheres are maintained during normal operation within prescribed pressure, temperature, and humidity limits by means of the contaimnent air recirculation system (VCS), the containment air filtration system (VFS), and the central chilled water system (VWS). The recirculation system cooling coils are provided with 45'F chilled water for temperature control. The filtration supply and exhaust subsystem can be utilized periodically to purge the containment air for pressure control. Periodic inspection and maintenance verify functional capability. 6.2.1.1.3 Design Evaluation l l Ty Wstinghouse-GOTHIC (WGOTHIC) computer code (Reference 1) is a computer , program lar modeling multiphase flow in a containment analysis. It solves the conservation
- equations in integral form for mass, energy, and momentum for multicomponent flow. The momentum conservation equations are written separately for each phase in the flow field (dropt liquid pools, and atmosphere vapor). The following terms are included in the momentum equation
- storage, cohvection, surface stress, body force, boundary source, phase i
interface source, and equipment source. l l l To model the passive cooling features of the AP600, several assumptions were made in I creating the plant decks. He external cooling water does not completely wet the containment i shell, therefore, both wet and dry sections of the she:1 were modeled in the WGOTHIC I analyses. The analyses assumed conservative coverage frardons that were selected based on I the duration of the transient. For example, a transient run extending 24 hours assumes the l coverages calculated at 24 hours for the entire transient. Table 6.'2.1.1-2 provides the I coverage fractions versus time calculated for the AP600. O Revision: S February 29,1996 6.2-2 [ W95tiligf10US8
- 6. Engineered S:f;ty Featrres O
I lieat conduction from the dry to wet section is not considered in the analysis, although calculations show this to be a benefit. Representative external cooling water flowrates, which I included the single failure assumption described earlier, are used for the wet sections. he I analyses also conservatively assume that the external cooling water is not initiated until 11 I minutes into the transient, allowing time to initiate the signal and to fill the headers and weirs I and develop the flow down the containment side walls. The effects of water flowing down i the shell from gravitational forces are explicitly considered in the analysis. The containment initial conditions of pressure, temperature, and humidity are provided in I Table 6.2.1.1-3. l For the LOCA events, two double-ended guillotine RCS pipe breaks are analyzed. The breaks I are postulated to occur in either a hot or a cold leg of the RCS. The hot leg break results in I the highest blowdown peak pressure. The cold leg break results in the higher post-blowdown peak pressure. The cold leg break analysis includes the long term contribution to containment pressure from the sources of stored energy, such as the steam generators. The LOCA mass and energy releases described in Subsection 6.2.1.3 are used for these calculations. For the MSLB event, a representative pipe break spectrum is analyzed. Various break sizes, power levels, and failure assumptions are analyzed with the WGOTHIC code. The MSLB mass and energy releases described in Subsection 6.2.1.4 are used for these calculations. I t] v l The results of the LOCA and MSLB postulated accidents are provided in Table 6.2.1.1-4. A comparison of the containment integrity analyses results to General Design Criterion 38 and I the Acceptance Criteria presented in the Standard Review Plan are also provided in I Table 6.2.1.1-4. An exception has been taken to the Standard Review Plan Acceptance I Criterion of reducing the containment pressure at 24 hours to less than 50 percent of the peak l calculated pressure. The exception is that 50 percent of the contaimnent design pressure is I being used instead of the peak calculated pressure. l The containment pressure response for the peak pressure steam line break case is provided in i Figure 6.2.1.1-1. The temperature response for this case is provided in Figure 6.2.1.1-2. I Figures 6.2.1.1-3 and 6.2.1.1-4 provide tne containment pressure and temperature response for l the peak temperature steam line break case. l The passive internal contamment heat sink data used in the WGOTHIC analyses is presented I in Tables 6.2.1.1-5 through 6.2.1.1-8. The contamment pressure and temperature responses I to a double-ended cold leg guillotine are presented in Figures 6.2.1.1-5 and 6.2.1.1-6 for the i distributed parameter model and Figures 6.2.1.1-7 and 6.2.1.1-8 for the lumped parameter 1 model. The lumped parameter model containment pressure and temperature response to a l double-ended hot leg guillotine break are presented in Figures 6.2.1.1-9 and 6.2.1.1-10. Data for both metallic and concrete heat sinks are presented. The physical properties of the l materials corresponding to the heat sink information is presented in Table 6.2.1.1-9. (3 i j! Revision: 5 [ W95tingh00S8 6.2-3 February 29,1996
ffEllEriW L Engineered Sity Features yg__ ; The containment shell temperature response is provided at three distinct elevations: the dome, el the spring line, and the operating deck level. Values for both wet and dry sections are i presented in Table 6.2.1.1 10. A discussion of the instrumentation provided inside containment to monitor and record the containment pressure and temperature is found in Section 7. I 6.2.1.1.4 External Pressure Analysis Certain design basis events and credible inadvertent systems actuation have the potential to result in containment external pressure loads. Evaluations of these events show that a loss of all ac power sources during extreme cold ambient conditions has the potential for creating the worst-case external pressure load on the containment vessel. This event leads to a reduction in the internal containment heat loads from the reactor coolant system and other active components, thus resulting in a temperature reduction within the containment and an accompanying pressure reduction. Evaluations are performed to determine the maximum external pressure to which the containment may be subjected during a postulated loss of all l ac power sources. For the loss of all ac power sources, ASME Service Level C limits are l l applicable and a containment external pressure of 3.0 psid is permitted. l The evaluations are performed with the assumption of a -40*F ambient temperature with a l l steady 48 mph wind blowing to maximize cooling of the containment vessel. The initial l internal containment temperature is conservatively assumed to be 120'F, creating the largest possible temperature differential to maximize the heat removal rate through the containment vessel wall. A negative 0.2 psig initial containment pressure is used for this evaluation. A I conservative maximum initial contaimnent relative humidity of 100% is used to produce the greatest reduction in containment pressure due to the loss of steam partial pressure by condensation. It is also conservatively assumed that no air leakage occurs into the containment dwing the transient. , i Evaluations are performed with conservatively low estimates of the containment heat loads and conservatively high heat removal through the containment vessel consistent with the i limiting assumptions stated above. Results of these evaluations demonstrate that at one hour I after the event the net external pressure is well within the 3.0 psid service level C limit. This I is sufficient time for operator action to prevent the containment pressure from dropping below l I the senice level C limit, based on the PAM's containment pressure indications (four I I containment pressure instruments) and the ability to mitigate the pressure reduction by l l opening either set of containment ventilation purge isolation valves, which are powered by the j l 1E batteries. ! I I The limiting case containment pressure transient is shown in Figure 6.2.1.1-11. O Revision: 5 February 29,1996 6.2-4 [ WB5tingh00S6 L _ _ _ _ _ _ _ _ _
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- 6. Engineered Satty Feat:res O
6.2.1.2 Containment Subcompartments 6.2.1.2.1 Design Basis Subcompartments within containment are designed to withstand the transient differential " pressures of a postulated pipe break. These subcompartments are vented so that differential pressures remain within structural limits. The subcompartment walls are challenged by the differential pressures resulting from a break in a high energy line. Therefore, a high energy line is postulated, with a break size chosen consiste ,t with the position presented in Section 3.6, for analyzing the maximum differential pressures across subcompartment walls. Section 3.6 describes the application of the mechanist ,c pipe break criteria, commonly referred to as leak-before-break (LBB), to the evaluation of 7,ipe mptures of pipes with a four inch or greater nominal diameter. This eliminates the need to consider the dynamic effects of postulated pipe breaks for pipes which qualify for LBB. However, the analyses of containment pressure and temperature, emergency core cooling, and environmental qualification of equipment are based on double-ended guillotine (DEG) reactor coolant system breaks and through-wall cracks. 6.2.1.2.1.1 Summary of Subcompartment Pipe Break Analyses Because LBB is applicable to pipes of four inches or greater in diameter, a postulated double
/7 V ended guillotine rupture of a three-inch line in the reactor coolant system (RCS)is analyzed to determine the maximum differential pressure across the subcompartment walls. The characteristics of the postulated rupture are determined in accordance with the methods and criteria of Section 3.6. Analyses are performed for a double ended guillotine break occurring in each of the subcompartments containing high energy piping, with the exception of the reactor vessel cavity where all piping is qualified to LBB (no high energy lines smaller than four inches are located in the reactor vessel cavity). Typical analysis models and results are I
described for breaks in a steam generator compartment, the pressurizer enclosure valve room, I maintenance fioor, and the operating deck. The steam generator compartment is analyzed for the effects of a three-inch double ended I guillotine break occurring in both the hot leg and cold leg pipe. In conformance with the Westinghouse mass and energy release methodology for subcompartment design, a 10 percent margin is applied to the releases for both postulated breaks. I The pressurizer enclosure valve room is analyzed for the effects of several different breaks. l
'Ihe three-inch double-ended cold-leg pipe release with 10 percent margin bounds other breaks l in this compartment.
I
'Ihe maintenance floor and operating deck regions are analyzed for the effects of several I
different breaks that could be postulated to occur in primary or secondary side piping. 'The i releases from a one square foot area rupture bound other postulated breaks in these I compartments. n h V Revision: 5 3 WBStiligh0USe 6.2-5 February 29,1996 7
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- 6. Engineered S1fity Featxres 1
l The CVS room is analyzed for the effects of several breaks that could be postulated to occur. 0!l I he three-inch double-ended cold-leg pipe release with 10 percent margin bounds other breaks I in this compartment. 1 The reactor vessel cavity pressurization loads are not considered, consistent with the position in Subsection 3.6.1.2. 6.2.1.2.2 Design Features The plant general arrangement drawings shown in Section 1.2 include descriptions of the l containment sub-compartments and surrounding areas. The general arrangement drawings are used in assembling the subcompartment analysis model. A detailed noding diagram of the i model is presented in Figure 6.2.1.2-1, Sheets 1 through 12. l To account for uncertainty between the as-built subcompartment configuration and the configuration modeled,40 percent margin is added to the calculated differential pressures, as discussed in Subsection 6.2.1.2.3.6. The subcompartment free volumes and the areas of the vent paths are presented in l Tables 6.2.1.2-1 through 6.2.1.2-3. Vent paths considered in the analyses are shown in the general arrangement drawings and consist of floor gratings and openings through walls. In the AP600 subcompartment analyses, no credit is taken for vent paths that become available only after the occurrence of the postulated break (such as blowout panels, doors, hinged panels and insulation collapsing). 6.2.1.2.3 Design Evaluation The TMD computer code (Reference 2) is used in the subcompartment analysis to calculate the differential pressures across subcompartment walls. The TMD code has been reviewed I by the NRC and approved for use in subcompartment differential pressure analyses. The methodology used to generate the short term mass and energy releases is described in Subsection 6.2.1.3.1. De initial atmospheric conditions used in the TMD subcompartment analysis are selected so that the calculated differential pressures are maximized. These conditions are chosen according to criteria identified in Subsection 6.2.1.2 of NUREG-0800 and include the maximum allowable air temperature, minimum absolute pressure, and zero percent relative I humidity. The initial conditions used in the analysis are tabulated in Tables 6.2.1.2-4. The containment and subcompartment atmospheres during normal operating conditions are maintained within prescribed pressure, temperature, and humidity limits by means of the containment air recirculation system (VCS), the containment air filtration system (VFS), and the central chilled water system (VWS). De recirculation system cooling coils are provided with 45'F chilled water to provide sufficient temperature control. He filtration supply and O Revision: 5 February 29,1996 6.2-6 [ Westillgl10USB
- 6. Engineered S f;ty Rctures u-
,m i %.;)
exhaust subsystem can be utilized to purge the containment air for pressure control. Periodic inspection and maintenance are performed to verify functional capability. 6.2.1.2.3.1 Flow Equation The flow equations used by the TMD code to calculate the flow between nodes are described in Reference 2. These flow equations are based on the unaugmented critical flow model, which demonstrate conservatively low critical flow velocity predictions compared to experimental test data. Due to the TMD calculation methods presented in Section 1.3.1 of l Reference. 2,100 percent entrainment results in the highest calculated differential pressures l and therefore this degree of entrainment is conservatively assumed in the subcompartment analysis. 6.2.1.2.3.2 Piping Systems l The subcompartment analysis for the steam generator compartment is performed assuming a l double-ended guillotine break in a three-inch inside diameter reactor cooling system hot leg l or cold leg pipe. The break is assumed to occur between the 84 foot, six inch elevation and I the 104 foot, three inch elevation of the steam generator compartment. Node 1 is the hot leg I break node of the TMD model and Node 2 is the cold leg break node (See Figure 6.2.1.2-1 for the noding diagram). Because the TMD code assumes homogeneous mixtures within a node, the specific location of the break within the node is not critical to the differential C'T pressure calculation. No flow restrictions exist that limit the flow out of the break. V l Re analysis for the pressurizer compartment pipe and valve room is performed assuming a I double-ended guillotine break in a three-inch inside diameter RCS hot leg pipe. This break I envelopes the branch lines that could be postulated to mpture in this area. The break is I as::umed to occur between the 107-foot,2-inch elevation and the 117-foot 6-inch elevation of I the pressurizer pipe and valve room compartment. Node 59 is the break node of the TMD l model (see Figure 6.2.1.2-1 for the noding diagram). I I The analysis for the steam generator vertical access area is performed assuming a double-I ended guillotine break in a three-inch inside diameter RCS cold-leg pipe, his break l envelopes the branch lines that could be postulated to rupture in this area. The break is I assumed to occur between the 83-foot,0-inch elevation and the 103-foot 5.5-inch eleva' ion I of the steam generator vertical access area compartment. Node 23 is the break node of the l TMD model (see Figure 6.2.1.2-1 for the noding diagram). l l The analysis for the maintenance floor and operating deck compartments are performed I assuming a one square foot rupture of a main steam line pipe. This break envelopes the I branch lines that could be postulated to mpture in these areas. The break is assumed to occur I between the 107-foot,2-inch elevation and the 135-foot,3-inch elevation of the maintenance I floor compartment and between the 135-foot,3-inch elevation and the 256-foot,2.375-inch I elevation of the operating deck region. Node 56 is the maintenance room break node and l Node 57 is the operating deck room break node of the TMD model (see Figure 6.2.1.2-1 for I the noding diagram). I,,I LJ Revision: 5 [ W85tingl100S8 6.2-7 February 29,1996
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- 6. Engineered Salty Feat:res l The analysis for the CVS room is performed assuming a double-ended guillotine break in a O
l three-inch diameter RCS cold-leg pipe. This break envelopes the branch lines that could be l postulated to rupture in this area. The breakis assumed to occur between the 91-foot,10-inch I elevation and the 105-foot,2-inch elevation of the CVS toom compartment. Node 66 is the l break node of the TMD model (see Figure 6.2.1.2-1 for the noding diagram). 6.2.1.2.3.3 Node Selection I The nodalization for the sub-compartments is analyzed in sufficient detail such that nodal boundaries are at the location of flow obstmetions or geometrical changes within the subcompartment. These discontinuities create pressure differentials between adjoining nodes. There are no significant discontinuities within each node, and hence the pressure gradient is negligible within any node. Details concerning the noding scheme are provided in Figure 6.2.1.2-1. 6.2.1.2.3.4 Vent Flowpath Flow Conditions The now characteristics for each of the subcompartments are tabulated in Tables 6.2.1.2-1 l through 6.2.1.2-3. These tables show that at no time during the transient does critical flow I exist through vent paths. The time-dependent mass and energy flow conditions are provided in Tables 6.2.1.3-2 and 6.2.1.3-3. 6.2.1.2.3.5 Vent Flowpath Flow Coefficients
! The subcompartment vent path data is tabulated in Tables 6.2.1.2-1 through 6.2.1.2-3. Loss coefficients are included in these tables.
6.2.1.2.3.6 Results l l The resultant maximum differential pressures of each node relative to the break node are I shown in Tables 6.2.1.2-5 through 6.2.1.2-9. The design of the sub-compartment walls for subcompartment pressurization is discussed in Subsection 3.8.3.3. I The results of the sub-compartment analysis demonstrate that the wall differential pressures I resulting from a high energy line break within the sub-compartments are well within the l design capability, even when a 40-percent margin is applied. I 6.2.1.3 Mass and Energy Release /.nalyses for Postulated Pipe Ruptures Mass and Energy releases are documented in this section for two different types of transients. l The first section describes the methodology used to calculate the releases for the subcompartment differential pressure analysis using the TMD code (referred to as the short term analysis). These releases are used for the subcompartment response in Subsection 6.2.1.2. O Revision: 5 February 29,1996 6.2-8 T Westirighouse
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- 6. Engineered Safety Featres g
Y l ne second section describes the methodology used to determine the releases for the containment pressure and temperature calculations using the WGOTIllC code (Reference 1) (referred to as the long term analysis). These releases are used for the containment integrity analysis in Subsection 6.2.1.1. The short term analysis considers only the initial stages of the blowdown transient, and takes i into consideration the application ofleak-before-break (LBB) methodology. LBB is discussed I in Subsection 3.6.3. Since LBB is applicable to RCS piping that is four inches in diameter I and greater, the mass and energy release analysis for sub-compartments postulates the l complete double-ended guillotine (DEG) severance of a three-inch pipe. The mass and energy I release postulated for a ruptured steam line is for a one square foot break. Conversely, the limiting break size for containment integrity analysis considers as its LOCA design basis the complete DEG severance of the largest reactor coolant system (RCS) pipe. The containment system receives mass and energy releases following a postulated rupture of the RCS. De release rates are calculated for pipe failure at two locations: the hot leg and the I cold leg. Rese break locations are analyzed for both the short-term and the long-term transients. Because the initial operating pressure of the RCS is approximately 2250 psi, the mass and energy are released extremely rapidly when the break occurs. As the water exits from the broken pipe, a portion ofit flashes to steam because of the differences in pressure and temperature between the RCS and containment. The RCS depressurizes rapidly since '(3 break flow exits both sides of the pipe in a DEG severance. V 6.2.1.3.1 Short Term Mass and Energy Release Data The AP600 short term LOCA mass and energy releases are predicted for the first ten seconds of the blowdown from a postulated DEG break of a three inch line in the RCS. The density of the fluid released from a postulated pipe rupture has a direct effect on the magnitude of the differential pressures that results scross subcompanment walls. A DEG mpture that is postulated in the cold leg piping is typically the most limiting scenario. This analysis provides mass and energy releases for a three inch DEG rupture in the cold leg and in the hot leg. The modified Zaloudek correlation (Reference 3) is used to calculate the critical mass flux from a three inch double-ended cold leg guillotine (DECLG) break and a three inch double-ended hot leg guillotine (DEHLG) break. This maximum mass flux is conservatively assumed to remain constant at the initial AP600 full power steady state conditions and the enthalpy is varied to determine the energy release rates. Conservative enthalpies are obtained from the SATAN-VI blowdown transients for ruptures of the largest RCS cold leg and hot leg piping in the AP600 design. This assumption maximizes the mass released, which is conservative for the subcompanment analysis. He initial conditions and inputs to the modified Zaloudek correlation are given in i Table 6.2.1.3-1. The short term LOCA mass and energy release data is provided in Tables b Revision: 5 [ Westillgh0tlSe 6.2-9 February 29,1996
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- 6. Engineered S:faty Features '
l 6.2.1.3-2 and 6.2.1.3-3. The short-term non-LOCA mass and energy release data are provided O' I in Table 6.2.1.34. 6.2.1.3.2 Long Term Mass and Energy Release Data A long term LOCA analysis calculational model is typically divided into fouc phases: I blowdown, which includes the period from the accident initiation (when the reactor is in a steady-state full power operation condition) to the time that the broken loop pressure equalizes to the containment pressure; refill, which is the time from the end of the blowdown to the time when the passive core cooling system (PXS) refills the vessel lower plenum; reflood, which begins when the water starts to flood the core and continues until the core is completely quenched; and post-reflood, which is the period after the core has been quenched and energy is released to the RCS primary system by the RCS metal, core decay heat, and the steam generators. l Re long-term analysis considers the blowdown, reflood, and post-reflood phases of the I transient. The refill period is conservatively neglected so that the releases to the containment are conservatively maximized. I he AP600 long-term LOCA mass and energy releases are predicted for the blowdown phase i for postulated DECLG and DEHLG breaks. The blowdown phase mass and energy releases are calculated using the NRC approved SATAN-VI computer code (Reference 4). The post I blowdown phase mass and energy releases are calculated considering the energy released from the available energy sources described below. The energy release rates are conservatively I modeled so that the energy is released quickly. The higher release rates result in a I conservative containment pressure calculation. De releases are provided in Tables 6.2.1.3-5 l and 6.2.1.3-6. 6.2.1.3.2.1 Energy Sources l De following energy sources are accounted for in the long-term LOCA mass and energy calculation:
- Decay heat
- Core stored energy RCS fluid and metal energy Steam Generator fluid and metal energy
- I
- Accumulators, core make-up tanks (CMTs), and the in-containment refueling water
! storage tank (IRWST) l
- Zirconium-water reaction l De methods and assumptions used to release the various energy sources during the blowdown i phase are given in Reference 4.
O Revision: 5 February 29,1996 6.2-10 3 WBStkighotise (
- 6. Engineered Safity Feat:res
,a
-( i s / The following items are used to conservatively analyze the energy release for maximum containment pressure: Maximum expected operating temperature Allowance in temperature for instmment error and dead band Margin in volume (+1.4 percent) Allowance in volume for thermal expansion (+1.6 percent) 100 percent full power operation Allowance for calorimetric error (+2.0 percent of full power)
- Conservatively modified coefficients of heat transfer Allowance in core stored energy for effect of fuel densification Margin in core stored energy (+15.0 percent)
Allowance in pressure for instrument error and dead band Margin in steam generator mass inventory (+10.0 percent) l
- One percent of the Zirconium surrounding the fuel is assumed to react 6.2.13.2.2 Description of Blowdown Model A description of the SATAN-VI model that is used to determine the mass and energy released from the RCS during the blowdown phase of a postulated LOCA is provided in Reference 4.
Significant correlations are discussed in this reference. ! ('i 6.2.13.23 Description cf Post Blowdown Model !V l The remaining RCS and SG mass and energy inventories at the end of blowdown are used to define the initial conditions for the beginning of the reflood portion of the transient. The broken and unbroken loop SG inventories are kept separate to account for potential differences in the cooldown rate between the loops. In addition, the mass added to the RCS from the IRWST is returned to containment as break flow so that no net change in system mass occurs. Energy addition due to decay heat is computed using the 1979 ANS standard (plus 2 sigma) decay heat table from Reference 4. The energy release rates from the RCS metal and steam generators are modelled using exponential decay rates. This modelling is consistent with analyses for current generation design analyses that are performed with the models described in Reference 4. l The accumulator, CMT, and IRWST mass flow rates are computed from the end of blowdown I to the time the tanks empty. The rate of RCS mass accumulation is assumed to decrease I exponentially during the reflood phase. More CMT and accumulator flow is spilled from the break as the system refills. The break flow rate is determined by subtracting the RCS mass addition rate from the sum of the accumulator, CMT and IRWST flow rates. l l Mass which is added to, and which remains in, the vessel is assumed to be raised to saturation. Therefore, the actual amount of energy available for release to the containment for a given time period is determined from the difference between the energy required to raise the temperature of the incoming flow to saturation and the sum of the decay heat, core stored C) Revision: 5 Y W65fingh00S8 6.2-11 February 29,1996
sm. -n.- l
- 6. Engineered Sity Features
.~ . - _
energy, RCS metal energy and SG mass and metal energy release rates. The energy release Ol rate for the available break flow is determined from a comparison of the total energy available I release rate and the energy release rate assuming that the break flow is 100-percent saturated steam. Saturated steam releases maximize the calculated containment pressurization. 6.2.13.2.4 Single Failure Analysis No single failure is assumed in the mass and energy release calculations. The safety injection system for the AP600 is passive, as opposed to active pumped safety injection systems for a conventional PWR. As a result, there is no single failure postulated for the mass and energy release analysis. The effects of a single failure are taken into account in the containment analysis of Subsection 6.2.1.1. 6.2.13.2.5 Metal-Water Reaction l Consistent with 10 CFR 50, Appendix K criteria, the energy release associated with the I zirconium-water exothermic reaction has been considered. The LOCA peak cladding I temperature analysis, presented in Chapter 15, that demonstrates compliance with the l Appendix K criteria demonstrat:s that no appreciable level of zirconium oxidation occurs. I his level of reaction has been bounded in the contaimnent mass and energy release analysis I by incorporating the heat of reaction from 1 percent of the zirconium smrounding the fuel. i His exceeds the level predicted by the LOCA analysis and results in additional conservatism I in the mass and energy release calculations. 6.2.13.2.6 Energy Inventories I Inventories of the amount of mass and energy released to containment during a postulated I LOCA are provided in summary Tables 6.2.13-2 through 6.2.13-6. 6.2.1.3.2.7 Additional Information Required for Confirmatory Analysis System parameters and hydraulic characteristics needed to perform confirmatory analysis are provided in Table 6.2.13-7 and Figures 6.2.13-1 through 6.2.13-4. I 6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary-System Pipe Rupture Inside Containment Steam line ruptures occurring inside a reactor containment structure may result in significant releases of high-energy fluid to the containment environment, possibly resulting in high containment temperatures and pressures. De quantitative nature of the releases following a steam line rupture is dependent upon the configuration of the plant steam system, the I containment design as well as the plant operating conditions and the size of the rupture. This section describes the methods used in determining the containment responses to a variety of postulated pipe breaks encompassing wide variations in plant operation, safety system performance, and break size. De spectrum of breaks analyzed is listed in Table 6.2.1.4-1. O Revision: 5 February 29,1996 6.2-12 T Westirighouse
1 i
- 6. Engineered Satty Features O
V 6.2.1.4.1 Significant Parameters Affecting Steam Line Break Mass and Energy Releases Four major factors influence the release of mass and energy following a steam line break: steam generator fluid inventory, primary-to-secondary heat transfer, protective system operation and the state of the secondary fluid blowdown. De following is a list of those plant variables which have significant influence on the mass and energy releases:
- Plant power level Main feedwater system design
- i Startup feedwater system design '
Postulated break type, size, and location Availability of offsite power
- Safety system failures Steam generator reverse heat transfer and reactor coolant system metal heat capacity.
He following is a discussion of each of these variables. 6.2.1.4.1.1 Plant Power Level Steam line breaks are postulated to occur with the plant in any operating condition ranging from hot shutdown to full power. Since steam generator mass decreases with increasing power level, breaks occurring at lower power generally result in a greater total mass release (3 to the plant containment. However, because ofincreased energy storage in the primary plant, d increased heat transfer in the steam generators and additional energy generation in the nuclear fuel, the energy released to the containment from breaks postulated to occur during power operation may be greater than for breaks occurring with the plant in a hot shutdown condition. Additionally, steam pressure and the dynamic conditions in the steam generators change with increasing power. They have significant influence on both the rate of blowdown and the amount of moisture entrained in the fluid leaving the break following a steam break event. Because of the opposing effects of changing power level on steam line break releases, no single power level can be identified as a worst case initial condition for a steam line break event. Therefore, several different power levels spanning the operating range as well as the hot shutdown condition are analyzed. 6.2.1.4.1.2 Main Feedwater System Design he rapid depressurization that occurs following a rupture may result in large amounts of water being added to the steam generators through the main feedwater system. Rapid closing isolation valves are provided in the main feedwater lines to limit this effect. De piping layout downstream of the isolation valves determine the volume in the feedwater lines that cannot be isolated from the steam generators. As the steam generator pressure decreases, some of the fluid in this volume will flash into the steam generator, providing additional secondary fluid that may exit out the mpture. t Q t
'J Revision: 5 3 Westingh00Se 6.2-13 February 29,1996 l
- f. Engineered S:f;ty Features The feedwater addition that occurs prior to closing of the feedwater line isolation valves e ;
influences the steam generator blowdown in several ways. First, the rapid addition increases the amount of entrained water in large-break cases by lowering the bulk quality of the steam generator inventory. Second, beenuse the water entering the steam generator is subcooled, it lowers the steam pressure, thereby reducing the flow rate out the break. Finally, the increased j flow rate causes an increase in the heat transfer rate from the primary-to-secondary system, j resulting in greater energy being released out the break. Since these are competing effects on the total mass and energy release, no worst case feedwater transient can be defined for all plant conditions. In the results presented, the worst effects of each variable have been used. For example, moisture entrainment for each break is calculated assuming conservatively small feedwater additions so that the entrained water is minimized or zero when support data are not available. Determination of total steam generator inventory, however, is based on l conservatively large feedwater additions, as explained in Subsection 6.2.1.43.2. 1 The unisolated feedwater line volumes between the steam generator and the isolation valves serve as a source for additional high-energy fluid to be discharged through the pipe break.
'Ihis volume is accounted for in the mass and energy release data presented in Subsection 6.2.1.43.2.
6.2.1.4.13 Startup Feedwater System Design Within the first minute following a steam line break, the startup feedwater system may be initiated on any one of several protection system signals. The addition of startup feedwater to the steam generators increases the secondary mass available for release to the containment, as well as the heat transferred to the secondary fluid. The effects on the steam generator mass are maximized in the calculation described in Subsection 6.2.1.43.2 by assuming full startup feedwater flow to the faulted steam generator starting at time zero from the safeguard system (s) signal or low steam generator level reactor trip and continuing until automatically terminated. 6.2.1.4.1.4 Postulated Break Type, Size and Location Postulated Break Type Two types of postulated pipe ruptures are considered in evaluating steam line breaks. First is a split rupture in which a hole opens at some point on the side of the steam pipe or steam header but does not result in a complete severance of the pipe. A single, distinct break area is fed uniformly by both steam generators until steam line isolation occurs. The blowdown flow rates from the individual steam generators are interdependent, since fluid coupling exists between the steam lines. Because flow limiting orifices are provided in each I steam generator, the largest split rupture can have an effective area prior, to isolation, that is no greater than the throat area of the flow restrictor times the number of steam generators. Following isolation, the effective break area for the steam generator with the broken line can be no greater than the flow restrictor throat area. O Revisicn: 5 February 29,1996 6.2-14 [ W65tiligh0US0 l l
- 6. E@.md Santy Features WR L G~
O The second break type is the double-ended guillotine rupture in which the steam pipe is completely severed and the ends of the break displace from each other. Guillotine ruptures are cinracterized by two distinct break locations, each of equal area, but being fed by different steam generators. The largest guillotine rupture can have an effective area per steam generator no greater than the throat area of one rteam line flow restrictor. Postulated Break Size Break area is alsoimportant when evaluating steam line breaks. It controls the rate of releases to the containment, and exerts significant influence on the steam pressure decay and the amount of entrained water in the blowdown flow. The data presented in this section include releases for three breaks at ea::h of four initial power levels. Included are two double-ended l~ ruptures and one split rupture, as follows: A full double-ended pipe rupture downstream of the steam line flow restrictor. For this case, the actual break area equals the cross-sectional area of the steam line, but the blowdown from the steam generator with the broken line is controlled by the flow restrictor throat area (1.388 square feet). The reverse flow from the intact steam generator is controlled by the smaller of the pipe cross section, the steam stop valve seat area, or the total flow restrictor throat area in the intact steam generator. The reverse flow has been conservatively assumed to be controlled by the flow restrictor in the intact loop steam generator. A small double-ended rupture having an area slightly larger than the area at which water entrainment ceases. Entraimnent is assumed in the forward direction only. Dry steam blowdown is assumed to occur in the reverse direction. I I In this case, the break areas analyzed for the small double-ended rupture are determined by two considerations. The first consideration is based on Reference (5) and the second consideration is based on the split break area that does not result in steam line isolation. I
'Ihe cases analyzed conservatively assume no water entrainment. 'Ihis means that the I
effluent is assumed to be dry saturated steam. Double <nded rupture areas larger and smaller than the split break area are presented. A split rupture representing the largest break which can neither generate a steam line isolation signal from the primary protection equipment nor result in moisture entrainment. Steam and feedwater line isolation signals are generated by high containment pressure signals for this type of break. Being a split rupture, the effective area seen by the faulted steam generator decreases by a factor of two, following steam line isolation. Moisture entrainment could occur at that time. However, since steam line isolation for these breaks generally does not occur before 20 to 60 seconds following such break, it is conservatively assumed that the pessure decreases sufficiently in the affected steam generator to preclude any moisture carryover. Table 6.2.1.4-1 lists the spectrum of secondary system pipe ruptures analyzed. O Revision: 5 3 WOElkighouse 6.2-15 February 29,1996
- 6. Engineered Saf;ty Feat res Postulated Break Location O
Break location affects steam line blowdown due to the pressure losses which occur in the length of piping between the steam generator and the break. De effect of the pressure loss is to reduce the effective break area seen by the steam generator. Although this reduces the rate of blowdown, it would not significantly change the total release of energy to the containment. Therefore, pipine loss effects are conservatively ignored in the blowdown I results. 6.2.1.4.1.5 Availability of Offsite Power The effects of the assumption of the availability of offsite power are enveloped in the analysis. Offsite power is assumed to be available where it maximizes the mass and energy released from the break because of the following: The continued operation of the reactor coolant pumps until automatically tripped as a result of core makeup tank (CMT) actuation. This maximizes the energy transfened from the reactor coolant system to the steam generator.
=
The continued operation of the feedwater pumps and actuation of the startup feedwater system until they are automatically terminated. This maximizes the steam generator inventories available for release. The AP600 is equipped with the passive safeguards system including the CMT and the i passive residual heat removal (PRHR) heat exchanger. Following a steam line mpture, these passive systems ar actuated when their setpoints are reached. This decreases the primary coolant temperatures. The actuation and operation of these passive safeguards systems do not require the availability of offsite power. l When the PRHR is in operation, the core-generated heat is dissipated to the in-containment refueling water storage tank (IRWST) via the PRHR heat exchanger. 'Ihis causes a reduction of the heat transfer from the primary system to the steam generator secondary system and causes a reduction of mass and energy releases via the break. Thus, the availability of ac power in conjunction with the passive safeguards system (CMT and PRHR) maximizes the mass and energy releases via the break. Therefore, blowdown occurring in conjunction with the availability of offsite power is more severe than cases where offsite power is not available. 6.2.1.4.1.6 Safety System Fnilures In addition to assuming a loss of system pressure, the following single active failures are considered: Failure of one main steam isolation valve Revision: 5 February 29,1996 6.2-16 3 Westinghouse
l
~-r l 6. Engineered Saf;ty Features o
Failure of one main feedwater isolation valve 6.2.1.4.1.7 Steam Generator Reverse Heat Transfer and Reactor Coolant System Metal Heat Capacity Once steam line isolation is complete, the steam generator in the intact steam loop becomes a source of energy that can be transferred to the steam generator with the broken line. His energy transfer occurs through the primary coolant. As the primary plant cools, the temperature of the coolant flowing in the steam generator tubes drops below the temperature of the secondary fluid in the intact unit, resulting in energy being returned to the primary coolant. This energy is then available to be transferred to the steam generator with the broken steam line. Similarly, the heat stored in the metal of the reactor coolant piping, the reactor vessel, and the reactor coolant pumps is transferred to the primary coolant as the plant cooldown progresses. This energy also is available to be transferred to the steam generator with the broken line. The effects of both the reactor coolant system metal and the reverse steam generator heat transfer are included in the results presented in this document. 6.2.1.4.2 Description of Blowdown Model (] y A description of the blowdown model used is provided in Reference 5. 6.2.1.4.3 Containment Response Analysis I ne.EGOTHIC Computer Code (Reference 1)is used to determine the containment responses following the steam line break. He containment response analysis is described in Subsection 6.2.1.1. 6.2.1.4.3.1 Initial Conditions l The initial containment conditions are di! cussed in Subsection 6.2.1.1. 6.2.1.4.3.2 Mass and Energy Release Data Using References 5 and 6 as a basis, masi and energy release data are developed to determine the containment pressure-temperature response for the spectrum of breaks analyzed. l Tables 6.2.1.4-2 and 6.2.1.4-3 provide the mass and energy release data for the cases that produce the highest containment pressure and temperature in the containment response analysis. Table 6.2.1.4-4 provides plant data used for the cases used in the mass and energy releases determination. De rate of startup feedwater addition represents the maximum runout flow rate to a fully depressurized steam generator. Actual isolation is dependent on signals generated by the l l protection and safety monitoring system. Feedwater isolation for the split breaks was based l Revision: 5 [ W85tingt100S8 6.2-17 February 29,1996
mii
- 6. Engine: red S:1;ty Frtures on the time required to reach the containment pressure setpoint that generates the isolation O
signal. The feedwater flow rates before automatic isolation assumed in the analyses are based I on input for the AP600 steam generator and main feed system design. 6.2.1.4.33 Containment Pressure-Temperature Results he results of the containment pressure-temperature analyses for the postulated secondary system pipe ruptures that produce the highest peak containment pressure and temperature are presented in Subsection 6.2.1.1. 6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies of Emergency Core Cooling System (PWR) The containment backpressure used for the AP600 cold leg guillotine and split breaks for the i emergency core cooling system (ECCS) analysis presented in Subsection 15.6.5 is described I herein. He minimum containment beckpressure for emergency core cooling system I performance during a loss-of-coolant accident has been computed using the WGOTHIC i computer code. Subsection 6.2.1.1 demonstrates that the AP600 containment pressurizes I significantly during large break LOCA events, and an analysis was performed to establish a containment pressure boundary condition applied to the WCOBRAfrRAC code (Reference I 10). A single-node containment model was used to assess containment pressure response. I Containment internal heat sinks used heat transfer correlations of 4 times Tagami during the l blowdown phase followed by 1.2 times Uchida. This analysis was performed for the first l 200 seconds of the accident. The calculated containment backpressure is provided in i Figure 6.2.1.5-1. Results of the WCOBRAfrRAC analyses demonstrate that the AP600 meets l 10 CFR 50.46 requirements. 6.2.1.5.1 Mass and Energy Release Data l The mass and energy releases to the containment during the blowdown portion only of the I double-ended cold-leg guillotine break (DECLG) transient are presented in Table 6.2.1.5-1, as computed by the WCOBRAfrRAC code. The mathematical models which calculate the mass and energy releases to the containment are described in Subsection 15.6.5. A break spectrum analysis is performed (see references in l Subsection 15.6.5) that considers various break sizes and Moody discharge coefficients for the i double-ended cold leg guillotines and splits. Mixing of steam and accumulator water injected I into the vessel reduces the available energy released to the contaimnent vapor space, thereby I minimizing calculated containment pressure. Note that the mass / energy releases during the I reflood phase of the subject break are not considered. This produces a conservatively low I contaminant pressure result for use as a boundary condition in the W COBRAfrRAC large i break LOCA analysis. O Revision: 5 February 29,1996 6.2-18 3 W65tkigt10llSe
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- 6. Engineered Safity Features KAE --
O 6.2.1.S.2 Initial Containment Internal Conditians I Initial containment conditions were biased for the ECCS backpressure analysis to predict a j l I conservatively low containment backpressure. Initial containment conditions included initial l pressure of 14.7 psia, initial temperature of 90 F, and a relative humidity of 99 percent. An i air annulus temperature of O'F was assumed and a linear temperature profile between 0 F and l 90*F was used in the containment shell, which separates the annulus from the containment i l volume. 1 6.2.1.5.3 Other Parameters i l Containment parameters, such as containment volume and passive heat sinks, were biased to ! I predict a conservative low containment backpressure. He containment volume used in the I { l calculations was conservatively set to 1.05 times the cold volume. Passive heat sink surface I I areas were approximately doubled from the heat sinks, presented in Tables 6.2.1.1-5 and I 6.2.1.1-6. Material propenies were biased high (density, conductivity, and heat capacity) as l l indicated in CSB 6-1 (Reference 10). To further minimize containment pressure, containment l l purge was assumed to be in operation at time zero, Two ten-inch diameter flo' paths were I provided in the containment model to simulate containment purge lines. Valves within these i lines were closed 5 seconds after 8 psig was reached. . l 6.2.1.7 Testing and Inspection l I (q) This section describes the functional testing of the containment vessel. Testing and in-service inspection of the containment vessel are described in Subsecdon 3.8.2.6. Isolation testing is described in Subsection 6.2.3. Leak testing is described in Subsection 6.2.5. Testing and inspection are consistent with regulatory requirements and guidelines. i i ne valves of the passive containmer,t cooling system are stroke tested periodically. ! l Subsection 6.2.2 provides a description of testing and inspection. ' The baffle between the containment vessel and the shield building is equipped with removable panels and clear observation panels to allow for inspection of the containment surface. See Subsection 3.8.2 for the requirements for in-service inspection of the steel containment vessel. l Subsection 6.2.2 provides a description of testing to be performed. Testing is not required on any subcompartment vent or on the collection of condensation from the containment shell. He collection of condensate from the containment shell and its use in leakage detection are discussed in Subsection 5.2.5. 6.2.1.8 Instrumentation Requirements Instrumentation is provided to monitor the conditions inside the containment and to actuate the appropriate engineered safety features, should those conditions exceed the predetermined levels. De instruments measure the containment pressure, containment atmosphere 1 l ii/D V Revision: 5 3 W8Silfigh00Se 6.2 19 February 29,1996 1
- 6. Engineered S:fety Features radioactivity, and containment hydrogen concentration. Instrumentation to monitor reactor O
coalant system leakage into containment is described in Subsection 5.2.5. De containment pressure is measured by four independent pressure transmitters. The signals are fed into the engineered safety features actuation system, as described in Subsection 7.3.1. Upon detection of high pressure inside the containment, the appropriate safety actuation signals are generated to actuate the necessary safety-related systems. Low pressure is alarmed but does not actuate the safety-related systems. De physically separated pressure transmitters are located outside the containment and connected to their sensors by filled and sealed hydraulic lines. Section 7.3 provides a description. The containment atmosphere radiation level is mor9tored by four independent area monitors located above the operating deck inside the containment building. The measurements are continuously fed into the engineered safety features actuation system logic. Section 11.5 provides information on the containment area radiation monitors. De engineered safety features actuation system operation is described in Section 7.3. The containment hydrogen concentration is measured by hydrogen monitors, as described in Subsection 6.2.4. Hydmgen concentrations are monitored by sensors distributed throughout containment to provide a representative indication of contaimnent hydrogen concentrations. I ne sensors also indicate the specific areas evaluated for potential hydrogen pockets. These O indications are used by the plant operators to control ignitors and monitor hydrogen I concentrations. High hydrogen concentration is alarmed in the main control room. Section 7.3 provides detailed information on the engineered safety features actuation system operation. 6.2.2 Passive Containment Cooling System The passive contamment cooling system (PCS) is an engineered safety features system. Its functional objective is to reduce the containment temperature and pressure following a loss of coolant accident (LOCA) or main steam line break (MSLB) accident inside the containment by removing thermal energy from the containment atmosphere. The passive containment cooling system also serves as the means of transferring heat to the safety-related ultimate heat sink for other events resulting in a significant increase in containment pressure and temperature. Finally, the passive containment cooling system limits releases of radioactivity (post-accident) by reducing the pressure differential between the containment atmosphere and the external environment, thereby diminishing the driving force for leakage of fission products from the containment to the atmosphere. This subsection describes the safety design bases of the safety-related containment cooling function. Nonsafety-related containment cooling, a function of the containment recirculation cooling system,is described in Subsection 9.4.6. O Revision: 5 February 29,1996 6.2-20 [ W65tiligh0USB
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- 6. Engineered Saf;ty Feat:res (nO) 1 l
6.2.2.1 Safety Design Basis '
=
he passive containment cooling system is designed to withstand the effects of natural i phenomena such as ambient temperature extremes, earthquakes, winds, tornadoes, or ! floods. Passive containment cooling system operation is automatically initiated upon receipt of I a Hi-2 containment pressure signal. 1 The passive containment cooling system is designed so that a single failure of an active component, assuming loss of offsite or onsite ac power sources, will not impair the capability of the system to perform its safety-related function. Active components of the passive containment cooling system are capable of being tested l during plant operation. Provisions are made for inspection of major components in accordance with the intervals specified in the AShE Code, Section XI. The passive containment cooling system components required to mitigate the consequences of an accident are designed to remain functional in the accident environment and to withstand the dynamic effects of the accident. I The passive containment cooling system is capable of removing sufficient thermal energy (] V including subsequent decay heat from the containment atmosphere following a design basis event resulting in containment pressurization such that the containment pressure remains below the design value with no operator action required for three days. The passive containment cooling system is designed to reduce containment pressure to less than one-halfits design pressure within 24 hours following a postulated loss of coolant accident. The passive containment cooling system is designed and fabricated to appropriate codes consistent with Regulatory Guides 1.26 and 1.32 and seismic Category I in accordance with Regulatory Guide 1.29 as described in Section 1.9. 6.2.2.2 System Design 6.2.2.2.1 General Description The passive contamment cooling system and components are designed to the codes and standards identified in Section 3.2; flood design is described in Section 3.4; missile protection is described in Section 3.5. Protection against dynamic effects associated with the postulated rupture of piping is described in Section 3.6. Seismic and environmental design and l equipment qualification are described in Sections 3.10 and 3.11. The actuation system is described in Section 7.3. {} V Revision: 5 [ WOStifigh00S8 6.2-21 February 29,1996 L
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- 6. Enginetred Sity Fe:tur s n-l 6.2.2.2.2 System Description 9 I he passive containment cooling system is a safety-related system which is capable of transferring heat directly from the steel containment vessel to the environment. This transfer ;
I of heat prevents the containment from exceeding the design pressure and temperature i l following a postulated design basis accident, as identified in Chapters 6 and 15. Containment pressure is further reduced to one-half the design pressure within 24 hours following the worst postulated loss of coolant accident. The passive containment cooling system makes use of the steel containment vessel and the concrete shield building surrounding the containment. The major components of the passive containment cooling system are: the passive containment cooling water storage tank (PCCWST) which is incorporated into the shield building structure above the containment; an air baffle, located between the steel containment vessel and the concrete shield building, which defines the cooling air flowpath; an air inlet and an air exhaust, also incorporated into the shield building stmeture; and a water distribution system, mounted on the outside surface of the steel containment vessel, which functions to distribute water flow on the containment. A recirculation path is provided to control the passive containment cooling water storage tank water chemistry and to provide heating for freeze protection. Passive containment cooling water storage tank filling operations and normal makeup needs are provided by the demineralized water transfer and storage system discussed in Subsection 9.2.4. De system piping and instrumentation diagram is shown in Figure 6.2.2-1. System parameters are shown in Table 6.2.2-1. A simplified system sketch is included as i Figure 6.2.2-2. 6.2.2.2.3 Component Description The mechanical components of the passive containment cooling system are described in this subsection. Table 6.2.2-2 provides the component design parameters. Passive Containment Cooling System Water Storage Tank - De passive containment cooling system water storage tank is incorporated into the shield building structure above the containment vessel. The inside wetted walls of the tank are lined with stainless steel plate. It is filled with demineralized water and has a useable volume of 400,000 gallons for passive containment cooling functions. The passive containment cooling system functions as the safety-related ultimate heat sink. The passive containment cooling water storage tank is seismically designed and missile protected. The tank also has redundant level measurement channels and alarms for monitoring the tank I water level and redundant temperature measurement channels to monitor and alarm for l l potential freezing. To maintain system operability, a recirculation loop that provides l chemistry and temperature control is connected to the tank. " J Re tank is constructed to provide sufficient thermal inertia and insulation such that draindown can be accomplished over a 72 hour period without heater operation. Revision: 5 February 29,1996 6.2-22 3 Westklgt10USS
l l l l
- 6. Engineered Saf;ty Features l
i [N Iv) ; In addition to its containment heat removal function, the passive containment cooling system ! l water storage tank also serves as a seismic Category I water storage reservoir for fire I protection following a safe shutdown earthquake. I De PCCWST suction pipe for the fire protection system (FPS) is configured so that actuation l l of the FPS will not hifringe on the 400,000 gallons volume allocated to the passive I containment cooling function. Herefore, the fire protection system suction pipe cannot reduce the water storage tank water volume below 400,000 gallons. Passive Containnw9t Cooling System Water Storage Tank Isolation Valves - The passive containment cool.ag system water storage tank outlet piping is equipped with two sets of redundant isolation valves. The air-operated butterfly valves are normally closed and open ' I upon receipt of a Hi-2 containment pressure signal. Dese valves fail-open provide a fail-safe position on the loss of air and/or loss of power. The normally-open motor-operated gate valves are located upstream of the butterfly valves. Hey are provided to allow for testing of the butterfly valves. he storage tank isolation valves, along with the passive contaimrent cooling water storage tank discharge piping and associated instrumentation between the passive containment cooling water storage tank and the downstream side of the isolation valves, are contained within a temperature-controlled valve room to prevent freezing. Valve room heating is provided by a locally installed electric unit heater to maintain the room temperature above 50*F. O () Flow Control Orifices - Orifices are installed in each of the three passive containment cooling system water storage tank outlet pipes. They are used, along with the different elevations of the outlet pipes, to control the flow of water from the passive containment cooling system water storage tank as a function of water level. De orifices are located within the temperature-controlled valve room. Water Distribution Bucket - A water distribution bucket is provided to uniformly deliver water to the outer surface of the containment dome. The redundant passive containment cooling water delivery pipes and auxiliary water source piping discharge into the bucket, below its operational water level, to prevent excessive splashing. A set of circumferentially-spaced distribution slots are included around the top of the bucket. The bucket is hung from the shield building roof and suspended just above the containment dome for optimum water delivery. Water Distribution Weir System - A weir-type water delivery system is provided to uniformly wet the containment shell during passive containment cooling system operation. He system includes channeling walls and collection troughs, equipped with distribution weirs. The distribution system is capable of functioning during extreme low- or high-ambient temperature conditions. Air Flow Path - An air flow path is provided to direct air along the outside of the containment shell to provide containment cooling. De air flow path includes a screened n shield building inlet, an air baffle that divides the outer and inner flow annuli, and a chimney (v) Revision: 5 T Westirigh0tlSB 6.2-23 February 29,1996
- 6. Engineered Sirety Featrres t
to increase buoyancy. Subsection 3.8.4.1.3 includes information regarding the air baffle. He O general arrangement drawings provided in Section 1.2 provide layout information of the air flow path. Chemical Addition Tank - The chemical addition tank is a small, vertical, cylindrical tank that is sized to inject a 30-percent-by-volume solution of hydrogen peroxide to maintain a passive containment cooling water storage tank concentration of 50 ppm for control of algae growth. Recirculation Pump - The recirculation pump is a 100 percent capacity centrifugal pump with wetted components made of austenitic stainless steel. De pump is sized to recirculate the entire volume of tank water once every week. Recirculation Heater - The recirculation heater is provided for freeze protection. He heater is sized based on heat losses from the passive containment cooling system water storage tank and recirculation piping at the minimum site temperature, as defined it. Section 2.3. 6.2.2.2.4 System Operation Operation of the passive containment cooling system is initiated upon receipt of two out of I four Hi-2 containment pressure signals. Manual actuation by the operator is also possible from either the main control room or remote shutdown workstation. System actuation consists of opening the passive containment cooling system water storage tank isolation valves. This allows the passive containment cooling system water storage tank water to be delivered to the top, external surface of the steel containment shell. The flow of water, provided entirely by the force of gravity, forms a water film over the dome and side walls of the containment structure. The flow of water to the containment outer surface is initially established at approximately 220 gpm for short-tenn containment cooling following a design basis loss of coolant accident. He flow rate is gradually reduced over a period of 72 hours to a value of approximately l 55 gpm. His flow provides the desired reduction in containment pressure over time and removes decay heat. The flow rate change is dependent only upon the decreasing water level in the passive containment cooling water storage tank. l To adequately wet the containment surface, the water is delivered to the distribution bucket above the center of the contamment dome which subsequently delivers the water to the containment surface. A weir-type water distribution system is used on the dome surface to j distribute the water for effective wetting of the dome and vertical sides of the containment shell. I The weir system contains radial arms and welts specifically located considering the effects of tolerances of the containment vessel design and construction. In addition, a corrosion-resistant paint or coating for the containment vessel is specified to enhance surface wetability, and film L formation. O' Revision: 5 l February 29,1996 6.2-24 [ Westingh0USB
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' \~J Re cooling water not evaporated from the vessel wall flows down to the bottom of the inner l containment annulus into floor drains. The redundant floor drains route the excess water to storm drains. The drain lines are always open (without isolation valves) and each is sized to accept maximum passive containment cooling system flow. The interface with the storm drain system is an open connection such that any blockage in the storm drains would result in the annulus drains overflowing the connection. A path for the natural circulation of air upward along the outside walls of the containment structure is always open. The natural circulation air flow path begins at the shield building inlet, where atmospheric air enters horizontally through openings in the concrete structure. Air flows past a set of fixed louvers and is forced to turn 90 degrees downward into an outer annulus. This outer shield building annulus is encompassed by the concrete shield building on the outside and a removable baffle on tne inside. At the bottom of the baffle wall, curved vanes aid in turning the flow upward 180 degrees into the inner containment annulus. This inner annulus is encompassed by the baffle wall on the outside and the steel containment vessel on the inside. Air flows up through the inner annulus to the top of the containment vessel and then exhausts through the shield building chimney. As the containment structure heats up in response to high containment temperature, heat is removed from within the containment via conduction through the steel containment structure, convection from the containment su1 face to the water film, convection and evaporation from the water film to the air, and radiation from the water film to the air baffle. As heat and (] water vapor are transferred to the air space between the containment structure and air baffle, C/ the air becomes less dense than the air in the outer annulus. This density difference causes an increase in the natural circulation of the air upward between the containment structure and the air baffle, with the air finally exiting at the top center of the shield building. The passive containment cooling system water storage tank provides water for containment wetting for 72 hours following system actuation. Operator action can be taken to replenish this water supply or to provide an alternate water source directly to the containment shell through installed safety-related and seismic piping connections. In addition, water sources used for normal filling operations can be used to replenish the water supply. The arrangement of the air inlet and air exhaust in the shield building structure has been elected so that wind effects aids the natural air circulation. The air inlets are placed at the top, outside of the shield building, providing a symmetrical air inlet that reduces the effect of wind speed and direction or adjacent structures. The air / water vapor exhaust structure is elevated above the air inlet to provide additional buoyancy and reduces the potential of exhaust air being drawn into the air inlet. The air flow inlet and chimney regions are both designed to protect against ice or snow buildup and to prevent foreign objects from entering the air flow path. Inadvertent actuation of the passive containment cooling system is terminated through operator action by closing either of the series isWtion valves from the main control room. I Subsection 6.2.1.1.4 provides a discussion >
.ne effects of inadvertent system actuation.
p V Revision: 5 [ W65tiftgfl00S8 6.2-25 February 29,1996
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- 6. Engineered S1ty Features ram-( !.2.3 Safety Evaluation O
- The safety-related portions of the passive containment cooling system are located within the shield building structure. 'Ihis building (including the safety-related portions of the passive containment cooling system) is designed to withstand the effects of natural phenomena such l as earthquakes, winds, tornadoes, or floods. Components of the passive containment cooling system are designed to withstand the effects of ambient temperature extremes.
1 Operation of the containment cooling system is initiated automatically following the receipt I of a Hi-2 containment pressure signal. De use of this signal provides for system acmation during transients, resulting in mass and energy releases to containment, while avoiding i unnecessary actuations. No other actuations are required to initiate the post-accident heat I removal function since the cooling air flow path is always open. Operation of the passive contamment cooling system may also be initiated from the main control room and from the remote shutdown work station. A description of the actuation system is contained in l Section 7.3. l Re active components of the passive containment cooling system, the isolation valves, are located in two redundant pipe lines. Failure of a component in one train does not affect the l operability of the other mechanical train or the overall system performance. The fail-open, air-operated valves require no power to move to their safe (open) position. De normally ( j open motor-operated valves are powered from separate redundant Class IE dc (battery) power I sources. Table 6.2.2-3 presents a failure modes and effects analysis of the passive containment cooling system. l' l l Capability is provided to periodically test actuation of the passive containment cooling system. Active components can be tested periodically during plant operation to verify operability. The system can be inspected during unit shutdown. Additional information is contained in Subsections 3.9.6 and 6.2.2.4, as well as in the Technical Specifications. ! He passive containment cooling system components located inside containment, the contamment pressure sensors, are tested and demonstrated to perform in a simulated design I basis accident environment. These components are protected from effects of postulated jet impingement and pipe whip in case of a high-energy line break. The containment pressure analyses demonstrate that the passive containment cooling system is capable of removing sufficient heat energy including subsequent decay heat from the containment atmosphere so that the peak pressure following the worst postulated loss of coolant accident is below the containment design pressure with no operator action for at least three days. Analyses also show that the containment pressure is reduced to below one-half of the design pressure within 24 hours following the most limiting design basis loss of coolant accident. I ne containment pressure analyses are based on an ambient air temperature of 115 F dry bulb and 80*F coincident wet bulb. De passive containment cooling system water storage tank water temperature basis is 120*F. Results of the analyses are provided in Subsection 6.2.1. Revision: 5 February 29,1996 6.2-26 T Westilighouse
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m-O 6.2.2.4 Testing and Inspection 6.2.2.4.1 Inspections
'Ihe passive contranment cooling system is designed to permit periodic testing of system readiness as specified in the Technical Specifications.
'Ihe portions of the passive containment cooling system from the isolation valves to the passive containment cooling system water storage tank are accessible and can be inspected during power operation or shutdown for leaktightness. Examination and inspection of the pressure retaining piping welds is performed in accordance with ASME Code, Section XI.
The design of the containment vessel and air baffle facilitates the inspection of the vessel dering plant shutdowns. 6.22.4.2 Preoperational Testing Preoperation testing for the passive containment cooling system is addressed in Sections 14.2.8.1.96 and 14.2.8.1.97 6.2.2.4.3 Operational Testing Operational testing is performed to:
) l
- Demonstrate that the sequencing of valves occurs on the initiation of Hi-2 containment pressure and demonstrate the proper operation of remotely operated valves.
Verify valve operation during plant operation. 'Ihe normally open motor-operated valves, in series with each normally closed air-operated isolation valve, are temporarily closed. This closing permits isolation valve stroke testing without actuation of the passive containment cooling sys'em. Verify water flow delivery, consistent with the accident analysis. Verify visually that the path for containment cooling air flow is not obstructed by debris or foreign objects. l
- Test frequency is consistent with the plant technical specifications (Section 16.3.6) and I inservice testing program (Section 3.9.6).
6.2.2.5 Instrumentation Requirements
'Ihe status of the passive containment cooling system is displayed in the main control room.
I The operator is alerted to problems with the operation of the equipment within this system during both normal and post-accident conditions. qO Revision: 5 T Westkighouse 6.2-27 February 29,1996
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Normal operation of the passive containment cooling system is demonstrated by monitoring 01 1
)
the recircu'ation pump discharge pressure, flow rate, passive containment cooling system water l storage tank level and temperature, and valve room temperature. Post-accident operation of the passive containment cooling system is demonstrated by monitoring the passive l containment cooling system water storage tank level, passive containment cooling system I cooling water flow rate, containment pressure and external cooling air discharge temperature. De activation signal-generating equipment fully meets IEEE Standard 279 guidelines for considerations such as operation, diversity, and separation of power supplies. Details are found in Chapter 7. The protection system providing system actuation is discussed in Chapter 7. 6.23 Containment Isolation System The major function of the containment isolation system of the AP600 is to provide contamment isolation to allow the normal or emergency passage of fluids tbtough the containment boundary while preserving the integrity of the containment boundary,if required. This prevents or limits the escape of fis,fon products that may result from postulated accidents. Containment isolation provisions are designed so that fluid lines which penetrate the primary containment boundary are isolated in the event of an accident. This minimizes the release of radioactivity to the environment. The containment isolation system consists of the piping, valves, and actuators that isolate the O containment. The design of the containment isolation system satisfies the requirements of NUREG 0737, as described in the following paragraphs. 6.23.1 Design Basis 6.23.1.1 Safety Design Basis A. The containment isolation system is orotected from the effects of natural phenomena, such as earthquake's, tornadoes, hurricanes, floods, and external missiles (General Design Criterion 2). B. The containment isolation system is designed to remain functional after a safe shutdown earthquake (SSE) and to perform its intended function following the postulated hazards of fire, internal missiles, or pipe breaks (General Design Criteria 3 and 4). C. The containment isolation system is designed and fabricated to codes consistent with the quality group classification, described in Section 3.2, assigned by Regulatory Guide 1.26 and the seismic category assigned by Regulatory Guide 1.29. The power supply and control functions are in accordance with Regulatory Guide 132. D. The containment isolation system provides isolation of lines penetrating the containment for design basis events requiring contahment integrity. Revision: 5 February 29,1996 6.2-28 y Westingh0use
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- 6. Engineered Saf;ty Feat:res
!O !V ! E. Upon failure of a main steam line, the containment isolation system isolates the steam generators as required to prevent excessive cooldown of the reactor coolant system or overpressurization of the contairunent. 1 l F. The containment isolation system is designed in accordance with General Design Criterion 54. i l G. l l Each line that penetrates the containment that is either a part of the reactor coolar.t l pressure boundary or that connects directly to the containment atmosphere, and does not i ' meet the requirements for a closed system (as defined in paragraph I below) except instrument sensing lines, is provided with containment isolation valves according to ) General Design Criteria 55 and 56. l H. Each line that penetrates the containment, that is neither part of the reactor coolant pressure boundary nor connected directly to the atmosphere of the containment, and that satisfies the requirements of a closed system is provided with a containment isolation
- valve according to General Design Criterion 57. A closed system is not a part of the reactor coolant pressure boundary and is not connected directly to the atmosphere of the containment. A closed system also meets the following additional requirements
Re system is protected against missiles and the effects of high-energy line break. l(]
- The system is designed to Seismic Category I requirements.
U The system is designed to ASME Code, Section Ill, Class 2 requirements. The system is designed to withstand temperatures at least equal to the containment design temperature. The system is designed to withstand the external pressure from the containment structural acceptance test. The system is designed to withstand the design basis accident transient and environment. I. Instrument lines penetrating the containment are provided with isolation valves according to General Design Criteria 55,56, and 57. Four containment pressure sensors are provided as sealed systems with bellow seals inside the containment, liquid filled capillaries between the seals, and the sensing element outside containment. 'Ihese instrument lines are closed systerns both inside and outside containment, are designed to
- withstand the containment pressure and temperature conditions following a loss of coolant accident, and are designed to withstand dynamic effects.
I J. The containment isolation system is designed so that no single failure in the containment isolation system prevents the system from performing its intended functions. n 1 ) . L./ Revision: 5 i W W85tiligh00S6 6.2-29 February 29. IW6
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- 6. Engineered Saf;ty Features K. Fluid penetrations supporting the engineered safety features functions have remote 0l manual isolation valves. These valves can be closed from the main control room or from the remote shutdown workstation,if required. ,
L. The containment isolation system is designed according to 10 CFR 50.34, so that the resetting of an isolation signal will not cause any valve to change position. ! 6.23.1.2 Power Generation Design Basis ne containment isolation system has no power generation design basis. Power generation design bases associated with individual components of the containment isolation system are discussed in the section describing the system of which they are an integral part. 6.23.1.3 Additional Requirements he AP600 containment isolation system is designed to meet the following additional requirements: A. The containment isolation elements are designed to minimize the number of isolation valves which are subject to Type C tests of 10 CFR 50, Appendix J. Specific requirements are the following:
. The number of pipe lines which provide a direct connection between the inside and outside of primary containment during normal operation are minimized.
Closed systems outside of containment that may be open to the containment atmosphere during an accident are designed for the same conditions as the containment itself, and are testable during Type A leak tests. ne total number of penetrations requiring isolation valves are minimized by appropriate system design. For example: In the component cooling system, a single header with branch lines inside of containment is employed instead of providing a separate penetration for each branch line. Consistent with other considerations, such as containment arrangement and exposure of essential safety equipment to potentially harsh environments, the equipment is located inside and outside of containment so as to require the smallest number of penetrations. Consistent with current practice, Type C testing is not required for pressurized water reactor main steam, feedwater, startup feedwater, or steam generator blowdown isolation valves. De steam generator tubes are considered to be a suitable boundary to prevent release of radioactivity from the reactor coolant system following an accident. De steam generator shell and pipe lines, up to and Revision: 5 February 29,1996 6.2-30 W Westinghouse
1 ! 6. Engineeral S:fity Featzres i l
/. m l 1
, C/ ; including the first isolation valve, are considered a suitable boundary to prevent l release of containment mdioactivity. I l l B. Personnel hatches, equipment hatches, and the fuel transfer tube are sealed by closures
- with double gaskets. {
C. Containment isolation is actuated on a two-out-of-four logic from, high-containment pressure signal, low-steamline pressure, and low T Provisions are provided for manual containment isolation from the main control room. l l D. Penetration lines with automatic isolation valves are isolated by eagineered safety l features actuation signals. I E. Isolation valves are designed to provide leaktight service only against the medium to l which the valves are exposed in the s'iort and long-term course of any accident. For I example, a valve is gas-tight if the valve is exposed to the containment atmosphere. F. Isolation valves are designed to have the capacity to close against the conditions that may exist during events requiring containment isolation. G. Isolation valve closure times are designed to lirnit the release of radioactivity to within regulation and are consistent with standard valve operators, except where a shorter
]Q/
closure time is required. H. Deleted. I. 'Ihe position of each power-operated isolation valve (fully closed or open), whether automatic or remote manual, is indicated in the main control room and is provided as input to the plant computer. Such position indication is based on actual valve position, 4 for example, by a limit switch which directly senses the actual valve stem position, rather than demanded valve position. J. Normally closed manual containment isolation valves have provisions for locking the valves closed. Locking devices are designed such that the valves can be locked only in the fully closed position. Administrative control provides verification that manual isolation valves are maintained locked closed during normal operation. Position locks provide confldence that valves are placed in the correct position prior to locking. K. Automatic contamment isolation valves are powered by Class 1E de power. Non-motor-operated valves fail in the closed position upon loss of a support system, such as instrument air or electric power. I L. Valve alignments used for fluid system testing during operation are designed so that either: containment bypass does not occur during testing, assuming a single failure; or exceptions are identified, and remotely operated valves provide timely isolation from the g control room. Containment isolation provisions can be relaxed during system testing. b Revision: 5 T Westinghouse 6.2-31 February 29,1996
- 6. Engineered Screty Featcres I
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The intent of the design is to provide confidence that operators are aware of any such O: condition and have the capability to restore containment integrity. 6.2.3.2 System Description 6.2.3.2.1 General Description Piping systems penetrating the containment have containment isolation features. These features serve to minimize the release of fission products following a design basis accident. SRP Section 6.2.4 and Regulatory Guide 1.141 provide acceptable alternative arrangements to the explicit arrangements given in General Design Criteria 55,56 and 57. Table 6.2.3-1 lists each penetration and provides a summary of the containment isolation characteristics. The Piping and Instrumentation Diagrams of the applicable systems show the functional arrangement of the containment penetration, isolation valves, test and drain connections. Section 1.7 contains a list of the Piping and Instrumentraie Diagrams. As discussed in Subsection 6.2.3.1, the AP600 containment isoir. tion design satisfies the NRC requirements including post-Three Mile Island requirements. Two barriers an provided -- one inside containment and one outside contaimnent. Usuelly these barriers ne valves, but in some cases they are closed piping systems not connected to the reactor coolant system or to the containment atmosphere. The AP600 has fewer mechanical containment penetrations (including hatches) and a higher percentage of normally closed isolation valves than current plants. 'Ihe majority of the penetrations that are normally open incorporate fail closed isolation valves that close automatically with the balance of the penetrations. Table 6.2.3-1 lists the AP600 containment mechanical penetrations and the isolation valves associated with them. Provisions for leak testing are discussed in Subsection 6.2.5. For those systems having automatic isolation valves or for those provided with remote-manual isolation, Subsection 6.2.3.5 describes the power supply and associated actuation system. I Power-operated (alt, motor, or pneumatic) containment isolation valves have position indication in the main control room. Two modes of valve actuation are considered in Table 6.2.3-1. The actuation signal that occurs directly as a result of the event initiating containment isolation is designated as the primary actuation signal. If a change in valve position is required at any time following primary actuation, a secondary actuation signal is generated which places the valve in an alternative position. The closure times for automatic containment isolation valves are provided in Table 6.2.3-1. The containment air filtration system is used to purge the containment atmosphere of airborne radioactivity during normal plant operation, as described in Subsection 9.4.7. 'Ihe system is I designed according to Branch Technical Position CSB 6-4 using 18-inch supply and exhaust lines and containment isolation valves. These valves close automatically on a containment isolation signal. ftevision: 5 February 29,1996 6.2-32 W Westinghouss ) 1
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- 6. Engineered S:f;ty Feat:res a
H-ii l H-(3 l Section 3.6 describes dynamic effects of pipe rupture. Section 3.5 discusses missile i protection, and Section 3.8 discusses the internal structures including any structure used as a protective device. Lines associated with those penetrations that are considered closed systems l inside the containment are protected from the effects of a pipe rupture and missiles. The actuators for power-operated isolation valves inside the containment are either located above l the maximum containment water level or in a normally nonflooded area. 'Ihe actuators are designed for flooded operation or are not required to function following containment isolation and designed and qualified not to spuriously open in a flooded condition. l Other defined bases for containment isolation are provided in SRP Section 6.2.4 and l Regulatory Guide 1.141. Conformance with Regulatory Guide 1.141 is provided to the extent specified in Section 1.9.1. 6.23.2.2 Component Description Codes and standards applicable to the piping and valves associated with containment isolation ' are those for Class B components, as discussed in Section 3.2. Containment penetrations are classified as Quality Group B and Seismic Category I. Section 3.11 provides the normal, abnormal, and post-loss-of-coolant accident environment that is used to qualify the operability of power-operated isolation valves located inside the containment. >n id The containment penetrations which are part of the main steam system and the feedwater system are designed to meet the stress requirements of NRC Branch Technical Position i MEB 3-1, and the classification and inspection requirements of NRC Branch Technical Position ASB 3-1, as described in Section 3.6. Section 3.8 discusses the interface between the piping system and the steel containment. i As discussed in Subsection 6.2.3.5, the instrumentation and control system provides the signals which determine when containment isolation is required. Containment penetrations are either normally closed prior to the isolation signal or the valves automatically close upon receipt of the appropriate engineered safety features actuation signal. 6.23.2.3 System Operation During normal system operation, approximately 25 percent of the penetrations are not isolated. These lines are automatically isolated upon receipt of isolation signals, as described in Subsections 6.2.33 and 6.2.3.4 and Chapter 7. Lines not in use during power operation are normally closed and remain closed under administrative control during reactor operation. l 6.233 Design Evaluation f A. Engineered safeguards and containment isolation signals automatically isolate process lines which are normally open during operation. The containment isolation system uses diversity in the parameters sensed for the initiation of containment isolation. (V) l Revision: 5 i W W85tingh0USS 6.2-33 February 29,1996 l l i
l l l
- 6. Engine: red S fity Features Table 6.2.3-1 identifies the signals that initiate closure of each penetration. He two O
redundant train-oriented containment isolation signals are generated by any of the following signals:
. Low pressurizer pressure
. Low steam-line pressure Low La
. High containment pressure Manual containment isolatiem actuation ne remainder of the containment isolation valves are closed on parameters indicative of the need to isolate.
B. Upon failure of a main steam line, the steam generators are isolated, and the main steam-I line isolation valves, main steam-line isolation bypass valves, power oprated relief block l valves, and the main steam-line drain arc closed to prevent excessive cooldown of the reactor coolant system or overpressurization of the containment. The two redundant train-oriented steam-line isolation signals are initiated upon receipt of any of the following signals:
. Low steam-line pressure
. High steam pressure negative rate
. High containment pressure
. Manual actuation l .
Uw La he main steam-line isolation valves, main steam line isolation valve bypass valves, main feedwater isolation valves, steam generator blowdown system isolation valves, and piping are designed to prevent uncontrolled blowdown from more than one steam generator. I ne main steam-line isolation valves and main feedwater isolation valves close fully I within 5 seconds after an isolation is initiated. he blowdown rate is restricted by steam flow restrictors located within the steam gewrator outlet steam nozzles in each blowdown path. For main steam-line breaks upseam of an isolation valve, uncontrolled blowdown from more than one steam generator is prevented by the main steam-line I isolation valves on each main steam line. Failure of any one of these components relied upon to prevent uncontrolled blowdown of more than one steam generator does not permit a second steam generator blowdown to occur. No single active component failure results in the failure of more than one main steam iso'ation valve to operate. Redundant main steam isolation signals, described in Section 7.3, are fed to redundant parallel actuation vent valves to provide isolation valve closure in the event of a single isolation signal failure. The effects on the reactor coolant system after a steam-line break resulting in single steam generator blowdown and the offsite radiation exposure after a steam line break Revision: 5 February 29,1996 6.2-34 [ West lr)ghouse
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- 6. Engineered Safety Featires n-
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outside cor"ainment are discussed in Chapter 15. The containment pressure transient following a main steam-line break inside containment is discussed in Section 6.2. C. He containment isolation system is designed accordirig to General Design Criterion 54. Leakage detection capabilities and leakage detection test program are discussed in Subsection 6.2.5. Valve operability tests are also discussed in Subsection 3.9.6. I Redundancy of valves and reliability of the isolation system are provided by the other safety design bases stated in Section 6.2. Redundancy and reliability of the actuation system are covered in Section 7.3. The use of motor-operated valves that fail as-is upon loss of actuating power in lines penetrating the containment is based upon the consideration of what valve position provides the plant safety. Furthermore, each of these valves, is provided with redundant backup valves to prevent a single failure from disabling the isolation function. Examples include: a check valve inside the containment and motor-operated valve outside the containment or two motor-operated valves in series, each powered from a separate engineered safety features division. D. Lines that penetrate the containment and which are either part of the reactor coolant pressure boundary, connect directly to the containment atmosphere, or do not meet the requirements for a closed system, except instrument sensing lines, are provided with one of the following valve arrangements conforming to the requirements of General Design Criteria 55 and 56, as follows: w/ One locked-closed isolation valve inside and one locked-closed isolation valve outside containment One automatic isolation valve inside and one locked-closed isolation valve outside contamment One locked-closed isolation valve inside and one automatic isolation valve outside containment. (A simple check valve is not used as the automatic isolation valve outside containment.) One automatic isolation valve inside and one automatic isolation valve outside containment. (A simple check valve is not used as the automatic isolation valve outside containment). Isolation valves outside containnient are located as close to the containment as practical. Upon loss of actuating power, air-operated automatic isolation valves fail closed. E. Each line penetrating the containment that is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere, and that satisfies the , requirements of a closed system, has at least one containment isolation valve. 'Ihis l containment isolation valve is either automatic, locked-closed, or capable of remote-manual operation. The valve is outside the containment and located as close to
! i lC/
Revision: 5 [ W95tingh0tlSe 6.2-35 February 29,1996
=:
- 6. Engineered Screty Features the containment as practical. A simple check valve is not used as the automatic isolation O
valve. His design is in compliance with General Design Criterion 57. F. Instrument lines penetrating the containment are provided with isolation valves according to General Design Criteria 55 and 56, and the containment pressure instrument lines are l designed according to Regulatory Guide 1.141. G. He containment isolation system is designed according to seismic Category I requirements as specified in Section 3.2. The cornponents (and supporting structures) of any system, equipment, or structure that are non-seismic Category I and whose collapse could result in loss of a required function of the containment isolation system through either impact or resultant floodmg are evaluated to confirm that they will not collapse when subjected to seismic loading resulting from a safe shutdown earthquake. Air-operated isolation valves fail in the closed position upon loss of air or power. Containment isolation system valves required to be operated after a design basis accident or safe shutdown earthquake are powered by the Class IE de electric power system. 6.2.3.4 Tests and Inspections Pre-operational testing is described in Chapter 14. De contamment isolation system is testable through the operational sequence that is postulated to take place following an accident, including operation of applicable portions of the protection system and the transfer between normal and standby power sources. He piping and valves associated with the containment penetration are designed and located to permit pre-service and in-service inspection according to ASME Section XI, as discussed I in Sections 3.9.6 and 6.6. Each line penetrating the containment is provided with testing features to allow containment leak rate tests according to 10 CFR 50, Appendix J, as discussed in Subsection 6.2.5. 6.2.3.5 Instrumentation and Control Application Instrumentation and control necessary for containment isolation, and the sensors used to determine that containment isolation is required, are described in Section 7.3. Containment isolation will be initiated by any of the high containment pressure signals, low T Iow steam-line pressure, or low pressurizer pressure signal using two out of four logic. Contamment isolation can also be initiated manually from the main control room. Containment isolation valves requiring isolation close automatically on a containment isolation signal. Containment isolation valves that are equipped witl power operators and are automatically I actuated may also be controlled individually from the main control room. Also, in the case of certain valves with actuators, a manual override of an automatic isolation signal is installed O Revision: 5 February 29,1996 6.2-36 3 W85tingh0Use l l
l i=u
- 6. Engineered S:f;ty Features q
l \J to permit manual control of the associated valve. ne override control function can be l ' performed only subsequent to resetting of the actuation signal. That is, deliberate manual I action is required to change the position of containment isolation valves in addition to resetting the original actuation signal. Resetting of the actuator signal does not cause any valve to change position. He design does not allow ganged reopening of the containmeni isolation valves. Reopening of the isolation valves is performed on a valve-by-valve basis, or on a line-by-line basis. Safety mjection signals take precedence over manual overrides of other isolation signals. For example, a safety injection signal causes isolation valve closure even though the high containment signal is being overridden by the operator. Containment I isolation valves with power operators are provided with open/ closed indication, which is I displayed in the main control room. ne valve mechanism also provides a local mechanical indication of valve position. Power supplies and control functions neces.sary for containment isolation are Class lE, as described in Chapters 7 and 8. 6.2.4 Containment Hydrogen Control System Following a loss of coolant accident (LOCA), hydrogen may be produced inside the reactor containment by reaction of the zirconium fuel cladding with water, by radiolysis of water, by corrosion of materials of construction, and by release of the hydrogen contained in the reactor coolant system. He containment hydrogen control system is provided to limit the hydrogen (] v concentration in the containment so that containment integrity is not endangered. Two situations are postulated, a design basis case and a severe accident case. In the design I basis case there is a limited reaction of less than one percent of fuel cladding zirconium with I water to form hydrogen. For this case there is an initial release of hydrogen due to the I reaction of fuel cladding with water and the release of hydrogen contained in the reactor coolant system. This initial hydrogen release to containment is not sufficient to approach the flammability limit of four v.olume percent. However, hydrogen continues to evolve to the containment due to radiolysis of water and the corrosion of materials in the containment, ne I flammability limit will eventually be reached unless mitigating action is taken. He function of the containment hydrogen control system is to prevent the hydrogen concentration from reaching the flammability limit. In the severe accident case it is assumed that 100 percent of the fuel cladding reacts with water. Although hydrogen production due to radiolysis and corrosion occurs, the cladding reaction with water dominates the production of hydrogen for this case. D e hydrogen generation from the zirconium-steam reaction could be sufficiently rapid that it may not be possible to prevent the hydrogen concentration in the containment from exceeding the lower flammability limit. ne function of the containment hydrogen control system for this case is I to promote hydrogen burning soon after the lower flammability limit is reached in the containment. Initiation of hydrogen burning at the lower level of hydrogen flammability prevents accidental hydrogen burn initiation at high hydrogen concentration levels and thus provides confidence that containment integrity can be maintained during hydrogen burns and m that safety-related equipment can continue to operate during and after the burns. (v) Revision: 5 [ W8Stingh00S8 6.2-37 February 29,1996
- 6. Engineered Saf:ty Features he containment hydrogen control system consists of the following functions:
e Hydrogen concentration monitoring i
- Hydrogen control during and followm' g a design basis LOCA (provided by passive I autocatalytic recombiners, PARS)
Hydrogen control during and following a degraded core or core melt (provided by hydrogen igniters). 6.2.4.1 Design Basis 6.2.4.1.1 Containment Mixing l Containment structures are arranged to promote mixing via natural circulation. He physical I mechanisms of natural circulation mixing that occur in the AP600 are discussed in l WCAP-14407 (Proprietary), WCAP-14408 (Nonproprietary) Reference 21, and summarized I as follows. For a postulated break low in the containment, buoyant flows develop through l the lower compartments due to density head differences between the rising plume and the I surrounding containment atmosphere, tending to drive mixing through lower compartments I and into the region above the operating deck. There is also a degree of mixing within the I region above deck, which occurs due to the introduction of and the entrainment into the I steam-rich plume as it rises from the operating deck openings. Rus, natural forces will tend I to mix the containment atmosphere. l l Two general characteristics have been incorporated into the design of the AP600 to promote l mixing and eliminate dead-end compartments. The campartments below deck are large open I solumes with relatively large interconnections, which promote mixing throughout the below I deck region. All compartments below deck are provided with openings through the top of the l compartment to elimmate the potential for a dead pocket of high-hydrogen concentration. In I addition, if forced containment air-circulation is deemed appropriate during post-accident I recovery, then nonsafety-related fan coolers are available for use by the operators i I In the event of a hydrogen release to the containment, passive autocatalytic recombiners I (PARS), act to recombine hydrogen and oxygen on a catalytic surface (see Section 6.2.4.2.2). I De enthalpy of reaction generates heat within a PAR, which further drives containment I mixing by natural circulation. Catalytic recombiners reduce hydrogen concentration at very I low hydrogen concentrations (<1 percent) and very high steam concentrations, and may also I promote convection to complement PCS natural circulation currents to inhibit stratification of I the containment atmosphere (Reference 18). De implementation of PARS has a favorable l impact on both containment mixing and hydrogen mitigation. 6.2.4.1.2 Survivability of System i he portion of the containn.ent hydrogen control system required for the design basis LOCA is designed to withstand the dynamic effects associated with postulated accidents, the Revision: 5 ) February 29,1996 6.2 38 [ Westingh0Use
- 6. Engineered S:f;ty Feat res EEn n-l C
environment existing inside the mntainment following the postulated accident, and a safe l shutdown earthquake. The containment hydrogen control equipment provided to mitigate severe accident conditions I (igniter subsystem) is designed to function under the event environment including the effects of combustion of hydrogen in containment. 6.2.4.1.3 Single Failure Protection The hydrogen monitoring function and the hydrogen recombination subsystem are designed to accommodate a single failure. The hydrogen ignition system, since it is provided only to I address a low-probability severe accident, is designed to accommodate probable component I and system failures. 6.2.4.1.4 Validity of Hydrogen Monitoring The hydrogen monitoring function monitors diverse locations within the containment to detect variations in hydrogen concentration. 6.2.4.1.5 Hydrogen Control for Design Basis Accident l De containment volume average hydrogen concentration is prevented from exceeding four O volume percent. His limit eliminates the potential for flammable conditions from being V rea:hed. 6.2.4.1.6 Hydrogen Control for Severe Accident l The containment hydrogen concentration is prevented from exceeding 10 volume percent I globally or locally. This limit, while allowing deflagration of hydrogen (burning of the hydrogen with flame front propagation at subsonic velocity), prevents the occurrence of I hydrogen detonation (burning of hydrogen with supersonic flame front propagation). De I distributed hydrogen ignition subsystem provide.s the control. 6.2.4.2 System Design 6.2.4.2.1 Hydrogen Concentration Monitoring Subsystem l The hydrogen concentration rnonitoring subsystem consists of two groups of eight hydrogen I sensors each. De sensors are placed in various locations throughout the containment free I volume including the upper dome and containment compartments. l l De system contains a total of three sensors designated as Class 1E and thirteen sensors I designated as Non-class 1E. The three Class 1E sensors are seismic Category 1 and serve to l provide a post accident monitoring function for design basis accidents. The sensors i designated as Non-class 1E provide a defense in depth function of monitoring local hydrogen I concentrations. ,U Revision: 5 [ W8Stingh0US8 6.2-39 February 29,1996
- 6. Engineered Sity Features I
O Each of the hydrogen sensors consists of a thermal conductivity detector and amplifier I powered by either a Class IE or non-Class IE power source. Sensor parameters are provided I in Table 6.2.4-1. Hydrogen concentration is continuously indicated in the main control room. Additionally, high hydrogen concentration alarms are annunciated in the main control room. l The hydrogen sensors consist of a thermal conductivity detector and amplifier. The detector includes a block with spropriately arranged cavities and gas passages. Four electrical elements are moanted in the cavities. Two elements are exposed to a sample atmosphere and the other two are exposed to a hydrogen-free reference gas of constant composition. T h e atmosphere sample reaches the thermal conductivity detector by diffusion through two porous metal barriers which act as flame arrestors between the detector elements and the atmosphere. The sensors are designed to provide a rapid response detection of changes in the containment hydrogen concentration. He response time of the sensor is 90 percent in 10 seconds. The four elements are electrically connected to form a bridge circuit through which the current is passed to heat the elements. The two elements exposed to the reference gas lose heat to the block at a constant rate and consequently have a stable temperature. The two elements exposed to the sample atmosphere dissipate heat at a rate that varies with the sample composition. Consequently, the temperature varies with variations in the sample hydrogen concentration. The element resistance changes with the temperature so that the bridge is electrically unbalanced. This unbalance is exhibited as a voltage that is proportional to the hydrogen concentration in the sample atmosphere. 6.2.4.2.2 Hydrogen Recombination Subsystem The hydrogen recombination subsystem is designed to accommodate the relatively slow I hydrogen production rate anticipated for a design-basis LOCA. He hydrogen recombination l subsystem consists of two passive autocatalytic recombiners installed inside the containment I above the operating deck at elevation 162' approximately 13 feet inboard from the I containment shell. The PARS are simple and passive in nature without moving parts and I independent of the need for electrical power or any other support system. The subsystem will I therefore operate independent of the availability of power following an accident resulting in I the generation of hydrogen. I Normally, oxygen and hydrogen recombine by rapid burning only at elevated temperatures I (greater than about 600*C). However, in the presence of catalytic materials such as the l l palladium group, this " catalytic burning" occurs even at temperatures below 0*C. Adsorption I l of the oxygen and hydrogen molecules occurs on the surface of the catalytic metal because I of attractive forces of the atoms or molecules on the catalyst surface. PAR devices use I palladium or platinum as a catalyst to combine molecular hydrogen with oxygen gases into l l water vapor. The catalytic process can be summarized by the following steps (Reference 16): l l
- 1) diffusion of the reactants (oxygen and hydrogen) to the catalyst; 2) reaction of the catalyst I (chemisorption); 3) reaction of intermediates to give the product (water vapor); 4) desorption I
of the product; and 5) diffusion of the product away from the catalyst. The reactants must O g Revision: 5 l February 29,1996 6.2-40 W WeStirigh0USS l
- 6. F;;gineered Satty Fe:tures o
V 'i I get to the catalyst before they can react and subsequently the product must move away from I the catalyst before more reactants will be able to react. I The PAR device consists of a stainless steel enclosure providing both the structure for the I device and support for the catalyst material. The enclosure is open on the bottom and top and i extends above the catalyst elevation to provide a chimney to yield additional lift to enhance i the efficiency and ventilation capability of the device. The catalyst material (pellet form) is 1 either constrained within screen cartridges or deposited on a metal plate substrate material and I supported within the enclosure. The spaces between the cartridges or plates serve as I ventilation channels for tne throughflow. During operation, the air inside the recombiner is I heated by the recombination process, causing it to rise by natural convection. As it rises, I replacement air is drawn into the recombiner through the bottom of the PAR and heated by I the exothermic reaction, forming water vapor, and exhausted through the chimney where the I hot gases mix with containment atmosphere. The device is a molecular diffusion filter and I thus the open flow channels are not susceptible to fouling. I PARS begin the recombination of hydrogen and oxygen almost immediately upon exposure I to these gases when the catalyst is not wetted. If the catalyst material is wet, then a short I delay is experienced in PAR startup (Ref. 20). The recombination process occurs at room I temperature during the early period of accidents prior to the buildup of flammable gas I concentrations. PARS are effective over a wide range of ambient temperatures, concentrations I of reactants (rich and lean, oxygen / hydrogen <1%) and steam inerting (steam concentrations ( l V) I
>50%). Although the PAR depletion rate reaches peak efficiency within a short period of time, the rate varies with hydrogen concentration and containment pressure.
I Reference 20 provides PAR pe:Jormance estimates appropriate (depletion rates) for a design I basis accident while a best-estimate depletion rate is appropriate for severe accident hydrogen I control scenarios where realistic estimates of system performance are appropriate due to the I low probability of occurrence. A conservative or lower bound estimate of depletion rate may I be used for a design basis accident analysis. The conservative depletion rate accounts for I effects such as instrumentation error, curve fitting, startup delay and a single failure. This rate I (with one PAR available) is used for the analysis results presented in Figure 6.2.4-1, " PAR I Sensitivity Study - Dry Conditions, Impact on Containment H2 Concentrations" l l The equations predicting the depletion rate are as follows: I i
- For H2 concentrations less than 0.2% depletion rate (kg/hr) = 0.0 I
l = ior H2 concentrations equal to or greater than 0.2% depletion rate (kg/hr) = 1 78,800 x [0.029883 x ([C-0.2]/100)2 + (0.001009 x [C-0.2]/100) x P]/(T + 273) l I where C = volume average H2 concentration at PAR inlet
- I P = total pressure (bars) j l T = gas temperature at PAR inlet ( C) i l,a
( ) ,v __ Revision: 5 Y W8Stiligl100S8 6.2-41 February 29,1996
- 6. Engineered Saf;ty Featrres l ne conditions under which the PARS are assumed to operate for a design basis accident for O I 1 defining the lower bound hydrogen depletion rate per reference 20 are:
1 1 l 1. Inlet gas temperatures ranging from 100 to 330 deg. F, l l l l 2. Pressures ranging from 1 to 4 bars, 'l l
- l. 3. Hydrogen concentrations up to 5 volume percent, ,
I l 4. Steam concentrations ranging from near zero to 75 percent, I l S. Condensing steam environment, and l l 1 6. No significant levels of potential catalyst poisons (e.g., iodine, carbon monoxide, cable i fire combustion products, tellurium). Such potential poisons would only be present at I significant levels during a postulated severe accident. l l De basis for defining the hydrogen depletion rate is testing conducted by Battelle Frankfurt I of both full scale and segment model NIS PAR units. De results of the tests and their use I in the definition of a hydrogen depletion rate equation appropriate for a design basis accident I is provided in references 19 and 20. De PAR testing and reporting of test data, conducted I under the NIS quality assurance program is appropriate for design certification. An evaluation I and summary of the QA program for the BMtelle tests is provided in reference 22. I he depletion rate assumed in the analysis is based on a generic PAR application as described I in Reference 20, and is expected to be representative of a number of vendor's recombiners. I he calcultted containment hydrogen concentration presented in Figures 6.2.4-1 and 6.2.4-2 l Is based on the assumptions and analysis discussed is Section 6.2.4.3. De results demonstrate I abundant margin for system performance. Further, the hydrogen concentration following an I accident with only one of the two available PARS operating within containment demonstrates I significant margin to maintaining hydrogen concentrations below the recommendations of l Regulatory Guide 1.7, Control of Combustible Gas Concentrations in Containment Following I a Loss-of-Coolant Accident. I Re recombiners are safety-related equipment. Eey are seismic Category I and are qualified I for the post-LOCA environment. De recombiners require no power supply and are self-l actuated simply by the presence of the reactants, hydrogen and oxygen. 1 A summary of component data for th^ hydrogen recombiners is provided in Table 6.2.4-2. 6.2A.2.3 Hydrogen Ignition Subsystem he hydrogen ignition subsystem is provided to address the possibility of an event that results in a rapid production of large amounts of hydrogen such that the containment hydrogen concentration exceed flammability limits before the hydrogen recombiners can be brought into use (and the rate of production exceeds the capacity of the recombiners even if they are Revision: 5 February 29,1996 6.2-42 3 Westinghouse
ni mL
- 6. Engineered Saf;ty Featrres l'D available). This massive hydrogen production is postulated to occur as the result of a degraded core or core melt accident (severe accident scenario) in which up to 100 percent of the zirconium fuel cladding reacts with steam to produce hydrogen.
The hydrogen ignition subsystem consists of one train containing 58 hydrogen igniters strategically distributed throughout the containment. Since the igniters are incorporated in the design to address a low-probability severe accident, the hydrogen ignition system is not Class 1E. The locations of the igniters are based on evaluation of hydrogen transport in the containment and the hydrogen combustion characteristics. Locations include compartmented areas in the containment and various locstions throughout the free volume, including the upper dome. For enclosed areas of the containment, at least two igniters are installed. The separation between igniter locations is selected to prevent the velocity of a flame front initiated by one igniter from becoming significant before being extinguished by a similar flame front l propagating from another igniter. I The igniter assembly is designed to maintain the surface temperature within a range of 1600 I to 1700*F surface at 1700*F in the anticipated containment environment following a LOCA. I A spray shield is provided to protect the igniter from falling water drops (resulting from condensation of steam on the containment shell and on nearby equipment and structures). {~} V Design parameter for the igniters are provided in Table 6.2.4-3. l 6.2.4.2.4 Containment Purge Containment purge is not part of the containment hydrogen control system. The purge I capability of the containment air filtration system (see subsection 9.4.7) can be used to provide contaimnent venting prior to post-LOCA cleanup operations. l 6.2.43 Design Evaluation (Design. Basis Accident) 6.2.43.1 Hydrogen Production and Accumulation Following a LOCA, hydrogen may be added to the reactor containment atmosphere by reaction of the zirconium fuel cladding with water, by radiolysis of water, by corrosion of materials of construction, and by release of the hydrogen contained in the reactor coolant system. The assumptions used in calculating the hydrogen release to containment are listed in Table 6.2.4-4. 7 Revision: 5 3 Westingh0use 6.2-43 February 29,1996
j mE
- 6. Engineered Sif;ty Feat res m-l 6.2A3.1.1 Zirconium-Water Reaction O
l Zirconium fuel cladding reacts with steam according to the following equation: Zr + 2 H2 0 - + ZrO2 + 2 H2+ heat There is 8.5 standard cubic feet (SCF) of hydrogen produced for each pound of zirconium that is reacted. He extent of the zirconium-water reaction is dependent on the effectiveness of the core cooling. An evaluation of the AP600 design shows that there is no zirconium-water I reaction during a design basis accident. The NRC model presented in Regulatory Guide 1.7 conservatively assumes that the cladding oxidizes to a depth of 0.00023 inch. For the 0.0225 inch cladding thickness used for AP600 fuel, this constitutes 1.09 percent of the zirconium. The hydrogen produced by the reaction of zirconium is 3000 SCF. His hydrogen is assumed to be released to the containment atmosphere at the beginning of the accident. 6.2.43.1.2 Radiolysis of Water Water radiolysis is a complex process involving reactions of numerous intermediates. However, the overall radiolytic process may be described by the equation: H O pt H + 10 2 2 2 2 Post-accident conditions in the containment create two distinct radiolytic environments. One environment exists inside the reactor vessel where radiolysis occurs due to energy emitted by decaying fission products in the fuel and absorbed by the cooling water. The second environment exists outside the reactor vessel, in the post-accident cooling solution itself, where radiolysis occurs due to the absorption of decay energy emitted by the fission products retained in the solution. The two basic differences between the core environment and the solution environment that affect the rate of hydrogen production are the fraction of energy I absorbed by the water and the type of flow regime. The rate of hydrogen production from radiolysis depends on the rate of energy absorption by the solution. Analysis of :nergy deposition in the reactor core where decaying fission products are retained in the fuel shows that beta radiation constitutes roughly 50 percent of the total decay energy. Since the beta radiation is absorbed by the fuel and the fuel clad, this energy is not available to the solution to contribute to the radiolysis of water. Additionally, most of the gamma radiation energy is absorbed by the fuel, fuel cladding, and other
- components; or it passes through the water without being absorbed. The solution in the i
reactor vessel would absorb approximately seven percent of the gamma radiation energy. However, consistent with Regulatory Guide 1.7, it is assumed that 10 percent of the core gamma energy is absorbed by the water.
- Revision
- 5 February 29,1996
) 6.2-44 W Westfrighouse l l
mm -
- 6. Engineered Saf;ty Felt:res m-V For the post-accident cooling solution, in which the fission products released from the core are assumed to be dissolved, energy is emitted directly into the solution. All of the beta radiation is assumed to be absorbed by the water. Since the mass of water is relatively large compared to the penetrating capability of gamma radiation, it is also assumed that 100 percent of the gamma radiation energy is absorbed by the water.
He radiolytic decomposition of water is a reversible reaction. In the reactor vessel, where the products of radiolysis are continuously flushed away by the circulation of cooling solution, there is little chance for hydrogen and oxygen to accumulate. Consequently, recombination of hydrogen and oxygen is assumed not to occur because significan; quantities of the two reactants are not available. He post-accident cooling solution in the sump, however, is a deep and relatively static environment where the products of radiolysis are lost from solution primarily by molecular diffusion. Tests simulating post-accident sump conditions demonstrate that there is significant reverse reaction in the sump. Hence, there is an apparent reduction in the quantity of hydrogen produced per unit energy absorbed by the water. He results of Westinghouse and Oak Ridge National Laboratory studies indicate maximum hydrogen yields of 0.44 molecules per 100 eV for core radiolysis and 0.3 molecules per l 100 eV for solution radiolysis. He results of these studies are published in References 12, i 13, and 14.
,a U)
I Re analysis performed for the AP600 assumes a hydrogen yield of 0.5 molecules per 100 eV for both the core and the solution radiolysis cases. His value is conservative relative to the referenced studies and is consistent with the guidance of Regulatory Guide 1.7. In a design basis LOCA there is expected to be no damage to the core and thus no release of activity from the core to the sump solution. The source term used for determining radiolysis l production of hydrogen is conservatively based on guidance of Regulatory Guide 1.7 which I states that 100 percent of noble gases,50 percent ofiodines, and 1 percent of other nuclides I are assumed to be released from the core even though it is inconsistent with the limited amount of fuel cladding reaction that is determined to take place. Appendix 15A provides the core fission product inventory at shutdown. Table 6.2.4-4 contains a summary of the assumptions used in the analysis of hydrogen produced from radiolysis. Production of hydrogen as a function of time is shown granhically I in Figure 6.2.4-3 and the accumulation of hydrogen is shown in Figure 6.2.4-4. 6.2.43.13 Corrosion of Metals In the environment that would exist inside the containment following a postulated LOCA, aluminum and zinc corrode to form hydrogen gas. Table 6.2.4-5 lists the inventory of l aluminum and zine inside the containment. I i s v Revision: 5 [ Westinghouse 6.2 45 February 29,1996
l
---g 1
- 6. Engineered Sity Feat:res Aluminum corrosion may be described by the overall reaction:
O ) 2 Al + 3 H O HA103 + 3 H2 2 About 21.4 SCF of hydrogen gas is produced for each pound of aluminum corroded. The corrosion of zine is described by the following reaction: Zn + 2 H 2O -+Zn(OH)2 + H2 About 5.9 SCF of hydrogen gas is produced for each pound of zine corroded. The corrosion rates for both aluminum and zinc are dependent on the post-accident temperature and pH conditions that the materials are subjected to. Table 6.2.4-5 provides the time-temperature cycle considered in the analysis of alumicm ."d zinc corrosion and also the corrosion rates for the metals at these temperatures. ' l Production of hydrogen as a function of time is shown graphically in Figure 6.2.4-3 and the I accumulation of hydrogen is shown in Figure 6.2.4-4. 6.2.43.1.4 Initial Reactor Coolant Hydrogen Inventory During normal operation of the plant, hydrogen is dissolved in the reactor coolant and is also contained in the pressurizer vapor space. Following a LOCA this hydrogen is assumed to be immediately released to the containment atmosphere. Table 6.2.4-4 lists the assumptions used for determining the amount of hydrogen from this source. The total hydrogen released to the , I containment as a result of this source is 1171 SCF. l 6.2.43.2 Hydrogen Mixing The AP600 is designed to prevent the accumulation of hydrogen in compartments. If there is the possibility of accumulation in compartments, venting is provided to allow the hydrogen l to escape to the larger containment volume. Mixing of the containment air mass is l accomplished through na9rd processes as a result of the passive cooling of the containment ! l that induces a recirculating air flow in the containment. The release rate for a desiga basis I accident is sufficiently slow that mixing is effectively assured. l 6.2A33 Hydrogen Recombination Assuming no hydrogen removal, the concentration of hydrogen in the containment atmosphere l l increases with time as shown in Figure 6.2.4-1. The curve shows that the flammability limit ' I of four volume percent is not reached until after 28 days. Hydrogen recombination begins I prior to reaching this limit. The passive autocatalytic recombiners are brought into service l by the presence of the reactants. The available PAR test data as discussed in reference 20 l supports PAR startup within 7 hours of reaching 1 volume percent hydrogen concentration in I containment. Subsequent to PAR startup the conservative lower bound equation for depletion I rates provided in reference 20 has been used to predict containment concentrations. Figure O Revision: 5 February 29,1996 6.2-46 [ Westingh0USB
== i
- 6. Engineered Saf.ty Featzres i \
V l 6.2.4-1 also shows the impact of operation of one of the two recombiners on containment I hydrogen concentration. He hydrogen concentration never exceeds 1.5 percent which I indicates ample margin in the hydrogen recombiner capacity. A further demonstration of the l PARS available capacity margin is provided by calculation of containment concentrations with I reduced depletion rates. Figure 6.2.4-2 provides the impact of one of two available PARS I operating at 20,10 and 1 percent of the conservative lower bound capacity. The curves l provide indication of the abundant hydrogen control margin. 6.2.4.4 Design Evaluation (Severe Accident) Although a severe accident involving major core degradation or core melt is not within the category of design basis accidents, the containment hydrogen control system contains design features specifically to address this potential occurrence. The hydrogen monitonng subsystem has sufficient range to monitor concentrations up to 20 percent hydrogen. The hydrogen ignition subsystem is provided so that hydrogen is burned off in a controlled manner, preventing the possibility of deflagration with supersonic flame front propagation which would result in large pressure spikes in the containment. The hydrogen released to the containment due to initial inventory of hydrogen in the coolant I would be the same as for the design basis case (see subsection 6.2.4.3.1.4). He hydrogen production due to corrosion of aluminum and zinc or to radiolysis of water is (l not of concern for evaluating the containment hydrogen control system for the severe accident () since hydrogen production from these sources takes place at a relatively slow rate and over a long period of time. It is assumed that 100 percent of the fuel cladding zirconium reacts with steam. This reaction may take several hours to complete. The igniters initiate hydrogen burns at concentrations less than 10 percent by volume and prevent the containment hydrogen erncentration from exceeding this limit. The evaluation of hydrogen control by the igniters is presented in the AP600 PRA. 6.2.4.5 Tests and Inspections 6.2.4.5.1 Hydrogen Monitoring Subsystem Functional and preopemtional testing is performed after installation and prior to plant startup to verify performance. The system is normally in service. Periodic testing and calibration are performed to provide ongoing confirmation that the hydrogen monitoring function can be reliably performed. 6.2.4.5.2 Hydrogen Recombination Subsystem l Functional and preoperational testing is performed prior to plant startup to verify performance. Periodic inspection and testing are performed to provide ongoing confirmation that the
, l hydrogen recombiners can operate reliably.
,! \ Q ,/ Revision: 5 [ W65tingh00S8 6.2-47 February 29,1996
-a l
- 6. Engineered Safety Frtures EAR -
l Each recombiner can be tested during plant shutdown to demonstrate operability. A sample O 1 of cartridges or plates are selected and removed from each PAR and surveillance bench tests I are performed on the removed specimens to confirm continued satisfactory performance, he I specimen is placed in a performance test apparatus and exposed to a known air / hydrogen i sample. The measured increase in temperature is used to indicate degradation in catalytic l action. 6.2.4.5.3 Hydrogen Ignition Subsystem Functional and preoperational testing is performed after installation and prior to plant startup to verify performance. Periodic inspection and testing are performed to provide ongoing confirmation that the hydrogen igniters can be reliably operated. l 6.2.4.6 Combined License Information l l Dis section has no requirement to be provided in support of the Combined License I application. 6.2.5 Containment Leak Rate Test System The reactor containment, containment penetrations and isolation barriers are designed to I permit periodic leak rate testing in accordance with General Design Criteria 52,53, and 54. He containment leak rate test system is designed to verify that leakage from the containment remains within limits established in the technical specifications, Chapter 16. 6.2.5.1 Design Basis Leak rate testing requirements are defined by 10 CFR 50 Appendix J, " Primary Reactor Contamment Leakage Testing for Water Cooled Power Reactors," which classifies leak tests as Types A, B and C. 6.2.5.1.1 Safety Design Basis ne containment leak rate test system serves no safety-related function other than containment isolation, and therefore has no nuclear safety design basis except for contamment isolation. I See subsection 6.2.3 for the containment isolation system. 6.2.5.1.2 Power Generation Design Basis he containment leak rate test system is designed to verify the leaktightness of the reactor containment. He specified maximum allowable containment leak rate is 0.12 weight percent of the containment air mass per day at the calculated peak accident pressure, P , identified in l subsection 6.2.1. He system is specifically designed to perform the following tests in accordance with the provisions of ANSI-56.8 (Reference 14): Revision: 5 I February 29,1996 6.2-48 3 Westklgflouse
- 6. Engineered Sar;ty Feat:res
!G')
Containment integrated leak rate testing (Type A): The containment is pressurized with clean, dry air to a pressure of P,. Measurunents of containment pressure, dry bulb temperature and dew point temperature are used to determine the decrease in the mass of air in the containment over time, and thus establish the leak rate. Local leak rate testing of containment penetrations whose design incorporates features such as resilient seals, gaskets, and expansion bellows (Type B): The leakage limiting boundary is pressurized with air or nitrogen to a pressure of P, and the pressure decay or the leak flow rate is measured. Local leak rate testing of containment isolation valves (Type C): ne piping test volume is pressurized with air or nitrogen to a pressure of P, and pressure decay or the leak flow rate is measured. For valves sealed with a fluid such as water, the test volume is pressurized with the seal fluid to a pressure of not less than 1.1 P,. He containment leak rate test system piping is also designed for use during performance of the containment structural integrity test. The instrumentation used for the stmetural integrity test may be different than that used for the integrated leak rate test. 6.2.5.1.3 Codes and Standards The containment leak rate test system is designed to conform to the applicable codes and (~} V standards listed in Section 3.2. Except as described in Table 6.2.5-1, the containment leak testing program satisfies Appendix J requirements. 6.2.5.2 System Description 6.2.5.2.1 General Description The containment leak rate test system is illustrated on Figure 6.2.51. Unless otherwise indicated on the figure, piping and instrumentation is permanently installed. Fixed test connections used for Type C testing of piping penetrations are not shown on Figure 6.2.5-1. These connections are not part of the containment leak rate test system and are shown on the applicable system piping and instrument diagram figure. Air compressor assemblies used for Type A testing are temporarily installed in the yard area near the Annex 11 building, and are connected to the permanent system piping. The number and capacity of the compressors is sufficient to pressurize the containment with air to a pressure of P, at a maximum containment pressurization rate of about five psi / hour. The compressor assemblies include additional equipment, such as air coolers, moisture separators and air dryers to reduce the moisture content of the air entering containment. Temperature and humidity sensors are permanently installed inside containment for Type A testing. Data acquisition hardware and instrumentation are permanently installed outside containment, in the auxiliary building. Instmmentation which is not required during plant ( operation for gross leak rate testing may be installed temporarily for the Type A tests. M) I Revision: 5 3 W85tifigh00se 6.2-49 February 29,1996
- 6. Engineered Safity Features The system is designed to permit depressurization of the containment at a maximum rate of elI 10 psi / hour.
{ Portable leak rate test panels are used to perform Type C containment isolation valve leak testing using air or nitrogen. The panels are also used for Type B testing of penetrations, for , I which there is no permanently installed test equipment. The panels include pressure regulators, filters, pressure gauges and flow instrumentation, as required to perform specific tests. 6.2.5.2.2 System Operation I Containment Integrated Leak Rate Test (Type A) l An integrated leak rate test of the primary reactor containment is performed prior to initial plant operation, and periodically thereafter, to confirm that the total leakage from the containment does not exceed the maximum allowable leak rate. The allowable leak rate specified in the test criteria is less than the maximum allowable containment leak rate, in accordance with Appendix J. Following construction of the containment and satisfactory completion of the stmetural I integrity test, described in subsection 3.8.2.7, a preoperational Type A test is performed as described in Chapter 14. Additional Type A tests are conducted during the plant life, at intervals in accordance with the technical specifications, Chapter 16.
- Pretest Requirements Prior to performing an integrated leak rate test, a number of pretest requirements must be satisfied as described in this subsection.
A general inspection of the accessible interior and exterior surfaces of the primary containment structure and components is performed to uncover any evidence of structural deterioration that could affect either the containment structural integrity or leaktightness. If there is evidence of structural deterioration, corrective action is taken prior to performing the Type A test. The structural deterioration and corrective action are reported in accordance with 10 CFR 50, Appendix J. Except as described above, during the period between the initiation of the containment inspection and the performance of the Type A test, no repairs or adjustments are made so that the containment can be tested in as close to the "as-is" condition as practical. Containment isolation valves are placed in their post-accident positions, identified in Table 6.2.3-1, unless such positioning is impractical or unsafe. Test exceptions to post-accident valve positioning are identified in Table 6.2.3-1 or are discussed in the test report. Closure of containment isolation valves is accomplished by normal operation and with no preliminary exercising or adjustments (such as tightening of a valve by manual handwheel after closure by the power actuator). Valve closure malfunctions or valve leakage that Revision: 5 e February 29,1996 6.2-50 W Westingholise l l I
ii =H
- 6. Engineered Sity Feat:res o
requires corrective action before the test is reported in conjunction with the Type A test report. Those portion 3 of fluid systems that are part of the reactor coolant pressure boundary and are open directly to the containment atmosphere under post-accident conditions and become an extension of the boundary of the containment, are opened or vented to the containment atmosphere prior to and during the test. Portions of systems inside containment that penetrate containment and could rupture as a result of a LOCA are vented to the containment atmosphere and drained of water to the extent necessary to provide exposure of the containment isolation valves to containment air test pressure and to allow them to be subjected to the full differential test pressure, except that: Systems that are required to maintain the plant in a safe condition during the Type A test remain operable and are not vented. Systems that are normally filled with water and operating under post-accident conditions are not vented. Systems which are not required to be vented and drained for Type A testing are identified in Table 6.2.3-1. De leak rates for the containment isolation valves in these systems, measured
, by Type C testing, are reported in the Type A test report.
V Tanks inside the containment are vented to the containment atmosphere as necessary to protect them from the effects of external test pressure and/or to preclude leakage which could affect the accuracy of the test results. Similarly, instrumentation and other components that could be adversely affected by the test pressure are vented or removed from containment. He containment atmospheric conditions are allowed to stabilize for a period of at least four hours prior to the start of the Type A test. The containment ventilation and cooling water systems are operated as necessary prior to, and during, the test to maintain stable test conditions.
- Test Method Be Type A test is conducted in accordance with ANSI-56.8, using the absolute method. The test duration is at least eight hours following the stabilization period. Periodic measurements of containment pressure, dry bulb temperatures and dew point temperatures (water vapor pressure) are used to determine the decrease in the mass of air in the containment over time.
A standard statistical analysis of the data is conducted using a linear least-squares-fit regression analysis to calculate the leak rate and the upper 95 percent corI:dence limit. De accuracy of the Type A test results is then verified by a supplemental test. De supplemental verification test is performed using either the superimposed leak method or the mass step change method, as described in ANSI-56.8. p. V Revision: 5 3 W8Stingt100Se 6.2-51 February 29,1996
u =__ v
- 6. Engineered Safety Festures Test criteria for the Type A test and the supplemental verification test are given in the O
technical specifications. If any Type A test fails to meet the criteria, the test schedule for subsequent tests is adjusted in accordance with 10 CFR 50, Appendix J . During the period between the completion of one Type A test and the initiation of the containment inspection for the subsequent Type A test, repairs or adjustments are made to components identified as exceeding individual leakage lindts, as soon as practical after such leakage is identified. Containment Penetration Leak Rate Tests (Type B) The following containment penetrations receive preoperational and periodic Type B leak rate tests in accordance with ANSI-56.8: Penetrations whose design incorporates resilient seals, gaskets or sealant compounds
. Air locks and associated door seals
- Equipment and access hatches and associated seals
. Electrical penetrations Containment penetrations subject to Type B tests are illustrated in Figure 6.2.5-1.
The fuel transfer tube penetration is sealed with a blind flange inside containment. The flanged joint is fitted with testable seals as shown in Figure 3.8.2-4. The two expansion bellows used on the fuel transfer tube penetration are not part of the leakage-limiting boundary of the containment. The personnel hatches (airlocks) are designed to be tested by internal pressurization. The doors of the personnel hatches have testable seals as shown in Figure 3.8.2-3. Mechanical and electrical penetrations on the personnel hatches are also equipped with testable seals. T h e natch cover flanges for the main equipment and maintenance hatches have testable seals as shown in Figure 3.8.2-2. Containment electrical penetrations have testable seals as shown in Figure 3.8.2-6. Type B leak tests are performed by local pressmization using the test connections shown on Figure 6.2.5-1. Unless otherwise noted in Table 6.2.3-1, the test pressure is not less than the calculated containment peak accident pressure, P,. Either the pressure decay or the flowmeter test method is used. These test inethods and the test criteria are presented below for Type C tests. Containment Isolation Valve Leak Rate Tests (Type C) Containment isolation valves receive preoperational and periodic Type C leak rate tests in accordance with ANSI-56.8. A list of containment isolation valves subject to Type C tests Revision: 5 February 29,1996 6.2 52 [ W85tiflgh00S8
! l
= i: l
- 6. Engineered Safety Featrres E m-U l
is provided in Table 6.2.3-1. Containment isolation valve arrangement and test connections provided for Type C testing, are illustrated on the applicable system piping and instrument diagram figure. Type C leak tests are performed by local pressurization. Each valve to be tested is closed by normal means without any preliminary exercising or adjustments. Piping is drained and vented as needed and a test volume is established that, when pressurized, will produce a differential pressure across the valve. Table 6.2.3-1 identifies the direction in which the differential pressure is applied. Isolation valves whose seats may be exposed to the containment atmosphere subsequent to a LOCA are tested with air or nitrogen at a pressure not less than P,. Valves in lines which are ; designed to be, or remain, filled with a liquid for at least 30 days subsequent to a LOCA are l' leak rate tested with that liquid at a pressure not less than 1.1 times P,. Isolation valves tested with liquid are identified in Table 6.2.3-1. I Isolation valves are tested using either the pressure decay or flowmeter method. For the pressure decay method the test volume is pressurized with air or nitrogen. The rate of decay of pressure in the known volume is monitored to calculate the leak rate. For the flowmeter method pressure is maintained in the test volume by supplying air or nitrogen through r. calibrated flowmeter. The measured makeup flow rate is the isolation valve leak rate. f'} The leak rates of penetrations and valves subject to Type B and C testing are combined in G' accordance with Appendix J. As each Type B or C test, or group of tests, is completed the combined total leak rate is revised to reflect the latest results. Thus, a reliable summary of containment leaktightness is maintained current. Leak rate limits and the criteria for the combined leakr A results are described in the technical specifications. Scheduling and Ret,orting of Periodic Tests l Schedules for the performance of periodic Type A, B, and C leak rate tests are in accordance with the technical specifications, Chapter 16. Provisions for reporting test results are described in the technical specifications. Type B and C tests may be conducted at any time that plant condition = permit, provided that the time between tests for any individual penetration or valve does not exceed the maximum allowable interval specified in the technical specifications, Chapter 16. Special Testing Requirements AP600 does not have a subatmospheric containment or a secondary containment. 'Ihere are no containment isolation valves which rely ot.1 luid f seal system. Thus, there are no special testing requirements. l l A Revision: 5 3 W85tingh00S8 6.2-53 February 29,1996
- =
- 0. Engineered S:f;ty Features 6.2.5.2.3 Component Description O
The system pressurization equipment is temporarily installed for Type A testing. In addition to one or more compressors, this hardware includes components such as aftercoolers, moisture separators, filters and air dryers. Although the hardware characteristics may vary from test to test, the pressurization equipment must meet the reqairements of Table 6.2.5-2. He flow cetrol valve in the pressurization line is a leaktight ball valve capable of throttling to a low flow rate. 6.2.5.2.4 Instrumentation Applications For Type A testing, instruments are provided to measure containment absolute pressure, dry bulb temperature, dew point temperature, air flow rate, and atmospheric pressure. Data acquisition equipment scans, processes and records data from the individual sensors. For Type B and C testing, instruments are provided to measure pressure, dry bulb temperature, and flow rate. He quantity and location of Type A instrumentation and permanently installed Type B instrumentation,is indicated on Figure 6.2.5-1. The type, make and range of test instruments may vary from test to test. The instmment accuracy must meet the criteria of Reference 14. 6.2.5.3 Safety Evaluation The containment leak rate test system has no safety related function, other than containment isolation and therefore requires no nuclear safety evaluation, other than containment isolation I which is described in subsection 6.2.3. 6.2.5.4 Inservice Inspection / Inservice Testing There are no special inspection or testing requirements for the containment leak rate test system. Test equipment is inspected and instruments are calibrated prior to testing in accordance with ANSI-56.8 criteria and the requirements of the test procedure. I 6.2.5.5 Combined License Information l l 'Ihis section has no requirements to be provided in support of combined license application. 6.2.6 References l 1. Kennedy, M., et al., "WGOTHIC Code Description and Validation," WCAP-14382, to I be issued in July.
- 2. " Ice Condenser Containment Pressure Transient Analysis Methods," WCAP-8077, March, 1973 (Proprietary), WCAP-8078 (Non-Proprietary), .
O Revision: 5 February 29,1996 6.2-54 3 Westingh0USS
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- 6. Engineered Sar;ty Feat res l / \
l i
- 3. Shepard, R. M., et. al., " Westinghouse Mass and Energy Release Data for Containment Design," WCAP-8264-P-A, June 1975 (Proprietary), and WCAP-8312-A, Revision 2 August 1975 (Non-Proprietary).
4
" Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version," WCAP-10325, May 1983 (Proprietary).
l l 5. Land, R. E., " Mass and Energy Releases Following A Steam Line Rupture," WCAP-8822 (Proprietary) and WCAP-8860 (Nonproprietary), September 1976.
- 6. Burnett, T. W. T., "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary) and WCAP-7907-A (Nonproprietary), June 1984.
l 7. Deleted l 8. Deleted l 9. 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cooled Nuclear Power Reactors," and Appendix K to 10 CFR 50, "ECCS Evaluation Model." I 10. Branch Technical Position CSB6-1, " Minimum Containment Pressure Model for PWR (^\ l ECCS Performance Evaluation." () I 11. Bajorek, S.M., Hochreiter, L.E., Young, M.Y., Dederer, S.I., Nissley, M.E., Tsai, C.K., Yeh, H.C., Chow, S.K., Takeuchi, K., Cunningham, J.P. and Stucker, D.L., {
" Code Qualification Document for Best Estimate Analysis," Volume 1, WCAP-12945-P, '
Revision 1, [ Proprietary], June 1992. I 12. Fletcher, W.D., Bell, M.J., and Picone, L.F., " Post-LOCA Hydrogen Generation in PWR Containments," Nuclear Technology 10. pp 420-427,1971. I 13. Zittel, H.E., and Row, T.H., " Radiation and Thermal Stability of Spray Solutions," Nuclear Technology 10. pp 436-443,1971. I 14. Allen, A.O., 'The Radiation Chemistry of Water and Aoucous Solutions. Princeton, N.J., Van Nostrand,1961. I 15. ANSI /ANS-56.8-1987, " Containment System Leakage Testing Requirements." l 16. 10 CFR 50, Appendix J (Draft Proposed Revision), " Containment Leak Rate Testing," January 10,1992. 1 17. Thomas C. L. Catalytic Processes and Proven Catalysts. Academic Press,1970. p. I 1
\.)
Revision: 5 1 3' #estinghouse 6.2-55 February 29,1996 s I
l l
- 6. Engineered Sirity Fe tures
\ l l 18. J. Rohde, et al., Hydrogen Mitigation by Catalytic Recombiners and Ignition During O l Severe Accidents," Third International Conference on Containment Design and ) I Operation, Canadian Nuclear Society, Toronto, Ontario, October 19-21, 1994. I 19. J. C. DeVine, Jr. " Passive Autocatalytic Recombiners for Combustible Gas Control ir ; I ALWR's," to Mr. James Wilson, April 8,1993. l l 20. EPRI Report, "NIS PAR Depletion Rate Equation for Evaluation of Hydrogen l Recombination During AP600 Design Basis Accident", EPRI ALWR Program, 1 November 15,1995 (attached). l I l 21. WCAP-14407 [ Proprietary] and WCAP-14408 [Non-Proprietary] W. GOTHIC Application I to AP600",7/95. I l 22. EPRI Report " Evaluation of Quality Assurance Applied to Battelle Tests of NIS Passive l Autocatalytic Recombiner", EPRI ALWR Program, October 1995. l l l O l l l l I t i O Revision: 5 February 29,1996 6.2-56 3 WBStiligt10Use
l
- 6. Engineered Saf;ty Fxtures j ol Table 6.2.1.1 1 l
SUMMARY
OF CALCULATED PRESSURES AND TEMPERATURES I Peak Peak i Pressure Available' Temperature Break (psig) Margin (psi) (*F) l Double-ended hot leg guillotine 40.6 4.4 338.6 I Double-ended cold leg guillotine 41.0 4.0 282.9 l 1.388 ft*, full DER,102% power, 40.5 4.5 328.1 MSIV failure I 1.388 ft', full DER,30% power, 43.6 1.4 320.2 MSIV failure
- 1. Design Pressure is 45 psig f
v o/ Revision: 5 [ W85tingh0USe 6.2-57 February 29,1996 l
- 6. Ezgineered Saf.ty Features O
Table 6 ? 1.1-2 COVERAGE FRACTION VS. TIME FOR AP600 Time Dome Dome Dome Cylinder Cylinder Cylinder Cylinder (hr) Top Middle Bottom Top Mid-Top Mid-Bottom Bottom 0.183 40% 40% 98 % 61 % 31 % 14 % 6% 2.167 40% 40% 100% 85 % 52 % 28 % 16 % 5.167 40% 40% 100% 100% 84 % 57 % 40% 5.667 40% 40% 85 % 48% 35 % 23 % 16 % 9.167 40% 40% 87 % 53% 40% 29 % 21 % 15.17 40% 40% 88 % 56 % 45 % 34 % 26% 21.17 40% 40% 87 % 66 % 45 % 35 % 28% 26.17 40% 40% 78 % 46 % 38% 30% 24 % O l l I i l l l i e Revision: 5 February 29,1996 6.2-58 3 W8Stingh0US8
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- 6. Engineered Saf;ty Features p
t
\
I Table 6.2.1.1-3 INITIAL CONDITIONS Internal Temperature ('F) . .. ....... . ...... ... . .. ...... . .... . . 120 Pressure (psia) . . . .. ...... ........ .. . ........ . . . . . . . . . .. ...... 15.7 i Relative Hamidity (%) . . . . .. ....... .. . ...... .. ....... . .. . . .. ... 0 Net Free Volume (ft') . . . .. . ..... ...... ........ ... .. ..... . . . 1.7 E+06 External Temperature (*F) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 15 dry bul b 80 wet bulb i i
/~s (N )i ,C \ _
Revision: 5 T Westinghouse 6.2-59 Feoruary 29,1996
l
= 1 jj 6. EEgineered Safety FGotures I
O Table 6.2.1.1-4 l l l RESULTS OF POSTULATED ACCIDENTS i l Distri. I I buted Lumped Lumped 102 % 30 % External l Acceptance DECLG DEHLG DECLG Power Power Pressuri-l Criterion LOCA LOCA LOCA MSLB MSLB zation I Criteriou Value Value Value Value Value Value Value l GDC 16 & GDC 50 < 45.5 psig 43 40.6 41.0 40.5 43.6 - l 10% Margin to l Design Pressure l GDC 38 < 22.5 psig - - 10.9 -- -- - I Rapidly Reduce I Containment Pressure l GDC 38 & 50 < 3 psid - - - - - 2.04 I External Pressure l 4 l GDC 38 & GDC 50 Most Severe One Train One One One One I Containment Heat ! of PCS Train of Train of Train of Train of ' l Removal Single Water PCS PCS PCS PCS I Failure Supply Water Water Supply Supply l ) Supply Supply Fails Fails I O Revision: 5 February 29,1996 6.2-60 3 W85tingl10US8
g_- t
- 6. Engineered Safety F;c.tures 1
(, l Table 6.2.1.1-5 (Sheet I of 7) - IIEAT SINK PROPERTIES i' Metallic Heat Sink Exposed Area Description Number 2 (ft ) Thickness (ft) Material Paint Reactor Cavity Containment Sump Pumps ! 1 18.84 0.083 SS NONE l RCDT 2 145.3 0.083 SS NONE l RCDT lleat Exchange 3 47.35 0.083 SS NONE i l HVAC Fans (2) 4 62.06 0.002 CS CZE l Platform El. 83'-0 5 360.8 0.00756 CS CZE l Platform E1.107*-2 6 1500.6 0.00756 CS CZE l Stairs El. 83'-0 to 107'-2 7 275.5 0.011 CS NONE l l Landing @ El. 92*-8 8 ,,,,, 0.00756 CS CZE l HVAC Duct 9 IN.0 0.0087 CS NONE Accumulator Cavity Southeast < ,/ ~3
) .G l Accumulator (01 A) 10 767.7 0.166 CS CZ l Platform El. 98*-0 11 23M.2 0.00756 CS CZE l Stairs El. 98'-0 to 107'-2 12 104.54 0.011 CS NONE Accumulator Cavity Northeast l Accumulator (01B) 13 767.7 0.166 CS CZ l Platform El. 98'-0 14 3360.4 0.00756 CS CZE l Stairs El. 98*-0 to 107-2 15 1N.54 0.011 CS NONE Steam Generator Room East l SG Lower Manway Platform (El.109'-7) 16 2240.1 0.0225 CS CZE l SG Tubesheet Platform (El. I13'-9) 17 1710.27 0.00756 CS CZE l Operating Deck Platform (El.135'-3) 18 1678.5 0.00756 C" CZE l Stairs El.104'-7 to 113'-g 19 IM.54 0.011 CS NONE . !l l
in -( ) w/ Revision: 5 3 W8Stingh00S6 6.2-61 February 29,1996
1
- 6. Engineered Safety Features l
e Table 6.2.1.1-5 (Sheet 2 of 7) HEAT SINK PROPERTIES Metallic Heat Sink Exposed Area Description Number (ft') Thickness (ft) Material Paint Steam Generator Room West l SG Support Column EL 83"-0 to 105'-71/2 20 181.39 0.13 CS CZE l SG Upper Support 21 281.03 0.405 CS CZE I SG Lower Manway Platform (El.104'-7) 22 2240.1 0.0225 CS CZE l SG Tubesheet Platform (El. I13'-9) 23 1710.27 0.00756 CS CZE l Operating Deck Platform (El.132'-0) 24 1678.5 0.00756 CS CZE l Stairs From El.104'-7 to 113'-9 25 104.54 0.011 CS NONE l SG Support Column 26 181.39 0.13 CS CZE l SG Upper Support 27 281.03 0.405 CS CZE CMT And CVS Room l Letdown Heat Exchanger 28 79.72 0.75 SS NONE l Mixed Bed & Cation Demineralizer (3) 29 480.86 0.75 SS NONE I Reactor Coolant Filters (2) 30 114.67 0.75 SS NONE l Support Steel 31 1449.32 0.0325 CS CZE l 5" Vertical Shield Plate 32 1252.95 0.405 CS CZE I 1" Vertical Shield Plate 33 992.87 0.07 CS CZE l 1" Vertical Shield Plate 34 292.88 0.25 CS CZE l 1" Horizontal Floor Plate 35 241 0.07 CS CZE l 3" Horizontal Floor Plate 36 227.94 0.25 CS CZE l Sump Pumps (2) 37 28.27 0.083 SS NONE l Platform El. 88*-2 & El. 95'-10 38 1328.4 0.00756 CS CZE 'l HVAC Ducts (2) 39 505.5 0.0087 CS NONE l Core Make-up Tanks (02A) & (028) 40 1848.8 0.405 CS CZE l Primary Containment Vessel 41 11063.75 0.13 CS CZE l l Columns (17) 42 3113.0 0.0325 CS CZE l Floor Framing Beneath El.135'-3 43 7563.38 0.03 CS CZE l Elevator Stair Tower 44 436.92 0.01632 CS CZE e Revision: 5 February 29,1996 6.2-62 [ Westingh00S8
__ . _ ~ -. am4
- 6. Engineered Safety Features v
l Table 6.2.1.1-5 (Sheet 3 of 7) HEAT SINK PROPERTIES i Metallic Heat Shik Exposed Area Description Number (ft') Thickness (ft) Material Paint l Maintenance Hatch 45 321.8 0.145 CS CZE l Platform El. I18'-6 46 22508.5 0.012 CS CZE l Stairway From El.107'-2 to 135'-3 47 320.15 0.011 CS NONE l Rail Handle Above Reactor Access 48 43.04 0.011 CS NONE Refueling Canal I Imwer Intemal Stand 49 386.9 0.042 SS NONE l Upper Internal Stand 50 546.9 0.042 SS NONE l Upender 51 79.481 0.3666 SS NONE l Refueling Canal Gate 52 2016 0.042 SS NONE l Platform El.130'-3 53 279.83 0.00756 CS CZE IRWST Room U]) l Metallic Wall & Stiffeners 54 7050.96 0.042 SS NONE l Upper East Steam Generator Compartment l Jib Crane 55 103.67 0.405 CS CZE I Feedwater Nozzle Platform (El.149*-7) 56 2806.45 0.00756 CS CZE l Upper Manway Platform (El.162'-1) 57 2806.45 0.00756 CS CZE l Jib Crane 58 77.39 0.405 CS CZE l Upper West Steam Generator Compartment
! Jib Crane 59 103.67 0.405 CS CZE l Feedwater Nozzle Platform El.149'-7 60 2806.45 0.00756 CS CZE l Upper Manway Platform El.162*-1 61 2806 45 0.00756 CS CZE l Jib Crane 62 77.39 0.405 CS CZE I Integrated Head Stand (1 side) 63 491.6 0.021 SS NONE I. !
w/ Revision: 5 [ W8stinghouse 6.2-63 February 29,1996 l l
wg-
- 6. Engineered Safety Features re-O Table 6.2.1.1-5 (Sheet 4 of 7)
HEAT SINK PROPERTIES Metallic Heat Sink Exposed Description Number Area (ft ) 2 Thickness (ft) Material Paint I Refueling Machine (I side) 64 742.2 0.07 CS CZE I Platform EL 147'-3 65 457.7 0.00756 [ CS CZE l South Inner Hall l Refueling Machine (1 side) 66 819.8 0.07 CS CZE ! l North Inner Hall ; l l Stairs El.135'-3 to EL 148'-0 67 290.7 i 0.011 CS NONE l landings Stairs El. 140*-3/16, El.142*-5, 68 147.6 0.00756 CS CZE ( El.144'-9/16 l l l Phiforms at El.149'-7 69 120.0 0.00756 CS CZE l Platforms at El.155' 10 70 54.61 0.00756 CS CZE { l Platforms at El.160'-6 71 1748.2 0.00756 CS CZE f l Platforms at EL 162*-1 72 54.61 0.00756 CS CZE l Platforms at EL 169'-0 73 1748.2 0.00756 CS CZE Stairs from EL 148'-0 to El.162*-1 and 74 399.95 0.011 CS NONE l from El.148'-0 to EL 169'-0 Landings at EL 149'-7, 75 262.2 0.00756 CS CZE El.154'-4 5/16 EL 165'-3 5/16 I l North Mid Quadrant l Elevator Support Structure 76 198.39 0.01632 CS CZE l Elevator Support Structure 77 342.32 0.01632 l CS CZE l Intemal Stiffener 78 356.675 0.07 CS CZE I Platform at EL 16.162'-l 79 798.7 0.00756 CS CZE l l Elevator Stair Tower 80 241.18 7.01632 CS CZE l Platform at El.178'-0 81 90.2 0.00756 CS CZE l HVAC Ring Duct 82 788.8 0.0087 CS NONE l Crane Girder 83 494.9 0.13 CS CZE l West Mid Quadrant l Integrated Head Stand (I side) 84 491.6 0.021 CS NONE l Platform at El.147'-3 85 457.7 0.00756 CS CZE O Revision: 5 February 29,1996 6.2-64 { West lg)ghouse
-5 &4.
- 6. Engineered Safety Features f
I ! v l Table 6.2.1.1-5 (Sheet 5 of 7) HEAT SINK PROPERTIES Metallic Heat Sink Exposed Description Number Area (ftr) Thickness (ft) Material Paint l Stairs at El.135'-3 to El.147*-3 86 136.8 0.011 CS NONE l Hydrogen Recombiner 87 204.75 0.0325 CS CZE l Containment Recirculation Unit 88 700.28 0.004 CS CZE l Internal Stiffener 89 356.675 0.07 CS CZ l Platforms at El.149'-7 90 1095.9 0.00756 CS CZE l Walkway El.162' 1 91 798.7 0.00756 CS CZE l Platforms at El.162'-l 92 1089.7 0.00756 CS CZE l Containment Recirculation Unit 93 265.9 0.004 CS CZE l HVAC Ring Duct 94 788.8 0.0087 CS NONE l Crane Girder 95 494.9 0.13 CS CZE l South Mid Quadrant l Internal Stiffener 96 356.675 0.07 CS CZE l Walkway El.162'-l 97 798.7 0.00756 CS CZE l HVAC Ring Duct 98 788.8 0.0087 CS NONE l Crane Girder 99 494.9 0.13 CS CZE l East Mid Quadrant l Hydrogen Recombiner 10'. 204.75 0.0325 CS CZE l Corainment Recirculation Unit 101 700.28 0.004 CS CZE l Internal 5tiffener 102 356.675 0.07 CS CZE l Platforms at El.149'-7 103 1095.9 0.00756 CS CZE l Walkway El.162'-l 104 798.7 0.00756 CS CZE l Platforms at El.162'-l 105 1089.7 0.00756 CS CZE l Containment Recirculation Unit 106 265.9 0.004 CS CZE l HVAC Ring Duct 107 788.8 0.0087 CS NONE l Crane Girder 108 494.9 0.13 CS CZE ,~ h w ./ - Revision: 5 Y W8Stingh00Se 6.2-65 February 29,1996
- 6. Engineered Safety Frtures l
e Table 6.2.1.1-5 (Sheet 6 of 7) HEAT SINK PROPERTIES Metallic Heat Sink Exposed Description Number Area (ft') Thickness (ft) Material Paint l North Outer Quadrant l Platform at EL 135'-3 109 183.3 0.00756 CS CZE l Internal Stiffener i10 402.1 0.007 CS CZE l Platform at El.162*-1 111 824.5 0.00756 CS CZE l HVAC Ring Duct 112 788.8 0.0087 CS CZE l Platform at El.178'-0 113 24.6 0.00756 CS CZE I Crane Girder 114 1101.1 0.13 CS CZE l West Outer Quadrant l Platform - Condensate Return Grating 115 824.35 0.00756 CS CZE l Internal Stiffener 116 402.1 0.07 CS CZE l Platform at El.162'-l 117 824.5 0.00756 CS CZE l HVAC Ring Duct 118 788.8 0.0087 CS NONE l Crane Girder 119 1101.1 0.13 CS CZE I South Outer Quadrant l Platform - Condensate Retum Grating 120 274.7 0.00756 CS CZE l Internal Stiffener 121 402.1 0.07 CS CZE l Platform at El.162'-l 122 824.5 0.00756 CS CZE l HVAC Ring Duct 123 788.8 0.0087 CS NONE l Crane Girder 124 1101.1 0.13 CS CZE l East Outer Quadrant l Main Equipment Hatch (1 side) 125 307.6 0.25 CS CZE l Intemal Stiffener 126 402.1 0.07 CS CZE l Platform at El.162'-l 127 824.5 0.00756 CS CZE l Main Equipment Hatch (1 side) 128 307.6 0.22 CS CZE l HVAC Ring Duct 129 788.8 0.0087 CS NONE l Crane Girder 130 1101.1 0.13 CS CZE I l Polar Crane Bridge and Motor Trolley 131 1996.0 0.145 CS CZE Revision: 5 February 29,1996 6.2-66 [ W85tiligh0USS
l
- 6. Engineered Safety Features
( ) (
'v' l
Table 6.2.1.1-5 (Sheet 7 of 7) HEAT SINK PROPERTIES Metallic Heat Sink Exposed l Description Number Area (ft') Thickness (ft) Material Paint l l North Inner Quadrant l l Polar Crane Bridae and Motor Trolley 132 1996.0 0.145 CS CZE l West Inner Quadrant Motor l _ Polar Crane Bridae and Meter Trollev 133 1996.0 0.145 CS C7F l South Quadrant Inner l Polar Crane Bridee and Motor Trollev 134 1906.0 0.145 CS C7F l Air Baffle / Containment Gap l Stairs; El. 242*-6 to 256'-4 142 157.662 0.011 CS NONE l Landings; El. 246*-4 to El. 253'-8 143 73.8 0.00756 CS CZE l Platforme E1. 241'-0 144 5207.0 0.00756 CS C7E l Air Baffle / Shield Building Gap l Platform Support Structure Beams and Hangers 151 1204.8 0.01632 CS CZ l Platform at El. 239'-0 152 6955.65 0.00756 CS CZE ('~ ! Platform Sunoort 153 14f 4.14 0.01632 CS CZE , k.s_) I l Lower Chimney l ' Support Columns 154 277.f1 0.0225 CS CZE l Stairs 155 110.24 0.011 CS NONE l Platform at El. 261'-0 156 142 106 0.00756 CS CZE l Top Chimney l Jon Chimney at El. 298'-6 157 78 925 0 00756 CS CZE l Notes to Table 6.2.1.15
- 1. Two types of materials will be used for metallic structures:
A. Carbon Steel (CS) B. Stainless Steel (SS)
- 2. Three coatings are used on metallic structures:
A. Epoxy (E) B. Carbo Zinc (CZ) C. Hot Dip Galvanizing ! l When Carbo Zinc and Epoxy are used together, the Carbo Zinc is applied as a primer then the Epoxy is applied as a topcoat. In addition, miscellaneous carbon steel items such as stairs, gratings, ladders, icilings, conduit, ducting and cable trays are hot l dip galvanized. In containment analysis, the galvanized structures will be treated as cat $on steel. ! 3. Thickness of paint: ! A. Epoxy: 4-8 Mils ! l B. Carbo Zinc: 2.5-6 Mils
.v I Revision: 5
[ W65tingh00S8 6.2-67 February 29,1996
- 6. Engineered Safety Fcctures l
Table 6.2.1.1-6 (Sheet I of 4) e CONCRETE HEAT SINKS Exposed Liner Liner Concrete Struct. Surface Area Thick Plate Plate Paint Paint Surface Number (ft2) (ft) Interior Exterior Interior Exterior REACTOR CAVITY l WALL 1 (2 sides) 1 219.50 4.00 CS CS CZE CZE l WALL 2 (2 sides) 2 693.00 4.00 CS CS CZE CZ l WALL 3 (2 sides) 3 218.28 4.00 CS CS CZE CZ l ' ROOF 1 (I side) 4 385.43 2.00 CS CS CZE NONE l WALL 4 (1 side) 5 164.50 4.00 CS CS CZE NONE l WALL 5 (1 side) 6 661.50 4.00 CS CS CZE CZ l WALL 6 (1 side) 7 488.35 4.00 CS CS CZE CZ I WALL 7 (1 side) 8 488.35 4.00 CS CS CZE CZ l WALL 8 (1 side) 9 214.50 2.00 CS CS CZE CZE l BULK 10 2429.00 4.00 CS NONE CZE NONE South East Accumulator Cavity l WALL 5V (I side) 11 245.46 4.00 SS CS NONE CZ l ROOF 3 (1 side) 12 718.50 2.00 CS NONE CZ E l BULK 13 2118.55 4.00 CS NONE CZE NONE North East Accumulator Cavity l WALL 8V (I side) 14 375.10 4.00 CS NONE CZ E I ROOF 4 (1 side) 15 993.00 2.00 CS NONE CZ E l BULK 16 2816.80 4.00 CS NONE CZE NONE East Steam Generator Room l WALL 1 (I side) 17 519.73 2.00 CS CS CZE CZE l WALL 2 (1 side) 18 472.94 2.00 CS CS CZE CZE l WALL 3 (I side) 19 1977.7 2.00 CS CS CZE CZE l WALL 5 (1 side) 20 758.16 4.00 SS CS NONE CZ l BULK 21 941.44 4.00 CS NONE E NONE O l Revision: 5 l February 29,1996 6.2-68 3 W95tlfigh00S8
I
- 6. Engineered Safety Fxtures
,Cp/ l Table 6.2.1.1-6 (Sheet 2 of 4) CONCRE'1E HEAT SINKS Exposed Liner Liner Concrete Struct. Surface Area Thick Plate Plate Paint Paint Surface Number (ft 2) (ft) Interior Exterior Interior Exterior West Steam Generator Room ! l WALL 1 (1 side) 22 707.33 2.00 CS CS CZE CZE l WALL 2 (1 side) 23 1626.37 2.00 CS SS CZ NONE l WALL 3 (1 side) 24 758.16 4.00 SS CS NONE CZ l BULK 25 1986.6 4.00 CS NONE CZE NONE CMT & CVS ROOM l ROOF 1 (1 side) 26 567.00 2.00 CS NONE CZE E l WALL 1 (1 side) 27 1721.28 2.00 CS SS CZ NONE l WALL 2 (1 side) 28 1630.00 4.00 SS CS NONE CZ ,,e l ROOF 2 (1 side) 29 533.4 2.00 CS NONE CZ E
\) . l ROOF 3 (1 side) 30 2472.78 2.00 CS NONE CZ E l ROOF 4 (1 side) 31 865.62 2.00 CS NONE CZ E l ROOF 5 (1 side) 32 631.92 2.00 CS NONE CZ E l ROOF 6 (1 side) 33 157.32 2.00 CS CS CZ CZ l WALL 3 (1 side) 38 262.50 2.00 CS CS CZE CZE l ROOF 11 (2 sides) 39 346.46 2.00 CS CS CZ CZ l BULK 40 5161.52 4.00 NONE NONE E NONE REFUELING ROOM l INT (2 sides) 41 776.37 2.00 SS NONE NONE NONE l WALL 22V (1 side) 42 1032.00 4.00 SS SS NONE NONE l WALL 23V (1 side) 43 673.99 4.00 SS CS NONE CZ l BULK 44 1539.00 4.00 SS NONE NONE NONE l
O d Revision: 5 g Westinghouse 6.2-69 February 29,1996
- 6. Engineered Safety FGctIrcs l
e Table 6.2.1.1-6 (Sheet 3 of 4) CONCRETE HEAT SINKS Exposed Liner Liner Concrete Struct. Surface Area Thick Plate Plate Paint Paint Surface Number 2 (ft ) (ft) Interior Exterior Interior Exterior IRWST ROOM l ROOF 6 (1 side) 45 569.78 2.00 SS NONE NONE E I ROOF 7 (1 side) 46 780.6 2.00 SS NONE NONE E l BULK 53 3585.20 4.00 SS NONE NONE NONE l UPPER EAST SG COMPARTMENT l WALLI (1 side) 54 216.75 2.0 CS CS CZE CZE l WALL 2 (1 side) 55 346.48 2.0 CS CS CZE CZE l WALL 3 (1 side) 56 346.48 2.0 CS CS CZE CZE l WALL 4 (1 side) 57 261.38 2.0 CS CS CZE CZE l WALL 5 (1 side) 58 70.13 2.0 CS CS CZE CZE l UPPER WEST SG COMPARTMENT l WALLI (1 side) 59 70.13 2.0 CS CS CZE CZE l WALL 2 (1 side) 60 346.48 2.0 CS CS CZE CZE l WALL 3 (1 side) 61 216.75 2.0 CS CS CZE CZE l WALL 4 (1 side) 62 345.48 2.0 CS CS CZE CZE l WALL 5 (1 side) 63 261.38 2.0 CS CS l WALLI (2 sides) 64 491.66 2.0 SS SS CZE CZE l NORTH INNER HALL l WALLI (2 sides) 65 1281.375 2.0 CS CS CZE CZE l WALLI (2 sides) 65 1005.0 2.0 CS CS CZE CZE El.135' - 3" to El.148'-0* l WEST MID QUADRANT l WALLI (2 sides) 67 491.66 2.0 SS SS NONE NONE e Revision: 5 February 29,1996 6.2-70 [ Westingh0USS
=-n
- 6. Engineered Safety Fectures A
I Table 6.2.1.1-6 (Sheet 4 of 4) CONCRETE HEAT SINKS Exposed Liner Liner Concrete Struct. Surface Area Thick Plate Plate Paint Paint Surface Number 2 (ft ) (ft) Interior Exterior Interior Exterior l AIR BAFFLE & l SHIELD BUILDING l UPPER GAP l ROOFI (1 side) 75 5227.65 3.0 CONCRETE I LOWER CHIMNEY l WALL 1 (2 sides) 76 13357.184 3.0 CONCRETE l FLOORI (2 sides) 77 2202.65 2.0 CONCRETE I ROOMI (2 sides) 78 1272.0 2.0 CONCRETE I FLOORI (2 sides) 79 23545.8 2.0 CONCRETE Notes: (,) (j l 1. ' Steel liner plates are either: A. Carbon Steel (CS) B. Stainless Steel (SS)
- 2. Paint on liner plates and concrete surfaces is escher.
A. Epoxy (E) l B. Carbo Zine (CZ) l When Carbo Zine and Epoxy are used together, the Carbo Zine is applied as a primer then the Epoxy is applied as a topcoat.
- 3. Thickness of paint is:
A. Epoxy: 4-8 Mils. I B. Carbo Zine: 2.5-6 Mils.
- 4. Floors at Elevations 107'-2 and 135'-3 have not steel liner plare.
- 5. Steel liner plates are assumed to be 1/2 inch thick in aU places.
G l Revision: 5 l E W8Stingh00Se 6.2-71 February 29,1996 l
- 6. Engineered Safety Features l
Table 6.2.1.1-7 (Sheet 1 of 4) e CONTAINMENT SHELL AND BAFFLE METAL PROPERTIES Metallic Heat Sink Exposed Thickness Description Number Area (ft 2) (ft) Material Paint STACK #1 l PCS @ 240.5' 1-a-1 (1 side) 395.63 0.13542 CS CZ I Baffle @ 240.5' l-b-1 (I side) 1284.00 0.00706 CS CZ l PCS @ 224.75' l-a-2 (1 side) 644.34 0.13542 CS CZ l Baftle @ 224.75' 1-b-2 (1 side) 595.10 0.00706 CS CZ l PCS @ 209' l-a-3 (1 side) 1653.63 0.13542 CS CZ l Baffle @ 209' 1-b-3 (1 side) 1804.57 0.00706 CS CZ l PCS @ 189.5' 1-a-4 (1 side) 980.29 0.13542 CS CZ l Baffle @ 189.5' l-b-4 (1 side) 997.00 0.00706 CS CZ l PCS @ 170' l-a-5 (1 side) 809.80 0.13542 CS CZ l Baffle @ 170' 1-b-5 (1 side) 823.61 0.00706 CS CZ l PCS @ 148' l-a-6 (1 side) 639.32 0.13542 CS CZ l Baffle @ 148' 1-b-6 (1 side) 650.22 0.00706 CS CZ . I PCS @ 135.25' 1-a-7 (1 side) 511.45 0.13542 CS CZ l Baffle @ 135.25' 1-b-7 (1 side) 520.18 0.00706 CS CZ STACK #2 l PCS @ 240.5' 2-a-1 (1 side) 395.63 0.13542 CS CZ l Baffle @ 240.5' 2-b-1 (1 side) 1284.00 0.00706 CS CZ l PCS @ 224.75' 2-a-2 (1 side) 644.34 0.13542 CS CZ l Baffle @ 224.75' 2-b-2 (1 side) 595.10 0.00706 CS CZ l PCS @ 209' 2-a-3 (1 side) 1653.63 0.13542 CS CZ l Baffle @ 209' 2-b-3 (1 side) 1804.57 0.00706 CS CZ l PCS @ 189.5' 2-a-4 (1 side) 980.29 0.13542 CS CZ l Baffle @ 189.5' 2-b-4 (1 side) 997.00 0.00706 CS CZ l PCS @ 170' 2-a-5 (1 side) 809.80 0.13542 CS CZ l Baffle @ 170' 2-b-5 (1 side) 823.61 0.00706 CS CZ l PCS @ 148' 2-a-6 (1 side) 639.32 0.13542 CS CZ l Baffle @ 148' 2-b-6 (1 side) 650.22 0.00706 CS CZ l PCS @ 135.25' 2-a-7 (1 side) 511.45 0.13542 CS CZ l Baffle @ 135.25' 2-b-7 (1 side) 520.18 0.00706 CS CZ O Revision: 5 February 29,1996 6.2-72 3 W85tillgl10USS
i l l
' 6. Engineered Safety Features p.3 lC j
l Table 6.2.1.1-7 (Sheet 2 of 4) CONTAINMENT SHELL AND BAFFLE METAL PROPERTIES
- Metallic Heat Sink Exposed Thickness
! Description Number Area (ft2) (ft) Material Paint STACK #3 l l l PCS @ 240.5' 3-a-1 (1 side) 395.63 0.13542 CS CZ t l Baffle @ 240.5' 3-b-1 (1 side) 1284.00 0.00706 CS CZ l PCS @ 224.75' 3-a-2 (1 side) 644.34 0.13542 CS CZ l Baffle @ 224.75' 3-b-2 (1 side) 595.10 0.00706 CS CZ l PCS @ 209' 3-a-3 (1 side) 1653.63 0.13542 CS CZ l Baffle @ 209' 3-b-3 (1 side) 1804.57 0.00'706 CS CZ l l PCS @ 189.5' 3-a-4 (1 side) 980.29 0.13542 CS CZ l Baffle @ 189.5' 3-b-4(1 side) 997.00 0.00706 CS CZ l PCS @ 170' 3-a-5 (1 side) 809.80 0.13542 CS CZ l Baffle @ 170' 3-b-5 (1 side) 823.61 0.00706 CS CZ l PCS @ 148' 3-a-6 (1 side) 639.32 0.13542 CS CZ ,s Ci)
' I Baffle @ 148' 3-b-6 (1 side) 650.22 0.007 % CS CZ l PCS @ 135.25' 3-a-7 (1 side) 511.45 0.13542 CS CZ l Baffle @ 135.25' 3-b-7 (1 side) 520.18 0.00706 CS CZ STACK #4 l PCS @ 240.5' 4-a-1 (1 side) 395.63 0.13542 CS CZ l Baffle @ 240.5' 4-b-1 (1 side) 1284.00 0.00706 CS CZ l PCS @ 224.75' 4-a-2 (1 side) 644.34 0.13542 CS CZ l Baffle @ 224.75' 4-b-2 (1 side) 595.10 0.00706 CS CZ l PCS @ 209' 4-a-3 (I side) 1653.63 0.13542 CS CZ l Baffle @ 209' 4-b-3 (1 side) 1804.57 0.00706 CS CZ l PCS @ 189.5' 4-a-4 (1 side) 980.29 0.13542 CS CZ l Baffle @ 189.5' 4-b-4 (1 side) 997.00 0.00706 CS CZ l PCS @ 170' 4-a-5 (1 side) 809.80 0.13542 CS CZ l Baffle @ 170' 4-b-5 (1 side) 823.61 0.00706 CS CZ l PCS @ 148' 4-a-6 (1 side) 639.32 0.13542 CS CZ l Baffle @ 148' 4-b-6 (1 side) 650.22 0.00706 CS CZ l PCS @ 135.25' 4-a-7 (1 side) 511.45 0.13542 CS CZ l Baffle @ 135.25' 4-b-7 (1 side) 520.18 0.00706 CS CZ ,9
__/ Revision: 5 T Westinghouse 6.2-73 February 29,1996
j 6. E;gineered Saf.ty Features l e Table 6.2.1.1-7 (Sheet 3 of 4) CONTAINMENT SHELL AND BAFFLE METAL PROPERTIES Metallic Heat Sink Exposed Thickness Description Nuniber 2 Area (ft ) (ft) Material Paint STACK #5 I PCS @ 240.5' 5-a-1 (1 side) 593.44 0.13542 CS CZ l Baffle @ 240.5' 5-b-1 (1 side) 1926.00 0.00706 CS CZ l PCS @ 224.75' 5-a-2 (1 side) 966.50 0.13542 CS CZ l Baffle @ 224.75' 5-b-2 (I side) 892.66 0.00706 CS CZ l PCS @ 209' 5-a-3 (1 side) 466.41 0.13542 CS CZ l . Baffle @ 209' 5-b-3 (1 side) 508.98 0.00706 CS CZ l PCS @ 189.5' 5-a-4 (1 side) 1150.77 0.13542 CS CZ l Baffle @ 189.5' 5-b-4 (1 side) 1170.40 0.00706 CS CZ l PCS @ 170' 5-a-5 (1 side) 1321.26 0.13542 CS CZ l Baffle @ 170' 5-b-5 (1 side) 1343.79 0.00706 CS CZ l PCS @ 148' 5-a-6 (1 side) 1491.74 0.13542 CS CZ l Baffle @ 148' 5-b-6 (1 side) 1517.18 0.00706 CS CZ l PCS @ 135.25' 5-a-7 (1 side) 1619.61 0.13542 CS CZ l Baffle @ 135.25' 5-b-7 (I side) 1647.22 0.00706 CS CZ STACK #6 i PCS @ 240.5' 6-a-1 (1 side) 593.44 0.13542 CS CZ l Baffle @ 240.5' 6-b-1 (1 side) 1926.00 0.00706 CS CZ l PCS @ 224.75' 6-a-2 (1 side) 966.50 0.13542 CS CZ l Baffle @ 224.75' 6-b-2 (1 side) 892.66 0.00706 CS CZ ! l PCS @ 209' 6-a-3 (1 side) 466.41 0.13542 CS CZ l Baffle @ 209' 6-b-3 (1 side) 508.98 0.00706 CS CZ l PCS @ 189.5' 6-a-4 (1 side) 1150.77 0.13542 CS CZ l Baffle @ 189.5' 6-b-4 (I side) 1170.40 0.00706 CS CZ l PCS @ 170' 6-a-5 (1 side) 1321.26 0.13542 CS CZ l Baffle @ 170' 6-b-5 (1 side) 1343.79 0.00706 CS CZ l PCS @ 148' 6-a-6 (I side) 1491.74 0.13542 CS CZ l Baffle @ 148' 6-b-6 (1 side) 1517.18 0.00706 CS CZ ! l PCS @ 135.25' 6-a-7 (1 side) 1619.61 0.13542 CS CZ l Baffle @ 135.25' 6-b-7 (1 side) 1647.22 0.00706 CS CZ ! Revision: 5' l February 29,1996 6.2-74 3 W85tiflgh0US8 l
l l r;;- -g 6.' Engineered Safety Features [ % bi o l l Table 6.2.1.1-7 (Sheet 4 of 4) CONTAINMENT SHELL AND BAFFLE METAL PROPERTIES i Metallic Heat Sink Exposed Thickness Description Number Area (ft') (ft) Material Paint STACK #7 l PCS @ 240.5' 7-a-1 (1 side) 593.44 0.13542 CS CZ l Baffle @ 240.5' 7-b-1 (1 side) 1926.00 0.00706 CS CZ l PCS @ 224.75' 7-a-2 (1 side) 966.50 0.13542 CS CZ l Baffle @ 224.75' 7-b-2 (1 side) 892.66 0..m6 CS CZ I l PCS @ 209' 7-a-3 (1 side) 466.41 0.13542 CS CZ I Baffle @ 209' 7-b-3 (1 side) 508.98 0.00706 CS CZ l PCS @ 189.5' 7-a4 (1 side) 1150.77 0.13542 CS CZ l Baffle @ 189.5' 7-b-4 (1 side) 1170.40 0.007 % CS CZ l PCS @ 170' 7-a-5 (1 side) 1321.26 0.13542 CS CZ l l Baffle @ 170' 7-b-5 (1 side) 1343.79 0.00706 CS CZ l PCS @ 148' 7-a-6 (1 side) 1491.74 0.13542 CS CZ [ l Baffle @ 148' 7-b-6 (1 side) 1517.18 0.00706 ' CS CZ l PCS @ 135.25' 7-a-7 (1 side) 1619.61 0.13542 CS CZ l Baffle @ 135.25' 7-b-7 (1 side) 1647.22 0.00706 CS CZ STACK #8 l PCS @ 240.5' 8-a-1 (1 side) 593.44 0.13542 CS CZ l Baffle @ 240.5' 8-b-1 (1 side) 1926.00 0.00706 CS CZ l PCS @ 224.75' 8-a-2 (1 side) 966.50 0.13542 CS CZ l Baffle @ 224.75' 8-b-2 (1 side) 892.66 0.00706 CS CZ l PCS @ 209' 8-a-3 (1 side) 466.41 0.13542 CS CZ l Baffle @ 209' 8-b-3 (1 side) 508.98 0.00706 CS CZ l PCS @ 189.5' ~ 8-a-4 (1 side) 1150.77 0.13542 CS CZ l Baffle @ 189.5' 8-b-4 (1 side) 1170.40 0.00706 CS CZ l PCS @ 170' 8-a-5 (1 side) 1321.26 0.13542 CS CZ l Baffle @ 170' 8-b-5 (1 side) 1343.79 0.00706 CS CZ l PCS @ 148' 8-a-6 (1 side) -1491.74 0.13542 CS CZ f l Baffle @ 148' 8-b-6 (1 side) 1517.18 0.00706 CS CZ l PCS @ 135.25' 8-a-7 (1 side) 1647.22 0.13542 CS CZ l Baffle @ 135.25' 8-b-7 (1 side) 1728.41 0.00706 CS CZ
.g-Revision: 5 T Westinghouse 6.2-75 February 29,1996
1
- =: l
- 6. Engineered Safety Features rm-O, l
Table 6.2.1.1-8 (Shtet 1 of 3) SHIELD BUILDING CONCRETE HEAT SINK PROPERTIES Exposed Liner Liner Concrete Struct. Surface Thick Plate Plate Paint Paint Surface Number Area (ft2) (ft) Interior Exterior Interior Exterior STACK #1 l Shield @ 240.5' ' 1-c-1 (1 side) 1284.0 3 CONCRETE l Shield @ 224.75' 1-c-2 (1 side) 1162.18 3 CONCRETE l Shield @ 209' 1-c-3 (1 side) 1895.06 3 CONCRETE I Shield @ 189.5' l-c-4 (1 side) 1046.15 3 CONCRETE I Shield @ 170' l-e-5 (1 side) 864.21 3 CONCRETE I Shield @ 148' 1-c-6 (1 side) 682.27 3 CONCRETE I Shield @ 135.25' 1-c-7 (1 side) 545.82 3 CONCRETE STACK #2 l Shield @ 240.5' 2-c-1 (1 side) 1284.0 3 CONCRETE I Shield @ 224.75' 2-c-2 (1 side) 1162.18 3 CONCRETE l Shield @ 209' 2-c-3 (1 side) 1895.06 3 CONCRETE I Shield @ 189.5' 2-c-4 (1 side) 1046.15 3 CONCRETE l Shield @ 170' 2-c-5 (1 side) 864.21 3 CONCRETE l Shield @ 148' 2-c-6 (1 side) 682.27 3 CONCRETE l Shield @ 135.25' 2-c-7 (1 side) 545.82 3 CONCRETE STACK #3 l Shield @ 240.5' 3-c-1 (1 side) 1284.0 3 CONCRETE l Shield @ 224.75' 3-c-2 (I side) 1162.18 3 CONCRETE l Shield @ 209' 3-c-3 (1 side) 1895.06 3 CONCRETE l Shield @ 189.5' 3-c-4 (1 side) 1G46.15 3 CONCRETE l Shield @ 170' 3-c-5 (1 side) 864.21 3 CONCRETE I Shield @ 148' 3-c-6 (1 side) 682.27 3 CONCRETE l Shield @ 135.25' 3-c-7 (1 side) 545.82 3 CONCRETF O Revision: 5 February 29,1996 6.2-76 3 W85tillgh0058
- 6. Engineered Safety Features V
l Table 6.2.1.1-8 (Sheet 2 of 3) SIIIELD BUILDING CONCRETE IIEAT SINK PROPERTIES Exposed Liner Liner Concrete Struct. Surface Thick Plate Surface Plate Paint Paint Number Area (ft') (ft) Interior Exterior Interior Exterior STACK #4 l Shield @ 240.5' 4-c-1 (1 side) 1284.0 3 CONCRETE I Shield @ 224.75' 4-c-2 (1 side) 1162.18 3 CONCRETE l Shield @ 209' 4-c-3 (1 side) 1895.06 3 CONCRETE l Shield @ 189.5' 4-c-4 (1 side) 1046.15 3 CONCRETE I Shield @ 170* 4-c-5 (1 side) 864.21 3 CONCRETE I Shield @ 148' 4-c-6 (I side) 682.27 3 CONCRETE l Shield @ 135.25' 4-c-7 (1 side) 545.82 3 CONCRETE STACKe-I Shield @ 240.5' 5-c-1 (1 side) 1926.0 3 CONCRETE I Shield @ 224.75' 5-c-2 (1 side) 1743.26 3 CONCRETE G y l Shield @ 209' 5-c-3 (1 side) 534.51 3 CONCRETE I Shield @ 189.5' 5-c-4 (1 side) 1228.08 3 CONCRETE I Shield @ 170' 5-c-5 (1 side) 1410.02 3 CONCRETE l Shield @ 148' 5-c-6 (1 side) 1591.96 3 CONCRETE l Shield @ 135.25' 5-e-7 (I side) 1728.41 3 CONCRETE STACK #6 l Shield @ 240.5' 6-c-1 (1 side) 1926.0 3 CONCRETE l Shield @ 224.75' 6-c-2 (1 side) 1743.26 3 CONCRETE I Shield @ 209' 6-c-3 (I side) 534.51 3 CONCRETE I Shield @ 189.5' 6-c-4 (1 side) 1228.08 3 CONCRETE l Shield @ 170' 6-c-5 (1 side) 1410.02 3 CONCRETE I Shield @ 148' 6-c-6 (1 side) 1591.96 3 CONCREl'E I Shield @ 135.25' 6-c-7 (1 side) 1728.41 3 CONCRETE v Revision: 5 3 W8Stingh00Se 6.2-77 February 29,1996
- 6. E:rgineered Saf;ty Fe tures l
e Table 6.2.1.1-8 (Sheet 3 of 3) SHIELD BUILDING CONCRETE HEAT SINK PROPERTIES Exposed Liner Liner Concrete Stnact. Surface Thick Plate Plate Paint Paint Surface Number Area (ft') (ft) Interior Exterior Interior Exterior STACK #7 l Shield @ 240.5' 7-c-1 (I side) 1926.0 3 CONCRETE I Shield @ 224.75' 7-c-2 (1 side) 1743.26 3 CONCRETE l Shield @ 209' 7-c-3 (1 side) 534.50 3 CONCRETE l Shield @ 189.5* 7-c-4 (1 side) 1228.08 3 CONCRETE l Shield @ 170' 7-c-5 (1 side) 1410.02 3 CONCRETE I Shield @ 148' 7-c-6 (1 side) 1591.96 3 CONCRETE l Shield @ 135.25' 7-c-7 (1 side) 1728.41 3 CONCRETE STACK #8 l Shield @ 240.5' 8-c-1 (1 side) 1926.0 3 CONCRETE l Shield @ 224.75' 8-c-2 (1 side) 1743.26 3 CONCRETE l Shield @ 209' 8-c-3 (1 side) 534.50 3 CONCRETE l Shield @ 189.5' 8-c-4 (1 side) 1228.08 3 CONCRETE I Shield @ 170* 8-c-5 (1 side) 1410.02 3 CONCRETE l Shield @ 148' 8<-6 (1 side) 1591.96 3 CONCRETE l Shield @ 135.25' 8-c-7 (1 side) 1589.69 3 CONCRETE O Revision: 5 February 29,1996 6.2-78 3 Westingh00S8
...sww.
- 6. Engineered Safety Features
- 4 I Table 6.2.1.1-9 PHYSICAL PROPERTIES OF PASSIVE HEAT SINKS Thermal Density Conductivity Specific Heat Dry Wet Material (Ibm /ft') (Btu /hr-ft *F) (Btu /lbm *F) Emis. Emis. l Epoxy 105 0.1875 0.35 0.81 0.95 l Carbon Steel 490.7 30 0.107 0.81 0.95 l Concrete 140. 0.83 0.19 0.81 0.95 l Stainless Steel 501. 9.4 0.12 0.81 0.95 l Carbo Zine 207.5 1.21 0.15 0.81 0.95 l Oxidized Carbo Zine 207.5 0.302 0.15 0.81 0.95 f% n LJ Revision: 5 3 W85tiflghouse 6.2-79 February 29,1996
Y
- 6. E:gineered Safsty Fe-tures 1
I Table 6.2.1.1 10 CONTAINMENT SIIELL TEMPERATURE PROFILES FOR COLD LEG BREAK I O [ Westinghouse Proprietary] [Provided under separate cover] l O Revision: 5 February 29,1996 6.2-80 W. Westinghouse
- 4manmutt:
- 6. Engineered Safety Features y
a lV) l Table 6.2.1.2-1 (Sheet 1 of 3) i l FLOW PATH DATA, HOOP FLOW FOR liOT LEG BREAK { l Hydr. D Length Flow A Equi. L A/A I Element K Factor F Factor inertial (ft.) (ft.) (sq. ft.) (ft.) l 1 4.1190E01 2.2000E02 83620E+00 3.7500E+00 3.8500Ev01 2.5700E+00 05.0000D01 l 2 4.0810E-01 2.2000E-02 83830E+00 5.6345E+00 6.9375E+01 2.6830E+00 5.0000E-01
! 3 5.4470E-01 22000E02 7.1220E+00 3.1920E+00 53150E+01 1.5640E+00 4.2600E-01 l 4 5.4460E-01 2.2000E02 6.6890E+00 3.1920E+00 53150E+01 1.4950E+00 4.2600E-01 I 5 4.0810E01 2.2000 5 02 83830E+00 5.6345E+00 6.9375E+01 2.6830E+00 5.0000E01 l 6 5.1290E-01 22000E02 8.1590E+00 3.8800E+00 3.2215E+01 2.7450E400 5.0000E-01 l 7 5.1360 5 01 2.2000E-02 8.1530E+00 5.1073E+00 5.7590E+01 2.7300E+00 5.0000 5 01 l 8 5.1630E01 2.2000E-02 8.1260E+00 4.8350E+00 53950E+01 2.7090E+00 5.0000E-01 l 9 5.1630E-01 22000E02 8.1260E+00 4.8350E+00 53950E+0! 2.7090E+00 5.0000E-01 l 10 5.1360E01 2.2000E-02 8.1530E+00 5.1073E+00 5.7590E+01 2.7300E+00 5.0000E01 j l 11 3.7930E-01 2 2000 5 02 1.0208E+01 3.9200E+00 33060E+0! 3.7540E+00 5.0000 5 01 L l 12 53690E01 22000E02 83560E+00 4.8090E+00 5.0445E+01 2.6560E+00 4.2620E.01 l 13 5.0000E01 2.2000E02 93120E+00 4.9700E+00 5.5735E+01 3.0780E+00 5.0000E01 l 14 5.0000E01 2.2000002 93120E+00 4.9700E+00 5.5735E+01 3.0780E+00 5.0000E01 l 15 53690E-01 2.2000E-02 8.5560E+00 4.R090E+00 5.0445E+01 2.6560E+00 4.2620E-01
! 16 1.1412E+00 2.2000E-02 6.5374E+00 52376E+00 3hl79E+0! 63750E+00 7.0670E01 l 17 1.1342E+00 2.2000E-02 6.6923E+00 6.9942E+00 7.1688E+0! 63750E+00 7.1060E01 l 18 1.1833E+00 2.2000E02 6.7260E+00 6.1880E+00 6.8308E+01 63750E+00 6.128050t l 19 1.1833E+00 2.2000E02 6.7260E+00 6.1880E+00 6.8308E+01 63750E+00 6.1280E-01 l 20 1.1342E+00 2.2000 5 02 6.6923E+00 6.9942E+00 7.1688E+01 63750E+00 7.1060E-01 l 21 .0000E+00 22000E02 7.0000E+00 1.0790E+01 1.4997E+02 7.0000E+00 1.0000E+00 l 22 13200E+00 2.20005 02 23665E+01 13430E+01 2.0460E+02 23665E+01 1.0000E+00 l 23 4.1400E01 2.2000E02 1.ll50E+01 1.6000E+01 4.0000E+02 1.0270E+01 1.0000E+00
,s
! ) LJ Revision: 5 T Westinghouse 6.2-81 February 29,1996
i
. . ,i
- 6. Engineered S^fety Fmtures l
Table 6.2.1.2-1 (Sheet 2 of 3) Gll l l FLOW PATH DATA, HOOP FLOW FOR HOT LEG BREAK l Hydr D Length Flow A Equi. L A/A l Element K-Factor F-Factor Inertial (ft.) (ft.) (sq. ft.) (ft.) l l 24 1.9790E40 2.2000E-02 1.1090E+01 1.0790E+01 1.4'97E+02 3.7900E+00 2.9000E-01 l 25 4.9380E-01 2.2000E-02 1.2814E+01 7.5456E+00 2.8032E+02 3.1736E+00 4.4960 6 01 l 26 5.1760 5 01 2.2000 5 02 1.1199E+01 7.2000E+00 2.4370E+02 3.4905E+00 4.6280E-01 l 27 7.9770E-01 2.2000 6 02 2.5879E+00 1.Il52E+00 5.8391E+00 2.Oll8E+00 1.000E+00 l 28 8.5900E01 2.2000E02 2.1460E+00 5.4220E01 3.4600E+00 2.0010E+00 1.0000E+00 l 29 7.9770E-01 2.2000 6 02 2.5879E+00 1.Il52E+00 5.8391E+00 2.Ull8E+00 1.0000E+00 l 30 5.6550E-01 2.2000 6 02 8.1013E+00 1.8143E+00 9.7682E+00 7.5290E+00 1.0740E-01 l 31 5.6630E-01 22000E02 8.0931E+00 1.7337E+00 6.5121E+00 7.5260E+00 1.0590E01 l 32 5.6550E-01 2.2000 6 02 8.1013E+00 1.8143E+00 9.7682E+00 7.5290E+00 1.0740E-01 l 33 5.6550 & O1 2.2000E02 8.1013E+00 1.8143E+00 9.7682E+00 7.5290E+00 1.0740E-01 l 34 5.6630 & O! 2.2000 5 02 8.0931E+00 1.7337E+00 6.5121E+00 7.5260E+00 1.0590501 l 35 5.6550E01 2.2000E02 8.1013E+00 1.8143E+00 9.7682E+00 7.5290E+00 1.0740E-01 l 36 9.7500 6 01 2.2000E-02 2. ISO 9E+00 1.2830E+00 5.8391E+00 2.0001E+00 1.0000E+00 l 37 9.8500 5 01 2.2000 & O2 2.0890E+00 5.4220 5 01 3.4600E+00 2.0000E+00 1.000 3 00 l 38 9.7500E-01 2.2000E-02 2.1509E+00 1.2830E+00 5.8391E+00 2.0001E+00 1.0000E+00 l 39 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 I 40 9.6900 6 01 2.2000 6 02 5.5970E+00 9.4770E-01 j 2.7680E+00 5.5000E+00 1.0000E+00 l 41 9.9930E-01 2.2000E-02 7.0150E+00 2.5000E+00 4.9087E+00 2.0000E+00 1.0000E+00 l 42 9.9930E-01 2.2000E-02 2.0150E+00 2.5000E40 4.9087E+00 2.0000E+00 1.0000E+00 l 43 9.9930&Ol 2.2000E-02 2.0150E+00 2.5000E+00 4.9087Fa00 2.0000E+00 1.0000E+00 l 44 9.9930E-01 2.2000E02 2.0150E400 2.5000E+00 4.9087E+00 2.0000E+00 1.0000E+00 l 45 1.4580E400 2.2000E02 4.6650E+00 4.8000E+00 3.6000E+01 4.0000E+00 5.7700 5 02 O Revision: 5 February 29,1996 6.2-82 [ Westinghnlise
- 6. Engineered Safety F::ctures
(~~) i ! V l Table 6.2.1.2-1 (Sheet 3 of 3) 1 I FLOW PATII DATA,IlOOP FLOW FOR IIOT LEG BREAK l Hydr. D Length Flow A Equi. L A/A l Element K. Factor F-Factor Inertial (ft.) (fL) (sq. ft.) (ft.) l 46 1.4580E+00 2.2000 6 02 4.6650E+00 4.800nE+00 3.6000E+01 4.0000E+00 5.7700E-02 l 47 8.8090E01 2.2000E02 7.8533E+00 2.7106E+00 1.4916E+01 7.5033E+00 1.0000E+00 l 48 9.1970E-Ol 2.2000 5 02 7.7355E+00 1.5289E+00 9.9463E+00 7.5014E+00 1.0000E+00 l 49 8.8090E01 2.2000 6 02 7.8533E+00 2.7106E+00 1.4916E+01 7.5033E+00 1.0000E+00 l 50 8.8090 & O! 2.2000 6 02 7.8533E+00 2.7106E+00 1.4916E401 7.5033E+00 1.0000E+00 l 51 9.1970E-01 2.2000 6 02 7.7355E+00 2.5289E+00 9.9463E+00 7.5014E+00 1.0000E+00 l 52 8.8090E01 2.2000 & O2 7.8533E+00 2.7106E+00 1.4919E+01 7.5033E+00 1.0000E+00 l 53 1.2570E+00 2.2000 & O2 4.2400E+00 4.1680E+00 2.5365E+01 1.2900E+00 1.5400E-01 I 54 3.0000 6 02 2.20006 02 7.1300E+00 1.2270E401 1.6464E+02 6.0280E+00 8.2600E-01 l 55 9.9300E-01 2.2000 6 02 5.9950E+00 63740E+00 6.8000E+01 2.5630E+00 3.9200&OI (S I 56 1.4800E+00 2.2000E-02 4.5440E+00 i
- 1.0000E+01 1.0000E+02 4.0000E+00 8.1400E.02
(/ # l 57 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 I l 58 1.0790E+00 22000E-02 7.8460E+00 9.4400E+00 3.4560E+02 63790E+00 73900E-01 l 59 4.2800E01 2.2000 6 02 1.ll20E+01 82300E+00 1.6670E+02 9.2400E+00 43400&Ol l 60 1.4340E+00 2.20006 02 33450E+00 13400E+00 1.6840E+01 2.0250E+00 1.0100E-01 l 61 1.0000E+00 2.2000E-02 6.1875E+00 137(X)E+00 83350E+01 9.7000E-01 5.0000E-01 l 62 2.0000E+00 2.2000E02 3.2400E+00 13700E+00 83350E+01 7.0600E-0 1.0000E+00 l 63 .0000E+00 2.2000E02 13041E+01 5.4740E+00 3.4680E+01 13041E+0! 1.0000E+00 l 64 2.4360E+00 2.2000E-02 1.5179E+01 5.7670E+00 4.1332E+01 13665E+01 3.3700 & O2 l 65 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 66 1.5000E+00 2.2000E02 6.S000E+00 5 7000E-02 93380E+01 5 3000 & 01 5.0000&OI
! 67 1.5000E+00 2.2000E-02 5.1900E-01 3,9960E+00 8.8800E-01 4.1700&Ol 1.0700E-02 l 68 1.7590E+00 22000E-02 1.4420E+01 4.9080E+00 3.1069E+01 13920E401 1.1900E-01 l
l
,/ 3
- ) i
%.J l Revision: 5 l
[ W85tingh00S8 6.2-83 February 29,1996 i
- 6. Engineered Safety Features l Table 6.2.1.2-2 (Sheet 1 of 4) e l
l FLOW PATH DATA, RADIAL FLOW FOR IIOT LEG BREAK l Hydr.D I Length Flow A Equl. L A/A I Elernent K. Factor F Factor Inertial (ft.) (ft.) (sq. ft.) (ft.) l 1 1.4980E-01 2.2000E-02 6.2581E+00 1.7535E+0! 3.1650E+02 4.7392E+00 1.0000E+00 l 2 2.2000E-02 2.2000E-02 1.3632E+01 1.0126E+01 2.0683E+02 1.3051E+01 1.0000E+00 I 3 .0000E+00 .0000E+00 .0000E+00 1.000E+00 1.0000E400 .0000E+00 .0000E+00 l 4 3.7800E-02 2.2000E-02 9.0000E+00 1.7535E+0! 3.1650E+02 9.0000E+00 1.0000E+00 l 5 2.2000E-02 2.2000E-02 1.3632E+01 1.0126E+01 2.0683E+02 1.3051E+01 1.0000E+00 l 6 6 8100E-02 2.2000E-02 7.0000E+00 7.8606E+00 9.3030F+01 5.6729E+00 1.0000E+00 l 7 1.0660E-01 2.2000E-02 1.5556E+01 5.7860E+00 5.6183E+01 1.4936E+01 7.6240E-01 l 8 1.3952E+00 2.2000E-02 3.6450E+00 4.1706E+00 1.8580E+01 2.5419E+00 1.5310E-01 l 9 1.7130E-01 2.2000E-02 1.1000E+01 7.8606E+00 9.3030E+01 1.1000E+01 7.6890E-01 l 10 1.0660E-01 2.2000E-02 1.5556E+01 5.7860E+00 5.6183E+01 1.4936E+01 7.6240E-01 l 11 1.2470E-01 2.2000E-02 6.7931E+00 1.2908E+01 1.9311E+02 5.2768E+00 1.0000E+00 l 12 3.7750E-01 2.2000E-02 1.5556E+01 8.2534E+00 1.0592E+02 1.4856E+01 6.0790E-01 l 13 0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 14 2.7060E01 2.2000E02 1.1000E+01 1.2908E401 1.9311E+02 1.1000E+01 6.7290E01 l 15 3.7750E-01 2.2000E-02 1.5556E+01 8.2534E+00 1.0592E+02 1.4856E+01 6.0790E-01 l 16 2.2280E-01 2.2000E-02 6.3333E+00 8.9074E+00 9.1444E+01 4.6778E+00 - 9.5630E-01 l 17 5.0680E-01 2.2000E-02 1.4500E+01 6.3812E+00 5.6854E+01 1.4500E+01 4.9550E-01 l 18 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E400 .0000E+00 .0000E+00 l 19 3.3180E-01 2.2000E-02 1.2000E+01 1.0062E+0! 1.1104E+02 1.2000E+01 6.2210E-01 l 20 5.0680E-01 2.2000E-02 1.4500E+01 6.3812E+00 5.6854E+01 !.4500E+01 4.9550E01 l 21 1.9790E+00 2.2000E-02 1.1090E+01 1.0790E+01 1.4997E+02 3.7900E+00 2.9000E01 1 22 .0000E+00 .0000E+00 .0000E400 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 e Revision: 5 February 19,1996 6.2 84 3 Westiflgh0US8
me
- 6. Engineered Safety Featt:res LJ ,
I Table 6.2.1.2-2 (Sheet 2 of 4) l l FLOW PATH DATA, RADIAL FLOW FOR HOT LEG BREAK l Hydr. D l Iength Flow A Equi. L A/A l Element K Factor F. Factor Inertial (ft.) (ft.) (sq. ft.) (ft.) I 23 1.3400E-01 2.2000E-02 1.9300E+01 1.0790E+01 1.4997E+02 1.4381E+01 7.3300E-01 l 24 .0000E+00 .0000E400 .0000E+00 1.0000E+00 1.0000E+00 .0000E400 .0000E+00 l l 25 l
.0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 )
l 26 2.6110E+00 2.2000E-02 5.3720E+00 4.1706E+00 1.8580E+01 2.7800E+00 2.5000E-02 l 27 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E400 l l 28 9.54 /0E-01 2.2000E-02 2.1070E+00 5.4220E-Ol 3.4600E+00 2.0000E+00 1.0000E+00 l 29 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 'X)00E+00 I l 30 3.6990E01 2.2000E02 1.4000E+01 6.1747E+00 3.9098E+01 1.4000E+01 5.9240E01 l 31 2.5000E-01 2.2000E-02 6.9988E+00 5.9448E+00 (- 3.6650E+01 5.0528E+00 1.0000E+00 (v) l 32 3.6990E-01 2.2000E-02 1.4000E+01 6.1747E+00 3.9098E+01 1.4000E+0! 5.9240E-01 I 33 2.S000E-01 2.2000E-02 6.9988E+00 5.9448E+00 3.6650E+01 5.0528E+00 5.0000E-01 l 34 4.3370E-01 2.2000E-02 2.2653E400 9.6430E-01 6.2800E+00 2.0073E+00 1.3260E-01 l 35 2.5000E-Ol 2.2000E-02 6.9988E+00 5.9448B+00 3.6650E+01 5.0528E+00 5.0000E-01 l l 36 .0000E+00 .0000E+00 .0000E400 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 37 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l l 38 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 39 .0000E+00 .0000E+00 .0000E400 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 40 9.9960E-01 2.2000E-02 5.6416E+00 9.4790E-01 2.7686E+00 5.5005E+00 1.0000E+00 l 41 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 42 8.9500E-01 2.2000E-02 2.1978E+00 2.5000E+00 4.9087E+00 2.0029E+00 1.0000E+00 l 43 8.9500E-01 2.2000E-02 2.1978E400 2.S000E+00 4.9087E+00 2.0029E+00 1.0000E+00
.I 44 8.9500E-01 2.2000E-02 2.1978E+00 2.5000E+00 4.9087E+00 2.0029E+00 1.0000E+00 f)
- Revision: 5 Y W8Stingh0US8 6.2-85 February 29,1996
- 6. EEgineered Safsty Features l
e Table 6.2.1.2-2 (Sheet 3 of 4) l I FLOW PATH DATA, RADIAL FLOW FOR HOT LEG BREAK l Hydr.D l Length Flow A Equi. L A/A l Eternent K-Factor F-Factor Inertial (ft.) (ft.) (sq. ft.) (ft.) l 45 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 46 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 47 .0000E+00 2.2000E-02 9.4248E+00 2.8638E+00 2.3745E+01 9.4248E+00 1.0000E+00 l 48 .0000E+00 2.2000E-02 7.8540E+00 2.86353+00 2.3745E+01 7.8540E+00 1.0000E+00 l 49 .0000E400 2.2000E-02 9.4248E+00 2.8638E+00 2.3745E+01 9.4248E+00 1.0000E40 l 50 .0000E+00 2.2000E-02 7.8540E+00 2.8638E+00 2.3745E+0! 7.8540E+00 1.0000E+00 l 51 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 52 .0000E+00 2.2000E02 7.8540B+00 2.8638E+00 2.3745E+01 7.8540E+00 1.0000E+00 l 53 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 54 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 55 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 56 .0000E400 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 57 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 58 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 59 4.3700E-01 2.2000E-02 6.3200E+00 7.5900E+00 6.1998E+01 1.9400E+00 3.8700E-01 l 60 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 61 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 62 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E400 .0000E+00 l 63 2.4360E+00 2.2000E-02 1.$179E+01 5.7670E+00 4.1332E+01 1.3665E+01 3.3700B-02 l 64 .0000E400 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 65 .0000E+00 .0000Ev 1 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 66 1.5000E+00 2.206 12 1.0070E+00 1.4600E+00 4.1330E+01 4.2200E-01 1.1070E-01 I 67 1.S000E+00 2.2000E-02 1.2360E+00 1.5300E+00 3.7334E+01 4.2800B-01 1.0000E-01 Revision: S February 29,1996 6.2-86 [ W85tingh0US8
_ . . . - . . . . . . ~ . -- .- _ _ _ _ . _ . - - _ . - . - . _ - . . . l l
- 6. Engineered Safety Fettures f3FFQ ~
Pr ' _ )q !n lV I Table 6.2.1.2-2 (Sheet 4 of 4)
- I I FLOW PATH DATA, RADIAL FLOW FOR HOT LEG BREAK l l Hydr. D Length Flow A Equi. L A/A I Element K-Factor F. Factor krtial (ft.) (ft.) (sq. ft.) (ft.)
I 68 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 69 .0000E+00 .0000E400 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 I 70 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000B+00 .0000E+00 l 71 1.5000E+00 2.2000E-02 4.9700E-01 8.5700E-01 3.0030E+00 4.1680E-01 1.0000E-02 l 72 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 73 .0000E+00 .0M0E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 i i l l N ! (O Revision: 5 [ W85tingh00S8 6.2-87 February 29,1996
.4ti iw p 6. Engineered Saf;ty Frtures l
e Table 6.2.1.2-3 (Sheet 1 of 4) I l FLOW PATH DATA, AXIAL FLOW FOR IIOT LEG BREAK l Hydr.D l Length Flow A Equi. L A/A l Element K-Factor F-Factor Inertial (ft.) (ft.) (sq. ft.) (ft.) l 1 1.4980E-01 2.2000E-02 6.2581E+00 1.7535E+01 3.1650E+02 4.7395E+00 1.0000E+00 l 2 4.8460E-01 2.2000E-02 2.2614E+00 1.ll52E+00 5.8391E+00 2.0007E+00 3.0700E-02 l 3 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 4 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 5 4.8460E-01 2.2000E-02 2.2614E+00 1.ll52E+00 5.8391E+00 2.0007E+00 3.0700E-02 l 6 6.8100E-02 2.2000E-02 7.0000E+00 7.8606E+00 9.3030E+01 5.6729E+00 1.0000E+00 l 7 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 8 .0000E+00 .0000E 40 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 9 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .000C3+00 l 10 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 11 1.2470E-01 2.2000E-02 6.7931E+00 1.2908E41 1.9311E+02 5.2768E+00 1.0000E+00 l 12 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E4 .0000E+00 .0000E+00 l 13 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 14 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 15 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 I 16 2.2280E-01 2.2000E-02 6.3333E+00 8.9074E+00 9.1444E+01 4.6778E40 9.5630E01 l 17 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 18 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 19 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 20 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E 40 .0000E+00 l 21 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 22 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l e Revision: 5 February 29,1996 6.2 48 [ Westiligh0tlSe
, . - . - -. - - - - - --- - -~^
l l p,- m l
- 6. Engineered Safety Features -1
[ j
- _ l h l Table 6.2.1.2-3 (Sheet 2 of 4)
I ! I FLOW PATH DATA, AXIAL FLOW FOR HOT LEG BREAK l l Hydr. D l l Length Flow A Equi. L A/A l Element K Factor F-Factor Inertial (ft.) (ft.) (sq. ft.) (ft.) l 23 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 24 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000Ed)0 .0000E+00 .0000E+00 l 25 .0000E+00 .00005?+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+ 00.0000E+0 l 26 3.4510E+00 2.2000E-02 7.8400E+00 3.2020E+00 1.0680E+01 3.5220E+00 1.8700E-02 l 27 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 3.5220E+00 1.8700E-02 l 28 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 29 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 30 4.7300E-01 2.2000E-02 2.1978E+00 2.5000E+00 4.9087E+00 2.0029E+00 5.3900E02 l 31 2.5000E-01 2.2000E-02 6.9988E+00 5.9448E+00 3.6650E+0! 5.0528E+00 1.0000EM)0 l 32 4.8430E-01 2.2000E-02 7.1416E+00 9.4790E-01 2.7686E+00 7.0005E+00 3.1500E-02 l 33 4.4660E-01 2.2000E-02 2.5879E+00 1.ll52E+00 5.8391E+00 2.0118E+00 1.0690E-01 l 34 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 35 4.4660E-01 2.2000E-02 2.5879E+00 1.1152E+00 5.8391E+00 2.0ll8E+00 1.0690E-01 l 36 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 37 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 38 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 39 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 40 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.00GOE+00 .0000E+00 .0000E+00 l 41 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 42 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 43 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 44 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l lA U Revision: 5 [ W9Stingh00S8 6.2-89 February 29,1996
- 6. Engineered Safity Feat:res l
Table 6.2.1.2-3 (Sheet 3 of 4) e t. [ l FLOW PATH DATA, AXIAL FLOW FOR HOT LEG BREAK l Hydr.D l Length Flow A Equi. L A/A I Element K Factor F. Factor Inertial (ft.) (ft.) (sq. ft.) (ft.) l 45 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 46 .0000E+00 .0000E+00 .0000C.00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 47 .0000E+00 .0000E+00 .0000'C+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 48 .0000E+00 2.2000E02 7.8540E+00 2.8638E+00 2.3745E+01 7.8540E+00 1.0000E+00 l 49 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l T .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 51 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 52 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 53 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 54 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 55 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 56 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 57 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 58 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 I 59 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 60 .0000E+00 .0000E400 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 61 .0000E+00 .0000E+00 .C000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l l 62 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 63 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 64 .0000E+00 .0000E+00 .0000E+00 1,0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 65 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 66 2.8600E-01 2.2000E-02 1.0090E+01 5.8220E+00 3.4460E+01 6.0400E+00 3.8800E-01 I e l "evision: 5 L February 29,1996 6.2-90 [ W65tlagh00S8
. . . . - . _ . ~ . ~ - . . - . . - _ - - - . ~ _-- -. . - . . . . . . . . . _ _
- 6. Engineered Safity F::etures
[f'73] [ y l Table 6.2.1.2-3 (Sheet 4 of 4) ! I I FLOW PATH DATA, AXIAL FLOW FOR HOT LEG BREAK l t Hydr.D l l Length Flow .A Equl. L A/A ; l Element . K-Factor F Factor Inertial (ft.) (ft.) (sq. ft.) (ft.) l 67 2.3100E-01 2.2000E-02 1.0090E+01 4.5090E+00 2.0668E+01 6.2060E+00 3.8800E-01 f l 68 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 69 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 70 .0000E+00 .0000E+00 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 71 1.5000E+00 2.2000E-02 6.2300E-01 9.3900E-01 7.6670E+00 4.1720E-01 2.5700E-02 l 72 .0000E+00 .0000E+00 .0000E @ 1.0000E+00 1.0000E+00 .0000E+00 .0000E+00 l 73 .0000E+00 .0000E400 .0000E+00 1.0000E+00 1.0000E+00 .0000E+00 0000+00 l 'l Il M Resision: 5 [ W96tingh00S8 6.2-91 February 29,1996
- 6. Engineered Stfety FGctures l
e Table 6.2.1.2-4 (Sheet 1 of 3) TMD MODEL NODE INFORMATION l Volume Steam Pressure Air Pressure Temperature l Element Number (cu. ft.) (psig) (psla) (F) 1 1.5557E+03 1.6930E+00 1.2607E+01 1.2000E+02 2 2.8594E+03 1.6930E+00 1.2607E+01 1.2000E+02 3 2.6862E+03 1.6930E400 1.2607E+01 1.2000E+02 4 2.6862E+03 1.6930E+00 1.2607E+01 1.2000E+02 5 2.8594E+03 1.6930E+00 1.2607E+01 1.2000E+02 6 5.9219E+02 1.6930E+00 1.2607E+01 1.2000E+02 7 1.0616E+03 1.6930E+00 1.2607E+01 1.2000E+02 8 1.0068E+03 1.6930E+00 1.2607E+01 1.2000E+02 9 1.0068E+03 1.6930E+00 1.2607E+01 1.2000E+02 10 1.0616E+03 1.6930E+00 1.2607E+01 1.2000E+02 11 1.3554E+03 1.6930E+00 1.2607E+01 1.2000E+02 12 2.4261E+03 1.6930E+00 1.2607E+01 1.2000E+02 13 2.2853E+03 1.593E+00 1.2607E+01 1.2000E+02 14 2.2853+03 1.6930E+00 1.2607E+01 1.2000E+02 15 2.4261E+03 1.6930E+00 1.2607E+01 1.2000E+02 16 6.4581E+02 1.6930E+00 1.2607E+01 1.2000E+02 17 1.2633E+03 1.6930E+00 1.2607E+01 1.2000E+02 18 1.3951E+03 1.6930E+00 1.2607E+01 1.2000E+02 19 1.3951E+03 1.6930E+00 1.2607E+01 1.2000E+02 20 1.2633E+03 1.6930E+00 1.2607E401 1.2000E+02 21 6.9962E+02 1.693E+00 1.260l+01 1.2000E+02 22 1.4997E+03 1.6930E+00 1.2607E+01 1.2000E+02 23 8.1840E+03 1.6930E+00 1.2607E+01 1.2000E+02 24 6.9962E+02 1.6930E+00 1.2607E4 01 1.2000E+02 e Revision: 5 F bruary 29,1996 6.2-92 W Westirighouse
I' t 14ttetit.::
- 6. Engineered Safety F:stures -
~
\
(Q l l Table 6.2.1.2-4 (Sheet 2 of 3) TMD MODEL NODE INFORMATION l Volume Steam Pressure Air Pressure Temperature I Element Number (cu. ft.) (psig) (psia) (F) 25 1.2647E+04 1.6930E+00 :.2607b+01 1.2000E+02 26 1.5507E+G4 1.6930E+00 1.2607E+01 1.2000E+02 27 2.3350E+01 1.6930E+00 1.2607E401 1.2000E+02 28 1.3820E+01 1.6930E+00 1.2607E+01 1.2000E+02 29 2.3350E+O 1.6930E+00 1.2607E+01 1.2000E+A2 30 6.3298E+02 1.6930E+00 1.2607E+01 1.2000E+02 31 2.4932E+02 1.6930E+00 1.2607E+01 1.2000E+02 32 6.3298E+02 1.6930E+00 1.2607E+01 1.2000E+02 33 6.3298E402 1.6930E+00 1.2607E+01 1.2000E+02 34 2.4932+02 1.6930E+00 1.2607E+01 1.2000E+02 35 6.3298E+02 1.6930E+00 1.2607E+01 1.2000E+02 36 2.3350E+01 1.6930E+00 1.2607E+01 1.2000E+02 37 1.3820E+0! 1.6930E+00 1.2607E+01 1.2000E402 38 2.3350E+01 1.6930E+00 1.2607E+01 1.2000E+02 39 3.8760E+01 1.6930E+00 1.2607E+01 1.2000E+02 40 3.0454E+01 1.6930E400 1.2607E+01 1.2000E+02 41 1.%30E+01 1.6930E+00 1.2607E401 1.2000E+02 42 1.%30E+01 1.6930E+00 1.2607E+01 1.2000E402 43 1.9630E+01 1.6930E+00 1.2607E+01 1.2000E+02 44 1.9630E+01 1.6930E+00 1.2607E+01 1.2000E+02 45 1.4443E+04 1.6930E+00 1.2607E+01 1.2000E402 46 1.1821E+04 1.6930E+00 1.2607E+0! 12000E+02 47 2.2385E+02 1.6930E+00 1.2607E+01 1.2000E+02 48 1.4924E+02 1.6930E+00 1.2607E+01 1.2000E+02 i l f) j\ l Revision: 5 T Westirighouse 6.2-93 February 29,1996
- 6. Engineered Safety Features i
Table 6.2.1.2-4 (Sheet 3 of 3) e TMD MODEL NODE INFORMATION l Volume Steam Pressure Air Premare Temperature i Element Namber (cu. ft.) (psia) (psia) (19 49 2.2385E+02 1.6930E+00 1.2607E+01 1.2000E+02 50 2.2385E+02 1.6930E+00 1.2607E+01 1.2000E+02 51 1.492E+02 1.6930E+00 1.2607E+01 1.2000E+02 52 2.2385E+02 1.6930E+00 1.2607E+01 1.2000E+C2 53 2.0080E+03 1.6930E400 1.2607E+01 1.2000E+02 54 1.0180E+03 1.6930E+00 1.2607E+01 1.2000E+02 55 1.7157E+03 1.6930E+00 1.2607E+01 1.2000E+02 56 1.4937E+05 1.6930E+00 1.2607E+01 1.2000E+02 57 1.1266E+06 1.6930E+00 1.2607E+01 1.2000E+02 58 5.9627E+03 1.6930E+00 1.2607E+01 1.2000E+02 59 3.6489E+03 1.6930E400 1.2607E+01 1.2000E+02 60 2.9590E+03 1.6930E+00 1.2607E+01 1.2000E+02 61 2.1393E+03 1.6930E+00 1.2607E+01 1.2000E+02 62 1.6532E403 1.6930E+00 1.2607E+01 1.2000E+02 63 3.5835E+02 1.6930E+00 1.2607E+01 1.2000E+02 64 4.9121E+02 1.6930E+00 1.2607E+01 1.2000E+02 65 6.0183E+02 1.6930E400 1.2607E+01 1.2000E+02 66 2.3814E+03 1.6930E+00 1.2607E+01 1.2000E+02 67 1.4394EM)3 1.6930E+00 1.2607E+01 1.2000E+02 68 2.2659E+03 1.6930E+00 1.2607E+01 1.2000E+02 69 1.8610E+03 1.6930E400 1.2607E+01 1.2000E+02 70 1.8612E+03 1.6930E400 1.2607E+01 1.2000E402 71 1.195E+03 1.6930E+00 1.2607E+01 1.2000E+02 1 72 1.7413E+03 1.6930E400 1.2607E401 1.2000E+02 73 4.4269E+092 1.6930E+00 1.2607E+01 1.2000E+02 l l l O ! Revision: 5 l February 29,1996 6.2-94 [ Westillgh0US8 1
xansmx
- 6. Engineered Safity Features ID G l Table 6.2.1.2-5 TMD MODEL NODE MAXIMUM DIFFERENTIAL PRESSURES RELATIVE TO NODE 1 (40% PRESSURE MARGIN NOT INCLUDED IN THESE VALUES) I Max AP (psi) Max AP (psi) Max AP (psi)
Node # Hot Leg Cold Leg Node # Hot Leg Cold Leg Node # Hot Leg Cold Iag 4 1 - - 21 1.M 0.96 41 1.05 1.06 2 0.87 0.64 22 1.05 1.02 42 1.05 1.08 3 1.04 0.86 23 1.05 1.08 43 1.05 1.08 4 1.04 1.03 24 1.05 1.08 44 1.05 1.08 5 0.87 0.96 25 1.05 1.08 45 1.05 1.08 6 0.95 0.97 26 1.05 1.08 46 1.05 1.08 7 1.M 0.86 27 1.00 1.03 47 1.05 1.08 8 1.05 1.N 28 0.60 0.82 48 , 1.05 1.08 j 9 1.05 1.07 29 1.00 O 0.47 49 .~ 1.05 1.08 V 10 1.N 1.06 30 1.05 1.03 50 1.05 1.08 l 11 1.05 1.08 31 1.03 1.05 51 1.05 1.08 12 1.05 1.07 32 1.05 1.08 52 1.05 1.08 13 1.05 1.08 33 1.05 1.08 53 1.05 1.08 14 1.05 1.08 34 1.05 1.08 54 1.05 1.08 15 1.05 1.08 35 1.05 1.08 55 1.05 1.08 j 16 1.05 1.08 36 1.05 1.08 56 1.05 1.08 i 17 1.05 1.08 37 I 1.05 1.08 57 1.05 1.08 j 18 1.05 1.08 38 1.05 1.08 58 1.05 1.08 ! 19 1.05 1.08 39 1.05 1.08 I-20 1.05 1.08 40 1.05 1.08 L Revision: 5 [ W85tingh00S8 6.2-95 February 29,1996
l
- 6. Engineered Saf;ty Fxtures I Table 6.2.1.2-6 RESULTS FOR MSLB IN NODE 56 MAXIMUM PRESSURE DIFFERENTIALS BETWEEN ELEMENT 56 AND ALL OTIIER ELEMENTS l Press Press Press Press l Elem Time Diff Elem Time Diff Elem Time Diff Elem Time DIN l 1 .243 .4606 2 .243 .4523 3 .243 .4691 4 .243 .4713 l 5 .243 .4660 6 .243 .4882 7 .243 .4867 .243 8 .4784 l 9 .243 .4904 10 .243 .4896 11 .243 .5233 12 .243 .5230 l 13 .243 .5217 14 .243 .5239 15 .243 .5229 16 .233 .5539 l 17 .233 .5548 18 .233 .5533 19 .233 .5537 20 .233 .5547 l 21 .229 .2682 22 .233 .2257 23 .023 .1353 24 .070 .2/.93 l 25 .244 .4563 26 .243 .5066 27 .243 .4410 28 .243 .4324 l 29 .243 .4323 30 .243 .4065 31 .243 .4058 32 .243 .4060 l 33 .243 .4072 34 .243 .4080 35 .243 .4073 36 .243 .4308 I 37 .243 .4200 38 .243 .4308 39 .068 .4210 40 .033 .2361 l 41 .243 .4600 42 .243 .4892 43 .243 .4897 44 .243 .4898 l 45 .049 .2821 46 .045 .2 643 47 .056 .3693 48 .056 .3693
- l. 49 .056 .3698 50 .056 .3699 51 .056 .3696 52 .056 .3696 l 53 .278 .3395 54 .035 .2454 55 .032 .2200 56 .000 .0000 l 57 .201 .5916 '58 .243 .5503 59 .064 .3149 60 .059 .3402 l 61 .229 .5593 62 .201 .5726 63 .068 .2282 64 .074 .1341 I 65 .085 .4306 66 .089 .4308 67 .086 .4145 68 .067 .3538 I 69 .068 .3628 70 .069 .3669 71 .087 .4226 72 .088 .4322 l 73 .045 .0654 O
Revision: 5 February 29,1996 6.2-96 T Westinghouse
ii= n
- 6. E*gineered Satty Frtures
!n
'%.)
l Table 6.2.1.2-7 RESULTS FOR BREAK IN PRESSURIZER VALVE ROOM (NODE 59) MAXIMUM PRESSURE DIFFERENTIALS BETWEEN ELEMENT 59 AND ALL OTIIER ELEMENTS l Press Press Press Press l Elem Time Diff Elem Time Diff Elem Time Diff Elem Time Diff I 1 .068 2.2458 2 .068 2.2438 3 .069 2.2461 4 .069 2.2472 I 5 .%9 2.2471 6 .%9 2.2480 7 .069 2.2469 8 .M9 2.2389 I 9 .069 2.2456 10 .069 2.2486 .069 11 2.2499 12 .069 2.2496 l 13 .069 2.2476 14 .069 2.2485 i5 .069 2.2500 16 .069 2.2484 l 17 .069 2.2484 18 .069 2.2480 19 .069 2.2481 20 .069 2.2484 I 21 .068 2.2287 22 .068 2.2181 '!3 .068 2.1917 24 .067 2.1230 l 25 .067 2.0856 26 .069 2.1184 :!7 .068 2.2384 28 .%8 2.2383 l 29 .068 2.2364 30 .%8 2.2278 .'1 .068
+
2.2305 32 .068 2.2284 1 33 .068 2.2009 34 .068 2.1944 .;5 .068 2.2007 36 .068 2.1416 I 37 .068 2.1599 38 .068 2.1415 39 .068 2.2343 40 .068 2.2198 l 41 .068 2.2377 42 .068 2.2380 43 .068 2.2245 44 .068 2.2244 I 45 .068 2.2394 46 .068 2.2367 47 .068 2.2346 48 .068 2.2357 fS I 49 .068 2.2347 50 .068 2.2295 51 .068 2.2281 52 .068 2.2295
! ! l 53 .068 2.2369 54 .068 2.2201 55 .068 2.2133 56 .%8 2.1926 I 57 .069 2.2468 58 .069 2.2072 59 .000 .0000 60 .011 1.0869 i 61 .079 2.0213 62 .074 2.0373 63 .008 .6942 64 .014 1.3688 l 65 .069 2.2519 66 .069 2.2510 67 .069 2.2497 68 .069 2.2471 l 69 .%9 2.2484 70 .069 2.2489 71 .069 2.2511 72 .069 2.2516 I 73 .068 2.2024 i ~g (V )
l Revision: 5 [ WOStingh0US8 6.2-97 February 29,1996
rem--
- 6. Engineered Scfety Fectures i
l Table 6.2.1.2-8 O l I RESULTS FOR 3" HL BREAK IN EAST STEAM GENERATOR COMP. (NODE 1) l 0 MAXIMUM PRESSURE DIFFERENTIALS BETWEEN I ELEMENT 1 AND ALL OTHER ELEMENTS I l Press Press Press Press l Elem Time DIN Elem Time Diff Elem Time DIN Elem Time Diff I 1 .000 .0000 2 .004 .8679 3 .005 1.0356 4 .005 1.0354 I 5 .004 .8642 6 .005 .9546 7 .005 1.0297 8 .005 1.0461 l l 9 .005 1.0461 10 .005 1.0295 11 .005 1.0451 12 .005 1.0466 I 13 .005 1.0468 14 .005 1.0468 15 .005 1.0466 16 .005 1.0468 l 17 .005 1.0468 18 .005 1.0468 19 .005 1.G468 20 .005 1.0468 l 21 .005 1.0199 22 .005 1.0454 23 .005 1.0468 24 .005 1.0468 l 25 .005 1.0168 26 .005 1.0468 27 .005 .9947 28 .003 .6293 1 29 .005 .995 30 .005 1.0461 31 .005 1.0335 32 .005 1.0461 I 33 .005 1.0468 34 .005 1.0468 35 .005 1.0468 36 .005 1.0468 I 37 .005 1.G168 38 .005 1.0468 39 .005 1.0468 40 .005 1.0168 l 41 .005 1.0467 42 .005 1.4167 43 .005 44 1.0468 .005 1.0468 l I 45 .005 1.0868 46 .005 1.0468 47 .005 1.0468 48 .005 1.G466 l 49 .005 1.0468 50 .005 1.0468 51 .005 1.0468 52 .005 1.0468 l I 53 .005 1.0168 54 .005 1.0468 55 .005 1.0468 56 .005 1.Gt68 l 57 .005 1.0468 58 .005 1.0468 59 .005 1.0468 60 .005 1.0468 I 61 .005 1.0468 62 .005 1.0468 63 .005 1.4568 64 .005 1.0468 I 65 .005 1.0468 66 .005 1.0468 67 .005 1.0468 68 .005 1.0468 I 69 .005 1.0168 70 .005 1.0168 71 .005 1.0468 72 .005 1.0468 I 73 .005 1.G468 l l 1 l Revision: 5 1 February 29,1996 6.2-98 3 W85tiligh00S8
ww 6 E:gineered Safety Features . n l l Table 6.2.1.2-9 RESULTS FOR 3" CL BREAK IN EAST STEAM GENERATOR COMP (NODE 2) 0 MAXIMUM PRESSURE DIFFERENTIALS BETWEEN ELEMENT 2 AND A'LL OTIIER ELEMENTS l Press Press Press Press I Elem Time din Elem Time Di# Elem Time din Elem Time din 1 .005 .5682 2 .000 .0000 3 .007 .7667 4 .008 .9032 5 .007 .8528 6 .007 .8670 7 .007 .7688 8 .008 .9136 9 .008 .9373 10 .008 .9276 11 .008 .9394 12 .008 .9344 13 .009 .9415 14 .009 .9425 15 .009 .9421 16 .009 .9425 17 .009 .9423 18 .009 .9426 19 .009 .9426 20 .009 .9426 21 .005 .6008 22 .007 .8722 23 .009 .940S 24 .009 .9424 25 .009 .9426 26 .009 .9426 27 .008 .9094 28 .006 .7522 29 .004 .4314 30 .008 .9140 31 .008 .9300 32 .009 .9415 33 .009 .9426 34 .009 .9425 35 .009 .9410 36 .009 .9426 37 .009 .9426 38 .009 .9423 39 .009 .9426 40 .008 .9328 41 .008 .9330 42 .009 .9424 43 .000 .9126 44 .009 .9422 45 .009 .9426 46 .009 .9426 47 .009 .9413 48 .009 .9421 49 .009 .9426 50 .009 .9426 51 .009 .9426 52 .009 .9425 () 53 57
.009
.009
.9426
.9426 54 58
.009
.009
.9426
.9426 55 59
.009 .9426 56 .009 .9426
.009 .9426 60 .009 .9426 61 .009 .9426 62 .009 .9426 63 .009 .9426 64 .009 .9426 65 .009 .9426 66 .009 .9426 67 .009 .9426 68 .009 .9426 69 .009 .9426 70 .009 .9426 71 .009 .9426 72 .009 .9426 73 .009 .9426
- s Revision: 5 l [ WBStilighouse 6.2 99 February 29,1996
r . 1 i t
- 6. E:gineered Safety F=tures !
i I i Table 6.2.1.3-1 O SHORT-TERM MASS AND ENERGY INPUTS i Vessel Outlet Temperature ('F) . . . . . . . . . . . . . . . . . . . . . . . . . . . ........ . . . . . . . . . . . . . . 597.0 Vessel Inlet Temperature (*F) . . . ................... ... .............. ..... .. . 528.6 Initial RCS Pressure (PSIA) ................................................2300.0 Zaloudek Coefficient (CK1) . . . . . . . . . . . . . . . . . . . . . . . . . . ................. .. ... . 1.018 Zaloudek Coefficient (C1) . . . . . . . . . . . . .. ...... .... . .. . ....... .... .. .. 0.9 O O Revision: 5 February 29,1996 6.2-100 W W85tingh00S8
I 70
- 6. Engineered Safety Fe:tures
'j l t . .hl O
V Table 6.2.1.3-2 SHORT TERM 3 INCH COLD LEG BREAK MASS AND ENERGY RELEASES Time Mass l Enein j (sec) (Ibm /sec) (Btu /sec) !. 0.0 0.0 0.0 l 0.001 3186.8 1.7084E+6 l 0.05 3186.8 1.7084E+6 l 1.000 3186.8 1.7084E+6 l 5.000 3186.8 1.6591E+6 l 7.000 3186.8 1.6225E+6 l 10.00 3186.8 1.6005E+6 LO l /'T (j Revision: 5 T Westinghouse 6.2-101 February 29,1996
- 6. Engi:eered Safety Features m __
Table 6.2.1.3-3 O SHORT. TERM 3 INCH HOT LEG BREAK MASS AND ENERGY RELEASES Time Mass Energy (sec) Obm/sec) (Btuisec) 0.0 0.0 0.0 l 0.001 2514.2 1.5623E+6 1 0.05 2514.2 1.5623E+6 l 1.000 2514.2 1.5640E+6 l 5.000 2514.2 1.6947E+6 l 7.000 2514.2 1.7966E M l 10.00 2514.2 1.8406E+6 O t O Revision: 5 February 29,1996 6.2-102 [ W85tingh00S9
- 6. Engineered Safety Fxtures s
LJ l Table 6.2.1.3-4 (Sheet 1 of 5) l I 2 SHORT TERM 1.0 FT MAIN STEAM LINE BREAK MASS AND ENERGY RELEASE l (BREAK IN COMPARTMENT 56; PIPE LENGTH = 100.0 FEET; 1 MAXIMUM TIME ASSUMPTION) I l Time Mass Energy l (sec) (Ibm /sec) (Btu /sec) l 0 0 531.05 l '0.05005 40981.47 531.05 I 0.10009 51582.38 531.06 I 0.15012 51914.07 531.26 l 0.20003 56324.09 531.69 l 0.25002 52193.14 531.8 l 0.30008 55751.79 532.27 l 0.35011 57991.96 532.5 I 0.40009 61908 532.92 l 0.45008 55661.48 532.94 l 0.50018 55396.21 533.19 1 0.5501 55799.85 533.36 l 0.60003 54505.29 533.57
-m l 0.65 55718.72 533.79 I
) l 0.70008 55560.81 533.97 I 0.75003 54278.71 534.05 I 0.80001 54863.3 534.35 1 0.85013 55133.12 534.59 l 0.90008 54317.39 534.78 l 0.95008 56022.1 535.13 1 1.00004 57188.84 535.55 i 1.10023 56590.82 536.24 l 1.20001 56403.73 537.07 l 1.3 57575.72 538.06 l 1.4 61043.11 538.06 l 1.5 58777.05 541.4 l 1.60004 58504.03 543.5 l 1.70002 57207.83 545.73 l 1.80018 55116.98 548.22 l 1.90003 53416.31 550.84
.I 2.00021 52195.53 553.49 l 2.10003 51119.43 556.25 l 2.20006 49678.89 559.07 I
'v' Revision: 5 T Westinghouse 6.2-103 February 29,1996
t
- : -_=__ :!.
- 6. FEgineered Safety Fcctures j sw --
l \ L l Table 6.2.1.3-4 (Sheet 2 of 5) O l I I SHORT TERM 1.0 FT2 MAIN STEAM LINE BREAK MASS AND ENERGY RELEASE i l (BREAK IN COMPARTMENT 56; PIPE LENGTH = 100.0 FEET; I l ' MAXIMUM TIME ASSUMPTION) l l Time Mass Energy I (sec) (Ibm /sec) (Btu /sec) I 2.30002 48656.4 561.59 l 2.4000 45212.87 563.94 l 2.50015 43459.59 565.89 1 2.60001 42131.39 567.77 l 2.70008 39891.28 569.9 I 2.80006 38753.61 571.77 I 2.90021 36262.46 573.9 l 3.00042 35552.01 575.36 1 3.10042 33986.79 576.85 I 3.20016 31634.74 578.54 I 3.3005 30089.11 580 l 3.40034 28141.72 581.74 l 3.50058 26615.49 583.42 I 3.6004 25359.76 585.02 1 3.70014 24380.64 ' 586.71 l 3.8003 23567.09 588.57 I 3.90024 22809.34 590.66 l 4.00033 22114.21 593.01 l 4.10004 21455.55 595.61 1 4.20027 21618.21 600.19 I 4.30039 20594.81 605.53 l l 4.40063 19988.71 610.29 I 4.50042 19239.33 611.04 l { 4.60017 18774.63 609.97 l 4.70017 18237.85 610.18
)
l 4.80048 17647.82 611.74 I 4.90038 17122.31 614.25 I 5.00008 16758.34 617.67 I 5.25057 15536.44 631.16 l 5.50006 14486.32 637.3 I 5.75009 13807.98 650.14 l 6.00029 13058.56 676.9 I 6.25027 13165.35 652.36 l 1 ' l L Revision: 5 February 29,1996 6.2-104 [ W95tingfl00S8
- 6. Ergineered Satty Features i
IR% !
/3 l
Table 6.2.1.3-4 (Sheet 3 of 5) I l 2 I SIIORT TERM 1.0 FT MAIN STEAM LINE BREAK MASS AND ENERGY RELEASE (BREAK IN COMPARTMENT 56; PIPE LENGTH = 100.0 FEET; I MAXIMUM TIME ASSUMPTION) l l Time Mass Energy I (sec) (Ibm /sec) (Btu /sec) I 6.50012 12507.45 655.4 I 6.75045 12365.93 638.72 l 7.00013 12253.41 628.34 l 7.25006 12334.64 618.73 l 7.50039 11509.17 644.64 I 7.75043 11775.96 615.22 l 8.00008 11674.54 606.9 I 8.25005 11216.94 615.66 l 8.50037 10459.3 641.56 I 8.75066 9975.453 671.02 l 9.0001 9755.574 657.37 l 9.25082 9640.006 651.94 I 9.50066 9493.376 640.31 1 9.75051 9252.547 648.42 l 10.00003 9069.055 648.55 l 10.25077 9143.009 624.93 l 10.50064 9211.633 605.15 l 10.75086 9301.821 591.84 l 11.0009 9380.36 585.87 l 11.25065 9379.361 584.17 l 11.50016 9300.834 585.7 l 11.75019 9142.256 589.39 l 12.00102 8898.41 594.36 l 12.25096 8892.234 604.59 l 12.500N 8480.493 616.05 l 12.7508 8179.785 628.66 l 13.00034 7882.542 635.05 i l Revision: 5 T Westingil00Se 6.2 105 February 29,1996
nnnumtr
- 6. E~gineered Saf ty Fectures l I
Table 6.2.1.3-4 (Sheet 4 of 5) ei l 8 l SIIORT TERM 1.0 FF MAIN STEAM LINE BREAK MASS AND ENERGY RELEASE I (BREAK IN COMPARTMENT 56; PIPE LENGT}i = 100.0 FEET; I MAXIMUM TIME ASSUMPTION) 1 I Time Mass Energy I (sec) (Ibm /sec) (Btu /sec) l 13.25039 7848.445 628.99 I 13.50008 7647.333 613.37 l 13.75066 7531.402 600.61 l 14.00103 7529.374 585.74 l 14.25131 7640.976 570.52 l 14.50035 7690.387 560.5 l 15.75015 7730.5 554.89 l 15.00028 7893.184 551.66 l 15.25174 7728.983 546.99 l 15.50094 7695.687 542.06 l 15.75006 7702.621 537.44 l 16.00055 7721.515 533.1 1 16.25114 7720.754 529.93 l 16.50039 7845.526 528.83 I 16.75008 7792.656 529.28 l 17.00015 7569.003 529.2 1 17.2509 7446.876 529.81 l 17.50015 7348.652 530.63 l 17.7509 7353.587 533.2 1 18.00035 7167.792 534.42
- l. 18.25022 6980.346 536.27 l 18.50029 6856.782 538.19 l 18.75026 6825.789 542.1 l 19.00113 6642.836 544.45 l 19.25092 6454.637 548.59 l 19.50153 6314.137 552.77 l 19.75094 6175.254 556.79 l
Revision: 5 e February 29,1996 6.2-106 3 Westingh00Se
- 6. Engineered Safety Features
)
{ \ l Table 6.2.1.3-4 (Sheet 5 of 5) l I 2 SHORT TERM 1.0 FT MAIN STEAM LINE BREAK MASS AND ENERGY RELEASE I (BREAK IN COMPARTMENT 56; PIPE LENGTII = 100.0 FEET; I MAXIMUM TIME ASSUMPTION) l l Time Mass Energy l (sec) (Ibm /sec) (Btu /sec) l 20.00045 6032.55 561.14 l 20.25099 6026.438 569.76 l 20.5011 5801.813 571.91 l 20.75082 5616.23 576.79 l 21.00125 5489.919 578.73 l 21.25023 5449.508 580.85 l 21.50078 5317.37 583.24 l 21.75106 5192.618 590.83 l 22.00106 5077.076 598.92 l 22.25069 4875.197 609.85 1 22.5012 4647.725 620.74 l 22.757068 4429.831 633.45 l 23.00052 4216.155 645.47 l 23.25065 4003.873 639.12 l 23.5009 (s) l 23.75058 3788.973 3586.032 631.33 622.47 l 24.00001 3411.624 611.28 l 24.2506 3243.65 598.33
'l 24.50004 3089.229 585.32 l 24.75027 2922.198 570.88 l 25.00053 2761.955 555.99 l 25.25037 2612.244 540.99 l 25.50013 2446.056 525.56 l 25.75087 2266.987 510.23 I 26.00013 2082.011 496.01 l 26.25065 1893.536 483.21 l 26.50021 1696.26 471.87 l 26.75015 1533.653 465.19 l 27.00075 1372.875 460.81 l 27.2509 1202.175 457.27 l 28.37 0 1167.4 1 30 0 1167.4 g
V Revision: 5 3 W95tingh00S8 6.2-107 February 29,1996
ghbh
- 6. Eiglieered S:f;ty Fez.tures mm-I Table 6.2.1.3-5 (Sheet 1 of 9)
O i I LONG-TERM DECLG BREAK l POST-BLOWDOWN MASS AND ENERGY RELEASES 1 l Time Mass Enthalpy Mass Enthalpy I (sec) (Ibm /sec) (Btu /lbm) (Ibm /sec) (Btu /lbm) l 0 0 0 0 0 I 0.05018 40807.8 531.04 0 0 l- 0.10006 51415.62 531.05 0 0 l 0.15 53605.18 531.28 0 0 l 0.20011 57605.54 531.7 0 0 l 0.25012 5649136 531.87 0 0 1 0.30008 55336.22 532.29 0 0 l 0.35016 52092.26 532.44 0 0 l 0.40003 60610.59 532 15 0 0 l 0.45018 64191.36 532.72 0 0 l 0.50003 63803.89 533.25 0 0 l 0.55013 62449.77 533.56 0 0 l 0.60012 61679.8 533.89 0 0 l 0.65001 61484.43 534.17 0 0 l 0.70003 61238 534.52 0 0 l 0.75004 59545.58 534.73 0 0 l 0.85006 59053.13 53535 0 0 l 0.9001 59119.19 535.67 0 0 l 0.95001 59525.57 536.08 0 0 l 1.00008 59848.18 536.5 0 0 l 1.10016 59770.71 537.41 0 0 l l 1.20009 59500.68 538.48 0 0 l
'l 1.30016 57917.15 539.87 0 0 l 1.40005 57134.81 541.49 0 0 l I 1.50018 56668.11 543.42 0 0 i l 1.60006 56262.47 545.51 0 0 l 1.70012 60543.26 547.95 0 0 l 1.80002 54369.46 550.25 0 0 l 1.90006 56577.14 552.62 0 0
I 2.00004 53516.96 555.65 0 0 l 2.10002 50010.8 558.31 0 0 0 Revision: 5 February 29,1996 6.2-108 y Westingh0use
su-w
- 6. Engineered Safety Fxtures
( N I Table 6.2.1.3-5 (Sheet 2 of 9) l l LONG-TERM DECLG BREAK l POST BLOWDOWN MASS AND ENERGY RELEASES l Time Mass Enthalpy Mass I Enthalpy (sec) (Ibm /sec) (Btu /lbm) (Ibm /sec) (Btu /lbm) I 2.20011 48952.83 561.05 0 0 1 2.3 47340.82 563.69 0 0 l 2.40023 45043.13 565.66 0 0 l 2.5002) 42988.27 567.51 0 0 l 2.60009 42921.87 568.95 0 0 l 2.7002 40175.77 570.82 0 0 l 2.8002 38195.66 572.61 0 0 l 2.90027 35992.53 574.28 0 0 l 3.00025 35135.79 575.41 0 0 I 3.10042 33091 576.57 0 0 l 3.20012 30285.49 577.93 0 0 l 3.30001 28275.31 579.17 0 0 [~N - l 3.40025 28461.7 581.32 0 0 l 3.50057 27581.74 583.7 0 0 l 3.60012 25860.66 586.19 0 0 I 3.70045 24041.52 587 0 0 l 3.80038 23092.48 585.67 0 0 l 3.90004 22838.94 583.67 0 0 l 4.00047 22839.6 581.73 0 0 l 4.10024 22952.33 579.68 0 0 l 4.20013 23328.38 577.08 0 0 I 4.30025 24186.82 574.68 0 0 l 4.40006 24673.8 574.99 0 0 I 4.50046 22714.09 579.43 0 0 l 4.60053 20355.21 584.6 0 0 l 4.70011 18616.09 585.43 0 0 l 4.80025 18906.58 581.82 0 0 l 4.90127 19231.26 582.52 0 0 I 5.00014 18868.84 587.31 0 0 l 5.25017 16910.84 600.57 0 0 l l 5.50024 16723.26 607.37 0 0
- .g~
\ 2 . 'J Revision: 5 T Westinghouse 6.2-109 February 29,1996
- 6. Ergineered Safety Features l
Table 6.2.1.3-5 (Sheet 3 of 9) e i I LONG TERM DECLG BREAK l POST BLOWDOWN MASS AND ENERGY RELEASES I Time Mass Enthalpy Mass Enthalpy I (sec) (ibm /sec) (Btu /lbm) (Ibm /sec) (Stu/lbm) I 5.75021 15630.08 622.46 0 0 I 6.00046 14927.82 619.58 0 0 l 6.25014 13819.42 622.69 0 0 l 6.50017 13146.93 625.96 0 0 l 6.75004 12839.71 614.37 0 0 l 7.00076 12718.63 606.73 0 0 l 7.25082 12457.77 603.1 0 0 I 7.50013 12436.33 615.31 0 0 l 7.75001 10397.57 700.28 0 0 l 8.00011 11225.92 628.48 0 0 l 8.25032 11341.57 607.45 0 0 l 8.50058 10746.42 624.1 0 0 l 8.75037 10070.86 649.75 0 0 I 9.00025 9625.501 668.I1 0 0 l 9.25024 9404.611 666.98 0 0 l 9.50026 9258.613 672.08 0 0 l 9.75046 8854.105 695.43 0 0 l 10.00011 8632.626 650.6 0 0 l 10.50024 8773.043 607.2 0 0 l 10.75008 8730.026 603.61 0 0 ! I 11.00065 8564.724 607.36 0 0 I i1.25054 8463.381 613.22 0 0 l 11.50005 8437.24 632.16 0 0 l 11.75003 7997.697 638.34 0 0 l 12.00047 7808.395 636.14 0 0 l 12.25084 7712.031 629.46 0 0
-l 12.50041 7661.505 620.78 0 0 l 12.75045 7589.224 613.65 0 0 l 13.00001 7481.231 608.13 0 0 l 13.25015 7395.79 602.47 0 0 l 13.50102 7568.571 598.98 0 0 0
Revision: 5 February 29,1996 6.2-110 [ W85tlRgh0US8
- 6. Engineered Safety Features t'
L}J l Table 6.2.1.3-5 (Sheet 4 of 9) l l LONG-TERM DECLG BREAK l POST. BLOWDOWN MASS AND ENERGY RELEASES l Time Mass Enthalpy Mass Enthalpy l (sec) (Ibm /sec) (Btu /!bm) (Ibm /sec) (Btu /lbm) I 13.75001 7353.851 590 0 0 l 14.00106 7316.183 577.08 0 0 l 14.25029 7384.659 565.58 0 0 l l' 14.50026 7487.738 555.09 0 0 l 14.75105 7572.562 547.96 0 0 l 15.00N1 7614.799 544.16 0 0 l 15.25128 7630.574 542.93 0 0 l 15.501 7574.608 543.15 0 0 l 15.75018 7592.818 545.88 0 0 l 16.00036 7622.272 550.2! 0 0 l 16.2512 7242.574 550.63 0 0 1 16.50058 7092.731 548.17 0 0 l
.C} ' 16.75056 7015.751 544.65 0 0 s/ l 17.00022 6958.681 541.57 0 0 l 17.25045 6907.833 539.49 0 0 l 17.50005 6836.467 548.93 0 0 l 17.75174 6741.548 540.05 0 0 l 18.00N9 6659.006 543.02 0 0 l 18.25102 6714.732 551.37 0 0 l 18.50013 6380.366 555.73 0 0 l 18.75045 6187.74 558.52 0 0 l 19.00098 6049.282 560.16 0 0 l 19.25078 5944.473 561.14 0 0 l 19.50057 5928.532 563.91 0 0 l 19.75015 5843.964 567.91 0 0 l 20.00187 5681.831 572.49 0 0 l 20.25095 5524.487 579.28 0 0 l 20.50135 5393.939 566.3 0 0 l
l 7 N Revision: 5 T Westingh0US8 6.2 111 February 29,1996
il5MQ h '
- 6. Engineered Safsty Frtures )
ra-1 I O Table 6.2.1.3-5 (Sheet 5 of 9) l l LONG-TERM DECLG BREAK l POST. BLOWDOWN MASS AND ENERGY RELEASES l Time Mass Enthalpy Mass Enthalpy ; I (sec) (thm/sec) (Btu /lbm) (Ibm /sec) (Btu /lbm) ! l 20.75026 3315.482 595.42 0 0 1 21.00026 3173.792 604.74 0 0 l 21.25062 5006.346 616.94 0 0 l 21.50157 4840.924 627.08 0 0 l 21.75061 4626.755 635.52 0 0 l 22.00119 4424.617 644.46 0 0 l 22.25015 4243.977 654.34 0 0 l 22.50057 4036.497 650.49 0 0 l 22.75069 3825.566 640.25 0 0 l 23.0005 3636.621 630.93 0 0 l 23.25033 3458.685 620 0 0 l 23.50017 3298.128 607.33 0 0 l 23.75109 3149.44 593.51 0 0 1 24.00023 3003.599 579.53 0 0 l 24.25004 2863.169 565.24 0 0 l 24.50091 '714.185 550.05 0 0 l 24.75038 2562.019 534.85 0 0 l 25.00082 2396.898 519.28 0 0 l 25.25019 2224.905 503.99 0 0 l 25.50001 2048.04 490.1 0 0 l 25.75086 1864.298 477.52 0 0 l 2ti.00029 1682.812 468.21 0 0 l 26.25073 1532.426 463.19 0 0 l 26.75052 12G4.403 455.84 0 0 l 27.00047 1021.557 453 0 0 l 27.25111 836.7731 455.97 0 0 l 27.50073 668.4349 480.86 0 0 l 27.75021 561.9642 588.82 0 0 l 28.00029 297.2754 425.97 0 0 l 28.25013 139.4778 291.71 0 0 l 28.37282 41.79308 292.48 0 0 0 Revision: 5 February 29,1996 6.2-112 3 Westlagt100Se
- 6. Engineered Safety Features
(~'N l U l l Table 6.2.1.3-5 (Sheet 6 of 9) I LONG TERM DECLG BREAK l POST-BLOWDOWN MASS AND ENERGY RELEASES I Time Mass Enthalpy Mass Enthalpy I (sec) (Ibm /sec) (Btu /lbm) (Ibm /sec) (Btu' Ibm) I 30 0 1167.4 0 1167.4 I 40 0 1167.4 0 1167.4 l 45 0 1167.4 0 1167.4 l 50 0 1167.4 0 1167.4 l 55 0 1167.4 0 1167.4 I 60 0 1167.4 0 1167.4 I 65 0 1167.4 0 1167.4 l 70 0 1167.4 23.9 1167.4 I 75 0 1167.4 19.27 1167.4 I 80 0 1167.4 91.88 1167.4 l 85 0 1167.4 154.36 1167.4 I 90 4.68 1103.15 203.45 1167.4
./'N I 95 52.9 302.21 201.5 1167.4 l 100 94.66 267.12 199.57 1167.4 l 105 130.61 254.41 197.88 1167.4 l 110 161.78 247.56 196.21 1167.4 l 115 89.9 294.05 200.47 1167.4 l 120 113.47 283.75 198.73 1167.4 l 125 133.99 277.32 197.01 1167.4
'I 130 151.88 272.79 195.3 1167.4 l 133.41 162.75 270.34 194.14 1167.4 I 135 167.49 269.32 193.6 1167.4 l 140 181.16 266.52 191.92 1167.4 l 145 193.13 264.16 190.25 1167.4 l 150 203.66 262.1 188.59 1167.4 l 151 205.58 261.73 188.29 1167.4 l 'O 220.78 258.7 185.68 1167.4 l li 234.15 255.8 182.82 1167.4 I
(s ! k Revision: 5 3 W8Stinghouse 6.2-113 February 29,1996
- 6. Engineered Saf;ty Fr.ctures l
Table 6.2.1.3-5 (Sheet 7 of 9) e l l LONG-TERM DECLG BREAK l POST-BLOWDOWN MASS AND ENERGY RELEASES l Time Mass Enthalpy Mass Enthalpy I (sec) (Ibm /sec) (Btu /lbm) (Ibm /sec) (Btu /lbm) l 180 244.76 253.2 180.01 11^7.4 l 190 253.29 250.81 177.24 1167.4 l 200 260.28 248.57 174.52 1167.4 l 220 108.55 335.13 183.76 1167.4 l 240 116.84 327.51 178.49 1167.4 i 260 123.57 320.72 173.41 1167.4 l 280 129.36 314.43 168.54 1167.4 l 300 134.55 308.5 163.84 1167.4 I 320 139.34 302.84 159.33 1167.4 I 340 143.84 297.41 154.97 1167.4 l 360 153.62 288.39 150.28 1167.4 1 380 57.19 391.15 158.75 1167.4 l 400 61.79 380.61 154.18 1167.4 1 401 78.13 360.2 137.84 1167.4 I 420 81.86 353.21 134.12 1167.4 l 440 130.9 300.33 125.09 1167.4 I 460 134.23 295.22 121.77 1167.4 I 480 100.69 324.07 122.3 1167.4 I 500 103.97 318.19 119.02 1167.4 l 520 77.49 347.79 119.51 1167.4 l 540 105.7 311.55 113.3 1167.4 l 541 123.58 294.37 111.42 1167.4 l 560 126.28 289.68 108.72 1167.4 l 580 38.2 418.67 117.8 1167.4 I 600 41.49 407.06 114.51 1167.4 1 650 48.02 384.45 107.98 1167.4 l 700 53.84 365.07 102.16 1167.4 l 750 59.06 347.92 96.94 1167.4 1 Revision: 5 e February 29,1996 6.2-114 W Westinghouse
- 6. Engineered Safety Features
- o l Table 6.2.1.3-5 (Sheet 8 of 9)
I I LONG-TERM DECLG BREAK l POST-BLOWDOWN MASS AND ENERGY RELEASES l l Tic._e Mass Enthalpy Mass Enthalpy l (sec) (Ibm /sec) (Bru/lbm) (Ibm /sec) (Btu /lbm) I 800 63.75 332.46 92.25 1167.4 I 850 69.36 316.72 86.64 1167.4 I 900 63.35 320.28 82.65 1167.4 l 950 68.27 305.57 77.73 1167.4 l 1000 72.81 292.11 73.19 1167.4 l 1035 74.85 284.27 71.15 1167.4 l 1100 52.84 313.35 69.16 1167.4 l 1200 58.66 288.13 63.34 1167.4 l 1300 63.59 266.44 58.41 1167.4 l 1400 67.81 247.55 54.19 1167.4 1 1404.11 67.97 246.82 54.03 1167.4 I 1405 68 246.67 54 1167.4
,/~] I 1500 71.45 230.97 50.55 1167.4 V I 2000 167.77 136.76 37.93 1167.4 I 3600 179.05 101.98 26.65 1167.4 I 4000 180.63 99.34 25.07 1167.4 l 6000 184.16 94.97 21.54 1167.4 I 7500 157.67 95.39 20.33 1167.4 I 80ra 129.25 96.72 19.75 1167.4 l 1000u 130.47 96.07 18.53 1167.4 l 15000 131.46 95.58 17.54 1167.4 l 16000 131.66 95.48 17.34 1167.4 l 20000 128.55 94.83 15.46 1167.4 l 25562 115.72 95.51 15.29 1167.4 l 30067 94.81 97.11 15.2 1167.4 I 36000 72.6 98.49 13.41 1 M 7.4 l 40000 0 98.49 15.4 1167.4 1 60000 0 98.49 13.8 1167.4 l
'n N) Revision: 5 [ W8Stingh00S8 6.2-115 February 29,1996
l l
- 6. Engteered Sdety Fmtures l
Tab!c 6.2.1.3-5 (Sheet 9 of 9) e i I LONG-TERM DECLG BREAK l POST. BLOWDOWN MASS AND ENERGY RELEASES l Time Mass Enthalpy Mass Enthalpy I (sec) (Ibm /sec) (Btu /lbm) (Ibm /sec) (Btu /lbm) l 80000 0 98.49 12.7 1167.4 l 100000 0 98.49 11.9 1167.4 l 150000 0 98.49 10.5 1167.4 l 200000 0 98.49 9.6 1167.4 1 400000 0 98.49 7.5 1167.4 l 600000 0 98.49 6.3 1167.4 l 800000 0 98.49 5.6 1167.4 l 1000000 0 98.49 5.1 1167.4 l 1500000 0 98.49 4.3 1167.4 l 2000000 0 98.49 3.8 1167.4 I 4000000 0 98.49 2.7 1167.4 l 4000000 0 98.49 2.7 1167.4 e Revision: 5 e February 29,1996 6.2-116 3 Westingh00S8
. . ~ .-. . -, _. . .. . . . - . .. .- . . - - -
- 6. Ergineered Safety Features '
s- _ _; C\ V l l Tame 6.2.13-6 (Sheet I cf 2) . l LONG-TERM DEHLG BREAK l BLOWDOWN MASS AND ENERGY RELEASES l l Time Mass Enthalpy l (sec) (Ibm /sec) (Btu /lbm) I I 0.0 0 0 l 0.05 74746.76 626.2 1 0.15 64195.07 637.108 l 0.25 51407.55 642.663 l 0.35 50632.16 639.027 l 0.45 49403.93 634.886 l 0.65 48846.71 626.2 1 0.85 ' 4P392.4 J'!0.241 i 1 47565.69 621.857 l 1.2 45516.28 620.039 l 1.4 45467.14 614.989 l 1.6 45724.46 610.242 l 1.8 45136.05 606 s l 2 45808 603.273
-l 2.5 44 % 3.51 587.315 I 3 40898.81 616.302 l 3.5 37852.99 632.26 l 4 35603.29 615.494 I 4.5 34478.13 601.455 l 5 32828.32 589.84 l- 5.5 30957.59 580.144 I 6 30023.08 570.145 l 6.5 25298.23 573.983 L l 7 23506.66 568.832 l 7.5 l 1770333 597.718 l 8 17168.7 573.175 i I 8.5 13906.76 599.839 l 9 12268.16 600.748 I ;
, L./ Revision: 5 T Westinghouse 6.2-117 February 29,1996
plaime i
- 6. Engineered Saf3ty Fcct:2es vn -
l l O l Table 6.2.1.3 6 (Sheet 2 of 2) I LONG-TERM DEHLG BREAK l BLOWDOWN MASS AND ENERGY RELEASES I l Time Mass Enthalpy I (sec) (Ibm /sec) (Btu /lbm) l 9.5 10473.05 633.775 I 10 9004.537 673.165
)
l 10.5 7813.659 723.564 l 11 6779.514 777.397 l 11.5 5822.029 834.361 l 12 4834.074 943.138 l 12.5 3701.439 1060.702 l 13 3289.87 1122.615 l 13.5 2937.427 1186.851 l 14 2457.231 1211.192 l 14.5 1957.462 1250.077 I I 15 1760 964 1246.946 l l 16 1335.38 1255.632 l 16.5 1173.669 1254.319 I l 17 1055.88 1251.895 l 17.5 958.9 % 5 1249.37 l 18 869.5154 1251.491 l 18.5 522.9105 1283.003 l 19 605.020* 1256.541 I 19.5 559.398 1263.005 l l 20 0 1285.276 l 20.5 76.64999 1307.546 l 21 73.53149 1306.435 l 21.5 51.45 1324.615 l 22 0 l 1324.615 O Revision: 5 February 29,1996 6.2-118 Y_/ W85tingh0USB
l
+itti ~ ~
- 6. Engineered Safety F=tures C\
V I Table 6.2.1.3-7 BASIS FOR LONG-TERM ANALYSIS l- Number of Loops 2 Active Core Length (ft) 12.0 Core Power, license application (MWt) 1933 Nominal Vessel Inlet Temperature (*F) 535.1 Nominal Vessel Outlet Temperature ('F) 600.0 l Steam Pressure (psia) 821.0 Rod Array 17 x 17 Accumulator Temperature (*F) 120.0 l Containment Design Pressure (psia) 59.7 O d .-( o ). .
~,
Revision: 5
'k W8stinghouS8 6.2-119 February 29,1996
MW
- 6. Fagineered Saf;ty Fe:tures sw-Table 6.2.1.4-1 O
SPECTRUM OF SECONDARY SYSTEM PIPE RUPTURES ANALYZED Power Level 102 % 70% 30% 0% Full DER, MSIV failure Full DER Full DER Full DER Full DER Full DER, MFWlV failure Full DER Full DER Full DER Full DER Small DER, MSIV failure (fS) 0.70 0.6 0.60 0.55 0.60 0.53 0.50 0.50 0.55 0.50 0.40 0.20 0.33 0.32 0.36 0.10 0.22 l Split Rupture (ft2) 0.37(a) o,41(a) o,44(a) 0.442(a) (a)As total area of two loops. DER = double ended rupture MSIV = main steam line isolation valve MFWIV = main feedwater isolation valve O l l l i 4 l 1 Revision: 5 February 29,1996 6.2-120 W We,stinghouse
- 6. E gineered Srfety Fetures p
o/ Table 6.2.1.4-2 (Sheet 1 of 31) MASS AND ENERGY RELEASE DATA FOR TIIE CASE OF MAIN STEAM LINE FULL DOUBLE ENDED RUPTURE FROM 30% POWER LEVEL WITII FAULTED LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES IIIGIIEST-CONTAINMENT PRESSURE Initial s tam generator mass ( lbm ) 167750 Mass added by feedwater flashing ( lbm ) : 9466 Mass added by unisolatable steam (lbm ) : 10082 Initial steam pressure ( psia ) : 9976.5 Feedwater line isolation at ( sec ) : 7.578 Steam line isolation at ( sec ) : 7.578 Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) 0.0 0.0 0.0 0.0 0.000 0.1 12438.1 14.845 1.244 1.485 < 0.2 12412.4 14.815 2.485 2.966 0.3 12393.0 14.793 3.724 4.445
/~T 0.4 12373.6 14.770 \
V] 0.5 12354.9 14.749 4.962 6.197 5.922 7.397 0.6 12336.6 14.728 7.431 8.870 0.7 12318.7 14.707 8.663 10.341 0.8 12300.6 14.686 9.893 11.809 0.9 12283.1 14.665 11.121 13.276 1.0 12265.9 14.645 12.348 14.740 l 1.1 12249.2 14.626 13.573 16.203 12 5396.4 6.443 14.112 16.847 1.3 5345.4 6.384 14.647 17.486 1.4 5305.4 6.337 15.177 18.119 1.5 5266.3 6.292 15.704 18.748 1.6 5227.9 6.247 16.227 19.373 1.7 5190.3 6.204 16.746 19.994 1.8 5153.5 6.161 17.261 20.610 1.9 5117.3 6.118 17.773 21.221
/G %.Y Revision: 5 3 W95tingh0Use 6.2 121 February 29,1996
mtetr
- 6. Engineered Sciety Fmtures Table 6.2.1.4-2 (Sheet 2 of 31) e l MASS AND ENERGY RELEASE DATA l
FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITII FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE TIIAT I PRODUCES IIIGHEST CONTAINMENT PRESSURE I I Time Mass Flow Energy Mow Integrated Mass Integrated Energy (sec) (thm/s) (10' Btu /s) (10' lbm) (10' Btu) l 2.0 5081.8 6.077 18.281 21.829 1 2.1 5046.9 6.036 18.786 22.433 l 2.2 5012.6 5.996 19.287 23.032 l 2.3 4978.8 5.957 19.785 23.628 l 2.4 4945.6 5.918 20.279 24.220 l 2.5 4913.4 5.880 20.771 24.808 l 2.6 4881.2 5.843 21.259 23.N2 I 2.7 4849.5 5.806 21.744 25.973 1 2.8 4818.3 5.769 22.226 26.550 l 2.9 4787.5 5.733 22.704 27.123 l l 3.0 4757.2 5.697 23.180 27.693 l 3.1 4727.8 5.663 23.653 28.259 l 3.2 4698.4 5.628 24.123 28.822 l 3.3 4669.4 5.594 24.590 29.381 l 3.4 4640.8 5.561 25.054 29.937 I 3.5 4612.6 5.528 25.515 30.490 l 3.6 4584.8 5.495 25.974 31.040 l 3.7 4558.0 5.463 26.429 31.586 l 3.8 4537.8 5.440 26.883 32.130 3.9 4516.5 5.415 27.335 32.671 Revisiois: 5 e February 29,1996 6.2-122 T Westilighet!Se
- 6. E gineered Sity Fe*tures ~
o G Table 6.2.1.4-2 (Sheet 3 of 31) I l h! ASS AND ENERGY RELEASE DATA l FOR THE CASE OF h1AIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) Obm/s) (10' Btu /s) (10' lbm) (10' Btu) ; I 4.0 4495.2 5.389 27.784 33.210 l 4.1 4474.1 5.365 28.232 33.747 l 4.2 4453.3 5.340 28.677 34.281 I 4.3 4432.8 5.316 29.120 34.812 l 4.4 4412.6 5.292 29.562 35.342 l 4.5 4392.6 5.269 30.001 35.869 I 4.6 4372.8 5.245 30.438 36.393 I 4.7 4353.3 5.222 30.873 36.915 j I 4.8 4334.0 5.200 31.307 37.435
'sj l 4.9 4315.0 5.177 31.738 37.953 l 5.0 4296.2 5.155 32.168 38.468 I l 5.1 4277.7 5.133 32.596 38.982 l 5.2 4259.4 5.112 33.022 39.493 l 5.3 4241.3 5.090 33.446 40.002 I 5.4 4223.5 5.069 '
33.868 40.509 l 5.5 4205.9 5.048 34.289 41.014
)
l 5.6 4188.6 5.028 34.708 41.516 l 5.7 4171.4 5.008 35.125 42.017 l 5.8 4154.5 4.988 35.540 42.516 I 5.9 4137.9 4.968 35.954 43.013 l l l ts I \ u_) Revision: 5 3 Westiflghouse 6 2-123 February 29,1996 l
wwn..
- 6. Eegineered Safety Frtures Table 6.2.1.4-2 (Sheet 4 of 31) e h
MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10'lbm) (10' Btu) l 6.0 4121.4 4.948 36366 43.508 l 6.1 4105.2 4.929 36.777 44.001 I 6.2 4089.1 4.910 37.186 44.492 l 6.3 4073.3 4.892 37.593 44.981 l- 6.4 4057.7 4.873 37.999 45.468 l 6.5 40423 4.855 38.403 45.954 I 6.6 4027.1 4.837 38.806 46.437 l 6.7 4012.1 4.819 39.207 46.919 l 6.8 3997.3 4.802 39.607 47399 I 6.9 3982.7 4.784 40.005 47.878 1 7.0 39683 4.767 40.402 48.354 l 7.1 3954.0 4.750 40.797 48.829 1 7.2 3940.0 4.734 41.191 49303 l 73 3926.1 4.717 41.584 49.775 I I 7.4 3912.4 4.701 41.975 50.245 l 7.5 3898.9 4.685 42.365 50.713 l 7.6 3885.5 4.669 42.753 51.180 l 7.7 38723 4.653 43.141 51.645 l 7.8 3859.2 4.6?3 43.526 52.109 I 7.9 38463 4.623 43.911 52.571 1 1 O Revision: 5 February 29,1996 6.2-124 T Westinghouse
g._ n
- 6. Engizeered Sity Features Q)
Table 6.2.1.4-2 (Sheet 5 of 31) I h1 ASS AND ENERGY RELEASE DATA 1 , FOR THE CASE OF h1AIN STEAM LINE FULL DOUBLE I l I ENDED RUPTURE FROh! 30% POWER LEVEL WITH FAULTED LOOP MAIN STEAh! LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT PRESSURE I i Time Mass Flow Energy Flow Integrated Mass l Integrated Energy I (sec) (Ibm /s) (10' litu/s) (10' lbm) (10' Btu) , I 8.0 3833.6 4.607 44.294 53.032 l 8.1 3821.0 4.592 44.677 53.491 l 8.2 3808.5 4.578 45.057 53.949 I 8.3 3796.2 4.563 45.437 54.405 1 8.4 3784.0 4.549 45.815 54.860 l 8.5 3771.9 4.534 46.193 55.314 I 8.6 3760.0 4.520 46.569 55.766 l 8.7 3748.2 4.506 46.943 56.216
'~N I 8.8 3736.5 4.492 47.317 56.666 l 8.9 3724.9 4.478 47.690 57.113 l 9.0 3713.4 4.465 48.061 57.560 !
l 9.1 3702.0 4.451 48.431 58.005 i l 9.2 3690.7 4.438 48.800 58.449 ! l 9.3 3679.5 4.424 49.168 58.891 l 9.4 3668.4 4.411 49.535 59.332 i t l 9.5 3657.4 4.398 49.901 59.772 l 9.6 3646.5 4.385 50.265 60.210 l 9.7 3635.6 4.372 50.629 60.648 l 9.8 3624.8 4.359 50.991 61.084 l 9.9 3614.1 4.346 51.353 61.518 A V Revision: 5 3 W95tingh00S8 6.2-125 February 29,1996
l##14111t#tt
- 6. Engineered Sif;ty Feitures Table 6.2.1.4-2 (Sheet 6 of 31) e l MASS AND ENERGY RELEASE DATA I
FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT PRESSURE I I Time Mass Fbw Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) I 10.0 3603.5 4.334 51.713 61.952 l 10.1 3592.9 4.321 52.072 62.384 I 10.2 3582.4 4.309 52.431 62.815 l 10.3 3571.9 4.296 52.788 63.244 l 10.4 3561.5 4.284 53.144 63.673 l 10.5 3551.2 4.271 53.499 64.100 l 10.6 3540.9 4.259 53.853 64.526 l 10.7 3530.8 4.247 54.206 64.950 1 10.8 3523.4 4.238 54.559 65.374 I 10.9 3515.2 4.229 54.910 65.797 l 11.0 3507.0 4.219 55.261 66.219 l 11.1 3498.7 4.209 55.611 66.640 I I i1.2 3490.5 4.199 55.960 67.060 l 11.3 3482.3 4.189 56.308 67.479 l 11.4 3474.1 4.180 56.655 67.897 l 11.5 3465.9 4.170 57.002 68.314 l 11.6 3457.6 l 4.160 57.348 68.730 l 11.7 3449.4 4.150 57.693 69.145 ( l 11.8 1707.9 2.055 57.864 69.350 l l 11.9 1704.7 2.051 58.034 69.555 O Revision: 5 February 29,1996 6.2-126 3 Westiligt100S8
i t?t W C. Ecgineered Satty Features I
,- s Table 6.2.1.4-2 (Sheet 7 of 31) l I
MASS AND ENERGY RELEASE DATA i I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l I ENDED RUPTURE FROM 30% POWER LEVEL WITII FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE TIIAT PRODUCES IIIGIIEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) I 12.0 1701.5 2.047 58.204 69.760 l 12.1 1698.3 2.044 58.374 69.964 l 12.2 1695.1 2.040 58.543 70.168 I 12.3 1691.9 2.036 58.713 70.372 l 12.4 1688.6 2.032 58.882 70.575 - l 12.5 1685.4 2.028 59.050 70.778 l 12.6 1682.1 2.024 59.218 70.980 l 12.7 1678.8 2.020 59.386 71.182 l 12.8 1675.6 2.016 59.554 71.384 V I 12.9 1672.3 2.012 59.721 71.585 l 13.0 1669.0 2.009 59.888 71.786 l 13.1 1665.7 2.005 60.054 71.987 l 13.2 1662.4 2.001 60.221 72.187 l 13.3 1659.0 1.997 60.387 72.386 l 13.4 1655.7 1.993 60.552 72.586 l 13.5 1652.4 1.989 60.717 72.784 l 13.6 1649.0 1.985 60.882 72.983 1 13.7 1645.7 1.981 61.047 73.181 l 13.8 1642.3 1.977 61.211 73.379 l 13.9 1638.9 1.973 61.375 73.576 l l (D N] Revision: 5 W W85tingh0USB 6.2-127 February 29,1996 i
HMi
- 6. Ergineered Saf;ty Fe:tures su-Table 6.2.1.4-2 (Sheet 8 of 31)
O l MASS AND ENERGY RELEASE DATA l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT PRESSURE I l Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (lbrn/s) (10' Btu /s) (10' lbm) (10' Btu) l I4.0 1635.5 1.969 61.538 73.773 I 14.1 1632.2 1.965 61.702 73.969 l 14.2 1628.8 1.960 61.865 74.165 l 14.3 1625.4 1.956 62.027 74.361 l 14.4 1622.0 1.952 62.189 74.556 l 14.5 1618.6 1.948 62.351 74.751 l 14.6 1615.1 1.944 62.513 74.945 l 14.7 1611.7 1.940 62.674 75.139 l 14.8 1608.3 1.936 62.835 75.333 l 14.9 1604.8 1.932 62.995 75.526 l 15.0 1601.4 1.928 63.155 75.719 l 15.1 1598.0 1.924 63.315 75.911 l 15.2 1594.5 1.919 63.475 76.103 l 15.3 1591.1 1.915 63.634 76.295 l 15.4 1587.6 1.911 63.792 76.486 l 15.5 1584.2 1.907 63.951 76.677 l 15.6 1580.7 1.903 64.109 76.867 l 15.7 1577.2 1.899 64.267 77.057 l 15.8 1573.8 1.895 64.424 77.246 l 15.9 1570.3 1.890 64.581 77.435 O Revision: 5 February 29,1996 6.2-128 [ W85tingt100Se
1 tn:nnme:::
- 6. Engineered Safety F=tures (b
Table 6.2.1.4-2 (Sheet 9 of 31) i MASS AND ENERGY RELEASE DATA l l FOR Tile CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 16.0 1566.8 1.886 64.738 77.624 l 16.1 1563.2 1.882 64.894 7', .o l 2 l 16.2 1559.7 1.878 65.050 78.000 l 16.3 1556.0 1.873 65.206 78.187 l 16.4 1552.4 1.869 65.361 78.374 l 16.5 1548.7 1.865 65.516 78.561 l 16.6 1544.9 1.860 65.670 78.747 l 16.7 1541.1 1.856 65.824 78.932 l 16.8 1537.3 1.851 65.978 79.117
'(/ I 16.9 1533.4 1.846 66.131 79.302 l 17.0 1529.5 1.842 66.284 79.486 l 17.1 1525.5 1.837 66.437 79.670 l 17.2 1521.5 1.832 66 539 79.853 l 17.3 1517.5 1.827 66.741 80.036 l 17.4 1513.5 1.822 66.892 80.218 l 17.5 1509.4 1.817 67.043 80.400 l 17.6 1505.3 1.813 67.194 80.581 l 17.7 1501.1 1.808 67.344 80.762 l l 17.8 1497.0 1.803 67.493 80.942 I 17.9 1492.8 1.798 67.643 81.122 l
l 1D. N-Revision: 5 Y WeStiligl10USS 6.2-129 February 29,1996
)
- 6. Engineered Sar;ty Fertvres Table 6.2.1.4-2 (Sheet 10 of 31) e l MASS AND ENERGY RELEASE DATA 1
FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITII FAULTED l } LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTA.INMENT PRESSURE l Time Mass Flow Energy Flow Integrated Mass Integrated Energy [ (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 18.0 1488.6 1.793 67.792 81.301 l 18.1 1484.4 1.787 67.940 i 81.480 ) 1 l 18.2 1480.1 1.782 68.088 81.658 l 18.3 1475.9 1.777 68.236 81.836 l 18.4 1471.6 1.772 68.383 82.013 I 18.5 1467.3 1.767 68.529 l 82.190 l 18.6 1463.0 1.762 68.676 82.366 { l 18.7 1458.6 1.757 68.822 82.541 l 18.8 1454.3 1.751 68.967 82.717 l 18.9 1449.9 1.746 69.112 82.891 l 19.0 1445.6 1.741 69.257 83.065 l 19.1 1441.2 1.736 69.401 83.239 l 19.2 1436.8 1.730 69.544 83.412 I 19.3 1432.5 1.725 69.688 83.584 l 19.4 1428.1 1.720 69.830 83.756 l 19.5 1423.7 1.715 69.973 83.928 I 19.6 1419.3 1.709 70.115 84.099 l 19.7 1415.0 1.704 70.256 84.269 l 19.8 1410.6 1.699 70.397 84.439 l 19.9 1406.2 1.694 70.538 84.608 l 1 l Revision: 5 e February 29,1996 6.2-130 3 Westinghotise \ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ - _ _ - _ _ _ - _ - - _ _ - - - - - - - --
- 6. Engineered S:fety Ittures n
(v) Table 6.2.1.4-2 (Sheet 11 of 31) 1 l MASS AND ENERGY RELEASE DATA l l FOR TIIE CASE OF MAIN STEAM LINE FULL DOUBLE ! I ENDED RUPTURE FROM 30% POWER LEVEL WITII FAULTED l l LOOP MAIN STEAM LINE ISG!.ATION VALVE FAILURE THAT l l PRODUCES IIIGHEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 20.0 1401.8 1.688 70.678 84.777 l 20.5 1388.8 1.673 71.373 85.614 f l 21.0 1367.7 1.647 72.,056 86.437 I 21.5 1346.3 1.622 72.730 87.248 l 22.0 1325.4 1.596 73.392 88.046 l 22.5 1304.7 1.572 74.045 88.832 I 23.0 1284.4 1.547 74.687 89.605 l 23.5 1264.4 1.523 75.319 90.367 I 24.0 1244.8 1.499 75.941 91.116 G' l 24.5 1225.5 1.476 76.554 91.855 l 25.0 12 % .7 1.453 77.157 92.581 l 25.5 1188.2 1.431 77.752 93.297 l 26.0 1170.1 1.409 78.337 94.001 l 26.5 1152.4 1.388 78.913 94.695 l 27.0 1135.1 1.367 79.480 95.379 l 27.5 1118.1 1.347 80.039 96.052 l 28.0 1101.5 1.326 80.590 96.715 l 28.5 1085.2 1.307 81.133 97.369 l 29.0 1069.3 1.288 81.667 98.013 l 29.5 1053.8 1.269 82.194 98.647 l f% Revision: 5 T W8Stingl100S8 6.2-131 February 29,1996
- mamm
- 6. Engineered Salty Fe-tures Table 6.2.1.4-2 (Sheet 12 of 31)
O l MASS AND ENERGY RELEASE DATA l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUI"fURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT PRESSURE I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 30.0 1038.5 1.250 82.714 99.272 l 30.5 1023.7 1.233 83.225 99.888 l 31.0 1009.1 1.215 83.730 100.496 l 31.5 994.9 1.198 84.227 101.095 l 32.0 981.0 1.181 84.718 101.685 l 32.5 967.5 1.165 85.202 102.268 l 33.0 954.2 1.148 85.679 102.842 l 33.5 941.2 1.133 86.149 103.408 l 34.0 928.5 1.117 86.614 103.967 l 34.5 916.2 1.102 87.072 104.518 l 35.0 904.0 1.088 87.524 105.062 l 35.5 892.2 1.074 87.970 105.599 l 36.0 880.6 1.059 88.410 106.129 I 36.5 869.3 1.046 88.845 106.651 l 37.0 858.3 1.032 89.274 107.168 l 373 847.6 1.019 89.698 107.677 l 38.0 837.1 1.007 90.116 108.181 l 38.5 826.8 0.994 90.530 108.678 l 39.0 816.9 0.982 90.938 109.169 I 39.5 807.1 0.970 91.342 109.654 O Revision: S February 29,1996 6.2-132 3 WBStinghouse
- 6. Engineered Safety Fect:res O
U Table 6.2.1.4-2 (Sheet 13 of 31) l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWEP LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT PRESSURE I l Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l l 40.0 797.6 0.959 91.741 ' 110.134 l 40.5 788.3 0.948 92.135 110.608 , I 41.0 779.2 0.937 92.524 111.076 l 41.5 770.4 0.926 92.909 111.539 l 42.0 761.7 0.915 93.290 111.997 l 42.5 753.1 0.905 93.667 112.449 l 43.0 744.8 0.895 94.039 112.897 l 43.5 736.6 0.885 94.408 113.339 G l 44.0 728.6 0.875 94.772 113.777 l 44.5 720.8 0.866 95.132 114.210 l 45.0 713.1 0.857 95.489 114.638 l 45.5 705.6 0.848 95.842 115.062 l 46.0 698.3 0.839 96.191 115.481 l 46.5 691.1 0.830 96.536 115.8 % l 47.0 683.9 0.821 96.878 116.307 I 47.5 677.0 0.813 97.217 116.713 l 48.0 670.1 0.804 97.552 117.116 l 48.5 663.4 0.796 97.884 117.514 l 49.0 656.9 0.788 98.212 117.908 l 49.5 650.4 0.781 98.537 118.298 O L] Revision: 5 3 W8Stingh00S8 6.2-133 February 29,1996
.3n--
- 6. Ezgineered Saf;ty Fe tures Table 6.2.1.4-2 (Simet 14 of 31)
O l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITII FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGIIEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy 8 (sec) (Ibm /s) (10' Btu /s) (10 lbm) (10' Btu) l 50.0 644.2 0.773 98.859 118.685 I 50.5 638.0 0.765 99.178 119.067 I 51.0 631.9 0.758 99.494 119.446 I 51.5 625.9 0.751 99.807 119.822 I 52.0 620.0 0.744 100.117 120.194 I 52.5 614.3 0.737 100.424 120.562 I 53.0 608.7 0.730 100.729 120.927 l 53.5 603.2 0.723 101.030 121.289 l 54.0 597.8 0.717 101.329 121.647 I 54.5 592.5 0.710 101.625 122.002 1 55.0 587.3 0.7M 101.919 122.354 l 55.5 582.2 0.698 102.210 122.703 l 56.0 577.2 0.692 102.499 123.049 I 56.5 572.3 0.686 102.785 123.392 I 57.0 567.5 0.680 103.069 123.732 i 57.5 562.8 0.674 103.350 124.069 I 58.0 558.2 0.669 103.629 124.4M l 58.5 553.7 0.663 103.906 124.735 I 59.0 549.2 0.658 104.181 125.064 l 59.5 544.9 0.653 104.453 125.391 O Revision: 5 February 29,1996 6.2-134 y West lnghouse
- tt:: ::tn
- 6. Ergineered Sity Features I
l Table 6.2.1.4-2 (Sheet 15 of 31) l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE i ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT PRESSURE I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (lbm/s) (10' Btu /s) (10' lbm) (10' Btu) I 60.0 540.6 0.648 104.723 125.715 l 60.5 536.5 0.642 104.992 126.036 l 61.0 532.4 0.637 105.258 126.354 l 61.5 528.3 0.633 105.522 126.671 l 62.0 524.4 0.628 105.784 126.985 l 62.5 520.5 0.623 106.044 127.2 % l 63.0 516.7 0.619 106.303 127.606 I 63.5 513.0 0.614 106.559 127.913 (' l 64.0 509.4 0.610 106.814 128.217
's I 64.5 505.8 0.605 107.067 128.520 l 65.0 502.3 0.601 107.318 128.820 l 65.5 498.9 0.597 107.567 129.119 I 66.0 495.5 0.593 107.815 129.415 l 66.5 492.3 0.589 108.061 129.710 l 67.0 489.0 0.585 108.306 130.002 l 67.5 485.9 0.581 108.549 130.293 I 68.0 482.8 0.577 108.790 130.582 l 68.5 479.6 0.574 109.030 130.868 l 69.0 476.6 0.570 109.268 131.153 69.5 473.7 0.566 109.505 131.437
(^)
\ .
Revision: 5 l [ W85tillghouse 6.2-135 February 29,1996
- 6. Engineered Sifety Fe'tures Table 6.2.1.4 2 (Sheet 16 of 31)
O l MASS AND ENERGY RELEASE DATA l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE TIIAT l PRODUCES IIIGHEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (19' lbm) (10' Btu) I 70.0 470.8 0.563 109.741 131.718 l 70.5 468.0 0.559 109.975 131.998 I 71.0 465.2 0.556 110.207 132.276 l 71.5 462.5 0.553 110.438 132.552 l 72.0 459.8 0.550 110.668 132.827 l 72.5 457.2 0.546 110.897 133.100 l 73.0 454.6 0.543 111.124 133.372 l 73.5 452.1 0.540 111.350 133.642 l 74.0 449.6 0.537 111.575 133.910 l 74.5 447.2 0.534 111.799 134.178 l 75.0 444.8 0.531 112.021 134.443 I 75.5 442.4 0.528 112.242 134.708 l 76.0 440.1 0.526 112.462 134.970 l 76.5 437.8 0.523 112.681 135.232 I 77.0 435.6 0.520 112.899 135.492 l 77.5 433.4 0.518 113.116 135.751 I 78.0 431.3 0.515 113.331 136.008 l 78.5 429.2 0.512 113.546 136.265 l 79.0 427.1 0.510 113159 136.520 l 79.5 425.1 0.508 113.972 136.773 O Revision: 5 February 29,1996 6.2-136 3 Westingliouse
- 6. Ecgineered Safaty Fatures
\
[G Table 6.2.1.4-2 (Sheet 17 of 31) I MASS AND ENERGY RELEASE DATA I I FOR TIIE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITII FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE TIIAT PRODUCES IIIGIIEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Blu) I 80.0 423.1 0.505 114.183 137.026 l 80.5 421.1 0.503 114.394 137.277 l 81.0 419.2 0.500 114.604 137.527 l 81.5 417.3 0.498 114.812 137.776 I 82.0 415.5 0.496 115.020 138.024 l 82.5 413.7 0.494 115.227 138.271 I 83.0 411.9 0.492 115.433 138.517 l 83.5 410.1 0.489 115.638 138.762 l 84.0 408.4 0.487 (-] 115.842 139.005 C/ l 84.5 406.7 0.485 116.045 139.248 l 85.0 405.0 0.483 116.248 139.490 l 85.5 403.4 0.481 116.450 139.730 l 86.0 401.8 0.479 116.651 139.970 I 86.5 400.2 0.477 116.851 140.209 l 87.0 389.6 0.476 117.050 140.446 I 87.5 397.1 0.474 117.248 140.683 1 88.0 395.6 0.472 117.446 140.919 l 88.5 394.1 0.470 117.643 141.154 l 89.0 392.7 0.468 117.840 141.388 l 89.5 391.3 0 467 118.035 141.621 O
)
Revision: 5 T Westingh0USe 6.2-137 February 29,1996
- 6. E:gineered Safety Fetures Table 6.2.1.4-2 (Sheet 18 of 31) e i l
MASS AND ENERGY RELEASE DATA ! l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT PRESSURE 1 I Time Mass Flow Energy Flow Integrated Mass Integrated Energy I (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 90.0 389.9 0.465 118.230 141.854 l 90.5 388.5 0.463 118.425 142.086 I 91.0 387.2 0.462 118.618 142.316 I 91.5 385.8 0.460 118.811 142.546 I 92.0 384.5 0.458 119.003 142.776 I 92.5 383.3 0.457 119.195 143.004 I 93.0 382.0 0.455 119.386 143.232 I 93.5 380.8 0.454 119.576 143.459 l 94.0 379.6 0.452 119.766 143.685 l 94.5 378.4 0.451 119.955 143.910 I 95.0 377.2 0.450 120.144 144.135 i I 95.5 376.0 0.448 120.332 144.359 l 96.0 374.9 0.447 120.519 144.583 l 96.5 373.8 0.445 120.706 144.805 l 97.0 372.7 0.444 120.893 145.027 l l 97.5 371.6 0.443 121.078 145.249 l 98.0 370.6 0.442 , 121.264 145.469 I 98.5 369.5 0.440 121.448 145.690 l 99.0 368.5 0.439 121.633 145.909 l 99.5 367.5 0.438 121.816 146.128 1 l O Revision: 5 February 29,1996 6.2-138 3 Westingh0US8
- 6. E gineered S:fety Fe:.tures A
I Table 6.2.1.4-2 (Sheet 19 of 31) I h1 ASS AND ENERGY RELEASE DATA I I FOR THE CASE OF h1AIN STEAh! LINE FULL DOUBLE I ENDED RUPTURE FROh! 30% POWER LEVEL WITH FAULTED l LOOP h1AIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) . (10' Btuh) (10' lbm) (10' Btu) I 100.0 366 5 0.437 122.000 146.346 l 101.0 365.0 0.435 122.365 146.781 l 10.1LO 363.1 0.433 122.728 147.214 l 103.0 361.3 0.430 123.089 147.644 l 104.0 359.5 0.428 123.449 148.072 l 105.0 357.8 0.426 123.806 148.498 l 106.0 356.1 OA24 124.162 148.922 l 107.0 354.5 0.422 124.517 149.344 r I 108.0 352.9 0.420 124.870 149.764 l 109.0 351.4 '418 125.221 150.183 l 110.0 349.9 0 #16 125.571 150.599 l 111.0 348.4 0.~ 5 125.919 151.014 l 112.0 347.0 u.a 3 126.256 151.427 l 113.0 345.7 0.411 126.612 151.838 l 114.0 344.3 0.410 126.956 152.248 I 115.0 343.1 0.408 127.299 152.656 l 116.0 341.9 0.407 127.641 153.063 l 117.0 340.6 0.405 127.982 153.469 l 118.0 339.5 0.404 128.321 153.872 l 119.0 338.3 0.403 128.660 154.275 A V Revision: 5 { Westinghouse 6.2-139 February 29,1996
- 6. Engineered Saf:ty F=tures Table 6.2.1.4-2 (Sheet 20 of 31) e' l
MASS AND ENERGY RELEASE DATA i FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT PRESSURE I I Thne Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (lbm/s) (10' Btu /s) (10' lbm) (10' Btu) l 120.0 337.2 0.401 128.997 154.676 l 121.0 336.2 0.400 129.333 155.076 l 122.0 335.1 0.399 129.668 155.475 l 123.0 334.1 0.397 130.002 155.872 l 124.0 333.1 0.396 130.336 156.269 l 125.0 332.2 0.395 130.668 156.664 l 126.0 3313 0.394 130.999 157.058 l 127.0 330.4 0.393 131.329 157.451 1 128.0 329.5 0.392 131.659 157.843 l 129.0 328.6 0.391 131.987 158.233 l 130.0 327.8 0.390 1I2.315 158.623 l 131.0 327.0 0.389 132.642 159.012 l 132.0 326.3 0.388 132.969 159.400 l 133.0 325.5 0.387 133.294 159.787 l 134.0 324.8 0.386 133.619 160.174 l 135.0 324.1 0.385 133.943 160.559 l 136.0 323.5 0.385 134.267 160.944 l 137.0 322.8 0.384 134.589 161.328 l 138.0 322.2 0.383 134.912 161.711 139.0 321.6 0.382 135.233 162.093 Revision: 5 e February 2a,1996 6.2-140 W Westirighouse
- 6. Engineered S:fety Fetures l
/ 'd l Table 6.2.1.4-2 (Sheet 21 of 31) i
! l MASS AND ENERGY RELEASE DATA l l l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l l ENDED RUI9'URE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT PRESSURE l l l Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (lbm/s) (10' Btu /s) (10' lbm) (16' Btu) l 140.0 321.0 0.382 135.554 162.475 l l I 141.0 320.5 0.381 135.875 162.856 l 142.0 319.9 0.380 136.195 163.236 l l 143.0 319.4 0.380 1 136.514 163.616 l l 144.0 318.9 0.379 136.833 163.995 l 145.0 318.4 0.379 137.151 164.373 l 146.0 318.0 0.378 137.469 164.751 l 147.0 317.6 0.377 137.787 165.129 I 148.0 317.2 f O) ! b/ l 149.0 316.8 0.377 138.104 165.506 0.377 138.421 165.882 l 150.0 316.4 0.376 138.737 166.258 l 151.0 316.0 0.376 139.053 166.634 l 152.0 315.7 0.375 139.369 167.009 l 153.0 315.4 0.375 139.684 167.384 l 154.0 315.1 0.375 140.000 167.759 l 155.0 314.9 0.374 140.314 168.133 l 156.0 314.6 0.374 140.629 168.507 I 157.0 314.4 0.374 140.943 168.881 l 158.0 314.2 0.373 141.258 169.254 l 159.0 314.0 0.373 141.572 169.627 L A L) Revision: 5 Y W85tingh0Use 6.2-141 February 29,1996
m rga
- 6. Engineered Sity Feitures 1
Table 6.2.1.4-2 (Sheet 22 of 31) O. I h1 ASS AND ENERGY RELEASE DATA l FOR THE CASE OF hIAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED ; I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I 1 PRODUCES HIGHEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Inte:ysted Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) I 160.0 313.8 0.373 141.885 170.000 l 161.0 313.7 0.373 142.199 170.373 l 162.0 313.5 0.373 142.513 170.746 l 163.0 313.4 0.372 142.826 171.118 l 164.0 313.3 0.372 143.139 171.490 , i 165.0 313.2 0.372 143.453 171.863
! l66.0 313.2 0.372 143.766 172.235 l 167.0 313.1 0.372 144.079 172.607 1 168.0 313.1 0.372 144.392 172.979 l 169.0 313.1 0.372 144.705 173.351 l 170.0 313.1 0.372 145.018 173.723 l 171.0 313.1 0.372 145.331 174.096 l 172.0 313.1 0.372 145.644 174.468 l 173.0 313.2 0372 145.958 174.840 I 174.0 313.3 0.372 146.271 175.212 l 175.0 313.3 0.372 146.584 175.585 l 176.0 313.4 0.372 146.898 175.957 l 177.0 313.5 0.373 147.211 176.330 l 178.0 313.6 0.373 147.525 176.702 l 179.0 313.8 0.373 147.839 177.075 O
Revision: 5 February 29,1996 6.2-142 [ W85tingfl00Se
l g..
- 6. Engine: red Saf;ty Factures j
jf3 Table 6.2.1.4-2 (Sheet 23 of 31) l MASS AND ENERGY RELEASE DATA l FOR THE CASE OF MAIN STEAM LINE FULL, DOUBLE I I l ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT PRESSURE I l l l Time Mass Flow Energy Flow Integrated Mass Integrated Energy l l (sec) Obm/s) (10' Btu /s) (10' lbm) (10' Btu) l 180.0 313.9 0.373 148.152 177.448 l l 181.0 314.0 0.373 148.466 177.822 l 182.0 314.2 0.373 148.781 178.195 l 183.0 314.4 0.374 149.095 178.569 l 184.0 314.5 0.374 149.410 178.942 I 185.0 314.7 0.374 149.724 179.316 l 186.0 314.9 0.374 150.039 179.691 l 187.0 315.1 0.375 150.354 180.065 l l 188.0 315.3 0.375 150.670 180.440 I U l 189.0 315.6 0.375 150.985 180.815 I l l 190.0 315.8 0.375 151.301 181.190 l 191.0 316.0 0.376 151.617 181.566 l 192.0 316.3 0.376 151.933 181.942 l 193.0 316.5 0.376 152.250 182.318
.I 194.0 316.7 0.377 152.566 182.695 l 195.0 317.0 0.377 152.883 183.072 I 196.0 317.2 0.377 153.201 183.449 l 197.0 317.5 0.377 153.518 183.826 l 198.0 317.8 0.378 153.836 184.204 l 199.0 318.0 0.378 154.1S4 184.582 i
~ t Revision: 5 T Westingh00S8 6.2-143 February 29,1996
G. Engineered Scf;ty Fectures Table 6.2.1.4-2 (Sheet 24 of 31) e l MASS AND ENERGY RELEASE DATA i FOR TIIE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT PRESSURE I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) I 200.0 318.3 0.378 154.472 184.960 l 202.0 318.7 0.379 155.110 185.718 I 204.0 319.2 0.379 155.748 186.477 l 206.0 319.8 0.380 156.388 187.237 l 208.0 320.4 0.381 157.029 187.999 l 210.0 320.9 0.382 157.671 188.762 l 212.0 321.5 0.382 158.313 189.527 I 214.0 322.0 0.383 158.957 190.293 I 216.0 322.5 0.383 159.603 191.060 l 218.0 323.0 0.384 160.249 191.828 I 220.0 323.5 0.385 160.896 192.597 I 222.0 324.0 0.385 161.544 193.368 1 224.0 324.4 0.386 162.193 194.139 l 226.0 324.9 0.386 162.842 194.912 l 228.0 325.2 0.387 163.493 195.685 l 230.0 325.6 0.387 164.144 196.459 l 232.0 325.9 0.388 164.7 % 197.234 l 234.0 326.1 0.388 165.448 198.010 l 236.0 326.4 0.388 166.101 198.786 l 238.0 326.6 0.388 166.754 199.563 o Revision: 5 February 29,1996 6.2-144 3 W85tingh00S6
- 6. Ergineered Saf;ty Features
,g)
Table 6.2.1.4-2 (Sheet 25 of 31) l I MASS AND ENERGY RELEASE DATA l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT PRESSURE I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) Obm/s) (10' Btu /s) 8 (10 lbm) (10' Btu) l 240.0 326.8 0.389 167.407 200.340 l 242.0 326.9 0.389 168.061 201.118 l 244.0 327.0 0.389 168.715 201.8 % I 246.0 327.0 0.389 169.369 202.673 l 248.0 327.0 0.389 170.023 203.451 l 250.0 327.0 0.389 170.677 204.229 l 252.0 326.9 0.389 171.331 205.007 l 254.0 326.8 0.389 171.985 205.784 l 256.0 326.7 0.388 172.638 206.561 'V l 258.0 326.5 0.388 173.291 207.337 l 260.0 326.3 0.388 173.944 208.113 l 262.0 326.0 0.388 174.596 208.889 l 264.0 325.7 0.387 175.247 209.663 l 266.0 325.4 0.387 175.898 210.437 l 268.0 325.0 0.386 176.548 211.210 l 270.0 324.6 0.386 177.197 211.982 l 272.0 324.2 0.386 177.846 212.753 l 274.0 323.8 0.385 178.493 213.523 l 276.0 323 3 0.384 179.140 214.292 l 278.0 322.8 0384 179.785 215.060 l (- N) Revision: 5 3 W85tingh011S8 6.2-145 February 29,1996 l
.-Jnamunisk 7
- 6. Engineered Sarity Futures Table 6.2.1.4-2 (Sheet 26 of 31) e l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l
LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT PRESSURE I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (thm/s) (10' Btu /s) (10' lbm) (10' Btu) I 280.0 3223 0.383 180.430 215.826 l 282.0 311.8 0383 181.074 216.592 l 284.0 3213 0.382 181.716 217356 l 286.0 320.7 0381 182.358 218.118 l 288.0 320.1 0381 182.998 218.879 l 290.0 319.5 0380 183.637 219.639 l 292.0 319.0 0.379 184.275 220.397 l 294.0 318.4 0378 184.912 221.154 l 296.0 317.8 0378 185.547 221.913 l 298.0 317.1 0377 186.181 222.664 I 300.0 316.5 0376 I86.815 223.416 1 302.0 315.9 0375 187.446 224.167 1 304.0 3153 0375 188.077 224.917 1 306.0 314.6 0.374 188.706 225.665 l 308.0 314.0 0373 189334 226.411 l 310.0 313.4 0.372 189.961 227.156 l 312.0 312.8 0372 190.587 227.899 I 314.0 312.1 0.371 191.211 228.641 I 316.0 311.5 L370 191.834 229382 l 318.0 310.9 0.369 192.456 230.121 l O l Revision: 5 February 29,1996 6.2-146 3 Westingh0Use i i
m
- 6. Engineered Sdety Fetures l
f~% l Table 6.2.1.4-2 (Sheet 27 of 31) l MASS AND ENERGY RELEASE DATA l l I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT PRESSURE I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) l (10' Btu) l 320.0 310.3 0.369 193.076 230.858 l 322.0 309.7 0.368 193.6 % 231.594 I 324.0 309.0 0.367 194.314 232.328 l 326.0 308.5 0.367 194.931 233.061 l l 328.0 307.9 0.366 195.546 233.793 l 330.0 307.3 0.365 196.161 234.523 l 332.0 306.7 0.364 196.774 235.252 l 334.0 306.1 0.364 197.386 235.979 ,,o) i( t V l I 336.0 338.0 305.5 0.363 197.997 236.705 304.9 0.362 198.607 237.430 1 I 340.0 304.4 0.362 199.216 238.153 l 342.0 303.8 0.361 199.824 238.874 1 344.0 303.3 0.360 200.430 239.595 l 346.0 302.7 0.360 201.036 240.314 l 348.0 302.2 0.359 201.640 241.032 l 350.0 301.6 0.358 202.243 241.748 I 352.0 301.1 0.358 202.845 242.464 1 354.0 300.5 0.357 203.446 243.177 l 356.0 300.0 0.356 204.046 243.890 l 358.0 299.5 0.356 204.645 244.601 f)
% .)
l Revision: 5 I W W9stingh00S8 6.2-147 February 29,1996 l
- 6. E gineered Saf;ty Fctures Table 6.2.1.4-2 (Sheet 28 of 31) e l MASS AND ENERGY RELEASE DATA I
FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT PRESSURE I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu's) (10' lbm) (10' Btu) l 360.0 298.9 0.355 205.243 245.311 l 362.0 298.5 0.354 205.840 246.020 l 364.0 297.9 0.354 206.436 246.728 l 366.0 297.4 0.353 207.031 247.434 l l 368.0 296.9 0.353 207.624 248.139 l l 370.0 296.4 0.352 208.217 248.843 l 372.0 295.9 0.351 208.809 249.546 I 374.0 295.4 0.351 209.400 250.247 l 376.0 294.9 0.350 209.989 250.948 I 378.0 294.4 0.350 210.578 251.647 ' l 380.0 294.0 0.349 211.166 252.345 I l 382.0 293.5 0.348 211.753 253.042 l l 384.0 293.1 0.348 212.339 253.~i38 l l 386.0 292.5 0.347 212.925 254.432 I 388.0 292.1 0.347 213.509 255.126 l 390.0 291.6 0.346 214.092 255.818 l 392.0 291.1 0.346 214.674 256.510 l 394.0 290.7 0.345 215.256 257.200 l 396.0 290.2 0.344 215.836 257.889 I 398.0 289.7 0.344 216.415 258.577 Revision: 5 e February 29,1996 6.2-148 [ W85tiligh0US8 l
r
- 6. E:gineered Safity Fectures l
l /'~h , l ! !,y Table 6.2.1.4-2 (Sheet 29 of 31) l MASS AND ENERGY RELEASE DATA I FOR TIIE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT PRESSURE l 1 I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (lbm/s) (10' Btu /s) 2 (10 lbm) (10' Btu) l 400.0 287.2 0.341 l 216.990 259.258 l l l 402.0 285.2 0.338 217.560 259.935 I 404.0 283.6 0.336 218.127 260.608 l 406.0 282.1 0.335 218.691 261.278 l l 408.0 280.7 0.333 219.253 261.944 ! l 410.0 279.4 0.331 219.812 262.607 I 412.0 278.2 0.330 220.368 263.267 l 414.0 276.8 0.328 220.922 263.923 l/^g I 416.0 275.3 0.327 221.472 264.577 l 418.0 273.7 0.325 222.020 265.226 l l 420.0 272.0 0.323 222.564 265.871 1 422.0 270.3 0.321 223.104 266.512 l 424.0 268.5 0.318 223.641 267.149 l 426.0 266.4 0.316 224.174 267.781 1 428.0 263.9 0.313 224.702 268.406 I 430.0 261.5 0.310 225.225 269.026 l 432.0 259.4 0.307 225.744 269.641 I 434.0 256.6 0.304 226.257 270.249 l 436.0 253.5 0.300 226.764 270.850 l 438.0 250.3 0.296 227.265 271.442 l l , /^g Revision: 5 Y W85tingh00S8 6.2-149 February 29,1996
- 6. Engineered Sity Fxtures Table 6.2.1.4-2 (Sheet 30 of 31) e l
MASS AND ENERGY RELEASE DATA l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT PRESSURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibnvs) (10' Btu /s) (10' lbm) (10' Bru) I 440.0 246.8 0.292 227.758 272.027 l 442.0 243.2 0.288 228.245 272.603 l 444.0 239.2 0.283 228.723 273.169 l 446.0 235.1 0.278 229.193 273.725 l 448.0 231.0 0.273 229.655 274.272 l 450.0 226.4 0.268 230.108 274.807 l 452.0 220.7 0.261 230.550 275.329 l 454.0 214.6 0.254 230.979 275.836 l 456.0 208.1 0.246 231.395 276.327 I 458.0 201.0 0.237 231.797 276.801 l 460.0 193.3 0.228 232.184 277.L7 l 462.0 185.1 0.218 232.554 277.694 l 464.0 176.3 0.208 232.907 278.109 l 466.0 166.8 0.196 233.240 278.501 l 468.0 156.4 0.184 233.553 278.869 l 470.0 145.2 0.170 233.843 279.210 l 472.0 133.6 0.157 234.110 279.523 l 474.0 123.3 0.144 234.357 279.811 l 476.0 111 3 0.130 234.580 280.071 1 478.0 98.7 0.115 234.777 280.301 Revision: S
- e. l February 29,1996 6.2-150 3 W8Stingh00S8 I I
- 6. Engineered Siftty F=tures i
Table 6.2.1.4-2 (Sheet 31 of 31) L MASS AND ENERGY RELEASE DATA I ! l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 30% POWER LEVEL WITH FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT PRESSURE I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 480.0 86.6 0.101 234.950 280.503 l 482.0 75.9 0.088 235.102 280.679 l 484.0 67.4 0.078 235.237 280.835 l 486.0 25.4 0.029 l 235.288 280.894 l 488.0 2.6 0.003 0 5.293 280.900 l 490.0 0.2 0.0003 235.293 280.901 ( l 492.0 0.0 0.0 235.293 280.901 C l
.n
( m_ Revision: 5 [ W85tinghouse 6.2-151 February 29,1996
l
- 6. Engineered S fety Fcr.tures l
I Table 6.2.1.4-3 (Sheet 1 of 27) l I MASS AND ENERGY RELEASE DATA l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE i l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT TEMPERATURE Initial steam generator mass ( lbm ) : 133650 Mass added by unisolatable steam ( lbm ) : 8419 , Mass added by feedwater flashing ( lbm ) : 8726 [ Initial steam pressure ( psia ) : 843.2 Feedwater line isolation at ( sec ) : 10.29 l Steam line isolation at ( sec ) 10.29 l Time Mass Flow Energy flow Integrated Mass Integrated Energy (sec) (thm/s) (10' Btu /s) (10' lbm) (10' Btu) l 0.0 0.0 0.0 0.0 0.000 l 0.1 10767.6 12.899 1.077 1.290 l 0.2 10754.4 12.884 2.152 2.578 I 0.3 10745.0 12.873 3.227 3.866 l 0.4 10735.7 12.862 4.300 5.152
- I 0.5 10726.5 12.851 5.373 6.437 l 0.6 10717.6 12.841 6.445 7.721 l 0.7 10708.8 12.830 7.516 9.004 1 0.8 10700.1 12.820 8.586 10.286 l 0.9 10691.6 12.810 9.655 11.567 l 1.0 10683.2 12.800 10.723 12.847 l 1.1 10675.0 12.791 11.791 14.126 l 1.2 4718.4 5.652 12.262 14.691 l 1.3 4665.1 5.589 12.729 15.250 l l 1.4 4626.8 5.544 13.192 15.805 l 1.5 4589.6 5.501 13.651 16.355 l 1.6 4553.7 5.458 14.106 16.901 l 1.7 4518.8 5.417 14.558 17.442 l 1.8 4485.0 5.378 15.006 17.980 l 1.9 4452.3 5.339 15.452 e8.514 O
Revision: 5 February 29,1996 6 2-152 W Westingh0use l l
l
- 6. Engineered Saftty Fcctures C'/\
Table 6.2.1.4-3 (Sheet 2 of 27) I h1 ASS AND ENERGY RELEASE DATA I FOR THE CASE OF A1AIN STEAh! LINE FULL DOUBLE ! l l ENDED RUPTURE FROh! 102% POWER LEVEL WITH FAULTED I LOOP MAIN STEAh! LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINh1ENT TEh1PERATURE l l Time Mass Flow Energy Flow Integrated Mass Integrated Energy l (sec) Obm/s) (10' Btu /s) (10' lbm) (10' Btu) l 2.0 4420.4 5.301 15.894 19.044 I 2.1 4389.5 5.265 16.333 19.571 l 2.2 4359.6 5.230 16.768 20.094 l 2.3 4330.3 5.195 17.202 20.613 l 2.4 4301.9 5.162 17.632 21.129 l 2.5 4274.3 5.129 18.059 21.642 l 2.6 4247.4 S.W / 18.484 22.152 l 2.7 4221.2 5.066 18.906 22.658
- e l 2.8 4105.6 5.036 19.326 23.162 b) l 2.9 4170.7 5.006 19.743 23.663 l 3.0 4146.3 4.978 20.157 24.160 l 3.1 4122.6 4.949 20.570 24.655 I 3.2 4999.4 4.922 20.979 25.148 l 3.3 4076.7 4.895 21.387 25.637 I 3.4 4056.1 4.871 21.793 26.124 1 3.5 4056.3 4.871 22.198 26.611 l 3.6 4049.8 4.863 22.603 27.098 l 3.7 4043.4 4.856 23.008 27.583 l 3.8 4037.4 4.849 23.411 28.068 l 3.9 4031.7 4.842 23.815 28.552 l
l (3 ,C _ Revision: 5 Y WBStingh00S8 6.2-153 February 29,1996
1
- t#tabast#:*
- 6. Ezgineered Sity Features l
1 Table 6.2.1.4-3 (Sheet 3 of 27) 0: l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT TEMPERATURE I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) Obm/s) (10' Btu /s) (10' lbm) (10' Btu) l 4.0 4025.7 4.835 24.217 29.036 1 4.1 4019.8 4.828 24.619 29.518 l 4.2 4013.9 4.821 25.021 30.000 l 4.3 4008.0 4.814 25.421 30.482 l 4.4 4002.1 4.807 25.822 30.963 1 4.5 3996.3 4.800 26.221 31.443 l 4.6 3990.4 4.793 26.620 31.922 l 4.7 3984.6 4.786 27.019 32.400 l 4.8 3978.9 4.779 27.417 32.878 I 4.9 3973.1 4.773 27.814 33.356 1 5.0 3967.5 4.766 28.211 33.832 1 5.1 3961.8 4.759 28.607 34.308 l 5.2 3956.3 4.753 29.002 34.783 i 5.3 3950.8 4.746 29.398 35.258 l 5.4 3945.3 4.740 29.792 35.732 l 5.5 3939.9 4.733 30.186 36.205 l 5.6 3934.6 4.727 30.580 36.678 i 5.7 3929.4 4.721 30.972 37.150 1 5.8 3924.2 4.715 31.365 37.622 1 5.9 3919.0 4.709 31.757 38.092 O Revision: 5 February 29,1996 6.2-154 3 Westingh0USS
- 6. E:gineered Saf2ty Fr.tzres 1
i
/
l Table 6.2.1.4-3 (Sheet 4 of 27) 1 I MASS AND ENERGY RELEASE DATA I i i FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED i ! I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT ( l PRODUCES IUGHEST CONTAINMENT TEMPERATURE 1 Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) Obm/s) (10' Btu /s) (10' lbm) (10' Btu) l 6.0 3914.0 4.703 32.148 38.563 I 6.1 3908.9 4.697 32.539 39.032 I 6.2 3904.0 4.691 32.929 39.501 i I 6.3 3899.1 4.685 33.319 39.970 l 6.4 3894.3 4.679 33.709 40.438 I 6.5 3889.5 4.674 34.098 40.905 l 6.6 3884.8 4.668 34.486 41.372 I 6.7 3880.1 4.662 34.874 41.838 p l- 6.8 3875.5 4.657 35.262 42.304 k, I 6.9 3870.9 4.652 35.649 42.769 I 7.0 3866.4 4.646 36.036 43.234 l 7.1 3861.9 4.641 36.422 43.698 l 7.2 3857.4 4.636 36.807 44.161 l 7.3 3852.9 4.630 37.193 44.624 1 7.4 3848.5 4.625 37.578 45.087 I 7.5 3844.1 4.620 37.962 45.549 l 7.6 3839.6 4.614 38.346 46.010 l 7.7 3835.2 4.609 38.729 46.471 I 7.8 3830.8 4.604 39.113 46.932 I 7.9 3826.3 4.599 39.495 47.392 l 1-m N Revision: 5 3 W95tingh00S8 6.2-155 February 29,1996 l l
wnamm ; C. E:gineered S faty Features i Table 6.2.1.4-3 (Sheet 5 of 27) l MASS AND ENERGY RELEASE DATA l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE TIIAT l PRODUCES HIGHEST CONTAINMENT TEMPERATURE I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' Ibm) (10' Btu) l 8.0 3821.8 4.593 39.877 47.851 1 8.1 3817.3 4.588 40.259 48.310 1 8.2 3812.8 4.583 40.640 48.768 l 8.3 3808.2 4.577 41.021 49.226 l 8.4 3803.6 4.572 41.402 49.683 l 8.5 3793.9 4.566 41.781 50.139 l 8.6 3794.1 4.560 42.161 50.595 l 8.7 3789.2 4.555 42.540 51.051 1 8.8 3784.3 4.549 42.918 51.506 l 8.9 3779.3 4.543 43.296 51.960 1 9.0 3774.2 4.537 43.674 52.414 I 9.1 3769.0 4.531 44.050 52.867 l 9.2 3763.7 4.524 44.427 53.319 l 9.3 3758.3 4.518 44.803 53.771 l 9.4 3752.7 4.511 45.178 54.222 I 9.5 3747.1 4.505 45.553 54.673 l 9.6 3741.3 4.498 45.927 55.122 1 9.7 3735.4 4.491 46.300 55.571 l 9.8 3729.3 4.483 46.673 56.020 I 9.9 3723.2 4.476 47.046 56.467 O Revision: 5 February 29,1996 6.2-156 [ W95tingh0US8
l
- 6. Engineered Sdety Features eDs G
Table 6.2.1.4-3 (Sheet 6 of 27) l MASS AND ENERGY RELEASE DATA l I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT TEMPERATURE I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (lbuds) (10' Btu /s) (10' lbm) (10' Btu) I 10.0 3716.9 4.469 47.417 56.914 l 10.1 3710.4 4.461 47.788 57.360 l 10.2 3703.8 4.453 48.159 57.806 l 10.3 3697.1 4.445 48.528 58.250 l l 10.4 3690.3 4.437 48.897 58.694 l 10.5 3684.4 4.430 49.266 59.137 l 10.6 3678.0 4.422 49.634 59.579 l 10.7 3671.4 4.415 50.001 60.021 l p l 10.8 3664.7 4.407 50.367 60.461
\b l 10.9 3657.8 4.398 50.733 60.901 l 11.0 3650.8 4.390 51.098 61.340 l 11.1 3643.6 4.381 51.462 61.778 l 11.2 3636.2 4.373 51.826 62.215 l 11.3 3628.8 4.364 52.189 62.652 l 11.4 3621.1 4.355
^
52.551 63.087 l 11.5 1855.2 2.231 52.737 63.310 l 11.6 1851.4 2.226 52.922 63.533 l 11.7 1847.5 2.221 53.107 63.755 I i1.8 1843.6 2.217 53.291 63.977 I 11.9 1839.6 2.212 53.475 64.198 l G Revision: 5 l Y W95tingh0US8 6.2-157 February 29,1996 l l
r.me. j- 6. Engineered Sciety Fe:tures Table 6.2.1.4-3 (Sheet 7 of 27) e l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT TEMPERATURE I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (see) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) I 12.0 1835.6 2.207 53.658 64.419 l 12.1 1831.5 2.202 53.842 64.639 l 12.2 1827.4 2.197 54.024 64.859 l 12.3 1823.2 2.192 54.207 65.078 l 12.4 1819.0 2.187 54.389 65.297 l 12.5 1814.8 2.182 54.570 65.515 l 12.6 1810.6 2.177 54.751 65.733 l 12.7 1806.3 2.172 54.932 65.950 l 12.8 1801.9 2.167 55.112 66.167 l 12.9 1797.6 2.162 55.292 66.383 l 13.0 1793.2 2.157 55.471 66.598 l 13.1 1788.8 2.151 55.650 66.814 l 13.2 1784.4 2.146 55.828 67.028 l 13.3 1779.9 2.141 56.006 67.242 l 13.4 1775.4 2.136 56.184 67.456 l 13.5 1770.9 2.130 56.361 67.669 l 13.6 1766.4 2.125 56.538 67.881 1 13.7 1761.9 2.119 56.714 68.093 l 13.8 1757.3 2.114 56.889 68.305 l 13.9 1752.8 2.109 57.065 68.516 Revision: 5 e February 29,1996 6.2-158 [ W65tiligh00S8
( mw
- 6. E gineered Sity Fe:tures l
l
")
(G l Table 6.2.1.4-3 (Sheet 8 of 27) I I hlASS AND ENERGY RELEASE DATA I I FOR THE CASE OF MAIN STEAh! LINE FULL DOUBLE l ENDED RUPTURE FROh! 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT TEMPERATURE I l Time Mass Flow Energy Flow Integrated Mass Integrated Energy l (sec) Obm/s) (10' Btu /s) (10' Ibm) (10' Btu) l l l 14.0 1748.2 2.103 57.240 68.726 ! l I4.1 1743.6 2.098 57.414 68.936 I 14.2 1739.0 2.092 57.588 69.145 l 14.3 1734.4 2.087 57.761 69.354 l l 14.4 1729.8 2.081 57.934 69.562 I 14.5 1725.2 2.076 58.107 69.769 i l 14.6 1720.6 2.070 58.279 69.976 l l 14.7 1716.0 2.065 58.450 70.183 i (~g i 14.8 1711.4 2.059 58.622 70.389 ib l 14.9 1706.8 2.054 58.792 70.594 1 15.0 1702.2 2.048 58.962 70.799 l 15.1 1697.6 2.043 59.132 71.003 l 15.2 1693.0 2.037 59.302 71.207 l 15.3 1688.4 2.032 59.470 71.410 l 15.4 1683.8 2.026 59.639 71.613 l 15.5 1679.3 2.021 59.807 71.815 l 15.6 1674.7 2.015 59.974 72.016 l 15.7 1670.1 2.010 60.141 72.217 l 15.8 1665.5 2.004 60.308 72.418 l 15.9 1660.8 1.999 60.474 72.618 l a G,r'%
- Revision
- 5 l [ WOStingl100S8 6.2-159 February 29,1996
1
- 6. Engineered Safity Fr.r.tures
[ i Table 6.2.1.4-3 (Sheet 9 of 27) O l MASS AND ENERGY RELEASE DATA I FOR TIIE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT TEMPERATURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy 5 (sec) (Ibm /s) (10' Btu /s) (10 lbm) (10' Btu) l 16.0 1656.1 1.993 60.639 72.817 l 16.1 1651.3 1.987 60.805 73.016 l 16 2 1646.6 1.982 60.969 73.214 l 16.3 1641.7 1.976 61.133 73.411 l 16.4 1636.9 1.970 61.297 73.608 l 16.5 1632.0 1.964 61.460 73.805 l 16.6 1627.0 1.958 61.623 74.001 l 16.7 1622.0 1.952 61.785 74.196 l 16.8 1617.0 1.946 61.947 74.391 l 16.9 1612.0 1.940 62.108 74.585 l 17.0 1606.9 1.934 62.269 74.778 l 17.1 1601.9 1.928 62.429 74.971 l 17.2 1596.7 1.922 62.589 75.163 l 17.3 1591.6 1.916 62.748 75.355 l 17.4 1586.4 1.910 62.906 75.546 l 17.5 1581.3 1.904 63.065 75.736 l 17.6 1576.1 1.897 63.222 75.926 ! I 17.7 1570.8 1.891 63.379 76.115 l 17.8 1565.6 1.885 63.536 76.303 1 l 17.9 1560.4 1.879 63.692 76.491 ! I l Revision: 5 February 29,1996 6.2-160 _ Westiligh00Se l l
- 6. Engineered Safity Fetures
!O Q 1 Table 6.2.1.4-3 (Sheet 10 of 27) i \ MASS AND ENERGY RELEASE DATA l l l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE ! l l ENDED RUITURE FROM 102% POWER LEVEL WITH FAULTED l l l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT ! i PRODUCES HIGHEST CONTAINMENT TEMPERATURE [ l l l Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) 8 (10 lbm) (10' Btu) ; l 18.0 1555.1 1.872 63.847 76.678 l 18.1 1549.9 1.866 64.002 76.865 l 18.2 1544.6 1.860 64.157 77.051 l 18.3 1539.3 1.853 64.311 77.236 I 18.4 1534.0 1.847 64.464 77.421 l 18.5 1528.8 1.841 64.617 77.605 l 18.6 1523.5 1.834 64.769 77.789 l 18.7 1518.2 1.828 64.921 77.971 l ,ew l 18.8 1512.9 1.822 65.072 78.154 l 18.9 1507.6 1.815 65.223 78335 l 19.0 1502.3 1.809 65.373 78.516 l 19.1 1497.0 1.803 65.523 78.696 l 19.2 1491.7 1.796 65.672 78.876 l 19.3 1486.5 1.790 65.821 79.055 l 19.4 1481.2 1.784 65.969 79.233 ! l 19.5 1475.9 1.777 66.117 79.411 l 19.6 1470.7 1.771 66.264 79.588 l 19.7 1465.4 1.765 66.410 79.765 l 19.8 1460.1 1.758 66.556 79.940 l 19.9 1454.9 1.752 66.702 80.116
.Cg Q
Revision: 5 [ W95tingh00S8 6.2-161 February 29,1996
Em _ ,,') j ;
- 6. Engineend Sity F=tures 1 i L .- 1 Table 6.2.1.4-3 (Sheet 11 of 27)
O l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT TEMPERATURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) Obm/s) (10' Btu /s) (10'lbm) (10' Btu) l 20.0 1449.7 1.746 66.847 80.290 l 20.5 1434.1 1.727 67.564 81.154 l 21.0 1409.0 1.697 68.268 82.002 l 21.5 1384.2 1.667 68.960 82.836 l 22.0 1359.7 1.638 69.640 83.655 l 22.5 1335.3 1.608 70.308 84.459 l 23.0 1311.8 1.580 70.964 85.249 l 23.5 1288.8 1.552 71.608 86.025 l 24.0 1266.4 1.525 72.24I 86.788 l 24.5 1244.5 1.499 72.864 87.537 l l 25.0 1223.2 1.473 73.475 88.274 l 25.5 1202.4 1.448 74.076 88.998 , l 26.0 1182.2 1.424 74.668 89.710 l 26.5 1162.5 1.400 75.249 90.410 l 27.0 1143.3 1.377 75.820 91.098 l 27.5 1124.5 1.354 l 76.383 91.775 l 28.0 1106.3 1.332 76.936 92.441 l 28.5 1088.5 1.311 77.480 93.097 l l 29.0 1071.3 1.290 78.016 93.742 l l 29.5 1054.8 1.270 78.543 94.377 l O Revision: 5 February 29,1996 6.2-162 T Westingh0use
l l l
.e s.
- 6. Engineered Sity F=tures l(l
] Table 6.2.1.4-3 (Sheet 12 of 27)
L MASS AND ENERGY RELEASE DATA i I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED t LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT TEMPERATURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Blu/s) (10' lbm) (10' Btu) l 30.0 1038.7 1.251 79.062 95.002 l 30.5 1023.1 1.232 79.574 95.618 i I 31.0 1008.0 1.214 80.078 96.2S I 31.5 993.3 1.196 80.575 96.823 l l 32.0 979.1 1.179 81.064 97.412 l 32.5 965.3 1.162 81.547 97.993 l 33.0 951.9 1.146 82.023 98.566 l 33.5 938.8 1.130 82.492 99.131 l 34.0 926.I 1.115 82.955 99.688 lV I 34.5 913.7 1.100 83.412 100.238 l 35.0 901.7 1.085 83.863 100.780 l 35.5 890.0 1.071 84.308 101.316 l 36.0 878.5 1.057 84.747 101.844 l 36.5 867.3 1.043 85.181 102.366 I 37.0 856.4 1.030 85.609 102.881 l 37.5 845.8 1.017 86.032 103.390 l 38.0 835.4 1.005 86.450 103.892 l 38.5 825.3 0.993 86.862 104.388 l 39.0 815.5 0.981 87.270 104.879 l 39.5 805.9 0.969 87.673 105.363 l
^LJ Revision: 5
[ W85tingh00S8 6.2 163 February 29,1996
m ir
- 6. Engineered Sity Fe:tures Table 6.2.1.4-3 (Sheet 13 of 27) e l MASS AND ENERGY RELEASE DATA l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I
ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT TEMPERATURE Time Mass Flow Energy Mow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 40.0 796.5 0.958 88.071 105.842 l 40.5 787.2 0.946 88.465 106.315 l 41.0 778.2 0.935 88.854 106.783 l 41.5 769.4 0.925 89.239 107.245 I 42.0 760.7 0.914 89.619 107.702 l 42.5 752.3 0.904 89.995 108.154 I 43.0 744.0 0.894 90.367 108.601 l 43.5 735.9 0.884 90.735 109.043 1 44.0 728.1 0.875 91.099 109.481 I 44.5 720.4 0.865 91.459 109.914 l 45.0 712.9 0.856 91.816 110.342 l 45.5 705.5 0.847 92.169 110.765 l 46.0 698.3 0.839 92.518 111.I85 l 46.5 691.3 0.830 92.863 111.600 l 47.0 684.4 0.822 93.206 112.011 l 47.5 677.6 0.814 93.544 112.417 l 48.0 671.0 0.8% 93.880 112.820 l 48.5 664.6 0.798 94.212 113.219 l 49.0 658.2 0.790 94.541 113.614 I 49.5 652.1 0.783 94.867 114.005 Revision: 5 e February 29,1996 6.2-1 64 T Westinghouse
g
- 6. Engineered Sity Fe:tures p
./7
( ) Table 6.2.1.4-3 (Sheet 14 of 27) I h1 ASS AND ENERGY RELEASE DATA l l FOR THE CASE OF h1AIN STEAh! LINE FULL DOUBLE l ENDED RUPTURE FROh! 102% POWER LEVEL WITH FAULTED I LOOP h1AIN STEAh! LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINh!ENT TEh1PERATURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibads) (10' Btu /s) (10' lbm) (10' Btu) l 50.0 646.1 0.775 95.190 114.393 I 50.5 640.2 0.768 95.510 114.777 I 51.0 634.4 0.761 95.828 115.158 l 51.5 628.7 0.754 96.142 115.535 l 52.0 623.2 0.748 96.454 115.909 l 52.5 617.7 0.741 96.762 116.279 l 53.0 612.4 0.735 97.069 116.646 1 53.5 607.3 0.728 97.372 117.011 I (] 54.0 602.2 0.722 97.673 117.372 V l 54.5 597.2 0.716 97.972 117.730 l 55.0 592.4 0.710 98.268 118.085 l 55.5 587.6 0.704 98.562 118.437 l 56.0 583.0 0.699 98.854 118.786 l 56.5 578.5 0.693 99.143 119.I33 l 57.0 574.0 0.688 99.430 119.477 1 57.5 569.7 0.683 99.715 119.819 I 58.0 565.5 0.678 99.997 120.157 l 58.' 561.3 0.673 100.278 120.494 l 59.0 557.2 0.668 100.557 120.827 I 59.5 553.3 0.663 100.833 121.159 T" L)T Revision: 5 Y W85tingh0US8 6.2-165 February 29,1996
wwww
- 6. Engineered Sity F=tures Table 6.2.1.4-3 (Sheet 15 of 27) e l
l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLP I ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED ! I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE TIIAT l PRODUCES HIGHEST CONTAINMENT TEMPERATURE I l l Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (thm/s) (10' Btu /s) (10' lbm) (10' Btu) l 60.0 549.4 0.658 101.108 121.4s9 l 60.5 545.6 0.654 101.381 121.815 l 61.0 541.9 0.649 101.652 122.139 I 61.5 538.2 0.645 101.921 122.461 l 62.0 534.7 0.640 102.I88 122.782 I l 62.5 531.2 0.636 102.454 123.100 l 63.0 527.8 0.632 102.718 123.416 I 63.5 524.5 0.628 102.980 123.730 l 64.0 521.2 0.624 103.241 124.042 l 64.5 518.0 0.620 103.500 124.352 l 65.0 514.9 0.616 103.757 124.660 l 65.5 511.8 0.613 104.013 124.966 l 66.0 508.9 0.609 104.267 125.271 I 66.5 506.0 0.605 104.520 125.573 l 67.0 503.1 0.602 104.772 125.874 l 67.5 500.3 0.599 105.022 126.I74 l 68.0 497.6 0.595 105.271 126.471 l 68.5 495.0 0.592 105.518 126.767 l 69.0 492.3 0589 105.764 127.062 l 69.5 489.8 0.586 106.009 127.355 O Revision: 5 February 29,1996 6.2-166 y Westinghouse j i
- 6. E:gineered Sat;ty Frtures O
l 1 Table 6.2.1.4-3 (Sheet 16 of 27) I MASS AND ENERGY RELEASE DATA 1 I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED I I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT TEMPERATURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibu'/s) (10' Btu /s) (10' lbm) (10' Btu) l 70.0 487.3 0.583 106.253 127.646 l 70.5 484.8 0.580 106.495 127.936 l 71.0 482.5 0.577 106.737 128.225 l 71 3 480.0 0.574 106.977 128.512 l 72.0 477.7 0.571 107.215 128.797 I 72.5 475.5 0.569 107.453 129.082 l 73.0 473.2 0.566 107.690 129.365 l 73.5 471.1 0.563 107.925 129.646 n l 74.0 469.0 0.561 108.160 129.926 l 74.5 466.9 0.558 108.393 130.206 I 75.0 464.9 0.556 108.626 130.483 l 75.5 462.9 0.553 108.857 130.760 l 76.0 460.9 0.551 109.088 131.036 l 76.5 459.0 0.549 109.317 131.310 l 77.0 457.1 0.546 109.546 131.583 l 77.5 455.3 0.544 109.773 131.855 l 78.0 453.5 0.542 110.000 132.126 l 78.5 451.7 0.540 110.226 132.396 l 79.0 45C D 0.538 110.451 132.665 l 79.5 448.3 0.536 110.675 132.933 i l J L' Revision: 5 3 WB5tlI1gh00S8 6.2-167 February 29,1996 l
- 6. Engineered Saftty Fe tures
, i Table 6.2.1.4-3 (Sheet 17 of 27)
I MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE i ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT TEMPERATURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 80.0 446.6 0.534 110.898 133.199 l 80.5 445.0 0.532 111.121 133.465 I 81.0 443.4 0.530 111.343 133.730 l 81.5 441.8 0.528 111.564 133.994 l 82.0 440.3 0.526 111.784 134.257 l 82.5 438.8 0.524 112.003 134.519 l 83.0 437.3 0.522 112.222 134.79.0 l 83.5 435.8 0.520 112.440 135.040 l 84.0 434.3 0.519 112.657 135.300 I 84.5 432.9 0.517 112.873 135.558 l 85.0 431.6 0.515 113.089 135.816 l 85.5 430.2 0.514 113.304 136.073 l 86.0 428.8 0.512 113.519 136.329 l 86.5 427.5 0.510 113.732 136.584 I 87.0 426.2 0.509 113.945 136.839 l 87.5 425.0 0.507 114.158 137.092 l 88.0 423.7 0.506 114.370 137.345 l 88.5 422.5 0.504 114.581 137.597 l 89.0 421.2 0.503 114.792 137.849 l l 89.5 420.1 0.501 115.002 138.099 l O Revision: 5 February 29,1996 6.2-168 3 Westinghouse 1
-mi
- 6. Ergineered Saf.ty Fe tures Mi i o
V Table 6.2.1.4-3 (Sheet 18 of 27) l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT TEMPERATURE i ! l l Tin:e Mass Flow Energy Flow Integrated Mass Integrated Energy l (sec) (Ibm /s) (10' Btu /s) (10' Ibm) (10' Btu) l 90.0 418.9 0.500 115.211 138.349 l 90.5 417.7 0.499 115.420 138.599 l 91.0 416.6 0.497 115.628 138.847 l 91.5 415.4 0.496 115.836 139.095 l 92.0 414.3 0.494 116.043 139.342 l 92.5 413.2 0.493 116.250 139.589 l 93.0 412.2 0.492 116.456 !39.835 I 93.5 411.1 0.491 116.661 140.080 l 94.0 410.0 ' (p) I 94.5 409.0 0.489 0.488 116.866 117.071 140.325 140.569 l 95.0 408.0 0.487 117.275 140.812 l 95.5 407.0 0.486 117.478 141.055 l 96.0 406.0 0.484 117.681 141.297 l 96.5 405.0 0.483 117.884 141.539 . l 97.0 404.0 0.482 118.086 141.780 1 l 97.5 403.1 0.481 118.287 142.020 l 98.0 402.1 0.480 118.488 142.260 l 98.5 401.2 0.479 118.689 142.500 l 99.0 400.3 0.477 118.889 142.738 l ' 99.5 399.4 0.476 119.089 l 142.976 i I l A t / V Revision: 5 3 W85tliigt100S8 6.2-169 February 29,1996
- 6. Engineered Screty Fectures
( Table 6.2.1.4-3 (Sheet 19 of 27) l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED I LOOP MA.C4 STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT TEMPERATURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy 5 (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) j l 100.0 398.4 0.475 119.288 143.214 l 101.0 397.1 0.474 119.685 143.688 ! l 102.0 395.4 0 472 120.081 144.159 l 103.0 393.7 0.469 120.474 144.629 I IM.0 392.0 0.467 120.866 145.096 s l 105.0 390.4 0.465 121.257 145.562 l 106.0 388.7 0.463 121.645 146.025 l 107.0 387.2 0.462 122.033 146.487 1 108.0 385.6 0.460 122.418 146.946 l 109.0 384.1 0.458 122.802 147.404 l 110.0 382.6 0.456 123.185 147.860 l 111.0 381.1 0.454 123.566 148.315 l 112.0 379.6 0.452 123.945 148.767 l 113.0 378.2 0.451 124.324 149.218 l 114.0 376.8 0.449 124.700 149.667 l 115.0 375.4 0.447 125.076 150.114 l 116.0 374.0 0.446 125.450 150.560 l 117.0 372.7 0.444 125.8 5 151.004 I i18.0 371.4 0.443 126.194 151.447 I 119.0 370.1 0.441 126.564 151.888 { i i I O Revision: 5 February 29,1996 6.2-170 g Westingflotise
1
- 6. Engine: red Safity Fe:tures I
t'D
^
Table 6.2.1.4-3 (Sheet 20 of 27) l MASS AND ENERGY RELEASE DATA l l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUITURE FROM 102% POWER LEVEL WITH FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT TEMPERATURE I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 120.0 368.8 0.439 126.933 152.327 l 121.0 367.5 0.438 127.300 152.765 l 122.0 366.3 0.436 127.667 153.201 l 123.0 365.1 0.435 128.032 153.636 l 124.0 363.9 0.433 128.3 % 154.070 l 125.0 362.7 0.432 128.758 154.502 1 126.0 361.6 0.431 129.120 154.932 l 127.0 360.4 0.429 129.480 155.362
- fw I 128.0 359.3 0.428 129.840 155.789
.- l 129.0 358.2 0.427 130.198 156.216 l 130.0 357.2 0.425 130.555 156 641 l 131.0 356.1 0.424 130.911 157.066 l 132.0 355.1 0.423 131.266 157.488 l 133.0 354.1 0.422 131.621 157.910 l 134.0 353.1 0.420 131.974 158.330 l 135.0 352.2 0.419 132.326 158.750 l 136.0 351.2 0.418 132.677 159.168 l 137.0 350.3 0.417 133.027 159.585 l 138.0 M 9.4 0.416 133.377 160.001 l 139.0 348.5 0.415 133.725 160.416 ir.
Revision: 5 [ W85tingh00S8 6.2-171 February 29,1996
- 6. Engineered Sity Fe tures
( I Table 6.2.1.4 3 (Sheet 21 of 27) l MASS AND ENERGY RELEASE DATA i FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE TIIAT I PRODUCES HIGHEST CONTAINMENT TEMPERATURE I I I Time Mass Flow Energy Flow Integrated Mass Integrated Energy I (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) l 140.0 347.7 0.414 134.073 160.829 l 141.0 346.8 0.413 134.420 161.242 l 142.0 346.0 0.412 134.766 161.654 l 143.0 345.2 0.411 135.111 162.065 l 144.0 344.5 0.410 135.455 162.475 l 145.0 343.7 0.400 135.799 162.884 l 146.0 343.0 0.408 136.142 163.292 1 147.0 342.4 0.407 136.485 163.700 l 148.0 341.7 0.407 13M26 164.106 l 149.0 341.0 0.406 .i e / El 164.512 I 150.0 340.3 0.405 137.508 164.917 l 151.0 339.7 0.404 137.847 165.321 l 152.0 339.1 0.404 138.186 165.725 l 153.0 338.6 0.403 138.525 166.128 l 154.0 338.0 0.402 138.863 166.530 l 155.0 337.5 0.402 139.200 166.931' l 156.0 336.9 0.401 139.537 167.332 I 157.0 336.4 0.400 139.874 167.'/33 l 158.0 336.0 0.400 140.210 168.132 1 159.0 335.5 0.399 140.545 168.531 O Revision: 5 February 29, '1996 6.2-172 [ W95tirigh00S8
._w
- 6. Ergineered Safity Fe:tures o
l N.) Table 6.2.1.4-3 (Sheet 22 of 27) I MASS AND ENERGY RELEASE DATA I I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT TEMPERATURE I l Time Mass Flow Energy Flow Integrated Mass Integrated Ec-rgy (rec) (IbtrJs) (10' Ltu/s) (10' lbm) (10' Btu) I 160.0 335.1 0.399 140.880 168.930 l 161.0 334.6 0.398 141.215 169.328 l 162.0 334.2 0.398 141.549 169.726 l 163.0 333.9 0.397 141.883 170.123 I 164.0 333.5 0.397 142.217 170.520 l 165.0 333.1 0.396 142.550 170.916 l 166.0 332.8 0.396 142.883 171.312 l 167.0 332.5 0.396 143.215 171.707 , pg l 168.0 332.2 0.395 143.547 172.102 l 169.0 332.0 0.395 143.879 172.497 l 170.0 331.7 0.395 144.211 172.892 l 171.0 331.5 0.394 144.542 173.286 l 172.0 331.3 0.394 144.874 173.680 l 173.0 331.0 0.394 145.205 174.074 l 174.0 330.9 0.394 145.536 174.468 l 175.0 330.7 0.393 145.866 174.861 l 176.0 330.5 0.393 146.197 175.254 l 177.0 330.4 0.393 146.527 175.647 l 178.0 330.3 0.393 146.857 176.040 l 179.0 330.1 0.393 147.187 176.432 l f' V Revision: 5 [ W85tingh00S8 6.2-173 February 29,1996 l
- 6. Ergineered S:fety Fe:.tures I
Table 6.2.1.4-3 (Sheet 23 of 27) O l l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT I PRODUCES HIGHEST CONTAINMENT TEMPERATURE I I Time Mass Flow Energy Mow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) I 180.0 330.0 0.393 147.518 176.825 l 181.0 329.9 0.392 147.847 177.217 l 182.0 329.9 0.392 148.177 177.61u l 183.0 329.8 0.392 148.507 178.002 l 184.0 329.7 0.392 148.837 178.394 l 185.0 329.7 0.392 149.166 178.786 l 186.0 329.6 0.392 149.496 179.178 l 187.0 329.6 0.392 149.826 179.570 l 188.0 329.6 0.392 150.155 179.962 1 189.0 329.6 0.392 150.485 180.354 l 190.0 329 6 0.392 150.814 180.746 l 191.0 329.6 0.392 151.144 181.138 l 192.0 329.6 0.392 151.474 181.530 l 193.0 329.6 0.392 151.803 181.922 l 194.0 329.6 0.392 152.133 182.314 l 195.0 329.7 0.392 152.463 182.706 l 196.0 329.7 0.392 152.792 183.099 l 197.0 329.7 0.392 153.122 182.491 l 198.0 329.8 0.392 153.452 183.883 l 199.0 329.8 0.392 153.781 184.275 , O Revision: 5 February 2>,1996 6.2-174 W Westinghouse
l l
- 6. Engi;eered Safety Features ex I
V Table 6.2.1.4-3 (Sheet 24 of 27) l l MASS AND ENERGY RELEASE DATA l FOR TIIE CASE OF MAIN STEAM LINE FULL DOUBLE , i ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE TIIAT I PRODUCES HIGHEST CONTAINMENT TEMPERATURE I Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Btu) I 200.0 329.9 0.392 154.111 184.667 I 202.0 329.9 1 0.392 154.771 185.452
! 204.0 330.1 0.393 155.431 186.237 l 206.0 330.1 0.393 156.092 187.023 ;
l 208.0 330.3 0.393 156.752 187.808 l 210.0 330.4 0.393 157.413 188.594 i l 212.0 330.5 0.393 158.074 189.381 l 214.0 330.6 0.393 158.735 190.167 gs l 216.0 330.7 0.393 159.3 % 190.954 V l 218.0 330.8 0.393 160.058 191.740 l 220.0 330.8 0.393 160.720 192.527 l 222.0 330.8 0.393 161.381 193.314 l 224.0 330.9 0.394 162.043 194.102 l 226.0 330.9 0.394 162.705 194M89 l 228.0 330.8 0.393 163.366 194.676 l 230.0 330.8 0.393 164.028 196.462 l 232.0 330.7 0.393 164.689 197.249 l 234.0 330.6 0.393 165.350 198.035 l 236.0 330.4 0.393 166.011
~
198.821 l 238.0 330.3 0.393 166.672 199.607 l l ~./ () I Revision: 5 3 Westingfl00Se 6.2-175 February 29,1996
rs. Engineered Sity Futures Table 6.2.1.4-3 (Sheet 25 of 27) e l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATsON VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT TEMPERATURE I i Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (Ibm /s) (10' Btu /s) (10' lbm) (10' Bio) l 240.0 330.0 0.393 167.332 200.392 l 242.0 329.8 0.392 167.991 201.177 l 244.0 329.6 0.392 167.650 201.961 l 246.0 329.3 0.392 169.309 202.744 I 248.0 329.0 0.391 169.967 203.5'.7 l 250.0 328.7 0.391 170.625 204.308 l 252.0 328.4 0.391 171.281 205.090 l 254.0 328.0 0.390 171.937 205.870 l 256.0 327.6 0.390 172.593 206.649 l 258.0 327.2 0.389 173.247 207.427 l 260.0 326.8 0.389 173.901 208.205 l 262.0 326.4 0.388 174.553 208.981 l 264.0 325.9 ' O.388 175.205 209.756 l 266.0 325.5 0.387 175.856 210.530 ; I 268.0 324.9 0.386 176.506 211.303 1 l 270.0 324.5 0.386 17,^.155 212.074 l l 272.0 324.0 0.385 177.803 212.845 l 774.0 323.4 0.385 178.450 213.614 l 276.0 322.9 0.384 179.095 214.382 l 278.0 322.4 0.383 179.740 215.148 1 I e Revision: 5 February 29,1996 6.2-176 [ West lDgh0US8
- mamam$t: :
- 6. Ergineered Saf1ty Futures V.CN Tele 6.2.1.4-3 (Sheet 26 of 27) l MASS AND ENERGY RELEASE DATA i
l FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE I ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED I LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT PRODUCES HIGHEST CONTAINMENT TEMPERATURE I Thne Mass Flow Energy Flow Integrated Mass l Integrated Energy l (sec) (ibm /s) (10' Btu /s) 8 (10 lbm) ' (10' Btu) I 280.0 321.8 0.383 180.384 215.913 l 282.0 321.2 0.382 181.026 216.677 l 284.0 320.7 0.381 181.667 217.440 l 286.0 320.1 0.380 182.308 218.201 l 288.0 319.6 0.380 182.947 218.960 l 290.0 318.9 0.379 183.584 219.719 l 292.0 318.4 0.378 184.7.21 220.476 ("s l 294.0 317.8 0.378 184.857 221.231
' I 296.0 317.2 0.377 185.491 221.985 l 298.0 316.6 0.376 186.124 222.738 I 300.0 316.0 0.376 186.756 223.489 l 302.0 315.4 0.375 187.387 224.239 1 304.0 314.9 0 374 188.017 224.987 I 306.0 314.2 0.373 188.645 225.734 l 308.0 312.1 0.371 189.270 226.476 I 310.0 307.0 0.365 189.884 227.206 l 312.0 303.1 0.360 190.490 227.926 I 314.0 298.6 0.355 191.087 228.635 l 316.0 293.9 0.349 191.675 229.333 l 318.0 288.9 0.343 192.253 230.019 l
1 f% Revision: 5 T Westirighouse 6.2-177 February 29,1996
- 6. Engineered Sity Fctures Table 6.2.1.4-3 (Sheet 27 of 27)
O l MASS AND ENERGY RELEASE DATA I FOR THE CASE OF MAIN STEAM LINE FULL DOUBLE l ENDED RUPTURE FROM 102% POWER LEVEL WITH FAULTED l LOOP MAIN STEAM LINE ISOLATION VALVE FAILURE THAT l PRODUCES HIGHEST CONTAINMENT TEMPERATURE Time Mass Flow Energy Flow Integrated Mass Integrated Energy (sec) (1bm/s) (10' Btu /s) 5 (10 lbm) (10' Btu) l 320.0 283.7 0.337 192.820 230.692 I 322.0 278.2 0.330 193.376 231.352 l 324.0 272.1 0.323 193.921 231.997 l 326.0 265.6 0.315 194.452 232.627 l 328.0 258.4 0.306 194.969 233.240 l 330.0 250.6 0.297 195.470 233.833 l 332.0 241.8 0.286 195.953 234.405 l 334.0 232.0 0.274 196.417 234.954 l 336.0 221.1 0.261 196.863 235.477 I 338.0 209.0 0.247 197.278 235.971 I 340.0 195.4 0.230 197.668 236.431 l 342.0 180.9 0.213 198.030 236.858 1 344.0 165.5 0.195 198.361 237.247 I 346.0 148.6 0.175 198.658 237.596 1 348.0 131.0 0.153 198.920 237.903 l 350.0 115.4 0.135 199.151 238.173 l 352.0 97.3 0.113 199.346 238.400 l 354.0 84.3 0.098 199.514 238.596 l 356.0 26.4 0.031 199.567 238.657 l 358.0 1.3 0.001 199.570 238.660 I 360.0 0.1 0.0001 199.570 238.660 l t l 362.0 0.0 0.0 199.570 238.660 O Revision: 5 l February 29,1996 6.2-178 [ W65tiflgl10US0 1
ir
- 6. Ergineered Safety Features '
Table 6.2.1.4-4 PLANT DATA USED FOR MASS AND ENERGY RELEASES DETERMINATION Plant data for all cases: Power, Nominal Rating (MWt) 1940 Nominal RCS Ibw (GPM) 194200 l t Nominal Full Load Ta vg ('F) 565.9 Nominal RCS Pressure (psia) 2250 I Nominal Steam Temperature (*F) 5183 Nominal Feedwater Enthalpy (BTU /lbm) 413.8 l l 1 l s__./
'0 ' ~~
Revision: 5 l W Westilighouse 6.2-179 February 29,1996
l ..
- 6. Engineered Saftty Fe:tures l
l O l Table 6.2.1.5-1 (Sheet 1 of 2) DECLG BREAK / MASS ENERGY RELEASES Time Mass Release Energy Release l (sec) (Ibm /sec) (Bru/sec) 0.01125 5099.869 2670001 0.51 51055.6 26504774 1.01 51273.84 26680200 1.51 49860.99 26033554 2.01 46206.26 24306452 2.51 41412.38 22093776 3.01 35793.19 19400192 3.51 30418.92 16703766 4.01 26853.12 15011002 l 4.51 246Gt.13 13976796 5.01 22754.77 13127462 5.51 20785.91 12302017 6.01 19052.57 11643622 6.51 18187.08 11179082 7.01 17869 10899948 7.51 17087.62 10497992 8.01 15744.36 9907718 8.51 14096.94 9156074 9.01 13299.04 8703237 9.51 12671.68 8305148 10.01 11523.11 7748078 11.01 10783.73 7150337 l l 12.01 9524.35 6427988 l 13.01 83 %.749 5768359 I i Revision: 5 e February 29,1996 6.2-180 3 Westinghouse
,r1 & u 4 ->a -.J . +
- 6. Ergineered Safity Fe:tures
- O \
L l \ Table 6.2.1.5-1 (Sheet 2 of 2) l DECLG BREAIUMASS ENERGY RELEASES Time Mass Release Energy Release l (sec) (Ibm /sec) (Btu /sec) 14.01 7543.894 5163321 15.01 6243.189 4418082 16.01 5291.53 3756248 17.01 4250.511 3093208 18.01 3642.385 2505180 19.01 2792.771 2093299 20.01 2480.567 1713555 21.011 2149.12 1480193 22.01 1839.571 1264136 23.01 1799.34 1082463 24.01 1651.465 973969.9 25.01 1452.628 820180.9 26.01 1458.903 741079.4 27.01 1062.866 589754.2 28.01 813.4185 475906 l 29.01 826.8542 395359.7 30.01 547.0674 303974.3 31.01 433.1182 234913.5 32.01 168.5994 132740.8 33.01 0. O. l^)
's_./
Revision: 5 Y W8stinghollse 6.2-181 February 29,1996 l
4MM-
- 6. Engineered Safety Features n-Table 6.2.2-1 PASSIVE CONTAINMENT COOLING SYSTEM PERFORMANCE PARAMETERS (Westinghouse Proprietary]
(Provided under separate cover] O Revision: 5 February 29,1996 6.2-182 N Westinghouse
l V .. .
- 6. Engineered S .fety Fe:tures .
o
.V Table 6.2.2-2 COMPONENT DESIGN PARAMETERS ,A [ Westinghouse Proprietary]
() [Provided under separate cover] ( jC9 l Revision: 5 [ WeStingh00$e 6.2-183 February 29,1996
??~S I "
- 6. Engineered Sity Fee.tures L__
Table 6.2.2-3 O FAILURE MODE AND EFFECTS ANALYSIS - PASSIVE CONTAINMENT COOLING SYSTEM ACTIVE COMPONENTS O [ Westinghouse Proprietary] [Provided under separate cover] d l l l I O' Revision: 5 February 29,1996 6.2-184 W Westinghouse ! { L l
m
- 6. E gi;eered S:fety Fe:tures Table 6.2.3-1 (@
Containment Mechanical Pene0 Containment Penetration System Sleeve LD. Sequence Line Flow Size or Closed Sys Qty Size Type Operato7 N o. Reg Guide IRC l CAS Pol C03 Service air in In 2" 56 No 1 2* Globe Manual 1 2" Check P02 CO2 Instrument air in 2" 56 No 1 2" Globe Air in i 2" Check CCS P03 C01 IRC loads in in 8" $6 No 8" 1 Gate Motor 1 8" Gate Motor l I" Check PO4 CO2 IRC loads out Out 8" 56 No 1 8" Gate Motor 1 8" Gate Motor l 1 I" Check CVS P05 COI Spent resin Out 2" 55 No 2" 1 Ball Manual flush out 1 2" Ball Manual l 1 1" Relief - l P06 CO2 letdown Out 2" 55 No i I" Relief - l 1 2" Globe Air 1 2" Globe Air P07 C03 Charging in 3" 55 No 1 3" Globe Motor (- x- l l P08 C04 H3 injection to in l' 55 No 1 1 3" 1" Globe Globe Motor Air RDS I 1" Check - P09 C10 Water to CMT in 2" 55 No 1 2" Globe Air and 1 2" Check - accumulators DWS P10 C01 Demin. water In 2" 56 No 2" Globe Manual l supply 1 1 2" Check - FHS Pil Fuel transfer N/A 36" 56 No 1 36" Blind N/A Range l FPS P12 C01 Pire protection In 4" 56 No 1 4" Gate Manual i standpipe sys. 1 4" Check - l W Westinghouse l --
pMIj%d l [ ""' Q[:d .hb 1 of 4) ns cf;d Isolation Valves Algo AVd k [ Apertufa L Isolation Device Test Actuation Mode Location Position Signal Closure Type l & Medium Direction N-S-A Primary Secondary Times Note ORC C-O-C None Manual None N/A C Air Forward IRC C-O-C None Self None N/A ORC O-O-C T Automatic Remote Manual std. C Air Forward IRC O-O-C None Self None N/A ORC O-O-C S Automatic Remote Manual std. C Air Forward IRC O-O-C S Automatic Remote Manual std. C-C-C None Self None C2.C O O-C S Automatic Remote Manual std. C Air Forward IRC 0 4-C S Automatic Remote Manual std. IRC C-C-C None Self None ORC C 4-C None Manual None N/A C Air Forward IRC C-C-C None Manual None N/A IRC C-C-C None Self None N/A IRC C-C-C None Self None N/A C Air Forward ORC O-O-C T Automatic Remote Manual std. IRC O-O-C T Automatic Remote Manual std. ORC 0-O-C TLP Automatic Remote Manual std. C Air Forward IRt O-O-C TLP Automatic Remote Manual std. OkC 04-C T Automatic Remote Manual std. C Air Forward IRC O-C-C None Self None N/A ORC C-C-C T Automatic Remote Manual std. C Air Forward IRC C-C-C None Self None N/A ORC C-O-C None Manual None N/A C Air Forward IRC C-O C None Self None N/A IRC C-O-C None N/A N/A N/A B Air Forward ORC C-C-C None Manual None N/A C Air Forward IRC C-C-C None Self None N/A l l l. 090 0700 4 . O Revision: 5 February 29,1996 6.2-185 n -.
e.# amusass
- 6. Engineered SJ;ty Fztures Table 6.2.3-1 (?
Containment Mechanical Penei Containment Penetration System Penetration P&ID GDC Slette I.D. Sequence Line Flow Size er Closed Sys Qty Size Type Operat N o. Ret Guide IRC PCS P13 C01 Cont. pressure N/A l' RGl.141 Yes I Globe l' Manud 1 Bellows N/A P14 CO2 N/A 1" RGl.141 Yes I 1" Globe Manud i Bellows N/A P15 C03 N/A 1" RGl.141 Yes i 1" Globe Manual 1 Bellows N/A P16 C04 N/A RGl.141 Yes I" Globe l' 1 Manud 1 Bellows N/A PSS Pl7 C01 RCS/PSX/CVS Out 3/8" 55 No 1 3/8" Globe Mr samples out 2 3/8" Globe Mr CO2 Cont. air Out 3/8" 56 No 3/8" Globe Mr I samples out 1 2 3/8" Globe Mr l C03 RCS/ Cont. air In 56 l' No 1 1" Globe Mr sample return i 1" Check - Spare N/A 3/8" 56 No 1 3/8" Cap N/A 1 3/8" Cap N/A PXS PIB C01 N 2to In 1" 55 No I 1" Globe Air accumulators 1 1" Check - RNS P19 COI RCS to RHR Out 10" 55 No ( pump 2 10" Gate Motor ' I 10" Gate Motor i 10" Gate Motor 1 3" Relief Self 1 Cate Manudl 1 3/8" " Globe Mannd I l P20 CO2 RHR pump to in 8" 55 No 1 8" Gate Motor RCS 1 8" Che k - SFS P21 C01 IRWST/Ref. In 4" 56 No i 4" Gate Motor cav. SFP pump I 4* Check - discharge l P22 CO2 IRWST/Ref. Out 6* 56 No 1 6" Gate Motor cav. punf. out i 6" Gate Motor l 1 1* Check - l i g: g-W- Westinghouse
. mage
W e-
%I4 et 2 of 4) tions crd Isolation Valves isolation Device Test Actuation Mode Location Position Signal Closure Type l & Medium Direction N-S-A Primary Secondary Times Note ORC O-04 None Manual None N/A A Air Forward IRC C-C-C N/A None N/A ORC O-O-O None Manual None N/A A Air Forward IRC C-C-C N/A None N/A ORC O-O-O None Manual None N/A A Air Forward IRC C-C-C N/A None N/A OrtC O-O-O None Manual None N/A A Air Forward IRC C-C-C N/A None N/A ORC C-C C T Automatic Remote Manual std. C Air Forward IRC C-C-C T Automatic Remote Manual std.
ORC C-C-C T Automatic Remote Manual std. C,4 Air Forward IRC C-C-C T Automatic Remote Manual std. ORC C-C-C T Automatic Remote Manual std. C Air Forward IRC C-C-C None Self None N/A ORC IRC C-C-C C-C-C N/A N/A N/A N/A N/A N/A N/A N/A A Air Forwas( W hh
- ORC O-O-C T Automatic Remote Manual std. C Air Forwwd 5 .
IRC C-C-C None Self None N/A IRC C-O-C None Remote Manual None std. C,4,b Air IRC C-O-C HR Automatic Remote Manual std.
-;, L J Reverse ..
~<.X
+0 0D ORC C-O-C HR Automatic Remote Manual std. Forward = y~
< ~'
IRC C <-C None Self None N/A - Reverse IRC C-C-C None Manual None N/A Forward IRC C-C-C None Manual None N/A Reverse ORC C-O-C HR Automatic Remote Manual std. C,4 Air Forward IRC C-O-C Nc.ne Self None N/A ORC C-O-C T Automatic Remote Manual std. C Air Forward IRC C-O-C None Self None N/A ORC C-O-C T Automatic Remote Manual std. C Air Forward IRC C-O C T Automatic Remote Muual std. IRC C-C-C None Self None N/A l C9old76on OP Revision: 5 i February 29,19% ! 6.2-187
p~
- 6. E23ineered Sity F :tures Table 6.2.3-1 (%
l Containment Mechanical Pened 1 Containsneat Penetration S Penetration P&lD GDC i Sleeve I.D. Sequence Line Flow Slee or Closed Sys Qty Size Yype O No. Res Guide IRC __ SGS P23 C01A Main steamline 01 Out 32" 57 Yes 1 32' Gate Pneu I 6" Gate May 3 8" Safety - I I 2" Globe Air i 1 3" Globe Air i P24 ColB Main steamline 02 Out 32" 57 Yes 1 32' Gate Pneu 1 6* Gate Motq 3 8" Safety -- 1 2" Globe Air i 1 3* Globe Air I P25 CO2A Main feedwater 01 In 16" $7 Yes 1 16" Gate Pneu P26 CO2B Main feedwater 02 In 16" 57 Yes 1 16" Gate Pneu l P27 CO3A SG blowdown 01 Out 4" 57 Yes 1 4" Globe Air l P28 C03B SG blowdown 02 Out 4" 57 Yes 1 4* Globe Air ' P29 C04 SG blowdown in 3" 57 Yes 1 3* Globe MI: recirculation l P31 COSA Startup feedwater 01 in 4" 57 Yes 1 4" Globe Mod I,s t l P33 C05B - Startup feedwater 02 In 4" 57 Yes i 4* Globe Mots
\ VFS P30 C01 Cont. air filter supply in 36" 56 No 1 18" Butterfly Air !
l (L./
- l. I 18" Butterfly Air i l 36" Flange N/A
! 36' Flange N/A; P32 CO2 Cont, air filter Out 36" 56 No I 18" Butterfly Air j l exhaust i 18" Butterfly Air : 36" Flange N/E 36" Flange N//A VWS P34 Col Fan Coolers out Out 8" 56 No 1 8* Butterfly Pneu I 8" Butterfly Pneu P35 CO2 Fan coolers in in 8' 56 No 1 10' Butterfly Pneu
! 10" Butterfly Pneu l
l f 1 j y%s ~ M
Mil beet 3 of 4) tio:::s cnd Isolation Valves Isolation Device Test Actuation Mode trator Location Position Signal Closure Type l & Medium Direction N-S-A Primary Secondary Times Note natic ORC O-C-C MS Automatic Remote Manual 5 sec A.2 N2 Forward r ORC O-O-C LSL Automatic Remote ManuaJ 5 see ORC C-C-C None Self None N/A ORC O-O-C MS Automatic Remote Manu J std. ORC C-C 4 MS Automatic Remote Manual std. natic ORC O-C-C MS Automatic Remote Manual 5 see A.2 N2 Forward r ORC O-O-C LSL Automatic Remote Manual 5 see ORC C-C-C None Self None N/A ORC O-O-C MS Automatic Remote Manual std. ORC C-C-C MS Automatic Remote Manual std. natic ORC O-C-C MF Automatic Remote Manual 5 see A.2 H09 Forward natic ORC 0-C-C MF Automatic Remote Manual 5 see A.2 Hy0 Forward ORC O-O-C PRHR Automatic Remote Manual std. A.2 H03 Forward ORC O-O-C PRHR Automatic Remote Manual std. A.2 H09 Forward ial ORC C-O-C None Manual None N/A A2 HO2 Forward r ORC C-O-C LTC Automatic Remote Manual std. A.2 H,0 Forward r ORC C-O-C LTC Automatic Remote Manual std. A.2 H,0 Forward ORC C-O-C T.HR Automatic Remote Manual N/A C Air Forward IRC C-O< T.HR Automatic Remote Manual N/A ORC C-C-C N/A N/A N/A IRC C-C-C N/A N/A N/A ORC CoC T Automatic Remote Manual 5 sec. C Air Forward IRC C-O-C T Automatic Remote Manual 5 sec. ORC C-C-C N/A N/A N/A N/A IRC C-C-C N/A N/A N/A N/A inatic ORC O-O-C T Automatic Remote Manual std. C,3,4 Air Forward rnatic IRC O-O-C T Automatic Remote Manual std. roatic ORC O-O-C T Automatic Remote Manual std. C,3,4 Air Forward tnatic IRC O-04 T Automatic Remote Manual std. ApERTORE y 4 pr a en m 99D lO700 @ O Revision: 5 February 29,1996 6.2-189
- 6. Ergi:eered Sity Fa.tures l
l Table 6.2.3-1 (S Containment Mechanical Penet Contalarnent Penetration 3 g,, Penetration P&lD GDC Sleeve I.D. Sequence Line Flow Size or Closed Sys Qty Size Type O No. Reg Guide IRC WLS P36 C01 Reactor coolant drain Out 2" 56 No 1 2" Globe Air tank out i 2" Globe Ai P37 CO2 Reactor coolant drain Both I" 56 No 1 1" Globe Ab tank gas 1 1" Globe Ab P38 C03 Normal cont. sump Out 2" 56 No 1 2" Globe Ab 1 2" Globe Air SPARE P39 N/A 12" 56 No 1 12" Flange N/- l 1 l SPARE P40 N/A 12" 56 No 1 12" Flange N/.- 1 l SPARE P41 N/A 12" 56 No 1 12" Flange N/, I l SPARE P42 N/A 12" 56 No I 12" Flange N/< l l SPARE P43 N/A 12" 56 No 1 12" Flange N/.- 1 CNS H01 N/A Main equipment hatch N/A 264" 56 No 1 Double Scaled Hatch H02 N/A Maintenance hatch N/A 192" 56 No I Double ; Scaled i Hatch H03 N/A Personnel hatch N/A 118" 56 No 1 Double i Scaled i Hatch H04 N/A Personnel hatch N/A 118" 56 No 1 Double Scaled Hatch l l l l W-Westinghouse I
11 _ w. et MY I gg La e u - - - Eni CARD of 4) fdso Availablo on s r.nd Isolation Valves AMMN Isolation Device Test Actuation Mode Loc: ties Posiden Signal Closure Type l & Medium Direction N-S-A Primary Secondary Times Note
- ORC O-O-C T Au'omatic Remote Manual std. C Air Forward IRC O-O< T Automatic Remote Manual std.
ORC C-C-C T Automatic Remote Manual std. C Air Forward IRC C-C-C T Automatic Remote Meiual std. , ORC C-C-C T Automatic Remote Manual std. C Air Forward IRC C-C-C T Autnmath Remote Manual std. ORC C-C-C N/A N/A N/A N/A B Air Forward IRC ORC C-C-C N/A N/A N/A N/A B Air Forward ORC C-C-C N/A N/A N/A N/A B Air Forward ORC C-C-C N/A N/A N/A N/A B Air Forward ORC C-C-C N/A N/A N/A N/A B Air Forward IRC C-C-C None Manual None N/A B Air Forward IRC C-C-C None Manual None N/A B Air Forward IRC C-C-C None Manual None N/A B Air Forward IRC C-C-C None Manual None N/A B Air Forward C: 90(0700%-cd Revision: 5 February 29,1996 6.2-191
- 6. Engineered Saf;ty F.ctures Table 6.'
Containment Mechanical Penei Explanation of Heading and ; System: Fluid system penetrating containment Containment Penetration: These fields refer to the penetration itself Penetration Sleeve I.D.: Actual penetration identification number P&ID Sequence No.: Penetration identification number used on the P& ids Line: Fluid system line Flow: Direction of flow in or out of containment Size.: Line size GDC or RG: Applicable general design criteria or Regulatory Guide Closed Sys IRC: Closed system inside containment as defined in SSAR Section 6.2.3.1.1 Isolation Device: These fields refer to the isolation devices for a given penetration Qty.: Number of subject devices per penetration Size: Device size Type: Device body type Operator- Operator type (for valves) Location: Device location inside or outside containment Position N-S-A: Device position for N (normal operation)
~
S (shutdown) r A (post-accident) ( Signal: Device closure signal MS: Main steamline isolation I LSL: Low steamline pressure M F: Main feedwater isolation i LTC: Low Tcold PRHR: Passive residual heat removal actuation T: Contaiument isolation TLP: Containment isolation coincident with low header pressure S: Safety Injection Signal I HR: High Containment Radiation Actuation Mode Primary / I Secondasy: Primary closure mode of operation / Secondary closure mode of operation Types: I I manual: manual manipulation at the valve (e.g. handwheel) l I self: self controlled valve (e.g. check or relief valve) I automatic: power operated valve automatically closes on a safety related I signal I remote manual: power operated valve requiring remote operator action (e.g. from l 6e MCR) l N/A: isolation devices without manipulation capability (e.g. flange) Closure Time: Required valve closure stroke time STD: Industry standard for valve type N/A: Not Applicable W W65tiligt100S6 == .manme
14#amm#tt; 1 ns and Isolation Valves lyms fzr Table 6.2.3-1 r,t: These fields refer to the penetration testing requirements pe: Required test type A: Integrated Leak Rate Test APERTURF" B: Local Leak Rate Test - penetration C: Local Leak Rate Test -- fluid systems C/4RD_ ite: See notes below Also Available on sdium: Test fluid on valve seat Aperture Card rection: Pressurization direction Forward: high pressure on containment side Reverse: high pressure on outboard side stes: , Containment leak rate tests are designated Type A, B, or C according to 10CFR50 Appendix J. The secondary side of the steam generator, including main steam, feedwater, startup feedwater, blowdown and sampling piping from the steam generators to the containment penetration, is considered an extension of the containment. These systems are not part of the reactor coolant pressure boundary and do not open i directly to the containment atmosphere during post-accident conditions. During type A tests, the ! secondary side of the steam generators is vented to the atmosphere outside containment to ensure that full test differential pressure is applied to this boundary. I The central chilled water system remains water-filled and operational during the Type A test in order to maintain stable containment atmospheric conditions. The containment isolation valves for this penetration are open during the Type A test to facilitate testing. Their leak rates are measured separately. The inboard butterfly valve is tested in the reverse direction. Upstream side of RNS hot leg suction isolation valves is not vented during LLRT to retain double isolation of RCS at elevated pressure. Valve is flooded during post accident operation. The inboard globe valve is tested in the reverse direction. The test is conservative since the test pressure tends to unseat the valve disc, whereas containment pressure would tend to seat the disc. CRo (d7mcW9 -O Revision: 5 February 29,1996 6.2 193 W- - .- - _. _ _ _ _ _ _ _ _ B
_ . . . .-. ~ .-. - - ~ . - . -. . . . - . - - _ - . . . . . - . -
- 6. Engineered Santy Fec;tures !
,~
/
4 t
\
Table 6.2.41 i COMPONENT DATA - HYDROGEN SENSORS (NOMINAL) Number . . . . . . . . . . . . . . . ...... .. .......... .. . . . . . . . . . . . . . . . . . . 16 (8 per train) Ruge (% hydrogen) .............. .. .-....... . ..... .. ...... . . . .. 0 - 20 i Response time . . . . . . . . . . . . . . . . . ......... ... ..... . .... . 90% in 10 seconds I I . ns i 1 1 t I I e Revision: 5 Y W95tingh00S8 6.2-195 February 29,1996 l J
- 6. Engineered Safety Features a __
O Table 6.2.4 2 COMPONENT DATA - HYDROGEN RECOMBINER (NOMINAL) Number . . . . . . . . ...................... ... ...... ... ........................ 2 l . Active Inlet Area (per unit) (ft2) ....... . . ............... ................... 2 10.7 (1 m ) i Inlet hydrogen concentration . . . . . . .. ............. ...........................0-4 range (volume percent) l Average efficiency (percent) . . . . . .. .......... .. ...... ...... ... .... ....... 85 I l Depletion Rate ......... .............. .................. . . . . . . . . . . . . Reference 20 0 Revision: 5 0 1 February 29,1996 6.2-196 W W8Sfingh00S8 i 1 1 1
- 6. E:gineered Sity Fe.tures Table 6.2.4-3 COMPONENT DATA - HYDROGEN IGNITER (NOMINAL)
Number . . . . . . . . . ..............
...........................................58 l Surface Temperature ('F) . . . . . . . . . . . . . . . . . ........... . . . . . . . . . . . . . . . . . . 1600 to 1700 l Power Consumption (W)
.... ........ ................ ........... .... . . . . 95 to 535 l
C t i a 4
+
j I' 4 g-J b S a i - e i t b Revision: 5 Y W86tlRgh0088 6.2-197 , February 29,1996 f
l 1 u=u l
- 6. Eegineered Safety Features ;
I M---- Table 6.2.4-4 (Sheet 1 of 3) ASSUMPTIONS USED TO CALCULATE HYDROOEN PRODUCTION FOLLOWING A LOSS OF COOLANT ACCIDENT General Core thermal power (MWt) . . . . . . . . . . . . . . . . . . . ..... ........ ... . . . . . . . . . . I ,972 3 Containment free volume (ft ).................. ..................... . . . 1.73 x 106 Zirconium. Water Reaction Weight of zirconium fuel cladding (Ib) . . . . . . . .. .................... .... .. . 34,788 Percent 6 conium-water reaction (%) ................... ....... .... ... .... 1.09 Radiolysis of Water in Reactor Vessel Percentage of core fission product inventory in core l Nob:e gases . . . . . . . . . . . . . . . . . .... ........ .. .. . .. ... ........ .0 I Iodines . . . . . . . . . . . . . .. .......... ... ... .......................50 l kemainder . . . . . . . . ....... .. ...... ........ ............. . . . . . . . 99 Energy absorption by core cooling solution Percent of gamma energy absorbed . ..... ......... ................ . . . . . 10 Percent of beta energy absorbed . . . ......... ... . .. .............. ....O Molecules of hydrogen produced per 100 eV . ....... . .. .... ....... .... . . . 0.5 energy absorbed by solution Radiolysis of Water in Sump Percentr.ge of core fission product inventory in the sump solution Noble gases . . . . . . . . . .. .............. . .... . ... .... ........... 0 I lodines ................. . . ... ................ ... ....... .. . 50 I Remainder . . . . . . . . . . . . . . . . . . . . .......... ... ............. .... .... .I Energy absorption by core cooling solution Percent of gamma energy absorbed . ... .............................. . 100 Percent of beta energy absorbed . . . . . . . . . . . . . . . ............. . . . . . . . . 100 Revision: 5 0 February 29,1996 6.2-198 W W85tillgh00S6
- 6. Engineered Saf;ty Features O
e s V l Table 6.2.4-4 (Sheet 2 of 3) I l J ASSUMPTIONS USED TO : CALCULATE HYDROGEN PRODUCTION I FOLLOWING A LOSS OF COOLANT ACCIDENT Molecules of hydrogen produced per 100 eV . . ........ ............. . ...... . 0.5 I energy absorbed by solution Corrosion of Materials Aluminum inventory in containment Weight Surface Component ! (lb) (sq. ft) ' 1 Excore detectors . . . . . . . . . ........... ... 25 . ........... ... .... . . .8 Flux mapping system ........... ....... . 120 ........ .......... . . . . 84 Miscellaneous valve parts . . . . . . . ........ 230 .. . .. ... ........86 RCDM connectors . . . . . . . . . . .. .... . . 190 . .... ... ... . . . . 42 l Paint . .......... ....... ....... . . 140 .. ........... 18,000 Contingency . . . . . . ............. .. . . 250 . .................... . 85 Other non.NSSS items . ........... ... .. 500.,.. .. . .......... .. 100 e m i Total aluminum . . . . . . . . . . . . . . . . . . . . . . . . . . 1,455 Q,) . l'
.f i %/
Revision: 5 y Westingh00S8 6.2-199 February 29,1996 i i.
, . bmi
- 6. Engineered Safety Features t
l t l 9 l Table 6.2.4-4 (Sheet 3 of 3) 1 1 l ASSUMPTIONS USED TO CALCULATE HYDROGEN PRODUCTION l t FOLLOWING A LOSS OF COOLANT ACCIDENT Zine inventory in containment Weight Surface l Component (Ib) (sq. ft) i Cable trays ..... .... .............. 310 .... ...... ..... . 2,100 Conduit . . . . . . . . . ... . ... ......... 500 . .. ... ... ... .. 3,500 Hangers ... .... . . .... . . . 24 ... . . ........ . 170 Junction boxes . ... . .... ...... ..... 100 .. . .. . ..... . . 730 Paint . . . . . . . ... ....... . .. .. . 1,200 . ... . . . . . . . . 72,000 Gratings . ... .. ... .... 680 .. ....... . . . . . . . 41,000 HVAC ductwork ... .... . . .... .. 840 ... . .. . . . 5,900 Stairs .. .. .. . .. .... . .... ... 13 . .... . . . . .. 800 Pipe supports . . . . . . . .... ........ . 510 . . . .. .. . . . 30,000 i Contingency . . . ..... . .. . .... . . 1.050 . ... . .. .. .. 39,000 Total zir ...... ........ . . .... ... . 5,227 Aluminum corrosion rate ....... . . . .... ... . . . . . . . . See Table 6.2.4-5 Zinc corrosion rate .. ... .. .......... . ....... .... . See Table 6.2.4-5 Containment temperature ... .. . . ........... .......... . . . . . Sec Table 6.2.4-5 Solution pH . . . . . . . . .. ..... .. .. .. ... . . ... .. . .. . ... 7 - 9.5 Initial Reactor Coolant Hydrogen Inventory Hydrogen concentration in reactor coolant (cc at STP per kg) .. .. .. ... . . . . . 40 l Reactor coolant mass (!b) . . . . . ................ ... ... . ...... . 353,000 l [ O Revision: 5 . February 29,1996 6.2-200 W West lrigt:0use
-wmt
- 6. Engineered Saf ty F atures "
O Table 6.2.4-5 POST-ACCIDENT CONTAINMENT TEMPERATURE AND ASSOCIATED CORROSION RATES FOR ALUMINUM AND ZINC Interval Temperature Al Corrosion Zn Corrosion (sec) (*F) (Ib/ft2-br) (Ib/ft2-hr) l l 0 - 25 300 0.066 0.00050 l 25 - 60 270 0.027 0.00031 l l 60 - 150 250 0.014 0.00022 i l 150 - 4000 270 0.027 l 0.00031 l 4000 - 9000 250 0.014 0.00022 l 9000 - 20,000 200 0.0023 0.000090 1 20,000 - 40,000 175 0.00084 0.000054 I >40,000 153 0.00033 0.000033 l
., )
Revision: 5 [ Westingh0use 6.2-201 February 29,1996 l
um:mmm::
- 6. Engineered Safety Features Table 6.2.5-1 (Sheet I of 2) e EXCEPTIONS TO 10 CFR 50 APPENDIX J LEAK TESTING REQUIREMENTS Appendix J Requirement AP600 Exception and Justification Type A tests are to be conducted in accordance with ANSI-56.8, which Paragraph III.A.I.(a)- Type A tests permits testing to proceed provided that the leak (s) can be isolated and that are required to be terminated if subsequent local leak rate testing is performed to demonstrate that the Type excessive leak paths, which would A test criteria are met. His approach can potentially reduce plant outage interfere with satisfactory time and is in accordance with ANSI-56.8 and indunry practice.
completion of the test, are identified. Type A tests are to be conducted in accordance with ANSI-56.8, which Paragraph III.A.3.(a) - Type A tests superseded ANSI N45.4. De NRC is proposing a rule change to delete the shall be conducted in accordance reference to any standard, and a companion Regulatory Guide that would with the provisions of ANSI N45.4- endorse ANSI-56.8. 1972. Type A tests are to be conducted for a minimum of 8 hours, in accordance Paragraph III.A.3.(a) - A Type A wi'h ANSI-56.8. Industry experience has shown that accurate test results test duration of 24 hours is can be achieved in less than 24 hours. The proposed NRC changes to 10 required. CFR 50, Appendix J (Reference 15) would reduce the minimum test duration from 24 to 8 hours. Type A tests are to be conducted at intervals not exceeding four years, Paragraph III.D.I.(a)- nree type A except that, if the test interval ends while containment integrity is not tests are to be performed at required or is required solely for cold shutdown or refueling activities, the approximately equal intervals test interval may be extendee indefinitely provided all deferred testing is during each 10. year service period. successfully completed prior to the time containment integrity is required. His exception is consistent with the proposed NRC changes to 10 CFR 50, Appendix J (Reference 15). Type B tests are to be conducted at intervals not exceeding 30 months, Paragraph III.D. - Type B tests are except that, if the test interval ends while containment integrity is not l required to be performed at required or is required solely for cold shutdown or refueling activities, the ; intervals not greater than 2 years. test interval may be extended indefinitely provided all deferred testing is ' successfully completed prior to the time containment integrity is required. ; This exception is consistent with the proposed NRC changes to 10 CFR 50, 1 Appendix J (Reference 15). ; l 1 O Revision: 5 February 29,1996 6.2-202 [ WB5tiflgfl00S8 1
l 1 j
- 5. Engineered Saf;ty Fe:tures n N
A Table 6.2.5-1 (Sheet 2 of 2) EXCEPTIONS TO 10 CFR 50 APPENDIX J LEAK TESTING REQUIREMENTS { Appendix J Requirement AP600 Exception and Justification Type B testing is to be conducted in accordance with ANSI-56.8, which Paragraph III.D.2.(b)(i) - Air locks permits testing of air locks which are not used during a 6-month period at shall be tested prior to initial fuel a pressure of Pa after the next usage rather than at 6 months. loading and at 6-month intervals I thereafter at a pressure not less than Type C tests are to be conducted at intervals not exceeding 30 months, Paragraph III.D.3 - Type C tests are except that, if the test interval ends while containment integrity is not required to be performed at required or is required solely for cold shutdown or refueling activities, the intervals not greater than 2 years. test interval may be extended indefinitely provided all deferred testing is successfully completed prior to the time containment integrity is required. This exception is consistent with the proposed NRC changes to 10 CFR 50, Appendi- J (Reference 15). l l 3 l (G l 1 i I l i
,7~%
v) Revision: 5 3 WBStingfl00S8 6.2-203 February 29,1996
ivi: s
- 6. Engineered Safety Features i
O Table 6.2.5 2 COMPONENT DATA - CONTAINMENT LEAK RATE TEST SYSTEM (NOMINAL VALUES) Air Compressors 4 Capacity (scfm. total) . . . . . . . ........................................10,500 Discharge pressure (psig) . . . . . . . . . . . . . . ............. .............. . . . . . . 100 Air Cooling and Drying Equipment Air pressure and flow capacity . . . . . . . . . . . . . . . . . . . . . .. . . . . . Consis.ent with compressor Dew point ('F at 100 psig) . ........ .... ............. .... .. . . . . . . . . 40 Air Filters (if twquired) Air pressure and flow capacity . . . ..... .......... ... . . . . . Consistent with compressor Mist and particulate removal efficiency . . . . . .. ............... .... . . . . . . . . . . . . 99 (minimum efficiency, in percent of particles larger than 3 microns) i l 1 l l
. Revision: 5 O.
February 29,1996 6.2-204 W W85tingh00S8
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- 6. Engineered Safity Featrres EM -
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b us S o ' ' ' ' l ' i i i I . , , , l , , , , 0 1 2 3 4 X18 Time (sec) 1 Figure 6.2.1.1 1 MSLB Containment Pressore vs. Time {%
'w 3 We#D8 6.2-205 Feb 2 ,1
- f. Engineered Safety Features e
S k . C -
~
5 k _ Ei . W k - 3 Eh _S 8 i i . , 1 . . i . I . . , , I , , , 0 1 2 3 4 X18 Time (SOC) l l l l i l Figure 6.2.1.1-2 MSLB Containment Temperature vs. Time O Revision: 5 February 29,1996 6.2-206 T Westinghouse
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- 6. Engineered S:f;ty Features j
"~
! (V l AP600 Pressure Comparison S E - 3m o co S
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, gm, MSLB Containment Pressure vs. Time N-] , Revision 5 W Westinghouse 6.2-207 February 29,1996 I-i
- 6. Engineered Safety Features sw O! !
S k - C
~
2 E b - ? g m - e g - i 8 " 8
. .i>>l . , i . I i ..l .,,,l ,,,,
0 1 2 3 4 5 X18 Time (sec)
! Figure 6.2.1.14 MSLB Containment Temperature vs. Time Revision: 5 February 29,1M 6.2-208 3 Westfrighouse
l
- 6. Engineered S:fety Features
,O LI , 1 hPressure(gaugeW805 level 12 S 3 2 - A . R - /~'3 - U : e o ....I . . i , ni! .......! .......I ,4 i i .iis 1 10 100 1000 1H04 1H05 Tene(sec) w open u amenessen l Figure 6.2.1.1-5 Distributed Model DECLG Containment Pressure vs. Time O .( Revision: 5 W Westingh00S8 6.2-209 February 29,1996
- 6. Engineered Safety Features e
ContenmentGasTemperatne CeI8851mel12 R . R1 .. E - R - R
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e I ~.......,: .......1 ........I . . . . . . . . . . ....n 1 10 100 1000 1e+04 1e+06 Tane(sec) wenea maaseen l l Figure 6.2.1.1-6 Distributed Model DECLG Cont 9inment Temperature vs. Time Revision: 5 February 29,1996 6.2-210 W WBStingh0US8
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- 6. Engimered S:fity Featrres FAE --
r ( 8 . o
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Tg. c'8 1 E : es h 81 V) i 1 - o o ......I ,,,,,,,,1 ,,,,,,,,i ,,,,,,,,i , , , , , , , , 1 10 100 1000 1e+04 1e+05 Time (sec) l Figure 6.2.1.1-7 Lumped Model DECLG Containment Pressure vs. Time
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( Revision: 5 l W
'lLiingliOUSe 6.2-211 February 29,1996
- 6. Engineered Safety Features O
8 -
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1 10 100 1000 1e+04 1e+05 Time (sec) l Figure 6.2.1.1-8 Lumped Model DECLG Containment Pressure vs. Time Revision: 5 l February 29,1996 6.2-212 3 Westingh00S8
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- 6. Engineered S fety Features I,,\
V S - o - w - g - b h e 3 h S 1
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(V3 : o o; . i,, i,,,,i....i..,, 0 10 20 30 40 ti0 Time (sec) l Figure 6.2.1.1-9 t,,N DEHLG Containment Pressure vs. Time
\,j Revision: 5 W Westinghouse 6.2-213 February 29,1996 l
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, External Pressure Analysis Containment Pressure vs. Time /
(_ r Revision: 5 W-Westingh0USe 6.2-215 February 29,1996 i
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- 6. Engineered Screty Features O
w s e au w n conme,w r l
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)b-% : th'-y % N1' Figure 6.2.1.2-1 (Sheet 1 of 12)
TMD Model Noding Diagram Revision: 5 February 29,1996 6.2-216 WB5tiligh0USB
- 6. Engineered S:f;ty Features g
~
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1 el 148 y h 6 SG g _ 1 Platform el 135 '-3" g O Platform ei 113'-4" 8 h Platform el 104 - 3" h h 10 g h j Q HL g '- g . ND D l i Figure 6.2.1.2-1 (Sheet 2 of 12) ' TMD Model Noding Diagram G('N Revision: 5 W Westinghouse 6.2-217 February 29,1996
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- 6. Engineered Safety Features p p, < n b'0V
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- 6. Engineered Saf2ty Features i
!O l V l l l 4_________g__ ___
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- 6. Engineered Safety Features i
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a Figure 6.2.1.2-1 (Sheet 5 of 12) TMD Model Noding Diagram Revision: 5 February 29,1996 6.2-220 W85tingh0USS
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- 6. Engineered Safety Features Ox lU
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- 6. Engine: red SIfety Features O
e 4 6
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ppoE k*l 9- ' 4 g @ [u[,, @y Figure 6.2.1.2-1 (Sbeet 7 of 12) TMD Model Noding Diagram Revision: 5 ! February 29,1996 6.2 222 Westinghouse
- 6. Engine: red S:fety Fe tures
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- 6. Engineered Safety Features O
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- 6. Engine: red Saf;ty Features m-O 4
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4
- 6. Engineered Safety Features n p ,i, O 1 1
I i l . A AA A_- H l 53 Av Lower concertment n- n H m n , 4e 47 52 1 50 9 w 8
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TMD Model Noding Diagram Revision: 5 February 29,1996 6.2 226 Westingh0US8 l
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- 6. Engineered Srfity Features O % Mb O i J ut r- -
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Revision: 5
@ Westinghouse 6.2 227 Febniary 29,1996
N3
- 6. Engineered Safety Features nu O
2.50E+06
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5 2 1.50E+06 # f
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5.00E+05 - 0.0 0.0 20000.0 40000.0 60000.0 80000.0 1.00E+05 10000.0 30000.0 50000.0 70000.0 90000.0 TIME (SEC) Figure 6.2.1.3-1 Integrated Mass Released for DECLG Ilreak Revision: 5 O February 29,1W 6.2-228 T Westirighouse
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'6. Engineered Saf;ty Features f)
' w.) 2 2.50E+09 i 1 1 l E 2.00E+09 w e
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i I 1 r l .0.0' -- ! 0.0 20000.0 40000,0 E9000.0 60000.0 1.00E+0L i 10000.0 30000.0 50000 ~ 70000.0 90000.0 T I ME (S'*:- l l l I i i Figure 6.2.1.3-2 Integrated Energy Released for DECLG Break iGI . Revision: 5
! T Westkighouse 6.2 229 Felsruary 29,1996
- 6. Engineered Safety Features re -
3.50E+05 , 3.00E+05 aE m 2.50E+05 2.00E+05
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50000.0 1 L l 0.0 < 0.0 5.0 10.0 15,0 20.0 25.0 l TIME (SEC) , I l I 1 1
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Figure 6.2.1.3-3 I Integrated Mass Released for DEllLG Break Revision: 5 O February 29,1996 6.2-230 3 Westinghouse
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- 6. Engineered Safety Features
.4 u- /,,,i N.
I 2.50E+08 l 3 2.00E+08 --- w 5 E 1.50E+08 1 w 4 7s i u 1.00E+08 / v) I - e
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0.0 0.0 5.0 10.0 15.0 20.0 25.0 TIME (SEC) l l l i l Figure 6.2.1.3-4 Integrated Energy Released for DEIILG Break
/m ~J Revision: 5 T Westinghouse 6.2-231 February 29,1996
- s F C' <
2E I 5i FAR SENSITIVITY STUDY - DRY CONDITIONS u o. g IMPACT ON CONTAINMENT H2 CONCENTRATIONS h G84 YAP 6X ATB O 3 c 5
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E e 4 No Recomt*ier c 0
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$ Time After Accident, days 9 8 1 $'
8 g 9 - - s N E w t i e 9 9
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PAR SENSITIVITY STUDY - CONDITIONS ! H2 CONCENTRATION FOR VARIOUS DEPLETION RATES b i GNYAP6Y ATB M : 2 if i R> 5 cf 5 gg e , _ $ 4 DR x 1/100 e g
% o
$ DR x 1/10 b
> & 3 h f DR x 1/5
- 5. E g g 2 - i ir h g 3C t Case PAR Depleton Ralr ,'Let) !
m j ~ F 8 ' s" ; E ; y 7 o . ... ....,ii i . . . . . . i i i i i { y, o 10 20 30 40 50 $p y.[. Time After Accident, days > $$ $~ $ & _e z a -
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- 6. Engineered Safety Features Ol i
- l l
l 4 3-f . l 2-1 ,. e
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n 0 $ 5 5 E E E E A 8 H rarmwioxsw.mp 1 1 1 i Figure 6.2.4-3 j Hydrogen Production Rate
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Revision: 5 February 29, IN 6.2-234 3 W65tingh00Se
4 J
- 6. Engine: red S:fety Fe tures I i
.O l l 4 i
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IME - l r N w M-
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o - D. ~ ~~ Nf43 0 0 b b b b b h k k k 3 ra rewiguxiw,np Figure 6.2.4-4 Hydrogen Accumulation in Containment - No Recombiner O) L. Revision: 5 [ W85tingh0056 6.2-235 February 29,1996
- 6. Engineered Safety features e
S S '- E s b 9 l . . . . . . . . . . . . g Time (sec) W_QOTHC 14 06h746162 9 r-Figure 6.2.5-1 O Revision: 5 February 29,1996 6.2-236 1,Y WBStingh00S8
i
- 15. Accid:nt Analyses [
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1 CHAPTER 15.0 ! ACCIDENT ANALYSES i 15.0.1 i Classification of Plant Conditions j The ANSI 18-2 (Reference 1) classification divides plant conditions into four categories ) according to anticipated frequency of occurrence and potential radiological consequences to the public. The four categories are as follows:
- Condition I: Normal operation and operational transients
. Condition II: Faults of moderate frequency j
- Condition III: Infrequent faults '
Condition IV: Limiting faults The basic principle applied ic relating design requirements to each of the conditions is that l the most probable occurrences should yield the least radiological risk and those extreme ' situations having the potential for the greatest risk should be these least likely to occur. i Where applicable, reactor trip and engineered safeguards functioning am assumed to the extent I allowed by considerations such as the single failure esiterion in fulfilling this principle. The ; evaluation models and parameters for the accident analysis radiological consequences are { l discussed in WCAP-14601 (Reference 11). t V 15.6.1.1 Condition 1: Normal Operation and Operational Transients Il Condition I occurrences are those that are expected to occur frequently or regularly in the O course of power operation, refueling, maintenance, or maneuvering of the plant. As such, l I Condition I occurrences are accommodated with margin between a plant parameter and the i value of that parameter that would require either automatic or manual protective action. ! Since Condition I events occur frequently, they must be considered from the point of view . of their effect on the consequences of fault conditions (Conditions II, III, and IV). In this regard, analysis of each fault condition described is generally based on a conservative set of
)
l initial conditions corresponding to adverse conditions that can occur during Condition I ; operation. A typical list of Condition I events follows. Steady-State and Shutdown Operations ! See Table 1.1-1 of Chapter 16. ' w/ Revision: 5 [ WeStingh00S8 15.0-1 February 29,1996 i
rs e=-. 15. Accident Analyses Operation with Permissible Deviations O Various deviations that occur during continued operation as permitted by the plant technical specifications are considered in conjunction with other operational modes. These deviations include the following: Operation with components or systems out cf service [such as an inoperable rod cluster control assembly (RCCA)]
- Leakage from fuel with limited clad defects
- Excessive radioactivity in the reactor coolant:
Fission products Corrosion products Tritium Operation with steam generator tube leaks
. Testing Opentional Transients a Flant heatup and cooldown Step load changes (up to 10 percent)
Ramp load changes (up to 5 percent / min)
=
Load rejection up to and including design full load rejection transient: 15.0.1.2 Condition II: Faults of Modemie Frequency These faults, at worst, result in a reactor trip with the plant being capable of returning to operation. By definition, these faults (or events) do not propagate to cause a more serious fault (Condition III or IV events). In addition, Condition II events are not expected to result in fuel rod failres, reactor coolant system failures, or secondary system overpressurization. The following faults are included in this category: l t l = Feedwater system malfunctions that result in a decrease in feedwater temperature (See ! Subsection 15.1.1) Feedwater system malfunctions that result in an increase in feedwater flow (See Subsection 15.1.2) Excessive increase in secondary steam flow (See Subsection 15.1.3) Inadvertent opening of a steam generator relief or safety valve (See Subsection 15.1.4) l O 1
'Revislom 5 l February 29,1996 15.0-2 T Westinghause l
l l
1 l
- 15. Accident Analyses l
l O) i l i
- Inadvenent operation of the passive residual heat removal heat exchanger (See Subsection 15.1.6) l Loss of external electrical load (See Subsection 15.2.2) !
1 Turbine trip (See Subsection 15.2.3) I
- l Inadvertent closure of main steam isolation valves (See Subsection 15.2.4)
Loss of condenser vacuum and other events resulting in turbine trip (See Subsection 15.2.5) Loss of ac power to the station auxiliaries (Sce Subsection 15.2.6) Loss of normal feedwater flow (See Subsection 15.2.7) Panial loss of forced reactor coolant flow (See Subsection 15.3.1) Uncontrolled rod cluster control assembly bank withdrawal from a subcritical or low power startup condition (See Subsection 15.4.1) Uncontrolled rod cluster control assembly bank withdrawal at power (See [] Subsection 15.4.2)
\)
Rod cluster control assembly misalignment (dropped full-length assembly, dropped full-length assembly bank, or statically misaligned assembly) (See Subsection 15.4.3) Startup of an inactive reactor coolant pump at an incorrect temperature (See Subsection 15.4.4) Chemical and volume control system malfunction that results in a decrease in the boron concentration in the reactor coolant (See Subsection 15.4.6) l
- Inadvertent operation of the passive core cooling system during power operation (See Subsection 15.5.1)
Chemical and volume control system malfunction that increased reactor coolant inventory i (See Subsection 15.5.2) l
- Inadvertent opening of a pressurizer safety valve (See Subsection 15.6.1)
Ereak in instrument line or other lines from the reactor coolant pressure boundary that I penetrate containment (See Subsection 15.6.2) (m)
\_/
Revision: 5 W.-. Westingh00Se 15.0-3 February 29,1996
m == r {-{
^
- 15. Accident Analyses 15.0.1.3 Condition III: Infnquent Faults O
Condition III events are faults which may occur infrequently during the life of the plant. They may result in the failare of only a small fraction of the fuel rods. The release of cadioactivity is not sufficient to interrupt or restrict public use of those areas beyond the exclusion area boundary,in accordance with the guidelines of 10 CFR 100. By definition, Condition III event alone does not generate a Condition IV event or result in a consequential loss of function of the reactor coolant system or containment barriers. The following faults are included in this category: Steam system piping failure (minor) (See Subsection 15.1.5) Complete loss of forced reactor coolant flow (See Subsection 15.3.2) Rod cluster control assembly misalignment (single rod cluster control assembly withdrawal at full power) (See Subsection 15.4.3) Inadvertent loading and operation of a fuel assembly in an improper position (See Subsection 15.4.7) Inadvertent operation of automatic depmssurization system (See Subsection 15.6.1) Loss-of-coolant accidents (LOCAs) resulting from a spectrum of postulated piping breaks within the reactor coolant pressure boundary (small break) (See Subsection 15.6.5) Gas waste management system leak or failure (See Subsection 15.7.1) Liquid waste management system leak or failure (Subsection 15.7.2) Release of radioactivity to the environment due to a liquid tank failure (See Subsection 15.7.3)
- Spent fuel cask drop accidents (See Subsection 15.7.5) 15.0.1.4 Condition IV: Limiting Faults Condition IV events are faults that are not expected to take place but are postulated because their consequences include the potential of the release of significant amounts of radioactive material. They are the faults that must be designed against and they represent limiting design cases. Condition IV faults are not to cause a fission product release to the environment resulting in doses in excess of the guideline values of 10 CFR 100. A single Condition IV event is not to cause a consequential loss of required functions of systems needed to cope with the fault, including those r,f the emergency core cooling system and the containment. The following faults are dassified in this category:
- Steam system piping fature (major) (See Subsection 15.1.5)
Revision: 3 February 29,1996 15.0-4 [ W85tingh0USC
- 15. Accident Analyses -
p) (~ Feedwater system pipe break (See Subsection 15.2.8) Reactor coolant pump shaft seizure (locked rotor) (See Subsection 15.3.3) Reactor calant pump shaft break (See Subsection 15.3.4) e Spechum of rod cluster control assembly ejection accidents (See Subsection 15.4.8) Steam generator tube rupture (See Subsection 15.6.3) Loss-of-coolant accidents resulting from a spectrum of postulated piping breaks within the reactor coolant pressure boundary (large break) (See Subsection 15.6.5) e Design basis fuel handling accidents (See Subsection 15.7.4) i 15.0.2 Optimization of Contml Systems A control system setpoint study is performed prior to plant operation to simulate performance of the primary plant control systems and overall plant performance. In this study, emphasis is placed on the development of the overall plant control systems that automatically maintains conditions in the plant within the allowed operating window and with optimum control system response and stability over the entire range of anticipated plant operating conditions. The p/ L control system setpoints are developed using the nominal protection system setpoints which are implemented in the plant. Where appropriate (such as in margin to reactor trip analyses), instrumentation errors are considered and are applied in an adverse direction with respect to maintaining system stability and transient performance. The accident analysis and control system setpoint study in combination show that the plant can be operated and meet both safety and operability requirements throughout the core life and for various levels of power operation. The centrol system setpoint study is comprised of analyses of the following control systems: plant control, axial offset control, rapid power reduction, steam dump (turbine bypass), steam generator level, pressurizer pressure, and pressurizer level. I 15.0.3 Plant Charactedstics and Initial Conditions Assumed in the Accident Analyses 15.0.3.1 Design Plant Conditions 1 Table 15.0-1 lists the principal power rating values assumed in the analyses performed. The thermal power output includes the effective thermal power generated by the reactor coolant pumps. The values of other pertinent plant parameters utilized in the accident analyses are given in i Table 15.0-3. A i i N) Revision: 5 Westingh0use 15.0-5 February 29,1996
l
- 15. AccidEt Analyses 15.0.3.2 Initial Conditions O
For most accidents that are departure from nucleate boiling (DNB) limited, nominal values of initial conditions are assumed. The allowances on power, temperature, and pressure are determined on a statistical basis and are included in the departure from nucleate boiling ratio (DNBR) safety analysis limit values (See subsection 4.4.1.1.2), as described in Reference 2. This procedure is known as the Revised Thermal Design Procedure (RTDP), and is discussed more fully in Section 4.4. For accidents that are not departure from nucleate boiling limited, or for which the revised thermal design procedure is not employed, the initial conditions are obtained by adding the maximum steady-state errors to rated values. The following conservative steady-state errors ; are assumed in the analysis: Core power 12 percent allowance for calorimetric error J l Average reactor coolant 6.5 'F allowance for controller deadband and ! system (RCS) temperature measurement errors l Pressurizer pressure 50 psi allowance for steady-state fluctuations and measurement errors Initial values for core power, average reactor coolant system temperature, and pressurizer pressure are selected to minimize the initial departure from nucleate boiling ratio unless otherwise stated in the sections describing the specific accidents. Table 15.0-2 summarizes the iniGal conditions and computer codes used in the accident analyses. l 15.0.3.3 Power Distribution The transient response of the reactor system is dependent on the 'nitial power distribution. The nucl;ar design of the reactor core minimizes adverse power distribution through the placement of fuel assemblies and control rods. Power distribution may be characterized by the nuclear enthalpy rise hot channel factor (Fs) and the total peaking factor (Fq ). Unless specifically noted otherwise, the peaking factors used in the accident analyses are those presented in Chapter 4. For transients that may be departure from nucleate boiling (DNB) limited, the radial peaking factor is important. 'Ihe radial peaking factor increases with decreasing power level due to - rod insertion. This increase in Fg is included in the core limits illustrated in I Figure 15.0.3-1. Transients that may be departure from nucleate boiling limited are assumed to begin with an Fg consistent with the initial power level defined in the technical specifications. The axial power shape used in the departure from nucleate boiling calculation is the 1.55 chopped cosine, as discussed in subsection 4.4.4.3, for transients analyzed at full power and O Revision: 5 February 29,1996 15.0-6 T Westinghouse
1 l .
- 15. Accident Analyses f l
l [ i t ) U the most limiting power shape calculated or allowed for accidents initiated at nonfull power or asymmetric rod cluster control assembly (RCCA) conditions. The radial and axial power distributions just described are input to the THINC code as described in Subsection 4.4.4.5. For transients which may be overpower limited, the total peaking factor (Fq) is important. Transients that may be overpower limited are assumed to begin with plant conditions including power distributions, which are consistent with reactor operation as defined in the technical specifications. For overpower transients that are slow with respect to the fuel rod thermal time constant (for example, the chemical and volume control system malfunction that results in a slow decrease in the boron concentration in the reactor coolant system as well as an excessive increase in secondary steam flow) and that may reach equilibrium without causing a reactor trip, the fuel rod thermal evaluations are performed as discussed in Subsection 4.4.4. For overpower transients that are fast with respect to the fuel rod thermal time constant (for example, the uncontrolled rod cluster control assembly bank withdrawal from suberitical or lower power startup and rod cluster control assembly ejection incident, both of which result in a large power rise over a few seconds), a detailed fuel transient heat transfer calculation is performed.
,a l 15.0.4 Reactivity Coefncients Assumed in the Accident Analysis The transient response of the reactor system is dependent on reactivity feedback effects, in ;
particular the moderator temperature coefficient and the Doppler power coefficient. These ' reactivity coefficients are discussed in Subsection 4.3.2.3. In the analysis of certain events, conservatism requires the use of large reactivity coefficient I values. He values used are given in Figure 15.0.4-1, which shows the upper and lower bound Doppler power coefficients as a function of power, used in the transient analysis. The justification for use of conservatively large versus small reactivity coefficient values is treated on an event-by-event basis. In some cases conservative combinations of parameters are used to bound the effects of core life, although these combina' ions may not represent possible realistic situations. I 15.0.5 Rod Cluster Contml Assembly Insertion Chametedstics The negative reactivity insertion following a reactor trip is a function of the acceleration of the rod cluster control assemblies (RCCAs) as a function of time and the variation in rod worth as a function of rod position. For accident analyses, the critical parameter is the time ofinsenion up to the dashpot entry, or approximately 85 percent of the rod cluster travel. In analyses where all of the reactor coolant pumps are coasting down prior to or simultaneous with RCCA insertion, a time of 1.8 seconds is used for insertion time to dashpot entry.
,a
_/ Revision: 5 [ W65dngt100S8 15.0-7 Fehmary 29,1996
?!= =E 1
- 15. Accident Analyses j l In Figure 15.0.5-1, the curve labeled " complete loss of flow transients" shows the RCCA O
position versus time normalized to 1.8 seconds assumed in accident analyses where all reactor coolant pumps are coasting down. In analyses where some or all of the reactor coolant pumps are running, the RCC A insertion time to dashpot is conservatively taken as 2.4 seconds. The I RCCA position versus time normalized to 2.4 seconds is also shown in Figure 15.0.5-1. The use of such a long insertion time provides conservative results for accidents and is intended to apply to all types of rod cluster control assemblies which may be used throughout plant life. Drop time testing requirements are specified in the technical specifications. I Figure 15.0.5-2 shows the fraction of total negative reactivity insertion versus normalized rod position for a core where the axial distribution is skewed to the lower region of the core. An axial distribution which is skewed to the lower region of the core can arise icom an unbalanced xenon distribution. This curve is used to compute the negative reactivity insertion versus time following a reactor trip, which is input to the point kinetics core models used in transient analyses. The bottom skewed power distribution itselfis not an input into the point kinetics core model. l There is inherent conservatism in the use of Figure 15.0.5-2 in that it is based on a skewed flux distribution, which would exist relatively infrequently. For cases other than those associated with unbalanced xenon distributions, significantly more negative reactivity is inserted than that shown in the curve, due to the more favorable axial distribution existing prior to trip. The normalized rod cluster control assembly negative reactivity insertion versus time is shown i in Figure 15.0.5-3. The curves shown in this figure were obtained from Figures 15.0.5-1 and i 15.0.5-2. A total negative reactivity insertion following a trip of four percent Ak is assumed in the transient analyses except where specifically noted otherwise. This assumption is conser-vative with respect to the calculated trip reactivity worth available as shown in Table 4.3-3. The normalized rod cluster control assembly negative reactivity insertion versus time curve I for an axial power distribution skewed to the bottom (Figure 15.0.5-3) is used in those transient analyses for which a point kinetics core model is used. Where special analyses require use of three-dimensional or axial one-dimensional core models, the negative reactivity insertion resulting from the reactor trip is calculated directly by the reactor kinetics code and is not separable from the other reactivity feedback effects. In this case, the rod cluster control I assembly position versus time of Figure 15.0.5-1 is used as code input. I 15.0.6 Trip Points and Time Delays to Tdp Assumed in Accident Analyses A reactor trip signal acts to open two trip breaker sets connected in series, feeding power to the control rod drive mechanisms. The loss of power to the mechanism coils causes the mechanisms to release the rod cluster control assemblies, which then fall by gravity into the l core. There are various instrumentation delays associated with each trip function including l delay in signal actuation, in opening the trip breakers, and in the release of the rods by the l mechanisms. The total delay to trip is defined as the time delay from the time that trip l Revisiom 5 i February 29,19% 15.0-8 Y Westingh0Use l
? E
- 10. Accident Analyses m
A V conditions are reached to the time the rods are free and begin to fall. Limiting trip setpoints assumed in accident analyses and the time delay assumed for each trip function are given in I Table 15.0 4. Referenca is made in that table to overtemperature and overpower AT trip i shown in Figure 15.0.3-1. The difference between the limiting trip point assumed for the analysis and the nominal trip point represents an allowance for instrumentation channel error and setpoint error. Nominal I ' trip setpoints are specified in the plant technical specifications. During plant startup tests,it is demonstrated that actual instrument time delays are equal to or less than the assumed vrbs. Additionally, protection system channels are calibrated and instrument response times are determined periodically in accordance with the plant technical specifications. I 15.0.7 Instrumentatkn Ddft and Calodmetric Eners, Power Range Neutmn Flux The instrumentation uncertainties and calorimetric uncertaintier used in establishing the power i range high neutron flux setpoint are presented in Table 15.0-5. The calorimetric uncertainty is the uncertainty assumed in the determination of core thermal N power as obtained from secondary plant measurements. The total ion chamber current (sum x of the top and bottom sections) is calibrated (set equal) to this measured power on a daily basis. I
/T The secondary power is obtained from measurement of feedwater flow, feedwater inlet
()' temperature to the steam generators, and steam pressure. Installed plant instramentation is used for these measurements, l 15.0.8 Plant Systems and Components Available for Mitigation of Accident Effects The plant is designed to afford proper protection against the possible effects of natural phenomena, postulated environmental conditions, and dynamic effects of the postulated accidents. In addition, the design incorporates features that minimize the probability and effects of fires and explosions. Chapter 17 discusses the quality assurance program that is implemented to provide confidence that the plant systems satisfactorily perform their assigned safety functions. The incorporation ! of these features in the plant, coupled with the reliability of the design, provide confidence ; I that the normally operating systems and components listed in Table 15.0-6 are available for ' mitigation of the events discussed in Chapter 15. l In determining which systems are necessary to mitigate the effects of these postulated events, the classification system of ANSI h18.2-1973 (Reference 1)is utilized. The design of safety related systems (including protection systems)is consistent with IEEE Standard 379-1988 and Regulatory Guide 1.53 in the application of the single-failure criterion. Conformance to Regulatory Guide 1.53 is summarized in Section 1.9.1. 1 (3,1 Revision: 5 Westinghouse 15.0 9 February 29,19%
Y
- 15. Accid:nt Analyses In the analysis of the Chapter 15 events, control system action is considered only if that action e.
results in more severe accident nesults. No credit is taken for control system operation if that operation mitigates the results of an accident. For some accidents, the analysis is performed both with and without control system operation to determine the worst case. I 15.0.9 Fission Product Inventories The sources of radioactivity for release are dependent on the specific accidert. Activity may be released from the primary coolant, from the secondary coolant, and from the reactor core if the accident involves fuel damage. The radiological consequences analyses utilize the I conservative design basis source terms identified in WCAP-14601 (Reference 11). I 15.0.10 Residual Decay Heat 15.0.10.1 Total Residual IIcat Residual heat in a suberitical core is calculated for the loss-of-coolant accident (LOCA) according to the requifements of 10 CFR 50.46, as described in References 3 and 4. The small break LOCA events utilize 10 CFR 50 Appendix K which assumes infinite irradiation time before the core goes suberitical to determine fission product decay energy. For all other accidents, the same models are used, except that fission product decay energy is based on core average exposure at the end of an equilibrium cycle. 15.0.10.2 Distribution of Decay Heat Following Loss-of-Coolant Accident During a LOCA, the core is rapidly shutdown by void formation, rod cluster control assembly insenion, or both; and, a large fraction of the heat generation considered comes from fission product decay gamma rap. This heat is not distributed in the same manner as steady-state fission power. Local peakmg effects, which are important for the neutron-dependent part of I the heat generation, do not apply to the gamma ray contribution. The steady-state factor, I which represents the fraction ofleTt generated within the clad and pellet, drops to 95 percent or less for the hot rod in a LOCA. For example, consider the transient resulting from the postulated double-ended break of the largest reactor coolant system pipe; one half second after the rupture about 30 percent of the heat generated in the fuel rods is from gamma ray absorption. The gamma power shape is less peaked than the steady-sta.e fission power shape, reducing tne energy deposited in the hot rod at the expense of adjacent colder rods. A conservative estimate of this effect on the hot rod is a reduction of 10 percent of the gamma ray contribution or three percent of the total heat. Since the water density is considerably reduced at this time, an average of 98 percent of the available heat is deposited in the fuel rods; the remaining two percent is absorbed by water, thimbles, sleeves, and grids. Combining the three percent total heat reduction from gamma redistribution with this two percent absorption produce as the net effect a factor of 0.95, which exceeds the actual heat production in the hot rod. The actual hot rod heat generation is computed during the AP600 large break LOCA transient as a function of core fluid conditions. Revision: 5 February 29,1996 15.0-10 3 Westingh0USS
- 15. Accident Analyses
, f3 ( ) I 15.0.11 Computer Codes Utilized Summaries of some of the principal computer codes used in transient analyses are given below. Other codes--in particular, very specialized codes in which the modeling has been I developed to simulate one given accident, such as those used in the analysis of the reactor coolant system pipe rupture (See Section 15.0.6) are summarized in their respective accident analyses sections. The codes used in the analyses of each transient are listed in Table 15.0- .. 15.0.11.1 FACTRAN Computer Code FACTRAN (Reference 5) calculates the transient temperature distribution in a cross section of a metal clad UO2 uel f rod and the transient heat flux at the surface of the clad using as input the nuclear power and the time-dependent coolant parameters (pressure, flow, temper-ature, and density). The code uses a fuel model which simultaneously exhibits the following features: A sufficiently large number of radial space increments to handle fast transients such as rod ejection accidents. Material properties which are functions of temperature and a sophisticated fuel-to-clad gap heat transfer calculation. [] U
=
The necessary calculations to handle post-departure from nucleate boiling transients: film boiling heat transfer correlations, zircaloy-water reaction, and partial melting of the materials. FACTRAN is further discussed in Reference 5. 15.0.11.2 LOFTRAN Computer Code The LOL RAN (Reference 6) program is used for studies of transient response of a pressurited water reactor system to specified perturbations in process parameters. LOFTRAN simulates a multiloop system by a model containing reactor vessel, hot and cold leg piping, steam generator (tube and shell sides), and pressurizer. The pressurizer heaters, spray, and safety valves are also considered in the program. Point model neutron kinetics, and reactivity effects of the moderator, fuel, boron, ano rods are included. The secondary side of the steam generator utilizes a homogeneous, saturated mixture for the thermal transients and a water I level correlation for indication and control. The protection and safety monitoring system is simulated to include reactor trips on high neutron flux, overtemperature AT, high and low pressure, low flow, and high pressurizer level. Control systems are also simulated including rod control, steam dump, feedwater control, and pressurizer level and pressure control. The emergency core cooling system, including the accumulators, is alsa modeled. LOFTRAN is a versatile program which is suited to both accident evaluation and control , studies as well as parameter sizing. !n , V) Revisiom 5 W W65tiligh00S8 15.0-11 February 29,1996
- 15. Accid;nt A alyses LOFTRAN also has the capability of calculating the transient value of departure from nucleate e
boiling ratio (DNBR) based on the input from the core limits illustrated in Figure 15.0.3-1. The cere limits represent the minimum value of departure from the nucleate boiling ratio as ! calculated for typical or thimble cell. LOFTRAN is further discussed in Reference 6. The LOFTRAN code is modified to allow the simulation of the passive residual heat removal I (PRHR) heat exchanger, core makeup tanks (CMT) and associated protection and safety I monitorir.g system actuation logic. A discussion of these models and additional validation is I presented in WCAP-14601 (Reference 11) and Reference 10. LOFITR2 (Reference 8) is a modified version of LOFTRAN with a more realistic break flow model, a two-region SG secondary side, and an improved capability to simulate operator actions during a steam generator tube rupture (SGTR) event. LOFITR2 is further discussed in Reference 8. The LOFITR2 code is modified to allow the simulation of the passive residual heat removal I heat exchanger, core makeup tanks (CMT) and associated protection system actuation logic. Tne modifications are identical to those made to the LOFTRAN code. A discussion of these I models is presented in WCAP-14601 (Reference 11) and Reference 10. 15.0.11.3 TWINKLE Computer Code The TWINKLE (Reference 7) program is a multidimensional spatial neutron kinetics code, which is patterned after steady-state codes presently used for reactor core design. The code uses an implicit finite-difference method to solve the two-group transient neutron diffusion equations in one, two, and three dimensions. The code uses six delayed neutron groups and contains a detailed multiregion fuel-clad-coolant heat transfer model for calculating pointwise Doppler and moderator feedback effects. The code handles up to 2000 spatial points and performs its own steady-state initialization. Aside from basic cross-section data and thermal-hydraulic parameters, the code accepts as input basic driving functions such as inlet temperature, pressure, flow, boron concentration, control rod motion, and others. Various edits are provided (for example, channelwise power, axial offset, enthalpy, volumetric surge, point-wise power, and fuel temperatures). The TWINKLE code is used to predict the kinetic behavior of a reactor for transients that I cause a major perturbation in the spatial neutron flux distribution. TWINKLE is further described in Reference 7. 15.0.11.4 THINC Computer Code l The THINC code is described in subsection 4.4.4.5. I O l Revision: 5 February 29,1996 15.0-12 T Westinghouse
l
- 15. Accident Analyses iA) l l 15.0.11.5 WESTAR Computer Code I
The WESTAR code is described in subsection 4.4. I 15.0.12 Component Failures 15.0.12.1 Active Failures SECY-77-439 (Reference 9) provides a description of active failures. An active failure results in the inability of a component to perform its intended function. An active failure is defined differently for different components. For valves, an active failure is the failure of a component to mechanically complete the movement required to perform its function. This includes the failure of a remotely-operated valve to change position on demand. T!c spurious, unintended movement of the valve is also considered as an active failure. Failure of a manual valve to change position under local operator action is included. Sprir.g-loaded safety or relief valves that are designed for and operate under single-phase fluid conditions are not considered for active failures to close when pressure is reduced below the valve set point. However, when valves designed for single-phase flow are challenged with two-phase flow, such as a steam generator or pressurizer safety valve, the failure to rescat is considered as an active failure.
!'O V For other active equipment such as pumps, fans, and rotating mechanical components, an active failure is the failure of the component to start or to remain operating.
For electrical equipment, the loss of power, such as the loss of offsite power or the loss of a diesel-generator, is considered as a single failure. In addition, the failure to generate an actuation signal, either for a single component actuation or for a system-level actuation, is also considered as an active failure. Spurious actuation of an active component is considered as an active failure for active components in safety-related passive systems. An exception is made for active components if specific design features or operating restrictions are provided that can preclude such failures (such as power lockout, confirmatory open signals, or continuous position alarms). A single incorrect or omitted operator action in response to an initiating event is also considered as an active failure. The error is limited to manipulation of safety-related equipment and does not include thought-process errors or similar errors that could potentially lead to common cause or multiple errors. 1 15.0.12.2 Passive Failures SECY-77-439 also provides a description of passive failures. A passive failure is the structural failure of a static component which limits the component's effectiveness in carrying f3 out its design fuaction. A passive failure is applied to fluid systems and consists of a breach V) Revision: 5 W WBStiligt100S8 15.0-13 February 29,1996
- 15. Accident Analyses in the fluid system boundary. Examples include cracking of pipes, spmng flanges, or valve e
packing leaks. Passive failures are not assumed to occur until 24 hours after the start of the event. Consequential effects of a pipe leak such as flooding, jet impingement, and failure of a valve with a packing leak must be considered. Where piping is significantly overdesigned or installed in a system where the pressure and temperature conditions are relatively low, passive leakage is not considered a credible failure mechanism. Line blockage is also not considered as a passive failure mechanism. 15.0.12.3 Limiting Single Failmes The most limiting single active failure (where one exists), as described in Section 3.1, of safety-related equipment, is identified in each analysis description. The consequences of this failure are described therein. In some instances, because of redundancy in protection equip-ment, no single failure which could adversely affect the consequences of the transient is I identified. The failure assumed in each analysis is listed in Table 15.0-7. I 15.0.13 Opemtor Actions l For events where the passive residual heat removal heat exchanger is actuated, the plant automatically cools down to the safe shutdown condition. Where a stabilized condition is I reached automatically following a reactor trip, it is expected that the operator may, following I event recognition, take manual control and proceed with orderly shutdown of the reactor in accordance with the normal, abnormal or emergency operating procedures. The exact actions taken and the time at which these actions occur depend on what systems are available and the plans for further plant operation. However, for these events, operator actions are not required to maintain the plant in a safe and stable condition. Operator actions typical of normal operation are credited for the inadvenent actuations of equipment in response to a Condition II event. I 15.0.14 Combined License Infonnation l This section has no requirement for additional information to be provided in suppon of the l Combined License application. 15.0.15 References
- 1. American National Standards Institute N18.2, " Nuclear Safety Criteria for the Design of Stationary PWR Plants," 1972.
- 2. Friedland, A. J. and Ray, S., " Revised Thermal Design Procedure," WCAP-11397-P-A (Proprietary) and WCAP-11398-A (Nonproprietary), April 1989.
e Revision: 5 Febnuuy 29,1996 15.0-14 3 Westingh00S8
- 15. ' Accide:t Analyses l
- /
- 3. Lee, N., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P A (Proprietary) and WCAP-10081 (Nonproprietary),
August 1985.
- 4. Bajorek, S. M., " Code Qualification Document for Best Estimate LOCA Analysis,"
WCAP-12945-P (Proprietary), August 1991.
- 5. Hargrove, H. G., "FACTRAN - A FORTRAN-IV Code for Thermal Transients in a UO 2
Fuel Rod," WCAP-7908 (Proprietary) and WCAP-7337 (Nonproprietary), June 1972.
- 6. Burnett, T. W. T., "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary) and WCAP-7907-A (Nonproprietary), April 1984.
- 7. Risher, D. H., Jr. and Barry, R. F., " TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Nonproprietary),
January 1975.
- 8. Lewis, R. N., "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," WCAP-10698-P-A (Proprietary) and WCAP-10750-A (Nonproprietary), August 1985.
- 9. Case, E. G., " Single Failure Criterion," SECY-77-439, August 17,1977.
( l
- 10. Carlin, E. L., "LOFTRAN and LOFTTR2 AP600 Code Applicability Document,"
l WCAP-14234, November 1994. I
- 11. Carlin, E. L., Kemper, R. M., Gresham, J. A., "AP600 Accident Analyses - Evaluation I Models," WCAP-14601 (Proprietary), February 1996.
t i in i
%)
Revision: 5 Y Westinghouse 15.0-15 February 29,1996
i i ____ x
- 15. Accidnt Analyses Table 15.0-1 e
NUCLEAR STEAM SUPPLY SYSTEM POWER RATINGS Thermal power output (MWt) 1940 Effective thermal power generated by the reactor coolant pumps (MWt) 7 Core thermal power (MWt) 1933 Revision: 5 February 29,1996 15.0-16 T Westinghouse
g '4 \ V V V Table 15.0-2 (Sheet I of 4) p
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED g 5 k N Reactivity Coefficients Assumed E go
=
InitialThermal ,$- g Computer Moderator Moderator Power Output Sect. $_ Codes Density Tempenture Assumed tion Faults Utilized (Ak/gm/cm3) (pem/*F) Doppler (MWt) 15.1 Increase in heat removal from the primary system Feedwater system mal- LOFTRAN 0.374 - Upper curve of 0 and 1940 functions that result in an Figure 15.0.4-1 increase in feedwater flow G w Excessive increase in LOFFRAN 0.0 and -- Upper & lower 1940 secondary steam flow 0.374 curves of Figure 15.0.4-1 l Inadvertent opening of a LOFTRAN Function of - See subsection 0 (subcritical) steam generator relief or moderator 15.1.4 safety valve density (see Figure 15.1.4-1) l Steam system piping LOFTRAN, Function of -- See subsection 0 (suberitical) failure THINC moderator 15.1.5 density (see y Figure 15.1.4-1) O' 2 l Inadvertent operation of LOFTRAN See subsection -- Upper curve 1940 i the PRHR 15.1.6.2.1 of Figure $y *g 15.0.4-1 .*: .Uf
-8 3- ll a tn
. n
-ime
yy Table 15.0-2 (Sheet 2 of 4) fh e 5-
SUMMARY
OF IMITIAL CONDITIONS AND COMPUTER CODES USED II' QR y* Reactivity Coefficients Assumed, ea m Initial Thermal Computer Moderator Moderater Power Output Sect. Codes Density Temperature Assumed tion Faults Utilized (Ak/gm/cm3) ( W F) Doppler (MWt) 15.2 Decrease in heat removal by the secondary system Loss of external electrical LOFIRAN 0.0 - Lower and 1940 load and/or turbine trip and upper cues of 0374 Figure 15.0.4-1 0 p Loss of nonemergency ac LOFIRAN 0.0 - Lower curve of 1978.8a g power to the station Figure 15.0.4-1 auxiliaries Loss of normal feedwater LOFIRAN 0.0 -- Lower curve of 1978.8a flow Figure 15.0.4-1 Feedwater system pipe LOFTRAN 0.374 - Lower curve of 1978.8a break Figure 15.0.4-1 15 3 Decrease in reactor coolant system flow rate
- Partial and complete loss LOFIRAN, 0.0 --
Lower curve of 1940 of forced reactor coolant FACTRAN, Figure 15.0.4-1
,E l flow WESTAR &
E. go Reactor coolant pump LOFIRAN, FACIRAN 0.0 -- Lower curve of 1978.8a E { shaft seizure (locked Figure 15.0.4-1 F; rotor) g ; a y a e G G
(]. %/ . Table 15.0-2 (Sheet 3 of 4) g
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED s IF 7 Reactivity Coefficients Assumed g liii {L Initial g Thermal [ y g Computer Moderator Moderator Power Output 2 Sect. Codes Density Temperature Assumed tion Faults Utilized (Ak/gm/cm3) (pcm/ F) Doppler (MWt) 15.4 Reactivity and power distribution anomalies Uncontrolled RCCA bank TWINKLE, - 0.0 Coefficient is 0 withdrawal from a sub- FACFRAN, consistent with a critical or low power THINC Doppler defect of startup condition -0.67%Ak !a Uncontrolled RCCA bank LOFTRAN 0.0 - Upper & lower 10%,60% & [ c withdrawal at power and curves of 100% of 1940 0.374 Figure 15.0.4-1 RCCA misalignment See NA - NA 1940 Section 4.3 Startup of an inactive LOFTRAN, 0.374 - Upper curve of 1358 reactor coolant pump at FACTRAN Figure 15.0.4-1 an incorrect temperature THINC Chemical and volume con- NA NA - NA 0 and 1940 trol system malfunction that results in a decrease m in the boron concentration 9 in the reactor coolant . 2
, Inadvertent loading and See NA --
NA 1940 g operation of a fuel Section 4.3 y g- assembly in an improper i:
-g position 4 ** 7 i
c"25 g Table 15.0-2 (Sheet 4 of o cr < l 2[ P.
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED i w ta Reactivity Coefficients Assumed
?
- Initial Thermal f Computer Moderator Moderator Power Output Sect- Codes Density Temperature Assumed tion Faults Utilized (Ak/gm/cm3) (pcm/*F) Doppler (MWt) 15.4 Spectrum of RCCA ejec- TWINKLE, Refer to sub- Refer to Coefficient 0 and 1978.8a tion accidents FACTRAN section 15.4.8 subsection consistent with 15.4.8 a Doppler defect of-0.67% AK at BOC and -0.63%
AK at EOC 15.5 Increase in reactor coolant inventory l Inadvertent operation of LOFFRAN 0 - Upper & lower 1940 l the emergency core cooling and [ o curves of Figure Q system during power operation 0.374 15.0.4 1 15.6 Decrease in reactor coolant inventory Inadvertent opening of a LOFTRAN 0.0 - Lower curve of 1940 pressurizer safety valve Figure 15.0.4-1 and inadvertent operation of ADS ' Steam generator tube LOFITR2 0.0 - Lower curve of 1978.8a failure Figure 15.0.41 LOCAs resulting from NOTRUMP See subsection - See subsection 1971.7 the spectrum of post- ,WLCOBRA/ 15.6.5 15.6.5 ? ulated piping breaks TRAC references g within the reactor references > g coolant pressure boundary 3. ho a 102% of rated thermal power. BOC - Beginning of Core Life g F h EOC - End of Core Life
}
e 9 9
- 15. Accident Analyses ~
~
- \
V Table 15.0-3 NOMINAL VALUES OF PERTINENT PLANT PARAMETERS UTILIZED IN ACCIDENT ANALYSES RTDP With 10% Without RTDP (a) Steam Genemtor Tube Plugging Without Steam With 10% Steam Generator Tube Genemtor Tube Plugging Plugging Thermal Output of NSSS 1940.0 1940.0 1940.0 (MWt) l Core Inlet Temperature (*F) 533.40 531.9 532.8 I Vessel Average Temperature 567.6 565.9 567.6 ( F) Reactor Coolant System 2250.0 2250.0 2250.0 Pressure (psia) l Reactor Coolant Flow Per 9.66 E+04 9.71 E+04 9.48 E+04 Loop (GPM) (3 'j
,. i Steam Flow From NSSS 8.43 E+06 8.43 E+06 8.43 E+06 (Ibm /hr) i Steam Pressure at Steam 794.0 801.0 794.0 Generator Outlet (psia) l Maximum Steam Moisture 0.10 0.10 0.10 Content (%)
Assumed Feedwater 435.0 435.0 435.0 Temperature at Steam Generator Inlet (*F) Average Core Heat Flux 1.43 E+05 1.43 E+05 1,43 E+05 (BTU /-hr-ft2) I a. Steady-state errors discussed in subsection 15.0.3 are added to these values to obtain initial conditions for transient analyses. (3 N._.] Revisiom 5 { [ WeStingh0Use 15.0-21 Febmary 29,1996
- 15. Accident Analyses Table 15.0-4 O'
TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES Limiting Trip Point Time Delays Trip Function Assumed in Analysis (s) Power range high neutron flux, high 118 % 0.5 setting Tower range high neutron flux, low 35 % 0.5 setting High neutron flux, P-8 84 % 0.5 Source range neutron flux NA 0.5 Overtemperature AT Variable (see Figure 15.0.3-1) 2.0 Overpower AT Variable (see Figure 15.0.3-1) 2.0 liigh pressurizer pressure 2460 psia 2.0 l Low pressurizer pressure 1800 psia (lower bound) 2.0 1 2030 psia (upper bound) 0.0 Low reactor coolant flow (from 87% loop flow 1.45 loop flow detectors) Reactor coolant pump under speed 90% nominal 0.767 Low steam generator level 0.0% of narrow range level 2.0 span High-2 steam generator level reactor 100% of narrow range level 2.0 trip span i O Revision: 5 February 29,1996 15.0-22 W Westingh0USB
l
- 15. Accident Analyses
(~) \ (! Table 15.0-5 (Sheet I of 2) DETERMINATION OF MAXIMUM POWER RANGE NEUTRON FLUX CHANNEL TRIP SETPOINT, BASED ON NOMINAL SETPOINT AND INHERENT INSTRUMENTATION UNCERTAINTIES Nominal setpoint (% of rated power) 109 Calodmetdc eners in the measurement of secondary system thennal power: Effect on Accuracy of Thennal Power Measurement Determination Vadable of Vadable (% of Rated Power) Feedwater temperature 13.*F Steam pressure (small correction 6 psi l Q) , (, . on enthalpy) Feedwater flow 0.5% Delta-P instrument span (two channels per steam generator) Assumed calorimetric error 2.0 (a)* Radial power distribution effects on 7.8 (b)* total ion chamber current Allowed mismatch between power range 2.0(c)* neutron flux channel and calorimetric measurement l l
!O 'V Revision: 5
{ Westinghouse 15.0-23 February 29,1996
5* E l E 15. Accident Analyses I Table 15.0-5 (Sheet 2 of 2) O DETERMINATION OF MAXIMUM POWER RANGE NEUTRON FLUX CHANNEL TRIP SETPOINT, BASED ON NOMINAL SETPOINT AND INHERENT INSTRUMENTATION UNCERTAINTIES I Calodmetdc eners in the measmement of secondary system thermal powen Effect on Accuracy of 1hennal Power Measurement Detennination Vadable of Variable (% of ruled nower) Instrumentation channel drift and 0.4% of instrument 0.84(d)*
; setpoint sproducibility span (120% power span)
Insuumentation channel 0.48(e)* temperature effects
- Total assumed error in setpoint 8.4 O
(% of rated power): [(a)2 + (b)2 + (c)2 + (d)2 + (e)2]I/2 Maximum power range neutron flux trip serpoint 118 assuming a statistical combination of individual uncertainties (% of rated power) O Revision: 5 February 29,1996 15.0-24 3 W65tingh00Se
- 15. Accident Analyses fN
\) Table 15.0-6 (Sheet I of 5) PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT ! AND ACCIDENT CONDITIONS 1 Reactor ESF Trip Actuation ESF & Incident Functions Functions Other Equipment l Section 15.1 1 Increase in heat removal from the primary system Feedwater system mal- Power range high flux, High-2 steam gener. Feedwater isolation functions that result in overtemperature AT, ator level produced valves an increase in feedwater overpower AT, nwual feedwater isolation flow and turbine trip Excessive increase in Power range high flux, -- -- l secondary steam flow overtemperature AT, overpower AT, manual .O Inadvertent opening of a Low pressurizer pres- Low pressurizer pres- CMT, feedwater steam generator safety sure, manual "S" sure, low Teold, isolation valves, valve Low-2 pressurizer steam line stop level valves Steam system piping "S", low pressurizer Low pressurizer pres- CMT, feedwater failure pressure, manual sure, low isolation valves, compensated steam main steam line line pressure, High-1 isolation valves containment pressure, (MSIVs), accu-low Tcold, manual mulators Inadvertent operation of Overpower AT, power Low pressurizer pres- CMT the PRHR range high neutron flux, sure, low Tcold. Iow pressurizer pressure, Low-2 pressurizer "S", manual level o Revision: 5 W W85tiflghouse 15.0-25 February 29,1996
m
=5 l
[ 5 15. Accident Analyses Table 15 A6 (Sheet 2 of 5) O' PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT , AND ACCIDENT CONDITIONS Reactor ESF Tdp Actuation ESF & J Incident Functions Functions Other Equipment l Section 15.2 Decrease in heat removal by the sec-ondary system Loss of external High pressurizer -- Pressurizer safety load / turbine trip pressure overtemp- valves, steam erature AT, overpower generator safety AT, manual valves Loss of non-emergency Steam generator low- Steam generator low PRHR, steam gener-ac power to the station narrow range level, high narrow range level ator safety valves, auxiliaries pressurizer pressure, co-incident with low pressurizer safety high pressurizer level, startup water flow, valves manual steam generator low wide range level Loss of normal feed- Steam generator low- Steam generator low PRHR, steam gener-water flow narrow range level, high narrow range level ator safety valves, pressurizer pressure, co-incident with low pressurizer safety high pressurizer level, startup water flow, valves manual steam generator low wide range level Feedwater system pipe Steam generator low Steam generator low PRHR, CMT, break narrow range level, high wide range level, low MSIVs, feedline pressurizer pressure, steam line pressure, isolation, pressurizer manual High-1 containment safety valves, steam pressure generator safety valves l 9 Revision: 5 - February 29,1996 15.0-26 [ Westiligh00S8 '
- 15. Accident Analyses
\ p t < Table 15.0-6 (Sheet 3 of 5) PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS Reactor ESF Trip Actuation ESF & Incident Functions Functions Other Equipment l Section 15.3 Decrease in reactor cooiant system flow rate . Partial and complete Low flow, anderspeed. -- Steam generator l loss of forced reactor manual safety valves, coolant flow pressurizer safety l valves Reactor coolant pump Low flow, manual, high -- Pressurizer safety (RCP) shaft seizure pressurizer pressure valves, steam (locked rotor) generator safety l
, valves l I i 'w) l Section 15.4 Reactivity and power distribution anomalies Uncontrolled RCCA Power range high flux - --
l bank withdrawal from a (low setpoint), source l subcritical or low power range high flux, startup condition intermediate range high flux, manual 1 Uncontrolled RCCA Power range high flux, -- Pressurizer safety I bank withdrawal at overtemperature AT, valves, steam power high pressurizer pres- generator safety sure, manual valves RCCA misalignment Overtemperature AT, -- -- manual Startup of an in-active Power range high flux, -- - reactor coolant pump at low flow (P-8 interlock), an incorrect temperature manual 1 (~s N.-] Revision: 5 { Westingh00Se 15.0-27 Febmany 29,1996
=,= ==
2
- 15. Accident Analyses Table 15.0-6 (Sheet 4 of 5)
O i l PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS Reactor ESF Tnp Actuation ESF & l Incident Functions Functions Other Equipmer.t l Section 15.4 l
- (continued)
CVS malfunction that Source range high flux, Source range flux Low insertion limit results in a decrease in overtemperature AT, doubling annunciators boron concentration in manual the reactor coolant Spectrum of RCCA Power range high flux, -- Pressurizer safety ejection accidents high positive flux rate, valves manual l Section 15.5 Increase in reactor coolant inventory Inadvertent operation of High pressurizer High pressurizer CMT, Pressurizer the ECCS during power pressure, manual, level, low Teold safety valves, CVS operation " safeguards" trip, high isolation, PRHR pressurizer level l Section 15.6 Decrease in reactor coolant inventory Inadvenent opening of a Low pressurizer Le v pressurizer pres- CMT, ADS, pressurizer safety valve pressure, overtemp- sure accumulator I or ADS path erature AT, manual O Revision: S February 29,1996 15.0-28 3 Westingh0US8 i
l
- 15. Accident Analyses i
l tm l Table 15.0-6 (Sheet 5 of 5) 1 PLANTSYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT l AND ACCIDENT CONDITIONS l 1 Reactor ESF l Tdp Actuation ESF & Incident Functions Functions Other Equipment l Section 15.6 (continued) Steam generator tube Low pressurizer Low pressurizer pres- CMT, PRHR, steam rupture pressure, overtempera- sure, high steam generator safety I ture AT, safeguards generator level, low and/or relief valves, , I ("S"), manual steam line pressure MSIVs, radiation l monitors (air removal, steamline and SG blowdown), startup feedwater isolation, CVS pump I isolation, pressurizer (3 I heater isolation, () I l steam generator PORV isolation LOCAs resulting from Low pressurizer High-1 containment CMT, accumulator, the spectrum of pressure, safeguards pressure, low ADS, Steam genera-postulat-d piping breaks ("S"), manual pressurizer pressure tor safety and/or I within the reactor relief valves, PRHR coolant pressure boundary l l 7m L) Revision: 5 Y W8Stiligh00S8 15.0-29 Febmary 29,1996 1 1
IS. Accident Analyses O1 Table 15,0-7 (Sheet 1 of 2) 1 SINGLE FAILURES ASSUMED IN ACCIDENT ANALYSES Event Description Failure ! Feedwater temperature reduction (a) _ Excessive feedwater flow One protection division Excessive steam flow One protection division inadvertent secondary depressurization One CMT discharge valve Steam system piping failure One CMT discharge valve i Inadvertent operation of the PRHR One protection division Steam pressure regulator malfunction (b) _ Loss of externalload One protection division Turbine trip One protection division Inadvertent closure of main steam isolation One protection division valve Loss of condenser vacuum One protection division Loss of ac power One PRHR discharge valve Loss of normal feedwate- One PRHR discharge valve Feedwater system pipe break One PRHR discharge valve ! Partial loss of forced reactor coolant flow One protection division I Complete loss of forced reactor coolant flow One protection division l Reactor coolant pump locked rotor One protection division Reactor coolant pump shaft break One protection division Rod cluster control assembly (RCCA) bank One protection division withdrawal from suberitical RCCA bank withdrawal at power One protection division Dropped RCCA, dropped RCCA bank One protection division Statically misaligned RCCA(c) _ Single RCCA withdrawal One protection division O Revision: 5 February 29,1996 15.0-30 W Westinghouse
- 15. Accident Analyses 7- ,
Y Table 15.0-7 (Sheet 2 of 2) SINGLE FAILURES ASSUMED IN ACCIDENT ANALYSES 1 Event Description Failure Inactive reactor coolant pump startup ' One protection division Flow controller malfunction (b) . Uncontrolled boron dilution One protection division Improper fuel loading (c) . RCCA ejection One protection division Inadvenent emergency core cooling system One protection division operation at power I Increase in reactor coolant system inventory One protection division Inadvertent reactor coolant system One protection division depressurization Failure of small lines carrying primary -- coolant outside containment (c) (r-)s Steam generator tube a2pture Faulted steam generator power operated relief valve fails opn Spectrum of loss-of-coolant accident Small breaks One 4th stage ADS valve Large breaks One CMT discharge valve i Double-ended CMT piping breaks One protection division (eliminates one 1st stage I and one 3rd stage ADS valve) (a) No protection action required. (b) Not applicable to AP600. (c) No transient analysis. () Revislom 5 T Westinghouse 15.0-31 February 29,1996
- 15. Accident Analyses e
Delta-T (Deg F) 80 p- kI m 1 \ . O v e r p owe r/
\ 'N \ I
\ \
/
Delta T Trip /
\ \ L \ Quality >
i
\ I I \'
\ \ \ Limit Lin 70r-Nominal i \
Operating g g g g g Conditions \ \ \
\ \ \ h
\ \ \ \ \
50 -
?A60 t
\
\
f\ \ ; \
\
psta \ \ \ \
\ \ \ \
2250 \ \ \ 50 - psia \ l\\ 2000 \ psia \ \ g g g 40 " 1875 k g /g Exit Boiling i i Limit Lines psia \ \ l
\ f 1 ----- O v e r t e m p e r a t u r e 1700 \ \ Delta T Trips 30 -
psia $/.I
) Locus of Points Where
/I SG Safety Valves Open I I I I I I 20 I I I I I I 530 550 570 590 610 630 650 Tavg (Deg F)
I I Figure 15.0.3-1 Illustradon of Overpower and Overtemperature AT Pmtection Revision: 5 February 29,1996 15.0-32 Westirighouse
- 15. Accidert Analyses ! j V !
l Gamma * ** D' <
-1 l Least Negative Doppler Only Power Defect = .595 % Delta K
-5 Co to 100 % Power)
-7 l
_g Most Negative Doppier On1y
'~
Power Defect = -1.18 % Deita'K) _qq C0 to 100 % Power)
-13 1
I
-15 O 20 40 60 80 100
% Power l
Figure 15.0.4-1 , ,,- Doppler Power Coefficient used in Accident Analysis ' i, / Revision: 5 [ Westinghouse 15.0-33 February 29,1996 1 l l
- 15. Accid:nt A:alyses e
Normalized RCCA Position (Distance Dropped / Distance to Top of Dashpot) 1.0 0.9 - 0.8 - 0.7 - AII or Some 0.6 - Reactor Coolant Pumps Running 0.5 - (Normalized to 0.4 - 2.4 Sec) g 0.3 Compiet_e Loss 0.2 - ' of Flow Transients 0.1 - (Norma l i zed to 1.8 Sec) I I I I I I 0 I O 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Normalized Drop Time (Time After Drop Begi ns/ Drop Time to Top of Dashpot) Figure 15.0.5-1 RCCA Position Versus Time to Dashpot Revision: 5 February 29,1996 15.0-34 Westinghouse
. .- - _ _. - ~ ...
l
- 15. Accidert Analyses '
!"$1 '
tm N) Normalized Reactivity Worth i i 1.O O.9 - 1 O.8 I O7 - l 0.6 - O.5 - r~.
't _) O.4 -
O.3 - 0.2 - O.1 - O ' O O.2 0.4 0.6 0.8 1O Rod Pos i t i on (Fract ion Inserted) Figure 15.0.5-2 f"\ Normalized Rod Worth Versus Position
' 'n Revision: 5 W gighym 15.0-35 February 29,1996 e
l.
b
- 15. Accid:nt Analyses e
l l Normalized RCCA Reactivity Worth 1.1 - 1.0 -- O.9 - Complete Loss O8 - Transients 0.7 - O6 - 05 - O.4 - Ali or some O Reactor Coolant O,3 - Pumps nunning 0.2 -
- 0. '1 -
I ' O I I I O O.5 1.0 1.5 2.0 25 30 3.5 Time After Drop Begins l Figure 15.0.5-3 I Nonnalized RCCA Bank Reactivity Worth vs. Drop Time Revision: 5 February 29,1996 15.0-36 l T Westinghouse 1
- 15. Accidcat Analyses o
N~Y 15.1 Increase in Heat Removal From the Primary System A number of eveas are postulated which could result in an increase in heat removal from the reactor coolant system (RCS). Detailed analyses are presented for the events that have been identified as limiting cases. Discussions of the following reactor coohnt system cooldown events are presented in this section: Feedwater system malfunctions causing a reduction in feedwater temperature Feedwater system malfunctions causing an increase in feedwater flow Excessive increase in secondary steam flow Inadvertent opening of a steam generator relief or safety valve
=
Steam system piping failure i = uadvenent operation of the passive residual heat removal heat exchanger The preceding events are Condition II events, with the exception of small steam system piping failures, which are considered to be Condition III and large steam system piping failure Condition IV events. Subsection 15.1 contains a discussion of classifications and applicable criteria.
^ The accidents in this section are analyzed. The most severe radiological consequences result / 'g from the main steam line break accident discussed in Subsection 15.1.5. Therefore, the radio-V logical consequences are reported only for that limiting case.
15.1.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature 15.1.1.1 Identification of Causes and Accident Reductions in feedwater temperature causes an increase in core power by decreasing reactor coolant temperature. Such transients are attenuated by the thermal capacity of the secondary i plant and of the reactor coolant system. The overpower /overtemperature protection (neutron overpower, overtemperature, and overpower AT trips) prevents any power increase that could lead to a departure from nucleate boiling ratio (DNBR) less than the safety analysis limit valves. A reduction in feedwater temperature may be caused by a low-pressure heater train or a high-pressure heater out of service. At power, this increased subcooling creates a greater load demand on the reactor coolant system. With the plant at no-load condit ions, the addition of cold feedwater may cause a decrease in reactor coolant system temperature and a reactivity insertion due to the effects of the negative moderator coefficient of reactivity. However, the rate of energy change is reduced as load and feedwater flows decrease, so the no-load transient is less severe than the full-power case, b
,m Revision: 5 T Westinghouse 15.1-1 February 29,1996
m .
- 15. Accident Analyses The nei effect on the reactor coolant system due to a reduction in feedwater temperature is e
similar to the effect of increasing secondary steam flow: That is, the reactor reached a new equilibrium condition at a power level corresponding to the new steam generator AT. A decrease in normal feedwater temperatum is classified as a Condition II event, tault of moderate frequency. The protection available to mitigate the consequences of a decrease in feedwater temperature is the same as that for an excessive steam flow increase, as discussed in Subsection 15.8 and listed in Table 15-6. 15.1.1.2 Analysis of Effects and Consequences l 15.1.1.2.1 Method of Analysis This transient is analyzed by computing conditions at the feedwater pump inlet following the removal of a low-pressure feedwater heater train or a high-pressure heater from service. These feedwater conditions are then used to recalculate a heat balance through the high-pressure heaters. This heat balance gives the new feedwater cos.d:.tions at the steam generator inlet. The following assumptions are made: Plant initial power level corresponding to 100 percent nuclear steam supply system thennal output ! = Isolation of one string of low-pressure feedwater heaters l-l Plant characteristics and initial conditions are further discussed in Subsection 15.3. 15.1.1.2.2 Results Isolation of a string of low-pressure feedwater heaters causes a reduction in feedwater temperature that increases the thermal load on the primary system. The calculated reduction l l in feedwater temperature is 45.4*F, resulting in an increase in heat load on the primary system j of less than 10 percent full power. Themfore, the transient results of this analysis are not i presented. 15.1.1.3 Cor.clusions The decrease in feedwater temperature transient is less severe than the increase in feedwater flow event and the increase in secondary steam flow event. (See Subsections 15.1.2 and 15.1.3.) Based on the results presented in Subsections 15.1.2 and 15.1.3, the applicable SRP Section 15.1.1 evaluation criteria for the decrease in feedwater temperature event are met. t e i Revision: 5 l February 29,1996 15.1-2 [ Westilighouse u- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - _ - - ---
- 15. Acddent Analyses
[t U i APbOO ! p
, t ,
15.1.2 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow t 15.12 1 Identification of Causes and Accident Description Addition of excessive feedwater causes an increase in core power by decreasing reactor coolant temperature. Such transients are attenuated by the thermal capacity of the seconda1y plant and the reactor coolant system. The overpower /overtemperature protection (neutron overpower, overtemperature, and overpower AT trips) prevents a power increase that leads to a departure from nucleate boiling ratio (DNBR) less than the safety analysis limit value. An example of excessive feedwater flow is a full opening of a feedwater control valve due to a feedwater control system malfunction or an operator error. At power, this excess flow causes a greater load demand on the reactor coolant system doe to increased subcooling in the steam generator. With the plant at no-load conditions, the addition of cold feedwater may cause a decrease in reactor coolant system temperature and a reactivity insertion due to the effects of the necative moderator coefficient of reactivity. Continuous addition of excessive feedwater is prevented by the steam generator High-2 water level signal trip, which closes the feedwater isolation valves and feedwater control valves and I trips the turbine, main feedwater pumps, and trips the reactor. (O (_/ An increase in normal feedwater flow is classified as a Condition II event, fault of moderate frequency. Plant systems and equipment available to mitigate the effects of the accident are discussed in Subsection 15.8 and listed in Table 15-6. 15.1.2.2 Analysis of Effects and Consequences 15.1.2.2.1 Method of Analysis The excessive heat removal due to a feedwater system malfunction transient is analyzed by using the detailed digital computer code LOFfRAN (Reference 1). This code simulates a multiloop system, neutron kinetics, pressurizer, pressurizer safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level. The system is analyzed to demonstrate plant behavior if excessive feedwater addition occurs because of system malfunction or operator error which allows a feedwater control valve to open fully. The following two cases are analyzed assuming a conservatively large negative j moderator temperature coefficient: Accidental opening of one feedwater control valve with the reactor just critical at zero-n load conditions i \ Revision: 5 [ W85tingh00S6 15.1-3 February 29,1996
l n -- n . f 15. Accidnt Analyses l \
+
Accidental opening of one feedwater control valve with the reactor in automatic control O at full power The reactivity insertion rate following a feedwater system malfunction is calculated with the following assumptions: For the feedwater control valve accident at full power, one feedwater control valve is assumed to malfunction resulting in a step increase to 115 percent of nominal feedwater flow to one steam generator For the feedwater control valve accident at zero-load condition, a feedwater control valve malfunction exurs, which results in a step increase in flow to one steam generator from 0 to 115 percent of the nominal full-load value for one steam generr. tor For the zero-load condition, feedwater temperature is at a conservatively low value of 40 F No credit is taken for the heat capacity of the reactor coolant system and steam generator thick metal in attenuating the resulting plant cooldown a The feedwater flow resulting from a fully open control valve is terminated by a steam generator High-2 level trip signal, which closes feedwater control and isolation valves, trips the main feedwater pumps. and trips the turbine Plant characteristics and initial conditions are further discussed in Subsection 15.3. Normal reactor control systems are not required to function. He reactor protection system may function to trip the reactor because of overpower or High-2 steam generator water level I conditions. No single active failure prevents operation of the protection and safety cnonitoring system. A discussion of anticipated transients without trip considerations is presented in Section 15.8. 15.1.2.2.2 Results In the case of an accidental full opening of one feedwater control valve with the reactor at zero power and the preceding assumptions, the maximum reactivity insertion rate is less than the maximum reactivity insertion rate analyzed in Subsection 15.4.1 for an uncontrolled rod cluster control assembly bank withdrawal from a suberitical or low-power startup condition. Therefore, the results of the analysis are not presented here. If the incident occurs with the unit just critical at no-load, the reactor may be tripped by the power range high neutron flux trip (low setting) set at approximately 25 percent nominal full power. The full-power case (maximum reactivity feedback coefficients, automatic rod control) results in the greatest power increase. Assuming the rod control system to be in the manual control mode results in a slightly less severe transient. O Revision: 5 February 29,1996 15.1-4 [ W65tingh0USB
- 15. Accide:t Analyses
,/3 e a When the steam generator water level in the faulted loop reaches the High-2 level setpoint, the feedwater control valves and feedwater pump discharge valves are automatically closed and the main feedwater pumps are tripped. This prevents continuous addition of the feedwater. In addition, a turbine trip and a reactor trip are initiated.
Transient results show the increase in nuclear power and AT associated with the increased thermal load on the reactor. (See Figures 15.1.2-1 and 15.1.2-2.) The depanure from nucleate boiling ratio does not drop below the limit value. Following the reactor trip, the plant approaches a stabilized and safe coadition; standard plant shutdo ve dures may then be followed to further cool down the plant. Since the power level rises to a maximum of about 4 percent during the excessive feedwater flow incident, the fuel temperatures also rises until after reactor trip occurs. The core heat flux lags behind the neutron flux response because of the fuel rod thermal time constant. Therefore, the peak value does not exceed 118 percent ofits nominal value (the assumed high neutron flux trip setpoint). The peak fuel temperature thus remains well below the fu:1 melting temperature. The transient results show that departure from nucleate boiling does not occur at any time during the excessive feedwater flow incident. Thus, the ability of the primary coolant to remove heat from the fuel rods are not reduced. The fuel cladding temperature therefore does not rise significantly above its initial value during the transient. p V The calculated sequence of events for this accident is shown in Table 15.1.2-1. 15.1.2.3 Conclusions The results of the analysis show that the departure from nucleate boiling ratios encountered for an excessive feedwater addition at power are above the limit value. The departure from nucleate boiling ratio design basis is described in Section 4.4. Additionally, the reactivity insertion rate that occurs at no-load conditions following excessive feedwater addition is less than the maximum value considered in the analysis of the rod withdrawal from suberitical condition analysis. See Subsection 15.4.1. 15.1.3 Excessive Increase in Secondary Steam flow 15.1.3.1 Identification of Causes and Accident Description An excessive increase in secondary system steam flow (excessive load increase incident) is a rapid increase in steam flow that causes a power mismatch between the reactor core power I and the steam generator load demand. The plant control system is designed to accommodate t a 10 percent step load increase or a 5 percent per minute ramp load increase in the range of 25 to 100 percent full power. Ar.y loading rate in excess of these values may cause a reactor i trip actuated by the protection and safety monitoring system. Steam flow increases greater
,3 than 10 percent are analyzed in Subsections 15.1.4 and 15.1.5.
b Revision: 5 [ W85tingh00S8 15.1-5 February 29,1996
- 15. Accidert Analyses
[ This accident could result from either an administrative violation such as excessive loading O by the operator or an equipment malfunction in the steam dump control or turbine speed control. During power operation, turbine bypass to the condenser is controlled by reactor coolant condition signals. That is, a high reactor coolant temperature indicates a need for turbine bypass. A single controller malfunction does not cause turbine bypass. Rather, ar interlock blocks the opening of the valves unless a large turbine load decrease or a turbine trip has oc-curred. Protection against an excessive load increase accident is provided by the following reactor protection system signals:
- Overpower AT
- Overtemperature AT
- Power range high neutron flux An excessive load increase incident is considered to be a Condition II event, as described in Subsection 15.1.
15.1.3.2 Analysis of Effects and Consequences 15.1.3.2.1 Method of Analysis This accident is analyzed using the LOFTRAN code (Reference 1). The code simulates the neutron kinetics, reactor coolant system, pressurizer, pressurizer safety valves, pressurizer spray, steam generator, steam generator safety valves and feedwater system. The code computes pertinent plant variables including tt uperatures, pressures and power level. Four cases are analyzed to demonstrate plant behavior following a 10 percent step load increase from rated load. These cases are as follows: Reactor control in manual with minimum moderator reactivity feedback Reactor control in manual with maximum moderator reactivity feedback Reactor control in automatic with minimum moderator reactivity feedback Reactor control in automatic with maximum moderator reactivity feedback For the minimum moderator feedback cases, the core has the least negative moderator temperature coefficient of reactivity; therefore, reductions in coolant temperature have the least impact on core power. For the maximum moderator feedback cases, the moderator temperature coefficient of reactivity has its highest absolute value. This results in the largest amount of reactivity feedback due to changes in coolant temperature. For the cases with automatic rod control, no credit is taken for AT trips on ove temperature or overpower in order to demonstrate the inherent transient capability of the plant. Under actual operating conditions, such a trip may occur, after which the plant quickly stabilizes. O Revision: 5 February 29,1996 15.1-6 [ W85tiflgh0US8 1
- 15. Accident Analyses o
A 10 percent step increase in steam demand is assumed, and each case is studied without credit being taken for pressurizer heaters. Initial reactor power, RCS pressure and temperature are assumed to be at their full power values. Uncertainties in initial conditions are included in the limit DNBR as described in Reference 2. Plant characteristics and initial conditions are fur *her discussed in Subsection 15.3. Normal reactor control systems and engineered safety systems are not required to function. l The protection and safety monitoring system is assumed to be operable; however, reactor trip is not encountered for most cases due to the error allowances assumed in the serpoints. No I single active failure prevents the protection and safety monitoring system from performing its intended function. 15.1.3.2.2 Results Figures 15.1.3-1 through 15.1.3-10 show the transient with the reactor in the manual control mode. For the minimum moderator feedback case there is a slight power increase, and the average core temperature shows a large decrease. This results in a departure from nucleate boiling ratio which increases above its initial value. For the maximum moderator feedback, manually controlled case there is a much faster increase in reactor power due to the moderator feedback. A reduction in the departure from nucleate boiling ratio is experienced, but the departure from nucleate boiling ratio remains above the safety analysis limit. O L' Figures 15.1.3-11 through 15.1.3-20 show the transient assuming the reactor is in the automatic control mode and no reactor trip signals occur. Both the minimum and maximum moderator feedback cases show that core power increases, thereby reducing the rate of decrease in coolant average temperature and pressurizer pressure. For both of these cases, the minimum departure from nucleate boiling ratio remains above the safety limit. I For the cases analyzed the plant power stabilizes at an increased power level. Nonnal plant operating procedures are followed to reduce power. Because of the measurement enors assumed in the setpoints, it is possible that reactor trip could actually occur for the automatic control cases. The plant then reaches a stabilized condition following the trip. The excessive load increase incident is an overpower transient for which the fuel temperature rises. Reactor trip may not occur for some of the cases analyzed, and the plant reaches a new equilibrium condition at a higher power level corresponding to the increase in steam flow. Since departure from nucleate boiling does not occur at any time during the excessive load increase transients, the capability of the primary coolant to remove heat from the fuel rod is not reduced. Thus, the fuel cladding temperature does not rise significantly above its initial value during the transient. The calculated sequence of events for the excessive load increase incident is shown in Table 15.1.2-1.
]
- (V Revision
- 5 3 W85tingh0USe 15.1-7 February 29,1996
m.t:: 1
- 15. Accid;nt Analyses 15.1.3.3 Conclusions e
The analysis just presented above shows that for a 10 percent step load increase the departure from nucleate boiling ratio remains above the safety analysis limit. The design basis for departure from nucleate boiling ratio is described in Section 4.4. The plant rapidly reaches a stabilized condition following the load increase. 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1.4.1 Identification of Causes and Accident Description The most severe core conditions resulting from an accidental depressurization of the main steam system are associated with an inadvertent opening of a single steam dump, relief, or safety valve. The analyses performed assuming a rupture of a main steam line are given in Subsection 15.1.5. The steam release, as a consequence of this accident, results in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the reactor coolant system causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. The analysis is performed to demonstrate that the following SRP Section 15.1.4 evaluation criterion is satisfied: h Assuming the most reactive stuck rod cluster control assembly (RCCA), with offsite power available, and assuming a single failure in the engineered safety features (ESF) system, there will be no consequential damage to the fuel or reactor coolant system after reactor trip for a steam release equivalent to the spurious opening, with failure to close, or the largest of any single steam dump, relief, or safety valve. This criterion is met by showing the departure from nucleate boiling (DNB) design basis is not exceeded. Accidental depressunzation of the secondary system is classified as a Condition II event as described in Subsection 15.1. The following systems provide the necessary protection against an accidental depressurization of the main steam system. Core makeup tank actuation on a safeguard signal from one of the following four signals: Two out of four low pressurizer pressure signals Two out of four low pressurizer level signals Two out of four low Tcold signals in any one loop Two out of four low steam line pressure signals in any one loop e Revision: 5 February 29,1996 15.1-8 W Westinghouse
- 15. Accident Analyses S j l 1
rh The overpower reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the safeguards signal
=
Redundant isolation of the main feedwater lines I Sustained high feedwater flow causes additional cooldown. Therefore, in addition to the I normal control action that closes the main feedwater valves following reactor trip, a safe-1 guards signal rapidly closes the feedwater control valves and feedwater isolation valves, I and trips the main feedwater pumps. 1 Redundant isolation of the startup feedwater system l Sustained high startup feedwater flow causes additional cooldown. Therefom, the low Tcold signal closes the startup feedwater control and isolation valves. Trip of the fast-acting main steam line isolation valves (designed to close in less than 10 seconds) on one of the following signals: Two out of four low steam line pressure signals in any one loop (above permissive P-11) Two out of four high negative steam pressure rates in any loop (below permissive ( P-il) Plant systems and equipment which are available to mitigate the effects of the accident I are also discussed in Subsection 15.0.8 and listed in Table 15.0.6. 15.1.4.2 Analysis of Effects and Consequences 15.1.4.2.1 Method of Analysis The following analyses of a secondary system steam release are performed: A full plant digital computer simulation using the LOFTRAN code (Reference 1) to determine reactor coolant system temperature and pressure during cooldown, and the effect of core makeup tank injection Analyses to determine that there is no damage to the fuel or reactor coolant system The following conditions are assumed to exist at the time of a secondary steam system release: End-of-life shutdown margin at no-load, equilibrium xenon conditions, and with the most reactive rod cluster control assembly stuck in its fully withdrawn position. Operation of rod cluster control assembly mechanical shim and axial offset banks during core bumup is restricted by the insertion limits so that shutdown margin requirements are satisfied .in r Revision: 5 T Westinghouse 15.1-9 February 29,1996 i i
- 15. Accid:nt Analyses The most negative moderator coefficient corresponding to the end-of-life rodded core e
with the most reactive rod cluster control assembly in the fully withdrawn position. The variation of the coefficient with temperature and pressure is included. The ke rr (considering moderator temperature and density effects) versus temperature at 1000 psi corresponding to the negative moderator temperature coefficient used is shown in Figure 15.1.4-1. The core power reactivity feedback is modeled as a function of core thermal power and core mass flow. The feedback calculations performed in LOFTRAN are discussed further in Subsection 15.1.5.2.1
=
Minimum capability for injection of boric acid solution corresponding to the most I restrictive single failure in the passive core cooling system. Low-concentration boric acid must be swept from the core makeup tank lines downstream of isolation valves I before delivery of boric acid (3400 ppm) to the reactor coolant loops. This effect has been accounted for in the analysis I = The case studied is a steam flow of 520 pounds per second at 1200 psia with offsite power available. This con.ervatively models the maximum capacity of any single steam dump, relief, or safety valve. Initial hot shutdown conditions at time zero are assumed since this represents the most conservative initial conditions Should the reactor be just critical or operating at powcr at the time of a steam release, the reactor is tripped by the normal overpower protection when power level reaches a trip point. Following a trip at power, the reactor coolant system contains more stored energy than at no-load, the average coolant temperature is higher than at no-load and there is appreciable energy stored in the fuel. Thus, the additional stored energy is removed via the cooldown caused by the steam release before the no-load conditions of reactor coolant system temperature and shutdown margin assumed in the analyses are reached. After the additional stored energy is removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis, which assumes no-load condition at time zero. However, since the initial steam generator water inventory is greatest at no-load, the magnitude and duration of the reactor coolant system cooldown are less for steam line release occurring at power. In computing the steam flow, the Moody Curve (Reference 3) for f(UD) = 0 is used Perfect moisture separation in the steam generator Offsite power is available, since this maximizes the cooldown Maximum cold startup feedwater flow Four reactor coolant pumps are initially operating Manual actuation of the passive residual heat removal system at time zero is conservatively assumed to maximize the cooldown Revision: 5 February 29,1996 15.1-10 3 W65tingh0USB
- 15. Accident Analyses i
m i
\ \'J 15.1.4.2.2 Results The results presented conservatively indicate the events that would occur assuming a secondary system steam release since it is postulated that the conditions just described occur simultaneously.
I Figures 15.1.4-2 through 15.1.4-13 show the transient results for a steam flow of 520 pounds per second at 1200 psia. The assumed steam release is typical of the capacity of any single steam dump, relief, or safety valve. Core makeup tanks injection and the associated tripping , ,he reactor coolant pumps are initiated automatically by low pressurizer pressure safeguard signal. Boron solution I at 3400 ppm enters the reactor coolant system, providing enough negative reactivity to prevent core damage. Later in the transient, as the reactor coolant pressure continues to fall, the I accumulators actuate and inject boron solution at 2600 ppm. The transient is conservative with respect to cooldown, since no credit is taken for the energy stored in the system metal other than that of the fuel elements and steam generator tubes, and the passive residual heat removal system is assumed to be actuated at time zero. Since the limiting portion of the transiem occurs over a period of about five minutes, the neglected stored energy is likely to have a significant effect in slowing the cooldown. O o The calculated time sequence of events for this accident is listed ia Table 15.1.2-1. a 15.1.4.3 Margin to Critical Heat Flux l Departure from nucleate boiling analysis is perfomied for the inadvenent opening of a steam generator relief or safety valve. That analysis demonstrates that the departure from nucleate boiling design basis, as described in Section 4.4, is met for the inadvertent opening of a steam generator relief or safety valve. 15.1.4.4 Conclusions The analysis shows that the criterion stated earlier in this subsection is satisfied. For an inadvertent opening of any single steam dump or a steam generator relief or safety valve, the departure from nucleate boiling design b..: sis is met. 15.1.5 Steam System Piping Failure 15.1.5.1 Identification of Causes and Accident Description The steam release arising from a rupture of a main steam line results in an initial increase in steam flow, which decreases during the accident as the steam pressure falls. The energy l removal from the reactor coolant system causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown ! g) (
- 2. ,
results in an insertion of positive reactivity. Revision: 5 k [ W8Stiligh0USe 15.1-11 Februany 29,1996
l i
- 15. Accid:nt Analyses If the most reactive rod cluster control assembly is assumed stuck in its fully withdrawn e
position after reactor trip, there is an increased possibility that the core becomes critical and returns to power. A retum to power following a steam line rupture is a potential problem l mainly because of the existing high-power peaking factors, assuming the most reactive rod j cluster control assembly to be stuck in its fully withdrawn position. The core is ultimately shut down by the boric acid solution delivered by the passive core cooling system. l The analysis of a main steam line rupture is performed to demonstrate that the following SRP Section 15.1.5 evaluation criterion is satisfied: Assuming the most reactive stuck rod cluster control assembly with or without offsite power and assuming a single failure in the engineered safety features (ESFs), the core cooling capa-bility is maintained. Radiation doses do not exceed the guidelines of 10 CFR 100. Departure from nucleate boiling and possi le clad perforation following a steam pipe rupture are not necessarily unacceptable. But, the following analysis, in fact, shows that the departure from nucleate boiling design basis is not exceeded for any rupture, assuming the most reactive assembly stuck in its fully withdrawn position. The departure from nucleate boiling ratio design basis is discussed in Section 4.4. A major steam line rupture is classified as a Condition IV event. Effects of minor seconda:v system pipe breaks are bounded by the analysis presented in this section. Minor secondary system pipe breaks are classified as Condition III events, as I described in Subsection 15.0.13. The major rupture of a steam line is the most limiting cooldown transient and is analyzed at zero power with no decay heat. Decay heat retards the cooldown, thereby reducing the likelihood that the reactor returns to power. A detailed analysis of this transient with the most limiting break size, a double-ended rupture, is presented here. Certain assumptions used in this analysis are discussed in Reference 4. Reference 4 also contains a discussion of the spectrum of break sizes and power levels arulyzed. The following functions provide the protection for a steam line rupture: (See Subsection 7.2.1.1.2.) Core makeup tank actuation from any of the following: l Two out of four low pressurizer pressure signals Two out of four High-1 containment pressure signals l Two out of four low steam line pressure signals in any loop i ' Two out of four low Tcold signals in any one loop Two out of four low pressurizer level signals O Revision: 5 February 29,1996 15.1-12 [ W8Stingh00S8
- 15. Accident Analyses
~]
- The overpower reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the safeguards signal Redundant isolation of the main feedwater lines l
l Sustained high feedwater flow causes additional cooldown. Therefore, in addition to the I normal control action that closes the main feedwater control valves, the safeguards signal I rapidly closes all feedwater control valves and feedwater isolation valves, and trips the I main feedwater pumps. i Redundant isolation of the startup feedwater system i i Sustained high startup feedwater flow causes additional cooldown. Therefore, the low I Tcold signal closes the stattup feedwater control and isolation valves. Fast-acting main steam Ane isolation valves (MSIVs) (designed to close in less than 10 l seconds) on any of the following: Two out of four High-1 containment pressure Two out of the four low steam line pressure signals in any one loop (above permissive P-II) / () - Two out of four high negative steam pressure rates in any one loop (below permissive P-ll) A fast-acting main steam isolation valve is provided in each steam line. These valves fully close within 10 seconds of actuation following a large break in the steam line. For breaks downstream of the main steam line isolation valves, closure of at least one valve in each line completely terminates the blowdown. For any break in any location, no more than one steam generator would experience an uncontrolled blowdown even if one of the main steam line isolation valves fails to close. A description of steam line isolation is included in Chapter 10. Flow restrictors are installed in the steam generator outlet nozzle, as an integral part of the steam generator. The effective throat area of the nozzles is 1.4 square feet, which is considerably less than the main steam pipe area; thus, the flow restrictors also serve to limit the maximum steam flow for a break at any location. Design criteria and methods of protection of safety-related equipment from the dynamic effects of postulated piping ruptures are provided in Section 3.6. g) (
\_-
Revision: 5 3 W8Stilighouse 15.1-13 February 29,1996
_ L--( ) 2 f 15. Accid =t Analyses l 15.1.5.2 Analysis of Effects and Consequences O 15.1.5.2.1 Method of Analysis The analysis of the steam pipe rupture is performed to detennine the following: The core heat flux and reactor coolant system temperature and pressure resulting from the cooldown following the steam line break. The LOFTRAN code (Reference 1) is used to determine the system transient. The thermal and hydraulic behavior of the core following a steam line break. A detailed thermal and hydraulic digital computer code, THINC, is used to determine if departure from nucleate boiling occurs for the core transient conditions computed by the LOFTRAN code. The following conditions are assumed to exist at the time of a main steam line break accident: End-of-cycle shutdown margin at no-load, equilibrium xenon conditions, and the most reactive rod control assembly stuck in its fully withdrawn position. Operation of the control rod mechanical shim and axial offset banks during core burnup is rastricted by the insertion limits so that shutdown margin mquirements are satisfied.
=
A negative moderator coefficient corresponding to the end-of-life rodded core with the most reactive rod cluster control assembly in the fully withdrawn position. The variation of the coefficient with temperature and pressure has been included. The kerr (considering moderator temperature and density effects) versus temperature at 1000 psia corresponding to the negative moderator temperature coefficient used is shown in Figure 15.1.4-1. The core power reactivity feedback is modeled as a function of thermal power and core mass flow. The core properties used in the LOFTRAN mode for feedback calculations are generated by combining those in the sector nearest the affected steam generator with those associated with the remaining sector. The resultant properties reflect a combination process that accounts for inlet plenum fluid mixing and a conservative weighing of the fluid properties from the coldest core sector. To verify the conservatism of this method, the power predictions of the LOFTRAN point kinetics modeling are confirmed by comparison with detailed core analysis for the limiting conditions of the cases considered. This core analysis explicitly models the hypothetical core ! configuration (i.e., stuck rod cluster control assembly, non-uniform inlet temperatures, ' pressure, flow, and boron concentration) and directly evaluates the total reactivity feedback including power, boron, and density redistribution in an integral fashion. The effect of void formation is also included. l l l Comparison of the results from the detailed core analysis with the LOFTRAN predictions ! verify the overall conservatism of the methodology. That is, the specific power, temperature, I Revision: 5 February 29,1996 15.1-14 3 W65tingh0US8 l l i
l Stu inn
- 15. Accident Analyses n :4 f
V and flow conditions used to perform the departure from nucleate boiling analysis are conservative. The portions of the passive core cooling system used in mitigating a steam line rupture are the core makeup tanks and the accumulators. There are no single failures that prevent core makeup tank injection. In modeling the core makeup tanks and the accumulators, conservative assumptions are used that minimize the capability to add borated water. Specifically, the core makeup tank injection line characteristics modeled reflect the failure of one core makeup tank discharge valve. Maximum overall fuel-to-coolant heat transfer coefficient, to maximize rate of cooldown Since the steam generators are provided with integral flow restrictors with a 1.4-square-foot throat area, any rupture in a steam line with a break area greater than 1.4 square feet, regardless of location, has the same effect on the primary plant as the 1.4-square-foot-double-ended rupture. The limiting case considered in determining the core power ; and reactor coolant system transient is the complete severance of a pipe, with the plant initially at no-load conditions, full reactor coolant flow with offsite power available. The results of this case clearly bound the loss of offsite power for the following four reasons: Loss of offsite power results in an immediate reactor coolant pump coastdown at the initiation of the transient. This reduces the severity of the reactor coolant [] V system cooldown by reducing primary-to-secondary heat transfer. The lessening of the cooldown, in turn, reduces the magnitude of the return to power. Following actuation, the core makeup tank provides borated water that injects into the reactor coolant system. Flow from the core makeup tank increases if the reactor coolant pumps have coasted down. Therefore, the analysis performed with offsite power and continued reactor coolant pump operation actually minimizes the rate of boron injection into the core and is conservative. In recognition of the preceding item, the protection system automatically provides a safety-related signal that initiates the coastdown of the reactor coolant pumps in parallel with core makeup tank actuation. Since this reactor coolant pump function is actuated early during the steam line break event (right after core makeup tank actuation), there is very little difference in the predicted departure from nucleate boiling ratio between cases with and without offsite power. Because of the passive nature of the safety injection system, the loss of offsite power does not delay the actuation of the safety injection system. ( Power peaking factors corresponding to one stuck rod cluster control assembly and nonuniform core inlet coolant temperatures are determined at the end of core life. The coldest core inlet temperatures are assumed to occur in the sector with the stuck rod. The power peaking factors account for the effect of the local void in the region of the stuck rod cluster control assembly during the return to power phase following the steam g] Revision: 5 [ W85tingh0US8 15.1-15 February 29,1996
T.2 i E
~
- 15. Accident Analyses line break. This void in conjunction with the large negative moderator coefficient par-O\
l tially offsets the effect of the stuck assembly. The power peaking factors depend upon ) the core power, temperature, pres,sure, and flow, and therefore may differ for each case studied. The analysis assumes initial hot standby conditions at time zero since this represents the most pessimistic initial condition. If the reactor is just critical or operating at power at the time of a steam line break, the reactor is tripped by the normal overpower protection system when power level reaches a trip point. Following a trip at power, the reactor coolant system contains more stored energy than at no-load, the average coolant temperature is higher than at no-load, and there is appreciable energy stored in the fuel. Thus, the additional stored energy reduces the cooldown caused by the steam line break before the no-load conditions of reactor coolant system temperature and shutdown margin assumed in the analyses are reached. After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no-load condition at time zero. In computing the steam flow during a steam line break, the Moody Curve (Reference 3) for f(IJD) = 0 is used. Perfect moisture separation in the steam generator. Maximum cold startup feedwater flow plus nominal 100 percent main feedwater flow. Four reactor coolant pumps are initially operating. Manual actuation of the passive residual heat removal system at time zero is conservatively assumed to maximize the cooldown. 15.1.5.2.2 Results The calculated sequence of events for the analyzed case is shown in Table 15.1.2-1. The results presented conservatively indicate the events that would occur assuming a steam line I rupture, since it is postulated that the conditions described occur simultaneously. 15.1.5.2.3 Core Power and Reactor Coolant System Transient Figures 15.1.5-1 through 15.1.5-14 show the reactor coolant system transient and core heat flux following a main steam line rupture (complete severance of a pipe) at initial no-load condition. Offsite power is assumed available so that, initially, full reactor coolant flow exists. During the course of the event, the reactor protection system initiates a coastdown of the reactor coolant pumps in conjunction with actuation of the core makeup tanks. The transient shown Revision: 5 February 29,1996 15.1-16 3 WestilighollSe
a _
- 15. Accident Analyses
\.j assumes an uncontrolled steam release from only one steam generator. Steam release from more than one steam generator is prevented by automatic trip of the fast-acting main steam isolation valves in the steam lines by high containment pressure signals or by low steam line i pressure signals. Even with the failure of one valve, release is limited to approximately 10 seconds for the other steam generator while the one generator blows down. The main steam isolation valves fully close in less than 10 seconds from receipt of a closure signal. As shown in Figure 15.1.5-3, the core attains criticality with the rod cluster control assemblies inserted (with the design shutdown assuming the most reactive rod cluster control assembly I stuck) before boron solution at 3400 ppm (from core makeup tanks) or 2600 ppm (from accumulators) enters the reactor coolant system. A peak core power significantly lower than the nominal full-power value is attained. The calculation assumes that the boric acid is mixed with and diluted by the water flowing in the reactor coolant system before entering the reactor core. The concentration after mixing depends upon the relative flow rates in the reactor coolant system and from the core makeup tanks or accumulators (or both). The variation of mass flow rate in the reactor coolant system due to weter density changes is included in the calculation. So is the variation of flow rate from the core makeup tanks or accumulators (or both) due to changes in the reactor coolant system pressure and temperature and the pressurizer level. The reactor coolant system and passive injection flow calculations include the line losses. (j] \ l l At no time during the analyzed steam line break event does the co:e makeup tank level approach the setpoint for actuation of the automatic depressurization. During non-LOCA I events, the core makeup tanks remain filled with water. The volume ofinjection flow leaving i the core makeup tank is offset by an equal volume of recirculation flow that enters the core I makeup tanks via the reactor coolant system cold leg balance lines. The passive residual heat removal system provides a passive, long-term mean of removing the core decay and stored heat by transferring the energy via the passive residual heat removal heat exchanger to the in-containment refueling water storage tank (IRWST). Normally the l passive residual heat removal heat exchanger is actuated automatically when the steam generator level falls below the low wide range level. For the main steam line rupture case I analyz.ed, the passive residual heat removal heat exchanger is conservatively actuated at time zero to rnaximize the cooldown. 15.1.5.2.4 Margin to Qitical Heat Mux The case presented in Subsection 15.1.5.2.2 conservatively models the expected behavior of the plant during a steam system piping failure. This includes the tripping of the reactor coolant pumps coincident with core makeup tank actuation. A departure from nucleate boiling analysis is performed using limiting assumptions that bound those of Subsection 15.1.5.2.2. I l Under the low flow (natural circulation) conditions actually present in the AP600 transient, the return to power is severely limited by the large negative feedback due to flow and power. f N) Revislom 5 [ W8Stingt100Se 15.1-17 February 29,1996
- 15. Accident Analyses 1
The results of the bounding analysis demonstrate that the depanure from nucleate boiling e ) design basis, as described in Section 4.4, is met for the steam system piping failure event. 15.1.5.3 Conclusions The analysis shows that the depanure from nucleate boiling design basis is met for the steam ! system piping failure event. Departure from nucleate boiling and possible clad perforation following a steam pipe rupture are not precluded by the criteria. The preceding analysis , shows that no departure from nucleate boiling occurs for the rupture assuming the most reac-tive rod cluster control assembly stuck in its fully withdrawn position. The radiological consequences of this limiting event are within the dose criteria of 10 CFR 100. ] j 15.1.5.4 Radiological Consequences The evaluation of the radiological consequences of a postulated main steam line break outside containment assumes that the reactor has been operating with the design basis fuel defect level (0.25 percent of power produced by fuel rods containing cladding defects) and that leaking steam generator tubes have resulted in a buildup of activity in the secondary coolant.
?ollowing the rupture, startup feedwater to the faulted loop is isolated and the steam generator is allowed to steam dry. Any radioiodines carried from the primary coolant into the faulted steam generator via leaking tubes are assumed to be released directly to the environment. It is conservatively assumed that the reactor is cooled by steaming from the intact loop.
15.1.5.4.1 Source Term The only significant radionuclide releases due to the main steam line break are the iodines that become airborne and are released to the environment as a result of the accident. Noble gases are also released to the environment but their impact is secondary since any noble gases entering the secondary side during normal operation are rapidly released to the environment. The accident conditions do not result in an increase in the releases of noble gases beyond the levels associated with normal operation. The analysis considers two different reactor coolant iodine source terms. both of which consider the iodine spiking phenomenon. In one case the initial iodine .ancentrations are assumed to be those associated with the design fuel defect level. The iodine spike is assumed to be initiated by the accident with the spike causing an increasing level of iodine in the reac-tor coolant. The second case assumes that the iodine spike occurs prior to the accident and that the maximum reactor coolant iodine concentration exists at the time the accident occurs. The reactor coolant noble gas concentrations are assumed to be those associated with the design fuel defect level. Revision: 5 e February 29,1996 15.1-18 [ WBStingh00S8
- 15. Accide:t Analyses l ((3 1
l The secondary coolant is assumed to have an iodine source term of 0.04 pCi/g dose equivalent I-131. This is 10 percent of the design basis primary coolant activity. l 15.1.5.4.2 Release Pathways There are two components to the accident releases: The secondary coolant in the steam generator of the faulted loop is assumed to be released out the break as steam. Any iodine contained in the coolant is assumed to be released.
=
The reactor coolant leaking into the steam generator of the faulted loop is assumed to be micased to the environment without any credit for iodine partitioning or iodine plateout onto the interior of the stenn generator. Credit is taken for decay of radionuclides until release to the environment. After release to the environment no consideration is given to radioactive decay or to cloud depletion by ground deposition during transport offsite. 15.1.5.4.3 Dose Calculation Models The models used to calculate thyroid doses and equivalent whole body doses resulting from (] inhalation of iodines and the model used to calculate whole-body doses due to immersion in U i the released noble gas activity are provided in WCAP-14601 (Reference 5). 15.1.5.4.4 Analytical Assumptions and Parameters The assumptions and parameters used in the analysis are listed in Table 15.1.5-1. 15.1.5.4.5 Identification of Conservatisms The assumptions and parameters used in the analysis contain a number of significant conservatisms: The reactor coolant activities are based on a fuel defect level of 0.25 percent whereas the expected fuel defect level is far less than this (See Section 11.1). The assumed leakage of 500 gallons of reactor coolant per day into each steam generator is conservative. The leakage is expected to be a small fraction of this during normal operation. The conservatively selected meteorological conditions are present only rarely. p U Revision: 5 [ W8stinghouse 15.1-19 February 29,1996
g = E
- 15. Accident Analyses l
15.1.5A.6 Dose Calculation Models O Using the assumptions from Table 15.1.5-1, the calculated release of activity due to the postu:4ted main steam line break is given in Table 15.1.5-2. The thyroid doses due to inha'ation of airbome iodines and the whole-body doses due to the combination ofimmersion in the noble gas cloud and the equivalent whole body dose resulting from inhalation ofiodines are analyzed for the site boundary (0 to 2 hour residence time) and for the low population zone outer boundary (0 to 8 hour residence time). The doses are provided in Table 15.1.5-3. The doses due to a main steam line break with a pre-existing iodine spike are within the guideline values of 10 CFR 100. The doses due to a main steam line break with an accident-initiated iodine spike are a small fraction (less than 10 percent) of the 10 CFR 100 limits. These doses are thus within the dose criteria defined in SRP Section 15.1.5, Appendix A. I 15.1.6 Inadvertent Operation of the Passive Residual Heat Removal Heat Exchanger 15.1.6.1 Identification of Causes and Accident Description l The inadvertent actuation of the passive residual heat removal heat exchanger causes an I injection of relatively cold water into the reactor coolant system. This produces a reactivity I insertion in the presence of a negative moderator temperature coefficient. The I overpower /overtemperature protection (neutron overpower, overtemperature, and overpower l AT trips) is intended to prevent a power increase that could lead to a departure from nucleate I boiling ratio less than the safety analysis limit. In addition, since the cold leg temperature is I reduced and the reactor coolant system depressurizes during this event, the low cold leg I temperature or low pressurizer pressure functions could generate a reactor trip. These i protection functions do not terminate operation of the passive residual heat removal heat I exchanger. The inadvertent actuation of the passive residual heat removal system could be caused by I operator error or a false actuating signal. Actuation of the passive residual heat removal heat I exchanger involves opening the isolation valves, which establishes a flow path from one I reactor coolant system hot leg, through the passive residual heat removal heat exchanger, and back into the associated steam generator cold leg plenum. l The passive residual heat removal heat exchanger is located above the core to promote natural circulation flow when the reactor coolant pumps are not operating. With the reactor coolant i pumps in operation, flow through the passive residual heat removal heat exchanger is I enhanced. The heat sink for the passive residual heat removal heat exchanger is provided by the in-containment residual water storage tank, in which the passive residual heat removal heat l exchanger is submerged. Since the fluid in the heat exchanger is in thermal equilibrium with I water in the tank, the initial flow out of the passive residual heat removal exchange'r is I significantly colder than the reactor coolant system fluid. Following this initial insurge, the I reduction in cold leg temperature is limited by the cooling capability of the passive residual I heat removal heat exchanger. Since the passive residual heat removal heat exchanger is O Revislom 5 February 29,19% 15.1-20 W Westinghouse
? l
$* t i
- 15. Accident Analyses d i
7 l 4 s / connected to only one reactor coolant system loop, the cooldown resulting from its actuation is asymmetric with respect to the core. l The response of the plant to an inadvertent passive residual heat removal heat exchanger actuation with the plant at no-load conditions is bounded by the analyses performed for the ' inadvertent opening of a steam generator relief or safety valve event (Subsection 15.1.4) and the steam system piping failure event (Subsection 15.1.5). Both of these events are conserva-I tively analyzed assuming passive residual heat removal heat exchanger actuation coincident l with the steam line depressurization. Therefore, only the response of the plant to an inadver-l tent passive residual heat removal initiation with the core at power is considered here. l The inadvertent actuation of the passive residual heat removal heat exchanger event is a { Condition II event, a fault of moderate frequency. Plant systems and equipment available to : I mitigate the effects of the accident are discussed in Subsection 15.0.8 and listed in i Table 15.0.6. The following reactor protection system functions are available to provide ' l protection in the event of an inadvertent passive residual heat removal heat exchanger actua- ! tion-1
=
The overpower reactor trips (neutron flux and AT) Two out of four low pressurizer pressure signals j
=
Two out of four low Tcold signals in any one loop
=
Two out of four low pressurizer level signals 7 15.1.6.2 Analysis of Effects and Consequences 15.1.6.2.1 Method of Analysis l The excessive heat removal due to an inadvertent passive residual heat removal heat i exchanger actuation transient is analyzed by using the digital computer code LOFrRAN (Reference 1). This code simulates a multiloop system, neutron kinetics, the pressurizer, pressurizer safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level. The system is analyzed to demonstrate plant behavior in the event of an inadvenent passive i residual heat removal heat exchanger actuation due to an operator error or a false actuation I signal that opens the valves that normally isolate the passive residual heat removal heat i exchanger from the remainder of the reactor coolant system. Both fuP wwer and zero-load I conditions are to be considered. The analyses for the inadvertent openmg of a steam gener-ator relief or safety valve event (Subsection 15.1.4) and the steam system piping failure event (Subsection 15.1.5) bound the results for the zero-power inadvertent passive residual heat I removal heat exchanger actuation transient. l The case considered here is the response of the plant to an inadvertent passive residual heat I removal heat exchanger initiation with the core initially operating at full power. The
- p (v )
Revision: 5 [ W85fingh00Se 15.1-21 February 29,1996
== =g
= m
~
- 15. Accident Analyses reactivity insertion transient arising from the inadvertent actuation of the passive residual heat O
I removal heat exchanger is calculated including the following assumptions: 1 = With the core at full power, the inadvertent passive residual 1, eat removal heat exchanger actuation occurs at 10 seconds. The LOFTRAN code explicitly models the performance I of the passive residual heat removal heat exchanger and the resulting cooldown transient expedenced by the reactor coolant system. l l = A conservative model for predicting the power excursion experienced by the core. I I This includes the use of a negative moderator coefficient corresponding to the end-of-life I rodded core. The variation of the coefficient with temperature and pressure has been i included in conjunction with a low level of power feedback. I l The core properties used in the LOFfRAN Code for feedback calculations are generated I by combining those in the sector nearest the loop with the passive residual heat removal i system with those associated with the remaining sector. The resultant properties reflect I a combination process that accounts for inlet plenum fluid mixing and a conservative I weighing of the fluid properties from the coldest core sector. I l To verify the conservatism of this method, the power predictions of the LOFfRAN point I kinetics modeling are confirmed by comparison with detailed core analysis for the i limiting conditions of the cases considered. This core analysis explicitly models the I hypotiietical core configuration (i.e., non-uniform inlet temperatures, pressure, flow and I boron) and directly evaluates the total reactivity feedback including power, boron and I density redistribution in an integral fashion. l l Comparison of the results from the detailed core analysis with the LOFTRAN predictions verify the overall conservatism of the methodology. That is, the specific power, I temperature, and flow conditions used to perform the depanure from nucleate boiling I analysis are conservative. I I
- The reactor trips on high neutron flux, ovenemperature and overpower AT trips are I
conservatively ignored. The analysis demonstrates that the applicable safety analysis I limits are met without a reactor trip being generated.
=
No credit is taken for the heat capacity of the reactor coolant system and steam generator thick metal in attenuating the resulting plant cooldown. l
=
Control systems are assumed to function only if their operation results in more severe I accident results. For the inadvenent passive residual heat removal heat exchanger I actuation event, both cases with and without automatic rod control are considered. Plant characteristics and initial conditions are further discussed in Subsection 15.0.3. No I single active failure prevents operation of the reactor protection system functions assumed in O l Revision: 5 February 29,1996 15.1-22 [ W85tingh00S8
- 15. Accident Analyses i vn) 1(
l l the analysis. A discussion of anticipated transients without scram considerations is presented in Section 15.8. 15.1.6.2.2 Results l
.I De system responses to an inadvertent passive residual heat removal heat exchanger actuation I
at power event, with manual rod control, are shown in Figures 15.1.6-1 through 15.1.6-10. The full-power case with manual rod control results in the greatest power increase. I The inadvertent operation of the passive residual heat removal heat exchanger incident is an i overpower transient for which the fuel temperature rises. Assuming a reactor trip does not 1 occur the plant reaches a new equilibrium condition at a higher power level corresponding to I the increase in power dc.nanded by the system. In the limiting case analyzed, the plant power I stabilizes at about 111 percent ofits nominal value. 1 Assuming the rod control system to be in automatic results in a slightly less limhing transient, as the control rods are inserted in response to a primary-to-secondary power mismatch. The results show the increase in nuclear power and AT associated with the inadvertent passive residual heat removal system actuation at full power. The departure from nucleate boiling ratio does not drop below the limit value. l l Since the power level rises during the inadvertent passive residual heat removal heat i l (O exchanger initiation, the fuel temperatures will also rise until after reactor trip occurs. The core heat flux lags behind the neutron flux response because of the fuel rod thermal time constant. The peak fuel temperature remains below the fuel melting temperature. The transient results show that departure from nucleate boiling does not occur at any time I during the inadvertent passive residual heat removal heat exchanger actuation event. So the ability of the primary coolant to remove heat from the fuel rods is not reduced. The calculated sequence of events for this accident is shown in Table 15.1.2-1. l The inadvertent operation of the passive residual heat removal heat exchanger is not included I among the design overpower transients considered in subsection 4.3. The conservative safety I analysis assumptions applied to this event do not credit a reactor trip to preclude the core i power from rising above 118 percent of rated thermal power. The nature of this excessive I cooldown transient dictates that the predicted power excursion is associated with very low I core inlet temperatures which can partially offset the penalties associated with the high power. 15.1.6.3 Conclusions The results of the analysis show that the departure from nucleate boiling ratios encountered I for an inadvertent actuation of the passive residual heat removal heat exchanger at power are l above the safety analysis limit values. (The departure from nucleate biling ratio design basis I is described in Section 4.4.) The results for an inadvertent passive r. sidual heat removal heat i exchanger actuation initiated from zero load conditions are bounded by the inadvertent A Revision: 5 I T W85tkighouse 15.1-23 February 29,1996 I
I I
- 15. Accident Analyses opening of a steam generator relief or safety valve event (Subsection 15.1.4) and the steam O\ i system piping failure event (Subsection 15.1.5).
I 15.1.7 Combined License Information ' l l This section has no requirement for additional information to be provided in support of the l Combined License application. l 15.1.8 References l
- 1. Bumett, T. W. T., et al., "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary),
and WCAP-7907 A (Nonproprietary), April 1984.
- 2. Friedland, A. J., and Ray, S., " Revised Thermal Design Procedure," WCAP-11397-P-A (Proprietary) and WCAP-11398-A (Nonproprietary), April 1989.
- 3. Moody, F. S., " Transactions of the ASME, Joumal of Heat Transfer," Figure 3, ,
page 134, February 1965. l
- 4. Wood, D. C., and Hollingsworth, S. D., " Reactor Core Response to Excessive Secondary ;
Steam Releases," WCAP-9226 (Proprietary) and WCAP-9227 (Nonproprietary), January 1978. { I S. Carlin, E. L., Kemper, R. M., Gresham, J. A., "AP600 Accident Analyses - Evaluation 1 Models," WCAP-14601 (Proprietary), Febmary 1996. I l O Revision: 5 February 29,1996 15.1-24 3 WOStingh0US8 1
i i t-+tm+++
- 15. Accident Analyses l
gg ; Table 15.1.2-1 (Sheet 1 of 3) TIME SEQUENCE OF EVENTS FOR INCIDENTS THAT RESULT IN AN INCREASE IN HEAT REMOVAL FROM TIIE PRIMARY SYSTF31 Accident Event Time (s) Feedwater system mal- One main feedwater control valve fails fully open 0.0 functions that result in an I increase in feedwater flow Minimum DNBR occurs 55.4 l Turbine trip /feedwater isolation on high steam 486.0 . generator level I Reactor trip on high steam generator level 490.0 l Feedwater isolation 498.0 Excessive increase in secondary steam
- 1. Manual reactor con- 10 percent step load increase 0.0 trol (minimum
[3 I moderator feedback) Equilibrium conditions reached (approximate time 220.
~C/ only)
- 2. Manual reactor con- 10 percent step load increase 0.0 trol (maximum mod-I erator feedback) Equilibrium conditions reached (approximate time 70.
only)
- 3. Automatic reactor 10 percent step load increase 0.0 control (minimum l moderator feedback) Equilibrium conditions reached (approximate time 140.
onip
- 4. . Automatic reactor 10 percent step load increase 0.0 control (maximum I moderator feedback) Equilibrium conditions reached (approximate time 50.
only) (g) v Revision: 5 W Westinghouse 15.1-25 February 29,1996
5
- 15. Accident Analyses Table 15.1.2-1 (Sheet 2 of 3) e TIME SEQUENCE OF EVENTS FOR INCIDENTS TIIAT RESULT IN AN INCREASE IN HEAT REMOVAL FROM THE PRIMARY SYSTEM Accident Event Time (s)
Inadvertent opening of a Inadvertent opening of one main steam safety or 0 steam generator relief or relief valve l Criticality attained 96.4 l Safeguards actuation signal on safeguards low T, 139.0 l Core makeup tank actuation 161.0 l Boron reaches core 176.7 Steam system piping fail- Steam line ruptures 0 ure Safeguards actuation signal on safeguards low 0.8 steam line pressure l Pressurizer empty 15.6 l Criticality attained 15.6 l Boron reaches core 32.0 0 Revision: 5 February 29,1996 15.1-26 3 Westinghouse
=-
- 15. Accid:nt Analyses o Table 15.1.2-1 (Sheet 3 of 3)
TIME SEQUENCE OF EVENTS FOR INCIDENTS TIIAT RESULT IN AN INCREASE IN IIEAT REMOVAL FROM THE PRIMARY SYSTEM Accident Event Time (s) Inadvertent operation of inadvertent actuation of the PRHR 10.0 the PRHR I Minimum DNBR occurs 24.2 1 Equilibrium condition reached (approximate time 100 l only) C 4 Revision: 5 [ Westingh00S8 15.1-27 February 29,1996 i
- 15. Accidaat Analyses Teble 15.1.5-1 e
PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE BREAK Reactor coolt.nt iodine activity Accident-initiated spike: Initial activity equal to the design basis reactor coolant activity of 0.4 pCi/g dose equivalent I-131 (see Table 11.1-2) with an assumed iodine spike that increases the rate of iodine release I from fuel into the coolant by a factor of 500 (see WCAP-14601 I [ Reference 5]) Preaccident spike: An assumed iodine spike that has resulted in an increase in the reactor coolant activity to 24 pCi/g of dose equivalent I-131 1 (see WCAP-14601 [ Reference 5]) Reactor coolant n)ble gas activity Design basis activity (See Table 11.1-2) Secondary coolant initial iodine activity 0.04 pCi/g dose equivalent 1131 (10% of design basis teactor coolant concentrations listed in Table 11.1-2) Secondary coolant mass (Ib) 3.58 E+05 Duration of accident (hr) 8 l Atmospheric dispersion factors As detailed in WCAP-14601 (Reference 5) { Steam generator in faulted loop Initial water mass (Ib) 1.79 E+05 Primary to secondary leak rate (Ib/hr) 130(a) Iodine partition coefficient 1.0 Steam generator in intact loop Primary to secondary leak rate (Ib/hr) 130(a) Steam released (Ib) 0 - 2 hr 3.07 E+05 2 - 8 hr 6.42 E+05 Iodine partition coefficient 0.C (a) Equivalent to 500 gpd at 561.5'F and 2250 psia i. e Revision: 5 February 29,1996 15.1-28 3 W85tirigh00Se
J -i!~
- 15. Accide t Analyses s
,7 Table 15.1.5 2 ACTIVITY RELEASED TO THE ENVIRONMENT DUE TO A STEAM LINE BREAK Iodines (accident-initiated iodine spike' Isotone 0 2 hr (Ci) 2 - 8 hr (Ci)
I-131 6.4 8.2 I-132 16.5 14.4 I-133 11.8 15.7 I-134 4.9 1.7 . I-135 8.9 11.2 Iodine (preaccident iodine spike) Isotone 0 - 2 hr (Ci) 2 - 8 hr (Ci) - I-131 7.8 12.5 I-132 12.0 6.0 g3 I-133 13.3 18.6
) I-134 2.4 0.4 \d I-135 8.6 8.9 Noble gases (both cases)
Isotone 0 - 2 hr (Ci) 2 - 8 hr (Ci) Xe-131m 0.1 0.4 Xe-133m 1.2 3.4 Xe-133 19 54 Xe-135m 0.04 0.0 Xe-135 0.05 1.1 Xe-138 0.007 0.0 Kr-85m 0.1 0.2 Kr-85 0.005 1.4 i Kr-87 0.5 0.02 i Kr-88 0.007 0.2 1 CT [V l Revision: 5 [ WOStingh00S8 15.1-29 February 29,1996
es ,
- 15. Accid nt Analyses l WAM -
Table 15.1.5-3 el RADIOLOGICAL CONSEQUENCES OF A STEAM LINE BREAK Thyroid doses (rem) Case 1 - Accident-initiated iodine spike S:;e bou ,4 (0 - 2 hr) 3.2 Low population zone (0 - 8 hr) 1.0 Case 2 - Preaccident spike Site boundary (0 - 2 hr) 3.9
, Low population zone (0 - 8 hr) 1.3 Whole body doses (rem)
Case 1 - Accident-initiated iodine spike Site boundary (0 - 2 hr) 0.10 Low population zone (0 - 8 hr) 0.03 Case 2 - Preaccident spike Site boundary (0 - 2 hr) 0.12 Low population zone (0 - 8 hr) 0.04 Revision: S e February 29,1996 15.1-30 W Westiligt100Se
- 15. Accid:nt Analyses O
i 1 l l. LOFT 4AP 1.0 X1995/04/27 17e42eSO.33 3859835198314 TP C1995/0*/i AP900 FEE 0LINE MALFUNCTICA QNCLR 0 0 0 12 I
.m 3 4:
X . h o 1- f \ Z w .
.. o 8- -
z . o - O O < oc w
.6- -
w -
~
Q w 4- - D . O C3. -
~
Oc
< .2- -
w ~ J O -
- l3 ~
2 ,,,,,,,,,.,,,,e ...e e i i r ' ' ' O 160 260 360 460 560 60 TlWE (S) 1 Figure 15.1.2-1 Feedwater Control Valve Malfunction Nuclear Power
\
i Revision: 5 Y W85tirighouse 15.1 31 February 29,1996 l [
N~
- 15. Accid:nt Antlyses WAE --
l LOFT 4AP 1.8 TP C1995/04/10 . X1998/04/37 17: 43 50 33 3858534394314 l APSOS FEE 0LINE MALFUNCTION 4 DELTAT 0 0 0 i 80 F 1 i n I
- 60- -
- w w .
ce c . w 1 Ca
- 40- -
1 J j w -
- Ca 1 a-1 o 20- -
O
-J .
4 l i -
' ' ' ' ' ' ' ' ' ' ' ' ' ' ~
4 0 l l l ,
- u 100 200 360 400 500 60
- TIME (S) i
.t 4 , Figure 15.1.2-2 Feedwater Control Valve M11 function Loop Delta T i O j Revision: 5 February 29,1996 15.1-32 { Westirigt100Se i 9
- 15. Accident Andyses l
i l l I l l LOFT 4AP 15 TP C1995/02/20 X1995/04/04 08:50:36.05 2983823717538 Case la Minimum Feedback (80L) with Manuel Rod Control m 1.4 d ~ O z1.2- _ o -
~1- '
u - 8- -
~
w v d , 6--- o m - o -
- c. .4--
w - o -
- 2- -
u . 2 - z g
. , , , , , , , , , . . , , , i 0 200 400 600 800 1000 Time (seconds)
Figure 15.1.31 Nuclear Power (Fraction of Nominal) vs. Time for 10 Percent Step Load Increase, Manual Control and Minimum Moderator Feedback V Revision: 5 3 W85tingh0t!se 15.1 33 February 29,1996
. .-. . .. - . -. - . - . ~. . - . . . . = =
- 15. Accid:nt Analys:s e
LOFT 4AP X1995/04/04 08 50:36 05 1.5 2983823717538 TP Cit 95/02/20 Case la Minimum Feedback (BOL) with Monuct Rod Control m 2600 0 . m - a _ w
, 2400 - -
o m - m
=
- 2200 --
~
~ ..
w 2000 - - 3 - m M . e w - Q- ' ' ' ' ' 1800 ' ' ' ' ' ' ' ' ' ' O 200 400 600 800 1000 Time (seconds) Figure 15.13-2 Pressurizer Pressure (psia) vs. Time for 10 Percent Step Load Increase, Manual Control and Minimum Moderator Feedback Revision: 5 February 29, IN 15.1-34 [ W85tlDgh0US8
- 15. Accidint Analyses k(
l l l l I LOFT 4AP 1.5 TP X1995/04/04 05:50:36 05 2983823717538 C1995/02/20 ) Case 1: Minimum Feedbock (BOL) with Wonuel Rod Control _ 1800 .
~
~
l i 1 .- u 1600 -- l
.o 3 .
o
] 1400 -
o - 1200 - . O ~ n - 1000 - - e
~
5 800 - - 9: o - w . 600 ' ' l l l 0 200 400 600 800 1000 Time (seconds) l l l l I' i Figure 15.1.3-3 Pressurizer Water Volume (ft') vs. Time for 10 Percent Step Load
. , _ Increase, Manual Control and Minimum Moderator Feedback Revision: 5 W Westingh0use 15.1 35 February 29,1996
. . . _ _ -. .= .
J =__ ' l
- 15. Accidint Aralyses
'AE =
0 E LOFT 4AP 15 TP C1995/02/20 X1995/04/04 08:50:35.05 2983823717536 Case 1: Minimum Feedback (SOL) with konuel Rod Control
- 600 _
w - 05 - e 590 - - o : 580 - - m. E o -
~
- 570 --
e : O, _ O -
- 560 - -
c 550 --- O
~
o 540 ' ' ' 0 200 400 600 800 1000 Time (seconds) Figure 15.1.3-4 Core Average Temperature ('? I vs. Time for 10 Percent Step Load Increase, Manual Control and Minimum Moderator Feedback j l Revision: 5 February 29,1996 15.1 36 [ W85tingh0USS l
- 15. Accidrnt An-Jyses !
A f% LOff4AP 1.5 TP C1995/02/20 X1995/04/04 Cose la 08:50:36.05 2983823717538 Minimum Feedback (80L) with Wonuoi Rod Control 4
- 3. 5 - -
O 3-o : c:: 25- _- ca :
- 2: -
o 2- - l.5-__-
~
1 0 2d0 4d0 6h0 8d0 1000 Time (seconds) l Figure 15.1.3-5 DNB Ratio vs. Time for 10 Percent Step Load Increase, Manual Control and Minimum Moderator Feedback CT U Revision: 5 [ WBSilligt10US8 15.1-37 February 29,1996
44; iiin.-.
- 15. Accident Analyses e
LOFT 4AP 1.5 TP C1995/02/20 II995/04/04 08:50:53.87 8237165382983 Case 2: Womimum feedback (EOL) with Manuel Rod Control m 1.4 ,
~
E o - z 1. 2 - .- I-2 : u -
.8--
m 6-m - o -
- n. .4- -
O -
= 2 .
u . s . z 0- ' ' ' ' ' ' ' ' ' ' ' ' ' ' 0 200 400 'a )? 860 1000 Time (sece,ds) Figure 15.1.3-6 Nuclear Power (Fraction of Nominal) vs. Time for 10 Percent Step Load Increase, Manual Control and Maximum Moderator Feedback Revision: 5 February 29,1996 15.1-38 [ W85tingl10US0
_ ... _ - - . ~ - - - , . _ - . . -
- 15. Accidsit Analyses NE f%
l l l l l i LOFT 4AP 9 /0 X 199 Cose 2 5/04goo4x i , 08 mum:50: 53.87 8237'65382983 Feedback (EOL) with Wonvol Rod Control l m 2600 o . m - l
- m. - .
, 2400 --
o m - m o L
,/~'s ' - 2200 - -
a. i )
%J -
o
~ .
[ I 2000 - - ! a - t m m . l u -
' 1800 ' ' ' ' ' , - ' ,, , , , , ,
0 200 400 600 800 1000 Time (seconds) l i l Figure 15.1.3-7 Pressurizer Pressure (psia) vs. Time for 10 Percent Step Load increase, Manual Control and Maximum Moderator Feedback Revision: 5 T West lDgh0US8 15.1-39 February 29,1996 l t l
+N-!!Mr"-
I
- 15. Accid:nt Analyses e
l l i LOFT 4AP X1995 ! Case 2 :/04/04 08:50:53 87 823i165382983 Monimum Feedstock (COL) with Manuel Rod Control
- 1800 m
,y 16 0 0 - -
o - 5
] 1400 --
m 1200 - - m -
, 1000 - -
5 800 -
=
U - 600 * ' 0 200 400 600 800 1000 Time (seconds) l Figure 15.13-8 Pressurizer Water Volume (ft') vs. Time for 10 Percent Step Load Increase, Manual Control and Maximum Moderator Feedback Revision: 5 February 29,1996 15.1-40 [ Westingh0!JS8 l l
. . - ... . , _ . ~ . _ . . _ . _ . _.. . - . _ ... . . . . - . . . - . . . . . . . -
l l l
-n-w
- 15. Accidnt Analyses i
iO l ' V t l l I i l LOFT 4AP 15 TP X1995/04/04 08:50 33.87 8237165382983 C1995/02/20 Case 2: Moximum Feedback (EOL) with Monuoi Rod Control i m 600 ^ 6 ! - I , en - e 540 -- 1 , o : l v 580 - i c-l r E o k~ l A 570 - V e en O -
- 560 --
e - 3 i l e 550- - i L. -
)
1 o -
~
O 540- ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' 0 200 400 600 800 1000
- Time (seconds) i Figure 15.1.3-9 Core Average Tempereture (*F) vs. Time for 10 Percent Step Load Increase, Manual Control and Maximum Moderator Feedback Revision
- 5 T Westinghouse 15.1 41 February 29,1996
- m- -
l
- 15. Accid:nt Analyses i
LOFT 4AP 1.5 TP C1995/02/20 X1995/04/04 Cose 2: 08:50:53.87 8237165382983 Wenimum Feedback (EOL) with Manuel Rod Control 4 . 3. 5 - __-
~
! O 3-9 .- o _~ cr 2 . 5 - - m z a
- O 2-_-
4 *
- 1. 5 -- _
O 200 400 600 800 1000 Time (seconds) i Figure 15.1.3-10 DNB Ratio vs. Time for 10 Percent Step Load Increase, Manual Control and Maximum Moderator Feedback Revision: 5 February 29,1996 15.1-42 3 W8Stingh0US8
l l
'h l 15. Accident Analys;s l
l [ W l i LOFT 4AP 1.5 TP X1995/04/04 08:51:08.63 2983823847538 C1995/02/20 Case 3: Minimum Feedback (80L) with Automatic Rod Control m 1.4
~
E o - se 1. 2 - - o 1-u . C 8-- 1 ' (.s. u
=
6-- m - o -
- c. .4- -
w - o - _= 2-- ' u _ m x 0 0 200 400 600 800 1000 Ilme (seconds) l Cigure 15.1.3-11 Nuclear Power (Fraction of Nominal) vs. Time for 10 Percent Step Load Increase, Automatic Control and Minimum Moderator Feedback A
'\
Revision: 5 W Westinghouse 15.1 43 February 29,1996
y 15. Accident Analyses e i LOFT 4AP 1.5 X1995/04/04 08:51:08 63 2983823847538 TP C1995/02/20 Case 3: Minimum Feedback (BOL) with Automatic Rod Control m 2600 a 0 - m -
- c. -
~
, 2400 --
m - a m -
- 2200 .
e
~ .
'E 2 0 0 0 - -
= l m
m - W u - 0- ' ' ' ' 1800 ' ' ' ' ' ' ' ' ' ' ' ' O 200 400 600 800 1000 Time (seconds) E Figure 15.1.3-12 4' l Pressurizer Pressure (psia) vs. Time for 10 Percent Step Load l Increase, Automatic Control and Minimum Moderator Feedback ! l Revision: 5
l
.. I
- 15. Accid 1nt Analyses i l
n%s l LOFT 4AP 15 TP X1995/0 04 08:51:08.63 2983823847538 C1995/02/20 Case 3 M imum Feedback (80L) with Automatic Rod Control 1800 C - u 1600 - l 3
~
~
1400 - -
. 1200 - -
e - t
% y A-
, 1000 - -
e
. ~
$ 800 -
e WB e - 600 ' ' - ' , . . 1 . 0 200 460 600 800 1000 Time (seconds) 1. Figure 15.1.3-13 Pressurizer Water Volume (ft') vs. Time for 10 Percent Step Load Increase, Automatic Control and Minimum Moderator Feedback
- (
(_- Revision: 5 [ W6Stingh0086 15.1-45 February 29,1996 i
. - . . - . - , ~ . .
- 4 _m.2.4~. ...m. . . _ . . . ...m _-., _.
l
- 15. AccidInt Analyses l
1 0 ! LOFf4AP 1.5 TP C1995/02/20 x1995/04/04 08:51:08.63 2983823847538 Case 3: Minimum Feedbeck (BOL) with Automatic Rod Control m 600 ,
- u. -
cn - e 590 -- ^ Q . 580 - . e -
~
570 -- o - m 560 - - e ~ 550-- o - u - o -
^
.o 540 ' ' '
l l l 0 200 400 600 800 1000 Time (seconds) i l Figure 15.1.3-14 l 1 Core Average Temperature (*F) vs. Time for 10 Percent Step Load l Increase, Automatic Control and Minimum Moderator Feedback < l
)
Revision: 5 February 29,1996 15.1-46 3 Westilighouse
.-. .~ - . . . . -- . .. . . . . . .
ii=:;
- 15. Accident Analyses M T l
I l F) NJ
\
l l i l l \ i l l l ! l LOFf4AP 15 TP C1995/02/20 Xt995/04/04 08:51 08.63 2983823847538 Case 3 Minimum Feedback (BOL) with Automatic Rod Control 4 1 l - 3.5- -
~
l _ s o 3_ 1 o _- __ _ l l O ce 2 . 5 --
~
CD z - c 2-- -
- 1. 5 - -
\ - 9 f f f t i f f f I
- 1 f 9 ' '
1 0 200 460 600 800 1000 Time (seconds) l l l Figure 15.1.3-15 DNB Ratio vs. Time for 10 Percent Step Load Increase, Automatic Control and Minimum Moderator Feedback O-Revision: 5 [ W85tingh0US8 15.1-47 February 29,1996 4
o-- ,
- 15. AccidInt An-lyses e
1 1 I LOFT 4AP 1.5 TP C1995/02/20 X1995/04/04 08:51: 22 86 8238845382983 Case 4 Mozimum Feedback (EOL) with Automatic Red Control m 1.4 - s -
- E12-{
O 1-2 . u _ o 8--- m - u 6- - m -
= -
O -
- n. .4- -
W*
-- . . - . . . . ._ ~
~~~~~~'~~~'
2-- _ u _
> ~
z o l l l l 0 200 400 600 800 1000 Time (seconds) Figure 15.1.3-16 Nuclear Power (Fraction of Nominal) vs. Time for 10 Percent Step Load Increase, Automatic Control and Maximum Moderator Feedback Revision: 5 February 29,1996 15.1-48 3 W85tingh0USS
m A - ea m m++ - =4 s am - I l
. :_-- i
- 15. Accident Analyses o\
\
J l 1 l j 4 LOFT 4AP 1.5 TP C1995/02/20 X1995/04/04 08:51: 22 86 8238845382983 Coss 4: Monimum Feedback (EOL) with Automatic Rod Control ) _ 2600 o - ! m -
- a -
v m 2400 - - o - m e
- 2200 - -
i lD o. w ~ e
~ -
~~
. 2000 - -
1 o _ m N -- ~ ... __,.O - e:r - -- - -- . . . . . . . . _ w - --.- ~ _ _ . . ,, u- 1800 ' l l I^ """" >
- - ~~_
o 200 400 600 800 1000 Time (seconds) l l l Figure 15.1.3 17 Pressurizer Pressure (psia) vs. Time for 10 Percent Step Load Increase, Automatic Control and Maximum Moderator Feedback Revision: 5 3 Westingh0use 15.1 49 February 29,1996 l
- 15. AccidInt Analyses e
LOFT 4AP 1.5 TP CIS95/02/20 X1995/ 04 08:51: 22 86 8238545382983 Case 4: imum Feedback (EOL) with Automatic Rod Control 1800
~
2 - w a 1600 - $
.o S I
] 1400--;
o - m 1200 -- sn -
, 1000 -- -
o
~
.2 ~
~
800 - - E - 600 ' ' ' ' ' > . , , , , , , , 0 200 400 6$0 800 1000 .. . - ~ ~ . . , . . . . Time (seconds)
-~
Figure 15.1.3-18 Pressurizer Water Volume (ft') vs. Time for 10 Percent Step Load Increase, Automatic Control and Maximum Moderator Feedback t Revision: 5 February 29,1996 15.1-50 VP Westinghouse
- 15. Accident An~ lyses
. (~x
\_
LOFT 4AP 1.5 TP X1995/04/04 08: bis 22 86 823 845382983 CIS95/02/20 Case 4: Mosimum Feedback ([0L with Automatic Rod Control m 600 L :
~
cm - e 590 - - m -
~
580 - - Q-E -t o - 570 -1 l[v )
- 7 :
I
- 560 --
o - o 550 - - u - o - C.) ' ' ' ' ' ' 540 ' l l 0 200 400 600 800 1000 Time (seconds) Figure 15.1.3-19 Core Average Temperature ( F) vs. Time for 10 Percent Step Load Increase, Automatic Control and Maximum Moderator Feedback b s.) Revision: 5 [ W85tingh0US9 15.1-51 February 29,1996 l l
- 15. Accidsnt Analyses i
h O l LOFT 4AP 1.5 TP Ct995/02/20 Il995/04/04 08:51:22.86 8238845382983 Case 4s Nesimum Feedback (EOL) with Automatic Rod Control 4 l l
- 3. 5 -- -
~
o 3
~
~
, o _ ! cc 2 . 5 -- l O $ z - a 2-_-
- 1. 5 -. -
~
j , , . . , , , , , , . , , , , , , , 0 200 400 600 800 1000 Time (seconds) Figure 15.1.3-20 DNB Ratio vs. Time for 10 Percent Step Load Increase, Automatic Control and Maximum Moderator Feedback
- Revision: 5 February 29,1996 15.1-52 3 Westingt10Use
i i
! l ! 15. Accident Anilyses I
d iO I 1 i s i 1.02 h i ! 1.01 - t ii 1 lt: 1.00 -
~
t O - 0.98 . 0.97
' i -
400 420 440 460 480 500 520 540 Core Average Temperature (T) Eigure 15.1.4-1 K,, vs. Core Average Temperature Steam Line Break Events Revision: 5 [ WOStingh00S6 15.1-53 February 29,1996
E%Cs *
- 15. Accident Analyses I rm-0 0.50 .
^ 0.45 h 2 . .
m - 1 0.40 ? l o . : 7 . 1 uo 035' : 8 030 h 0.25 h l : 45 0.20 F r S 0.15 r I0 10 r 3 : Z 0.05 F 0.00 O 100 200 300 400 500 Tune (s) i Figure 15.1.4-2 Nuclear Power Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 February 29, im 15.1-54 y Westifigt10US6 e
5 . . . . ! 15. Accident Anilyses 4
- O e
1 l l 0.50 . 1
\
0.45 r i 0.40 r o - 2; : y 035 o 0.30 .- ! c . 0.25 l k 0.20 - O .f 0.15 i 0.10 - U 0.05 .'- . A .. .. .t............ 0 100 200 300 400 500 Tune (s) Figure 15.1.4-3 Core Heat Flux Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 [ W85tingh0US8 15.1-55 February 29,1996
.~ , . . _ . . . . . .- _.. . ~ .. .~. _ ... -. --........ .. .. .. -.- -.. . -..
I
- 15. Accident Analyses e i I
i 3000 g s - I ~
.: a 2000 3
5 l "
~
g a n 0 1000 - T l : u . l . 61 4 A
- 0 -
X *s A% I t! 3 ' f [-1000 $ g5 . ~ ' f) -
.. . l . 1 i. t 0 100 200 300 400 500 Time (s) i l
l Figure 15.1.4-4 Reactivity Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 February 29, IM 15.1-56 W Westinghouse i
ml
- 15. Accident Andyses n-O a.
550 1 C :
" 500
[ - 450 r* E :
# 400 :
o : O j o 8 300 :- O : 250 i, , . , 0 100 200 300 400 500 Tune (s) Figure 15.1.4-5 Core Average Temperature Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 T Wesdnghouse 15.1-57 February 29,1996
- 15. Accid:nt Analyses l e
g - Fadied Intact G 550 ww
, ====== ...
8 500 7 i : ........' p 450 : - so
.s
. ~.,**s'..
C -
= 400 .- ...,
u Q
~
eo . S 350 - a t) 8 - a: 300
~
g _- - i f , , , 0 100 200 300 400 500 Tune (s) Figure 15.1.4-6 Reactor Vessel Inlet Temperature Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 February 29,1996 15.1 58 [ Westirigh00S8
i l ver 1
- 15. Accid:nt An . lyses 21
, 3. . ;.O 4
s! 22 - J k
- i. 2000 -
I
- m -
t e . 1 *a ch - !
- 1500 - I l
1 2 n - m l M - b
; $ 1000 O h i
E : i 500 - i - l 4
- 0 -
0 100 200 300 400 500 1 1 Tune (s) d 4 i i j Figure 15.1.4-7 RCS Pressure Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve i:O, Revision: 5
- T Westinghouse 15.1-59 February 29,1996 4
i 1
d Wh$ l 15. Accident Analyses 1 O 1 4 l ! 1200 j p 1000 ~ !- 5 - l 8 - j - B 800 -
$w -
i S 8 )0 - - v hu - j' O - i
.R 14* :
i -
~
I r J M.200 - i 0 i 1 0 100 200 300 400 300 ! he(s) i i Figure 15.1.4-8 Pressurizer Water Volume Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 February 29,1996 15.1-60 W Westinghouse
. ~ - . - . . - . . . . . - . - . . _ . . . . . . . . . - .. ~ . . . - - - . . . . . . . . . . . - - . . - . ~ . . . . . - . - - . - - . - . . - - .
4 i ::_:. i 15. Accident Analyses - l f I
- O 1 v i
i i I ^ l 1 i i } 1.4 - p 1.2 m - Y j mE 1.0 . ~ 8
.a 0.8 h U _
j v 0.6 O ' O.4 h
~
O 0.2 h
' ' ' ' ' =
0.0 0 100 200 300 400 500 Time (s) Figure 15.1.4-9 Core Flow Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 Y W96tiligh0US8 15.1-61 February 29,1996
__ .. _ _ _ . _ _ _ _ . - . _ _ _ _ . . _ . _ . . - _ _ _ _ _ _ . . . ~ . . . . _ _ _ . . - _ _ . _ . _ . _ . _ _ _ _ . . _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . . _ _ .
- 15. Accidtnt Analyses e
2.00 1 : 2 1.75 7 3 . Z : g 1.50 1.25 m :
" 1.00 -
$ i e
g 0.75 h m g 0.50 7 5 25 1 ) E ' ' ' 0.00 O 100 200 300 400 500 Time (s) Figure 15.1.4 10 Feedwater Flow Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 February 29,1996 15.1-62 T Westirghouse
. _ _ _ _ . . ~ . _ _ _ _ . . - - . _ . _ . _ . _ _ _ . . . . . . _ _ _ _ . . _ _ _ . _ _ . _ , . . _ . . _ . . . _ _ . . _ . _ _ . _ . _ _ . . . _ . _ . _ _ . _ - . _ _ . _ . . _ . _ . _
f
- 4.4_.
- 15. Accident Analyses
- l 1
iO i i I l 1 i i k, ' I i 1 i ! l i 500 l
~
I 400 - (
^ .
,,300 f M
i -
~
O u g200 : 100 0 O 100 200 300 400 500 - Time (s) Figure 15.1.4-11 Core Boron Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 [ W8Stifigh00S8 15.1-63 February 29,1996
ii-- L
- 15. Accident .Andyses 0
l 1200 I
- Fauhed Intact 1000 - imp LOOP Q . .
*E 800 -
S E
= -
E 600 - g _
~
m
]*m -
O
^ ' ' '
O O 100 200 300 400 500 Tune (s) Figure 15.1.4-12 Steam Pressure Transient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5 February 29,1M 15.1-64 [ W85tiligh0USB
. - . ~ . . . ~ _ . - - . . _ . . - . . . - - . - . . . . . . - . . - . . . . - . . . ~ . _ . . . . . _ - - . - -
l
- 15. Accident Analyses !
o i l 250 i FaultedIntact Lap W l i 200 - -.. n -. - b.
, 150 -
C l 5 o E ; 100 - i i E O .e. to i I 50 - : I I
.i s. . . i 0 100 200 300 400 500 Time (s) l Figure 15.1.4-13 Steam Flow TL.aient Inadvertent Opening of a Steam Generator Relief or Safety Valve Revision: 5
[ W85tingh0USS 15.1-65 February 29,1996
FTM"J
- 15. Accident Analyses 0
0.50 .
^ 0.45 F
*1 0.40
~
o : Z 0.35 r o : 8 0.30 - 0.25 h i v : b 0.20 h S 0.15 :
- O g
- 0.10 [-
8 : Z 0.05
" .~.
.. . i. . .. ..m .i. . .
0.00 0 100 200 300 400 500 600 700 Tune (s) Figure 15.1.5-1 Nuclear Power Transient Steam System Piping Failure O Revision: 5 February 29,1996 15.1-65 3 W85tingh0USS
1 i
.T-...' -
- 15. Accident Analyses l [
O l I 1 l l l 1 0.50 .
^ 0.45 i 1 l
'E 0.40 h o -
Z w 035 : o - c -
=g o 030 :
O E *" ? g 0.15 2 : 2 0 10 r o - 0 0.05
^
0.00 .t. .v. .,. 0 100 200 300 400 500 600 700 Time (s) Figure 15.1.5-2 Core Heat Flux Transient Steam System Piping Failure O Revision: 5 g g gse 15.1-67 February 29,1996
._. .. . . . . . -.- . _ ~ . . . _ . _ . _ . - . - ~ . - . . . - - - . - . . - _ . - . . _ . . - . . . - - - . . - - . . . _ . . . - . .
t
- 15. Accident Analyses '
sm-F O 3000 - 2000 h p 1000 - 3 :
? 0 ?
3 ' 8 W -1000
-2000 h.
~. .t. .t. .t. .t. .t. .t. i 0 100 200 300 400 500 600 700 l Time (s)
Figure 15.1.5 3 Reacthity Transient Steam System Piping Failure O Revision: 5 February 29,1996 15.1-68 [ WBStiflghouse
1
- 15. Accident Analyses E if )
l O l O. . 550 - (C y 5m : a : E : ES 400 o - 32 h O a ; e o sw : - U . g .f. .t. .t. .t. .t. .i. . 0 100 200 300 400 500 600 700 Time (s) Figure 15.1.54 Core Average Temperature Transient Steam System Piping Failure Revision: 5 T Westinghouse 15.1-69 February 29,1996
- .... - . _ ~ -...,_.. -. . -- -..- .~ .-- -.... -...- - - ... - - -..-.-.. -.._ - _ - _ _ - _ l t'
- 15. Accident Analyses nn-l l
I 600 1 g 550 Faulted intact ' L LMP lo0P g 5@ ' E .
' S.,
450 - G -
"E a 350 } .
s, g
~
.& S .: 's y .
a 250 ? ~.... - ....
, ==.,
.i. .. .i. .t. .i. .i. .i,.
0 100 200 300 400 500 600 700 Tune (s) Figure 15.1.5-5 Reactor Vessel Inlet Temperature Transient Steam System Piping Failure Revision: 5 February 29,1996 15.1-70 [ W8Stiligh00S8
- 15. Accident Antlyses .
o 2500 > 2000 m i v 1500 - E a , t 1000 - O g M~
, .. .t. .i. .t. .r. .i.
0 100 200 300 400 500 600 700 Time (s) Figure 15.1.5-6 Reactor Coolant System Pressure Transient Steam System Piping Failure O Revision: 5 Y W85tiligh00S8 15.1-71 February 29,1996
- 15. Accidsnt Analyses e
sm 700 - k vg, aw 5a 5 S 400 h 3a e J! '" 100
"^^' ' '
0 O 100 200 300 400 500 600 700 Tune (s) Figure 15.1.5-7 Pressurizer Water Volume Transient Steam System Piping Failure Revision: 5 : February 29,1996 15.1-72 W Westingh00S8
- 15. Accident Analyses
] i' o
t 1.4 . 1.2 - n - 3
~
] 1.0 f N -
.f0.8 0
h -: v0 - O j.6 : 0 U 84 h 0.2 L
,,_ ,3, .. .d .e. . w..i.
0.0 -- - 0 100 200 300 400 2 @ M Time (s) Figure 15.1. J Core Flow Transient Steam System Piping F-; are O Revision: 5 W US8 15.1-73 February 29,1996 i
F
- 15. Accident Analyses e i l
I m 2.00 - 1 1
*2 1.75 1 3 .
Z :
'tg 1.50 :
~E 1.25 h j
N 1 l
- 1.00 -
I
.$ 1 g 0.75 o
~
hi lC 0.50 :
~
$ 25 1 l l
S I
- "1. ... ... .. .. . .. ...
. 0.00 l 0 100 200 300 400 500 600 700 Time (s) l I l l i i Figure 15.1.5-9 Feedwater Flow Transient Steam System Piping Failure 1 0!l l Revision: 5 February 29,1996 15.1-74 [ W95tingh00S8 l i_______________________.________
saw-
- 15. Accident Analyses O
500 .
~
i 1 3= i - 7 CD O l# o - 100
~
1
~
.t. .t. .t. .t. .t. ..
0 100 200 300 400 500 600 700 Time (s) Figure 15.1.5-10 Core Boron Transient Steam System Piping Failure Revision: 5 OT Westinghouse 15.1-75 February 29,1996
... . . . _ - - . . - . - - . - - - . - . - _ . - . - - . - - . . - . _ - - _ . - . . _ - - . - . ~ . . - . . _ - . _ - . . . . . _ . . .
![5 (( . .
i
- 15. Accident Analyses O.
1200 Paulsed intact 1000 - Lwp bp 2 y800-e o E 600 " . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , M '. e n <= : q M .
' ' " ' ~
0 0 100 200 300 400 500 600 700 Time (s) Figure 15.1.5-11 Steam Pressure Transient Steam System Piping Failure O~ Revision: 5 February 29,1996 15.1-76 [ WBStingt10US8
. - . _ . - - - _ . . . . . . _ . . - . - . - . ~ . _ - - - . . _ . . - . . - - . _ - - . - - - _ . . . - - . . . . -
;mussm:
- 15. Accident Analyses lO l
3500
.~ Fadted Intact 3000 - Loop Ioop T 2500 -
a v 2000 - o E 1500 - un 1000 j I M I
.e........i.........!....... .f. ....... ......... ...
0 100 200 300 400 500 600 700 Tune (s) Figure 15.1.5-12 Steam Flow Transient Steam System Piping Failure O Revision: 5 l l W M W65tingh00S8 15.1 77 February 29,1996
1 j~
'+'
- 15. Accident Analyces n-O M. .
180 m - g 160 - 140 : C< ik 120 - o - E a 100 . o -
*D ~
O '
*e
~
O\ 60 .- l u 40 F 20 ~ - 0' '- '- '- '- - >- 0 100 200- 300 400 l 500 600 700 i Tune (s) Figure 15.1.5-13 l Core Makeup Tank Injection Flow Steam System Piping Failure Revision: 5 i February 29,1996 15.1-78 T Westinghouse i I ..
. . - - -.- . . . . . . . ..-.. ~ . . ... - .- - ..~ . . -.. . - - . -. - .. .-.. _ ... -... . ... . . -. - .. .._.
i mm+ - _ 15. Accident An: lyses ' 1
\
O . n50 - 1 M . O
~
4 v - 1
- 3 g1750 4
3 - l w 1500 s - g1250 O : i 1000 - yg ..t. ..t. .t. .s. .g, .,_ 0 100 200 300 400 500 600 700 Time (s) - Figure 15.1.5-14 1 Core Makeup Tank Water Volume Steam System Piping Failure 4 Revision: 5 [ W8Stingh00S8 15.1-79 February 29,1996
M... ' ili.M..i. t::
- 15. Accident Analyses
, i, 4 O t 16 - c - s 14 -- 2 :
- 1. 2 -
.C I-
~. .
= 8- -
$. 2
- 6- :
- a. 4- - -
W , i
.= 2-- -
1 o
= -
1
* ~
0 ' ' . , , , , , ; O 200 460 600 8d0 1000 TIME (S) l
)
i I l I l 1 l l Figure 15.1.6-1 i i Nuclear Power Transient Inadvertent Operation of the PRHR l Revision: 5. February 29,1996 15.1-80 W
==
W8Stingh00Se I j
- 15. Accidst Analyses
(\J~) _ 1.6 -
- 1. 4 -
g E :
- 1. 2 - -
- 1-:
v
- : 8-_ -
( D 3 r ~
.6- -
w
.4- -
2-7., - 0 l l l l 0 200 400 600 800 1000 TIME (S) Figure 15.1.6-2 Core Heat Flux Transient Inadvertent Operation of the PRHR U,n Revision: 5 W Westinghouse 15.1-81 February 29,1996
ii-u
- 15. Accident A:trJyses e
_ .IE-02 - x ~ l o
- 8 E-0 3 --
m - c -
- 6 E-0 3 --'
T 4 E-0 3 - f 2 E O ce o
. W 0-'.
k o ~
.2E-03 l l O 200 400 600 800 1000 TIME (S)
Figure 15.1.6-3 Reactivity Transient Inadvertent Operation of the PRHR Revision: 5 February 29,1996 15.1-82 3 Westingh0US0 i
R I
- 15. Accident An.Jyses f;l . l O !
1 G 1 l l m 600 . i
~
- 2. _
550- _-
- 1
! 500 -- -
450-- i 3 _
= 400--_
- ~
.: 350 ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '
l l 0 200 400 600 800 1000 TIME (S) Figure 15.1.6-4 [ Core Average Moderator Temperature Transient ' -O Int.dvertent Operation of the PRHR , U: Revision: 5 T Westinghouse 15.1-83 February 29,1996
1
,,m- +~
- 15. Accident Andyses 1
1 0\ l l l l 1 l l m 2500 ~ 0 m F a ~
- 2000 -- ,
l c - w a - a i 1 m 1500 -- l c _ w 1 ~ w 1000 -- w . -
=
e 500 -- . ! m . ! c . i w _ ' ' 0 0 200 400 600 800 1000 TIME (S) i Figure 15.1.6-5 Pressurizer Pressure Transient Inadvertent Operation of the PRHR O Revision:' 5 February ' ,1996 15.1-84 [ W85tiligh00S8
- 15. Accident Analyses
-f~
x w
- 1.2 E
i 1-_r
~
8- - 6- - ' fQ 5 4-V ; E 2- f 0 0 0 2A0 4d0 60 8d0 10'00 TIME (S) Figure 15.1.6-6 RCS Flow Transient Inadvertent Operation of the PPdfR
/
t
\s Revisfor: 5 3 W8Stiflgh00S8 15.1-85 Febniary 29,1996
..........g l"I
=
g;
- 15. Accid 2nt Analysss
.i ..
O P 3.5 3 : - 2.5-j c:: 2-
.In :
z . c 1. 5 - : 1- b 5 -3 _~ O ~ ' ' ' ' ' ' ' ' ' ' ' ! 0 280 i . , 400 600 800 1000 TIME (S) 1 Figure 15.1.6-7 ' DNBR Transient Inadvertent Operation of the PRIIR ! O 1 eb in 29,1996 15.1-86 [ Westifigt100SB
- 15. Accident Analyses
!(% 'Q) l l l l 1 I i l
,, 12 o -
c .
'so 10 -- -
= _
~
% 3 -
I c o _- 1 u 6-- o l 7~. I l*() m 4-- o i 2 . ce
- c Ctl:
0 l l l o 200 400 600 800 1000 TIME (S) l l 1 l t Figure 15.1.6-8 PRHR Flow Transient Inadvertent Operation of the PTJIR fg
'%.]
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- 15. Accident Analyses -
(J D i l n 1200 1 3 m 7 o 1000- - a
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i l 6 Revision: 5 T Westingh00S8 15.1-89 February 29,1996
I l-
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'15. Accident Analyses i
15.2 Decrease in Heat Removal by the Secondary System j- ' A number of transients and accidents are postulated which could result in a reduction of the { capacity of the secondary system to remove heat generated in the reactor coolant system. ! Detailed analyses are presented in this section for the following events which are identified I l . as more limiting than the others: Steam pressure regulator malfunction or failure that results in decreasing steam flow l
- l
' Loss of external electrical load i Turbine trip I Inadvertent closure of main steam isolation valves
. Loss of condenser vacuum and other events resulting in turbine trip i Loss of ac power to the station auxiliaries Loss of normal feedwater flow t Feedwater system pipe break i t
The above items are considered to be Condition II events, with the exception of a feedwater l j system pipe break, which is considered to be a Condition IV event. '
- The radiological consequences of the accidents in this section are bounded by the radiological
! consequences of a main steam line break (See Subsection 15.1.5). I l 15.2.1 Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steamflow p There are no steam pressure regulators in the AP600 whose failure or malfunction cause a steamflow transient. 15.2.2 Loss of External Electrical Load 15.2.2.1 Identification of Causes and Accident Description L l A major load loss on the plant can result from loss of electrical load due to some electrical system disturbance. Offsite ac power remains available *o operate plant components, such as the reactor coolant pumps; as a result, the standby onsite diesel generators do not function for this event. Following the loss of generator load, an immediate fast closure of the turbine L control valves occurs. The automatic turbine bypass system accommodates the excess steam ; L generation. Reactor coolant temperatures and pressure do not significantly increase if the turbine bypass system and pressurizer pressure control system are functioning properly. If the condenser is not available, the excess steam generation is relieved to the atmosphere. l Additionally, main feedwater flow is lost if the condenser is not available. For this situation, feedwater flow is maintained by the startup feedwater system. l For a loss of electrical load without subsequent turbine trip, no direct reactor trip signal is generated, and the plant is expected to trip from the reactor protection system if a safety limit is approached. A continued steam load of approximately five percent exists after total loss y of extemal electrical load, because of the steam demand of plant auxiliaries.
- (
1 Revision: 5 L Y WBEllRghouse 15.2-1 February 29,1996
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- 15. Accident Analyses 1
In the event that a safety limit is approached, protection is provided by high pressurizer O pressure, high pressurizer water level, and overtemperature AT trip. Voltage and frequency relays associated with the reactor coolant pump provide no additional safety function for this I event. Following a complete loss of external electrical load, the maximum turbine overspeed I is not expected to damage the voltage and frequency sensors. Any degradation in their performance is ascertained at that time. Any increased frequency to the reactor coolant pump i motors results in a slightly increased flow rate and subsequent additional margin to safety limits. For postulated loss of load and subsequent turbine-generator Sverspeed, the i overfrequency condition is not seen by the reactor protection system egi.pment, or other safety-related loads. Safety-related loads and the reactor protection sysNm equipment are supplied from the 120-volt ac instrument power supply system, which in tum, is supplied from the inverters. The inverters are supplied from a de bus energized from batteries or by a regulated ac voltage. In the event that the steam dump valves fail to open following a large loss of load, the steam generator safety valves may lift and the reactor may be tripped by the high pressurizer pressure signal, the high pressurizer water level signal or the overtemperature AT signal. The steam generator shell side pressure and reactor coolant temperature increase rapidly. The pressurizer safety valves and sterun generator safety valves are, however, sized to protect the reactor coolant system (RCS) and steam generator against overpressure for load losses, without assuming the operation of the turbine bypass system, pressurizer spray, or automatic RCCA control. The steam generator safety valve capacity is sized to remove the steam flow at the guaranteed nuclear steam supply system thermal rating from the steam generator, without exceeding 110 percent of the steam system design pressure. The pressurizer safety valve capacity is sized ! to accommodate a complete loss of heat sink with the plant initially operating at the maximum { turbine load along with operation of the steam generator safety valves. The pressurizer safety l valves are then able to relieve sufficient steam to maintain the reactor coolant system pressure within 110 percent of the reactor coolant system design pressure. l A more complete discussion of overpressure protection can be found in Reference 1. 1 A loss of external load is classified as a Condition II event, fault of moderate frequency. A loss-of-extemal-load event results in a plant transient that is bounded by the turbine trip ! event analyzed in Subsection 15.23. Therefore, a detailed transient analysis is not presented ' for the loss-of-extemal-load event. l l The primary side transient is caused by a decrease in heat transfer capability from primary to i secondary due to a rapid termination of steam flow to the turbine, accompanied by an l automatic reduction of feedwater flow. (Should feedwater flow not be reduced, a larger heat l sink is available and the transient is less severe.) Reduction of steam flow to the turbine I following a loss of external load occurs due to automatic fast closure of the turbine control valves. Following a turbine trip event, termination of steam flow occurs via turbine stop valve closure, which occurs in approximately 0.15 seconds. The transient in primary pressure, Revision: 5 February 29,1996 15.2-2 W Westinghouse l l
=mm
- 15. Accidelt Analyses 1
l temperature, and water volume is less severe for the loss of extemal load than for the turbine trip due to a slightly slower loss of heat transfer capability, j i The protection available to mitigate the consequences of a loss of extemal load is the same as that for a turbine trip, as listed in Table 15.0-6. 15.2.2.2 Analysis of Effects and Consequences l Refer to Subsection 15.2.3.2 for the method used to analyze the limiting transient (turbine trip) I in this grouping of events. The results of the turbine trip event analysis bound those expected l for the loss of external load, as discussed in Subsection 15.2.2.1. Plant systems and equipment which may be required to function to mitigate the effects of a complete loss of load are discussed in subsection 15.0.8 and listed in Table 15.0-6. i The reactor protection system may be required to terminate core heat input and to prevent departure from nucleate boiling. Depending on the magnitude of the load loss, pressurizer safety valves and/or steam generator safety valves may open to maintain system pressures l below allowable limits. No single active failure prevents operation of any system required i to function. Normal reactor control systems and engineered safety systems are not required to function. The passive residual heat removal system may be automatically actuated following a loss of main feedwater. This further mitigates the effects of the transient. () 15.2.2.3 Conclusions I Based on results obtained for the turbine trip event and considerations described in l Subsection 15.2.2.1, the applicable SRP Section 15.2.1 evaluation criteria for a { loss-of-external-load event are met. (See Sub-section 15.2.3) l l 15.2.3 Turbine Trip 15.2.3.1 Identification of Causes and Accident Description ! l The turbine stop valves close rapidly (about 0.3 seconds) on loss of trip fluid pressure , actuated by one of a number of possible turbine trip signals. Turbine trip initiation signals l include: ' Generator trip Low condenser vacuum Loss of lubricating oil Turbine thrust bearing failure
- Turbine overspeed
- Manual trip Reactor trip Ci G
Revision: 5 T Westinghouse 15.2 3 February 29,1996
i
- 15. Accid:nt Analyses 1
Upon initiation of stop valve closure, steam flow to the turbine stops abmptly. Sensors on ' the stop valves detect the turbine trip and initiate turbine bypass. The loss of steam flow results in an almost immediate rise in secondary system temperature and pressure, with a resultant primary system transient described in Subsection 15.2.2.1 for the loss of external load event. A slightly more severe transient occurs for the turbine trip event due en the more rapid loss of steam flow caused by the more rapid valve closure. The automatic turbine bypass system accommodates up to 40 percent of rated steam flow. Rear. tor coolant temperatures and pressure do not increase significantly if the turbine bypass system and pressurizer pressure control system are functioning properly. If the condenser is not available, the excess steam generation is relieved to the atmosphere, and main feedwater flow is lost. For this situation, feedwater flow is maintained by the startup feedwater system to provide adequate residual and decay heat removal capability. Should the turbine bypass system fail to operate, the steam generator safety valves may lift to provide pressure control. See subsection 15.2.2.1 for a fmther discussion of the transient. A turbine trip is classified as a Condition II event, fault of moderate frequency. A turbine trip is more limiting than loss of external load, loss of condenser vacuum, and other events which result in a turbine trip. As such, this event is analyzed in detail. Results and discussion of the analysis are presented in subsection 15.23.2. 15.2.3.2 Analysis of Effects and Consequences 15.2.3.2.1 Method of Analysis In this analysis, the behavior of the unit is evaluated for a complete loss of steam load from 100 percent of full power, without rapid power reduction, primarily to show the adequacy of the pressure-relieving devices, and also to demonstrate core protection margins. The turbine is assumed to trip without actuating the rapid power reduction system. This assumption delays reactor trip until conditions in the reactor coolant system result in a trip due to other signals. Thus, the analysis assumes a worst transient. In addition, no credit is taken for the turbine bypass system. Main feedwater flow is terminated at the time of turbine trip, with no credit taken for startup feedwater or the passive residual heat removal system (except for long-term recovery) to mitigate the consequences of the transient. ! The turbine trip transients are analyzed by employing the computer program LOFTRAN (Reference 2). The program simulates the neutron kinetics, reactor coolant system, pressurizer, pressurizer safety valves, pressurizer spray, steam generator and steam generator j safety valves. The program computes pertinent plant variables, including temperatures, pressures and power level. The LOFTRAN code is modified to incorporate the specific passive safeguards system features for the AP600. A description of these modifications are I presented in WCAP-14601 (Reference 6). O Revision: 5 February 29,1996 15.2-4 3 Westinghouse
- 15. Accident Analyses
,_x I ) I v' l The major assumptions used in the analysis are summarized below:
Initial Operating Conditions The accident is analyzed using the revised thermal design procedure. Initial core power, i reactor coolant temperature, and pressure are assumed to be at their nominal values consistent ' with steady-state full power operation. Uncertainties in initial conditions are included in the departure from nucleate boiling ratio (DNBR) limit as described in WCAP-11397 (Reference 3). Reactivity Coefficients Two cases are analyzed: Minimum Reactivity Feedback - A least negative moderator temperature coefficient and a least negative Doppler-only power coefficient are assumed. (See Figure 15.0.4-1) Maximum Reactivity Feedback - A conservatively large negative moderator temperature coefficient and a most negative Doppler-only power coefficient are assumed. (See Figure 15.0.4-1) Reactor Control tO V From the standpoint of the maximum pressures attained, it is conservative to assume that the reactor is in manual control. If the reactor is in automatic control, the control rod banks move prior to trip and reduce the severity of the transient. Steam Release No credit is taken for the operation of the turbine bypass system or steam generator power-operated relief valves. The steam generator pressure rises to the safety valve setpoint where steam release through safety valves limits secondary steam pressure at the setpoint value. Pressurizer Spray Two cases for both the minimum and maximum reactivity feedback cases are analyzed.
=
Full credit is taken for the effect of pressurizer spray in reducing or limiting the coolant i pressure. Safety valves are also available, with maximum capacity. No credit is taken for the effect of pressurizer spray in reducing or limiting the coolant I pressure. Safety valves are operable, with maximum capacity. I t n
' usl>
Revision: 5 l W
==
Westinghouse 15.2-5 February 29,1996 l I
n=u f 15. Accid:nt Analyses
]
Feedwater Flow O Main feedwater flow to the steam generators is assumed to be lost at the time of turbine trip. I No credit is taken for startup feedwater flow or the passive residual heat removal, heat I exchanger, since a stabilized plant condition is reached before startup feedwater initiation or passive residual heat heat exchanger removal is normally assumed to occur. The startup feedwater flow or passive residual heat removal removes core decay heat following plant stabilization. Reactor Trip i Reactor trip is actuated by the first reactor trip setpoint reached, with no credit taken for the rapid power reduction on the turbine trip. Trip signals are expected due to high pressurizer pressure, overtemperature AT, high pressurizer water level, and low steam generator water level. Plant characteristics and initial conditions are further discussed in Subsection 15.0.3. Plant systems and equipment which may be required to function to mitigate the effects of a turbine trip event are discussed in Subsection 15.0.8 and listed in Table 15.0-6. I The protection and safety monitoring system may be required to function following a turbine trip. Pressurizer safety valves and/or steam generator safety valves may be required to open to maintain system pressures below allowable limits. No single active failure prevents operation of any system required to function. Normal reactor coolant system and engineered safety systems are not required to function. However, cases are analyzed both with and without the operation of pressurizer spray to determine the worst case for presentation. 15.2.3.2.2 Results The transient responses for a turbine trip from 100 percent of full-power operation are shown for four cases: two cases for minimum reactivity feedback and two cases for maximum reactivity feedback (Figures 15.2.3-1 through 15.2.3-24). The calculated sequence of events for the accident is shown in Table 15.2-1. Figures 15.2.3-1 through 15.23-6 show the transient responses for the total loss of steam load with minimum reactivity feedback, assuming full credit for the pressurizer spray and pressurizer safety valves. No credit is taken for the steam bypa.s. The reactor is tapped by the high pressurizer pressure trip channel. The minimum DNBR remains well above the safety analysis limit values. The steam generator safety valves limit the secondary steam conditions to saturation at the safety valve setpoint. Figures 15.2.3-7 through 15.2.3-12 show the responses for the total loss of steam load with maximum reactivity feedback. All other plant parameters are the same as the above. The I minimum depasture from nucleate boiling ratio remains well above the safety analysis limit i values. The pressurizer safety valves and steam generator safety valves prevent Revision: 5 February 29,1996 15.2-6 T Westinghouse
4Whaber W
- 15. Accide t Analyses i s V
overpressurization in primary and secondary systems. The rise in the reactor coolant system average temperature causes a large reduction in neutron flux due to reactivity feedback effects, resulting in a decrease in pressurizer pressure. The turbine trip accident is also studied assuming the plant to be initially operating at 100 percent of full pwer with no credit taken for the pressurizer spray or the turbine bypass system. The reactor is tripped on the high pressurizer pressure signal. Figures 15.2.3-13 through 15.2.3-18 show the transients with minimum reactivity feedback. The neutron flax remains essentially constut at 100 percent of full power until the reactor is tripped. The I minimum departure from nucleate boiling ratio remains well above the safety analysis limit I values. In this case, the pressurizer safety valves are actuated and maintain reactor coolant system pressure below 110 percent of the design value. Figures 15.2.3-19 through 15.2.3-24 show the transients with maximum reactivity feedback, I with the other assumptions being the same as in the preceding case. Again, the minimum I departure from nucleate boiling ratio remains well above the safety analysis limit values and the pressurizer safety valves are actuated to limit primary pressure. Reference 1 presents additional results of analysis for a complete loss of heat sink, including loss of main feedwater. This analysis shows the overpressure protection that is afforded by the pressurizer and steam generator safety valves. O V 15.2.3.3 Conclusions I Results of the analyses, including those in Reference 1, show that the plant design is such that a turbine trip presents no challenge to the integrity of the reactor coolant system or the main steam system. Pressure-relieving devices incorporated in the two systems are adequate to limit the maximum pressures to within the design limits. The anulyses show that the departure from nucleate boiling ratio does not decrease below the safety analysis limit at any time during the transient. Thus, the departure from nucleate boiling design basis, as described in Section 4.4, is met. 15.2.4 Inadvertent Closure of Main Steam Isolation Valves Inadvertent closure of the main steam isolation valves results in a turbine trip with no credn taken for the turbine bypass system. Turbine trips are discussed in Subsection 15.2.3. 15.2.5 Loss of Condenser Vacuum and Other Events Resulting in Turbine Trip Loss of condenser vacuum is one of the events that can cause a turbine trip. Turbine trip-1 initiating events are described in subsection 15.2.3. A loss of condenser vacuum p> events *.he use of steam dump to the condenser; however, since steam dump is assumed to be unavailable in the turbine trip analysis, no additional adverse effects result if the turbine trip is caused by loss of condenser vacuum. Therefore, the analysis results and conclusions contained in
-7 Subsection 15.2.3 apply to the loss of the condenser vacuum. In addition, analyses for the
)
J Revision: 5 3 W8Stiflgh00S8 15.2-7 February 29,1996 s
- 15. Accid:nt Aralyses I
other possible causes of a turbine trip, listed in subsection 15.2.3.1, are covered by e Subsection 15.2.3. Possible overfrequency effects due to a turbine overspeed condition are discussed in Subsection 15.2.2.1 and are not a concern for this type of event. 15.2.6 Loss of ac Power to the Plant Auxiliaries 15.2.6.1 Identification of Causes and Accident Description The loss of power to the plant auxiliaries is caused by a complete loss of the offsite grid accompanied by a turbine-generator trip. The on-site standby ac power system remains available but is not credited to mitigate the accident. This transient is more severe than the turbine trip event analyzed in Subsection 15.2.3 because for this case the decrease in heat removal by the secondary system is accompanied by a reactor coolant flow coastdown which further reduces the capacity of the primary coolant to remove heat from the core. The reactor will trip: Upon reaching one of the trip setpoints in the primary and secondary systems as a result of the flow coastdown and decrease in secondary heat removal Due to the loss of power to the control rod drive mechanisms as a result of the loss of power to the plant. Following a loss of ac power with turbine and reactor trips, the sequence described below O occurs: Plant vital instruments are supplied from the Class IE and UPS. As the steam system pressure rises following the trip, the steam generator power-operated I relief valves may be automatically opened to the atmospnere. The condenser is assumed not to be available for turbine bypass. If the steam flow rate through the power-operated relief valves is not available, the steam generator safety valves may lift to dissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in the reactor. As the no-load temperature is approached, the steam generator power-operated relief valves (or safety valves, if the power-operated relief valves are not available) are used to dissipate the residual decay heat and to maintain the plant at the hot shutdown condition if the startup feedwater is available to supply water to the steam generators. The onsite standby power system, if available, supplies ac power to the selected plant permanent nonsafety loads. l 1
- If startup feedwater is not available, the PRHR is actuated. The PRHR heat exchanger l transfers the core decay heat and sensible heat to the IRWST and provides an l uninterrupted core heat removal capability following any loss of normal and startup feedwater.
Revision: 5 February 29,1996 15.2-8 3 W85tingh00S8
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. _. .. l i
- 15. Accidelt Aralyses n
I The startup feedwater system, if available, is started automatically when low level occurs in either steam generator. During a plant transient, core decay heat removal is normally accomplished by the startup feedwater system. If that system is not available then emergency core decay heat removal is provided by the passive residual heat removal heat exchanger. The passive residual heat I removal heat exchangers is a C-tube heat exchanger (HX), connected through inlet and outlet headers to the reactor coolant system. The inlet to the heat exchanger is from the reactor I coolant system hot leg and the return is to the SG outlet plenum. The heat exchanger is located above the core to provide natural circulation flow when the RCPs are not operating. The in-containment refueling water storage tank provides the heat sink for the HX. The passive residual heat removal heat exchanger, in conjunction with the passive containment cooling system (PCS), keeps the reactor coolant subcooled indefinitely. After the in-containment refueling water storage tank water reaches saturation (in about two hours), steam starts to vent to the containment atmosphere and the condensation which collects on the containment steel shell (cooled by passive containment cooling system) returns to the in-containment refueling water storage tank, maintaining fluid level for the passive residual heat removal heat exchanger heat sink. Without any recovery of condensate, the in-containment refueling water storage tank inventory is sufficient to provide the passive residual heat removal heat exchanger operation for 72 hours. Upon the loss of power to the reactor coolant pumps, coolant flow necessary for core cooling Q and the removal of residual heat is maintained by natural circulation in the reactor coolant and () PRHR loops. A loss of ac power to the plant auxiliaries is a Condition II event, a fault of moderate frequency. This event is more limiting with respect to long-term heat removal than the turbine-trip-initiated decrease in secondary heat removal without loss of ac power, which is discussed in Subsection 15.2.3. A loss of offsite power to the plant auxiliaries, can also result in a loss of normal feedwater if the condensate pumps lose their power supply. Following the reactor coolant pump coastdown caused by the loss of ac power, the natural circulation capability of the reactor coolant system removes residual and decay heat from the core, aided by the passive residual heat removal system. An analysis is presented here to show that the natural circulation flow in the reactor coolant system following a loss of ac power event is sufficient to remove residual heat from the core. The plant systems and equipment available to mitigate the consequences of a loss of ac power event are discussed in Subsection 15.0.8 and listed in Table 15.0-6. 15.2.6.2 Analysis of Effects and Consequences 15.2.6.2.1 Method of Analysis A detailed analysis using a modified version of the LOFTRAN code (Reference 2) described I n in WCAP-14601 (Reference 6)is performed to simulate the system transient following a plant (v) Revision: 5 T Westinghouse 15.2-9 February 29,1996 L -
g-9 i
- 15. Accidut Analyses loss of offsite power. The simulation describes the plant neutron kinetics and reactor coolant e l system, including the natural circulation, pressurizer, and steam generator system responses. I The digital program computes pertinent variables, including the steam generator level, pressurizer water level, and reactor coolant average temperature.
l The loss of ac power to the station auxiliaries is evaluated to demonstrate the adequacy of the ! I reactor protection, the PRHR heat exchanger, and RCS natural circulation capability in I removing long term decay heat and pmventing excessive heatup of the RCS with possible RCS overpressurization or loss of RCS water. The assumptions used in this analysis minimize the energy removal capability of the system and maximize the possibility of water relief from the coolant system by maximizing the coolant system expansion. The assumptions used in the analysis are similar to the loss of normal feedwater flow accident (see Subsection 15.2.7) except that power is assumed to be lost to the reactor coolant pumps at the time of the reactor trip.
'Ihe assumptions used in the analysis are as follows:
=
The plant is initially operating at 102 percent of the design power rating with initial I reactor coolant temperature 6.5=F above the nominal value and the pressurizer pressure 50 psi above the nominal value. Core residual heat generation is based on ANSI 5.1 (Reference 3). ANSI 5.1 is a conservative representation of the decay energy release rates. Reactor trip occurs on steam generator low level (narrow range). Offsite power is assumed to be lost at the time of reactor trip. This is more conservative than the case in which offsite power is lost at time zero, because of the lower steam generator water mass at the time of the reactor trip. 1 - A heat transfer coefficient is assumed in the steam generator associated with reactor coolant system natural circulation flow conditions following the reactor coolant pump coastdown. l
- The passive residual heat removal heat exchanger is actuated by the low steam generator water level (narrow range) coincident with a low startup feedwater flow rate (startup feedwater is assumed unavailable).
Conservative PRHR heat transfer coefficients (low) associated with the low PRHR flow rate caused by the RCP trip are assumed. For the loss of ac power to the station auxiliaries, the only safety function required is I core decay heat removal. That is accomplished by the PRHR heat exchanger. The worst I single failure is assumed to occur in the PRHR heat exchanger. Revision: 5 February 29,1996 15.2-10 3 W85tiligt100S8
wm
- 15. Accidert Analyses I
'w I
De actuation of the PRHR heat exchanger requires the opening of one of the two fail open valves arranged in parallel at the PRHR discharge Since no single failure can be assumed i that impairs the opening of both valves, the failure of a single valve is assumed. Moreover, !
- l. only one out of the two PRHR HXs is assumed.
Secondary system steam reliefis achieved through the steam generator safety valves. De pressurizer safety valves are assumed to function. Plant characteristics and initial conditions are further discussed in Subsection 15.0.3. Plant systems and equipment which are necessary to mitigate the effects of a loss of ac power to the station auxiliaries are discussed in Subsection 15.0.8 and listed in Table 15.0-6. j Normal reactor control systems are not required to function. De reactor protection system l 1 is required to function following a loss of ac power, as analyzed here. He PRHR heat exchanger is required to function with a minimum heat transfer capability. No single active ; failure prevents operation of any system required to function. l 15.2.6.2.2 Results De transient response of the reactor coolant system following a loss of ac power to the plant auxiliaries is shownin Figures 15.2.6-1 through 15.2.6-11. He calculated sequence of events !W for this event is listed in T.ble 15.2-1.
<I De LOFTRAN code results show that the natural circulation flow and the passive residual i
heat removal system are sufficient to provide adequate core decay heat removal following reactor trip and reactor coolant pump coastdown. Immediately following the reactor trip, the PRHR heat transfer capability and the steam generator heat extraction rate are sufficient to slowly cool down the plant. At about ! 1-200 seconds following reactor trip, the decrease in the steam generator water inventory results I in a decrease in steam generator heat transfer rate and consequently in a slow heatup of the RCS. 1 At about 1600 seconds following reactor trip, the PRHR heat transfer rate overcomes the core L decay heat and the plant starts a slow, steady cooldown.
- l. A low T,,i, " S" signal is eventually reached: the RCPs are tripped and the CMTs start l l injecting cold borated water in the RCS. PRHR capacity is then lowered and the RCS starts L 1 to heat up, i
l l Pressurizer safety valves open to discharge steam to containment and reclose later in the !' I transient when PRHR capacity exceeds the decay heat production rate. 1 I De capacity of the PRHR is sufficient to avoid water relief through the pressurizer safety I valves. w) Revision: 5 W Westinghouse 15.2 11 February 29,1996 l
__7
; ', 15. Accidrt Analyses ! i I
1 The calculated sequence of events for this accident is listed in Table 15.2-1. As shown in e Figures 15.2.6-5 and 15.2.6-6, in the long term, the plant starts a slow cooldown driven by I the PRHR system. Plant procedures may be followed to funher cool down the plant. 15.2.6.3 Conclusions Results of the analysis show that for the loss of ac power to plant auxiliaries event all safety criteria are met. Since DNBR remains above the safety analysis limit values, the core is not adversely affected. PRHR heat removal capacity is sufficient to prevent water relief through the pressurizer safety valves. The analysis demonstrates that sufficient long-term reactor coolant system heat removal capability exists via natural circulation and the passive residual heat removal system following reactor coolant pump coastdown to prevent fuel or clad damage and so that the reactor coolant system is not overpressurized. 15.2.7 Loss of Normal Feedwater Flow 15.2.7.1 Identification of Causes and Accident Description A loss of normal feedwater (from pump failures, valve malfunctions, or loss of ac power sources) results in a reduction in the capability of the secondary system to remove the heat generated in the reactor core. If an alternative heat sink such as startup feedwater or the PRHR is not supplied to the plant, core residual heat following reactor trip heats the primary system water to the point where water relief from the pressurizer occurs, resulting in a substantial loss of water from the reactor coolant system. Since the plant is tripped well before the steam generator heat transfer capability is reduced, the primary system variables do not approach a departure from nucleate boiling condition. A small secondary system break can affect normal feedwater flow control causing low steam generator levels prior to protective actions for the break. This scenario is addressed by the assumptions made for the feedwater system pipe break (see Subsection 15.2.8). The following occurs upon loss of normal feedwater (assuming main feed vater pump failures or valve malfunctions): The steam generator water inventory decreases as a consequence of the continuous steam supply to the turbine. The mismatch between the steam flow to the turbine and the feedwater flow eventually leads to the reactor trip on a low steam generator water level signal. The same signal also actuates startup feedwater system. As the steam system pressure rises following the trip, the steam generator power-operated relief valves are automatically opened to the atmosphere. The condenser is assumed to be unavailable for turbine bypass. If the steam flow path through the power-operated relief valves is not available, the steam generator safety valves may lift to dissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in the reactor. Revision: 5 February 29,1996 15.2-12 3 W85tingh00Se
2 -- u
- 15. Accide:t Analyses
,y i
Ns)
- As the no-load temperature is approached, the steam generator power-operated relief valves (or safety valves, if the power-operated relief valves are not available) are used to dissipate the residual decay heat and to maintain the plant at the hot shutdown condition if the startup feedwater is used to supply water to the steam generator.
I
- If startup feedwater is not available, the PRHR is actuated on either a low steam generator water level (narrow range) coincident with low startup feedwater flow rate I
signal or a low-low steam generator water level (wide range) signal. The PRHR heat I exchanger transfers the core decay heat and sensible heat to the IRWST so that core heat removal is uninterrupted following a loss of normal and startup feedwater. A loss of normal feedwater is classified as a Condition II event, a fault of moderate frequency. The reactor trip on low narrow range water level in either steam generator provides the necessary protection against a loss of normal feedwater. The startup feedwater system is started automatically, as discussed in Subsection 15.2.6.1. I If startup feed is unavailable then the PRHR heat exchanger is started as discussed in Subsection 15.2.6. An analysis of the system transient is presented below to show that following a loss of normal i feedwater the passive residual heat removal heat exchanger is capable of removing the stored (]
\d and residual decay heat, thus preventing either overpressurization of the reactor coolant system or loss of water from the reactor coolant system, and returning the plant to a safe condition.
15.2.7.2 Analysis of Effects and Consequences 15.2.7.2.1 Method of Analysis A detailed analysis using a modified version of the LOFTRAN code (Reference 2), described in WCAP-14601 (Reference 6), is performed to obtain the plant transient following a loss of normal feedwater. The simulation describes the plant neutron kinetics, reactor coolant system (including the natural circulation), pressurizer, and steam generators. The program computes pertinent variables, including the steam generator level, pressurizer water level, and reactor coolant average temperature. The assumptions used in the analysis are as follows: The plant is initially operating at 102 percent of *he design power rating. Reactor trip occurs on steam generator low (narrow range) level. 1 Since for the loss of normal feedwater mitigatica, the only safety function required is the I core decay heat removal, that is carried by the PRHR heat exchanger, the worst single I failure is assumed to occur in the PRHR heat exchanger. The actuation of the PRHR r n i I heat exchanger requires the opening of one of the two fail open valves arranged in L.) Revision: 5 WOStiflgh0US8 15.2-13 February 29,1996 l l
- 15. AccidInt Analyses l
parallel at the PRHR discharge. Since no single failure can be assumed that impairs the O ! I opening of both valves, the failure of a single valve is assumed. l I = The passive residual heat removal heat exchanger is actuated by the Low-low steam i generator water level wide range signal.
=
Secondary system steam relief is achieved through the steam generator safety valves. I = The initial reactor coolant average temperature is 6.5'F higher than the nominal value, and initial pressurizer pressure is 50 psi higher than nominal. The loss of normal feedwater analysis is performed to demonstrate the adequacy of the reactor i protection and the PRHR heat exchanger in removing long-term decay heat and preventing excessive heatup of the reactor coolant system with possible resultant reactor coolant system overpressurization or loss of reactor coolant system water. As such, the assumptions used in this analysis minimize the energy removal capability of the system and maximize the possibility of water relief from the coolant system by maximizing the coolant system expansion. For the loss of normal feedwater transient, the reactor coolant volumetric flow remains at its normal value, and the reactor trips via the low steam generator narrow range level trip. The reactor coolant pumps continue to run until they are automatically tripped when the CMTs an: actuated. Plant characteristics and initial conditions are further discussed in Subsection 15.0.3. Plant systems and equipment which are necessary to mitigate the effects of a loss of normal I feedwater accident are discussed in subsection 15.0.8 and listed in Table 15.0-6. Normal reactor control systems are not required to function. The reactor protection system is required I to function following a loss of normal feedwater, as analyzed here. The PRHR heat I exchanger is required to function with a minimum heat transfer rate capability. No single active failure p.ever ts operation of any system to perform its required function. A discussion of anticipated tracsients without scram considerations is presented in Section 15.8. 15.2.7.2.2 Results Figures 15.2.7-1 through 15.2.7-10 show the significant plant parameters following a loss of normal feedwater. Following the reactor and turbine trip from full load, the water level in the steam generators falls due to the reduction of steam generator void fraction. Steam flow through the safety valves continues to dissipate the stored and core decay heat. O Revision: 5 February 29,1996 15.2-14 [ W85tingh0USB
I
- 15. Accident Analyses i
! \
l
\v)
The capacity of the PRHR, when the reactor coolant pumps are operating, is much larger than the decay heat and in the first part of the transient the RCS is cooled down and the pressure i decreases. ' i A low Teold "S" signal is eventually reached; the RCPs are tripped and the CMTs start l injecting cold borated water in the RCS. PRHR capacity is then lowered and RCS starts to heat up. Pressurizer safety valves open to discharge steam to the containment and reclose later in the transient when PRHR capacity exceeds the decay heat production rate. The capacity of the PRHR is sufficient to avoid water relief through the pressurizer safety valves. The calculated sequence of events for this accident is listed in Table 15.2-1. As shown in Figure 15.2.7-3 & 4, in the long term, the plant starts a slow cooldown driven by the PRHR system. Plant procedures may be followed to further cool down the plant. 15.2.7.3 Conclusions Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the reactor coolant system, or the steam system. The PRHR heat removal capacity is O such that reactor coolant water is not relieved from the pressurizer safety valves. DNBR b always remains above the safety analysis limit values and RCS and steam generator pressures remain below 110% of their design values. 15.2.8 Feedwater System Pipe Break 15.2.8.1 Identification of Causes and Accident Description A major feedwater line rupture is a break in a feedwater line large enough to prevent the addition of sufficient feedwater to the steam generators to maintain shell-side fluid inventory in the steam generators. If the break is postulated in a feedwater line between the check valve and the steam generator, fluid from the steam generator may also be discharged through the I break. (A break upstream of the feedwater line check valve would affect the plant only as a loss of feedwater. This case is covered by the evaluation in Subsections 15.2.6 and 15.2.7.) Depending upon the size of the break and the plant operating conditions at the time of the break, the break could cause either a reactor coolant system cooldown (by excessive energy discharge through the break) or a reactor coola" system heatup. Potential reactor coolant system cooldown resulting from a secundary pipe rupture is evaluated in Subsection 15.1.5. Therefore, only the reactor coolant system heatup effects are evaluated for a feedwater line mpture. The feedwater line rupture reduces the ability to remove heat generated by the core from the
/
es reactor coolant system for the following reasons: s ( ) v Revision: 5 W WBStiflgh0US8 15.2-15 February 29,1996
} nwm
- 15. Accid:nt Analyses
( Feedwater flow to the steam generators is reduced. Since feedwater is subcooled, its loss e may cause reactor coolant temperatures to increase prior to reactor trip. Fluid in the steam generator may be discharged through the break and would then not be available for decay heat removal after trip. I
- The break may be large enough to prevent the addition of main feedwater after trip.
l The passive residual heat removal heat exchanger functions to: Prevent substantial overpressurization of the reactor coolant system (less than 110 percent of design pressures). Maintain sufficient liquid in the reactor coolant system so that the core remains in place and geometrically intact with no loss of cor: cooling capability. A major feedwater line rupture is classified as a Condition IV event. The severity of the feedwater line mpture transient depends on a number of system parameters, including break size, initial reactor power, and the functioning of various control and safety related systems. Sensitivity studies presented in Reference 4 illustrate that the most limiting feedwater line rupture is a double-ended rupture of the largest feedwater line. The main feedwater control system is assumed to malfunction due to an adverse environment. The water levels in both steam generators are assumed to decrease equally until the Low-low steam generator level reactor trip setpoint is reached. After reactor trip, a double-ended rupture of the largest feedwater line is assumed. These assumptions conservatively bound the most limiting feedwater line rupture that can occur. Analysis is performed at full power assuming the loss of offsite power at the time of the reactor trip. This is more conservative than the case where power is lost at the initiation of the event. The case with offsite power available is not presented since, due to the fast CMT actuation (on a "S" signal generated by the low steam line pressure), the RCPs are tripped by the protection system a few seconds after the reactor trip. The only difference between the cases with and without offsite power available is the RCPs operating status. The following provides the protection for a main feedwater line mpture: A reactor trip on any of the following four conditions: High pressurizer pressure Overtemperature AT Low steam generator water level in either steam generator Safeguards signals from either of the following: Two out of four low steam line pressure in either steam generator. Two out of four high containment pressure (High-1). Refer to Chapter 7 for a description of the actuation system. Revision: 5 February 29,1996 15.2-16 3 Westinghouse
. .. . . - . - - - ~ . - . . - - - - . . ~ . - . - - . . - . - - - . . . .-
1 L 15. Accident Analyses l' i t L a
- m. -
L& LQ p.-- .I The PRHR heat exchanger (HX) provides a passive method for decay heat removal. The !
.I heat exchanger is a C-tube type, located inside the IRWST. The HX is above the RCS 3
I to provide natural circulation of the reactor coolant. Operation of the PRHR heat I ] exchanger is initiated by the opening of one of the two parallel power operated valves 1 at the PRHR cold leg. Refer to Subsection 6.3.2.2.5 for a description of the PRHR. E .. i
- 15.2.8.2 Analysis of Effects and Consequences 15.2.8.2.1 Method of Analysis l
A detailed analysis using a modified version, described in WCAP-14601 (Reference 6), of the LOFTRAN code (Reference 2) is performed in order to determine the plant transient following a fMwater line rupture. The code describes the plant thermal kinetics, reactor coolant system . (including natural circulation), pressurizer, steam generators, and feedwater system responses l and computes pertinent variables, including the pressurizer pressure, pressurizer water level, and reactor coolant average temperature. l The cases analyzed assume a double-ended rupture of the largest feedwater pipe at full power. 1 L Major assumptions used in the analysis are as follows: The plant is initially operating at 102 percent of the design plant rating.
- j. I e Initial reactor coolan; average temperature is 6.5 F above the nominal value, and the.
- - initial pressurizer pressure is 50 psi above its nominal value.
L ' No credit is taken for pressurizer spray. Initial pressurizer level is at a conservative maximum value and a conservative initial steam generator water level is assurned in both steam generators. i, No credit is taken for the high pressurizer pressure reactor trip. Main feedwater to both steam generators is assumed to stop at the tirne the break occurs. (all main feedwater spills out through the break.) L .* A double-ended break area of 1.12 square feet is assumed. This maximizes the L blowdown discharge rate following the time of trip, which maximizes the resultant L heatup of the reactor coolant. A conservative feedwater line break discharge quality is assumed prior to the time the ! reactor trip occurs, thereby maximizing the time in which the trip setpoint is reached. After the trip occurs, a saturated liquid discharge is assumed until all the water inventory is discharged from the affected steam generator. This minimizes the heat removal capability of the affected steam generator. '( Revision: 5 T Westk'.ghouse 15.2 17 February 29,1996
. E--e i
- 15. Acciant Analyses Reactor trip is assumed to be initiated when the low steam generator narrow range level e
setpoint is reached on the ruptured steam generator. I
- The passive residual heat removal heat exchanger is actuated by the low steam generator water level (wide range) signal. A 17-second delay is assumed following the low level signal to allow time for the alignment of passive residual heat removal valves.
No credit is taken for heat energy deposited in reactor coolant system metal during the reactor coolant system heatup. No credit is taken for charging or letdown. Steam generator heat transfer area is assumed to decrease as the shell-side liquid inventory decreases. Conservative core residual heat generation is assumed based upon long-term operation at the initial power level preceding the trip (Reference 3). No credit is taken for the following four potential protection logic signals to mitigate the consequences of the accident: High pressurizer pressure Overtemperature AT High pressurizer level High containment pressure Receipt of a low steam generator water level narrow range signal in at least one steam generator starts the motor driven startup feedwater pumps, which in turn initiate the startup feedwater flow to the steam generators. The PRHR is initiated if the steam generator water level drops to the low steam generator level (wide range) or if a low startup feedwater flow is concomitant to a low steam generator water level (narrow range) signal. Similarly, receipt of a low steam line pressure signal in at least one steam line initiates a steam line isolation signal which closes all main steam line and feed line isolation valves. This signal also gives a safeguard "S" signal which initiates flow of cold borated water from the CMTs to the RCS. Plant characteristics and initial conditions are further discussed in Subsection 15.0.3. I The plant control system is not assumed to function to mitigate the consequences of the event. I The protection and safety monitoring system is required to function following a feedwater line rupture as analyzed here. No single active failure prevents operation of this system. The engineered safety features assumed to function are the passive residual heat removal system, core makeup tank and steam line isolation valves. For the case without offsite power, there is a flow coastdown until flow in the loops reaches the natural circulation value. The natural circulation capability of the reactor coolant system Revision: 5 February 29,1996 15.2 18 3 W65tingh00Se
i 1
- 15. AccideTt Analyses n
(V) 1 is shown (see subsection 15.2.6) to be sufficient to remove core decay heat following reactor I trip for the loss of ac power transient. Pump coastdown characteristics are demonstrated in I , subsections 15.3.1 and 15.3.2 for single and multiple reactor coolant pump trips, respectively. J A cetailed description and analysis of the core makeup tank is described in l l Subsection 6.3.2.2.1. The passive residual heat removal heat exchanger is described in l Subsection 6.3.2.2.5. 15.2.8.2.2 Results I Calculated plant parameters following a major feedwater line rupture are shown in Figures 15.2.8-1 through 15.2.8-10. The calculated sequence of events for the case analyzed is listed in Table 15.2-1. The results presented in Figures 15.2.8-5 and 15.2.8-7 show that pressures in the RCS and main steam system remain below 110 percent of the respective design pressure. Pressurizer I pressure decreases after reac*ar trip on the low steam generator water level (83.0 seconds) due to the loss of heat input. In the first part of the transient, due to the conservative analysis assumptions, the system response following the feedwater line rupture is similar to the loss of ac power to the station auxiliaries (Subsection 15.2.6). n (,l I A few seconds after the trip, the CMTs are actuated (111.2 seconds) on low steam line pressure in the ruptured loop while the PRHR is actuated on a low steam generator water level I wide range (107.1 seconds). I The addition of the PRHR and the CMT flow rates aids in cooling down the primary system and helps to provide sufficient fluid to keep the core covered with water. In the long term, pressurizer safety valves open again due to the mismatch between decay heat and PRHR heat transfer capability. In the first part of the transient, there is a strong cooling effect due to the CMTs that inject cold water into the RCS and receive hot water from the cold leg. In the long term, this effect is much lower due to the heatup of the CMTs. Also, the injection driving head is lowered. RCS temperatures are low (below 500 F at 7000 seconds.) and in this condition, the PRHR is not able to remove the entire decay heat. RCS temperatures tend to increase until an equilibrium between decay heat power and heat absorbed by the PRHR is reached. Finally, after about five hours the PRHR heat transfer capability exceeds the decay heat power and the RCS temperatures, pressure and pressurizer water volumes start to steadily decrease. As previously stated, core cooling capability is maintained throughout the transient since RCS inventory is increasing due to CMT injection. n i ! G! Revision: 5 T Westinghouse 15.2-19 February 29,1996
- 15. Accident An: lyses 15.2.8.3 Conclusions e l Results of the analyses show that for the postulated feedwater line rupture, the passive residual heat removal system capacity is adequate to remove decay heat, to prevent overpressurizing the reactor coolant system, and to maintain the core cooling capal,ility. Radioactivity doses from the postulated feedwater lines rupture are less than those presented for the postulated main steam line break. The SRP Section 15.2.8 evaluation criteria are therefore met.
1 1 15.2.9 Combined License Information
)
l l This section has no requirement for additional information to be provided in support of the l l Combined License application. l l 15.2.10 References
- 1. Cooper, L., Miselis, V., and Starek, R. M., " Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, Revision 1, June,1972. (Also letter NS-CE-622, C. Eicheldinger (Westinghouse) to D. B. Vassallo (NRC), additional information on WCAP-7769, Revision 1, April 16,1975).
- 2. Bumett, T. W. T., et al., "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary) and WCAP-7907-A (Nonproprietary), April 1984.
- 3. "American National Standard for Decay Heat Power in Light Water Reactors,"
ANSl/ANS-5.1-1979, August 1979. O
- 4. Lang, G. E., and Cunningham, J. P., " Report on the Consequences of a Postulated Main Feedline Rupture," WCAP-9230 (Proprietary) and WCAP-9231 (Nonproprietary), January 1978.
5 Friedland, A. J., and Ray, S., " Revised Thermal Design Procedure," WCAP-il397-P-A (Proprietary) and WCAP-11398-A (Nonproprietary), April 1989. I 6. Carlin, E. L., Kemper, R. M., Gresham, J. A., "AP600 Accident Analyses - Evaluation 1 Models," WCAP-14601 (Proprietary), February 1996. O Revisiom 5 i February 29,1996 15.2-20 W W85tingh00S8 i
,n.
4-
- 15. Accident Analyses
,f~\
k, Table 15.2-1 (Sheet 1 of 5) TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH RESULT IN A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Accident Event Time (s) I. Turbine Trip A. With pressurizer control Turbine trip; loss of main feedwater 0.0 (minimum reactivity I feedback) High pressurizer pressure reactor trip point 7.1 reached l Rods begin to drop 9.1 l Minimum DNBR occurs 10.5 l Peak pressurizer pressure occurs 11.0 1 Initiation of steam release from steam 13.0 generator safety valves O () I B. With pressurizer control (maximum reactivity feedback) Turbine trip; loss of main feedwater flow Minimum DNBR occurs 0.0 3.5 l High pressurizer pressure reactor trip setpoint 7.3 reached 1 Rods begin to drop 9.3 l Peak pressurizer pressure occurs 10.5 l Initiation of steam release from steam 13.0 generator safety valves i f% w ,-)
\
i Revision: 5 [ W85tingl10US8 15.2-21 February 29,1996 l
- 15. Accidint Ar.nlyses Table 15.21 (Sheet 2 of 5) e TIME SEQUENCE OF EVENTS FOR INCIDENTS WIIICII RESULT IN A DECREASE IN HEAT REMOVAL BY TIIE SECONDARY SYSTEM Accident Event ' Time (s)
C. Without pressurizer control Turbine trip; loss of main feedwater flow 0.0 (minimum reactivity I feedback) High pressurizer pressure reactor trip point ' 5.8 reached i Rods begin to drop 7.8 Minimum DNBR occurs 9.0 Peak pressurizer pressure occurs 9.5 l Initiation of steam release from steam 13.0 generator safety valves D. Without pressurizer control Turbine trip; loss of main feedwater flow 0.0 (muimum reactivity feeGack) Minimum DNBR occurs 4.0 High pressurizer pressure reactor trip 5.7 Rods begin to drop 7.7 Peak pressurizer pressure occurs 9.0 I Initiation of steam release from steam 13.0 generator safety valves l l 1 Revision: 5 e February 29,1996 15.2-22 3 Westirigt100Se
1 e :
- 15. Accidert Analyses l
,,mx l t 1 i V i Table 15.2-1 (Sheet 3 of 5)
TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH ; RESULT IN A DECREASE IN HEAT REMOVAL BY j THE SECONDARY SYSTEM J Accident Event Time (s) I l II. Loss of ac Power to the Plant Feedwater lost 10.0 ; I Auxiliaries l t Low steam generator water level reactor trip 83,8 l l reached i l Rods begin to drop, ac power is lost, reactor 85.8 l l coolant pumps start to coastdown { l Pressurizer safety valves open 86.6 l Steam generator safety valves open 87.3 1 Maximum pressurizer pressure reached 87.8 i Pressurizer safety valves reclose 90.9 I Maximum pressurizer water volume reached 97.2
/O l \ PRHR actuation on low SG water level 151.0 l (narrow range) coincident with low SFW l
l Pressurizer safety valves open 806.0 l Pressurizer safety valves reclose 1,455.0 l SG safety valves reclose I 7,902.0 i CMT actuation on low Tcoid "S" signal 8,002.0 i Steamline isolation on low Teoid "S" signal 8,002.0 l Pressurizer safety valves open 18,180.0 l Pressurizer safety valves reclose 22,082.0 I PRHR extracted heat matches decay heat - 24,000.0 i l f 1 m Revision: S [ W8Mingh00S8 15.2-23 February 29,1996
J.- ;;. i
= i
- 15. Accident Analyses AP+,o l
O l Table 15.2-1 (Sheet 4 of 5) TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH RESULT IN A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM . 1 Accident Event Time (s) l III. Loss of Normal Feedwater Flow Feedwater lost 10.0 l Low steam generator water level (narrow 83.8 l range) reactor trip reached l Rods begin to drop 85.8 l Pressurizer safety valves open 86.6 l SG safety valves open 873 1 Maximum pressurizer pressure reached 87.8 l Pressurizer safety valves reclose 89.1 1 Maximum pressurizer water volume reached 89.2 1 PRHR actuation on low SG water level (wide 143.0 1 range) l SG safety valves reclose 188.0 l CMTs actuation on low T w "S" signal 568.0 l Steamline isolation on low T.u "S" signal 568.0 l RCP trip on low T. "S" signal 583.0 l Pressurizer safety valves open 8,670.0 l Prer,surizer safety valves reclose 18.574.0 l PRHR extracted heat matches decay beat - 25,000.0 0 Revision: 5 February 29,1996 15.2-24 [ WC5tingh0USS
i
- 15. Accidert Analyses IB i l I'~ l
- (
l Table 15.2-1 (Sheet 5 of 5) TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH ! RESULT IN A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Accident Event Time (s) l IV. Feedwater System Pipe Break Main feedline break occurs 10.0 I l Pressurizer safety valves open 40.5 I i Low Steam Generator water level (narrow 83.0 l , l range) reactor trip reached l I i I Rods begin to drop 85.0 ) I i I Loss of offsite power occurs 85.0 I l Low steamline pressure setpoint 89.2 l l Pressurizer safety valves close 96.0 l l O I All steam and feedline isolation valves close 101.2 U l PRHR actuation on low SG water level (wide 107.1 l range) l l CMT valves fully opened 111.24 1 1 Faulted Steam Generator empties 111.0 l l Intact Steam Generator safety valves open 144.0 l l Intact Steam Generator safety valves reclose 558.0 1 I Pressurizer safety valves open ~3640.0 I
'l PRHR extracted heat matches decay heat ~11550.0 i
l l G' Revision: 5 Y W8Stingh00S8 15.2-25 February 29,1996
=
- 15. Accide;t Analyses mu-l l
1 l NucIear Power CFraction of NominalJ ! 1.4 1.2 - 1.0 - 1 0.8 - 0.6
$l 0.4 -
0.2 - 0.0 I I I I I I ' O 10 20 30 40 50 60 70 80 90 100 Time (Sec) Figure 15.2.3-1 l Nuclear Power (Fraction of Nominal) vs. Time for Turbine Trip Accident with Pressurizer Spray and Minimum Moderator Feedback Revision: 5-l February 29,1996 15.2-26 [ Westingh00S8
M
- 15. Accident Analyses ..
,es) \s Pressurizer Pressure (psia) 2600 2500 -
2400 - 2300 2200 -
.o
(,) 2100 - 2000 - 1900 -
' I I 1800 I I I O 10 20 30 40 50 60 70 80 90 100 T i me ( Sec)
Figure 15.23-2 Pressurizer Pressure (psia) vs. Time for Turbine Trip
,q Accident with Pressurizer Spray and Minimum Moderator Feedback ,O Revision: 5
[ W8Stingil00S8 15.2-27 February 29,1996 i
- =:: '
- 15. Accident Analyses
*l l
Pressur izer Water Vo l ume (Cub ic Feet) 1700 1 1600 - 1500 - 1400 - 1300 - 1200 - 1100 -
~
h 1000 - 900 - 800 - I I ' 700 O 1 2 3 4 5 6 7 8 9 10 Time (Sec) Figure 15.2.3-3 Pressurizer Water Volume (ft') vs. Time for Turbine Trip Accident with Pressurizer Spray and Minimum Moderator Feedback Revision: 5 February 29,1996 15.2 20 W Westinghouse
- 15. Accident Analyses o
%/
Core I n iet Temperature (Deg F) 600 590 580 570 - 560 -- 550 - ( 540 - 530 520 - 510 - 500 ' ' ' O 10 20 30 40 50 60 70 80 90 100 Time (Sec) Figure 15 2.3-4 Core Inlet Temperature ( F) vs. Time for Turbine Trip Accident ,(T with Pressurizer Spray and Minimum Moderator Feedback L..) Revision: 5 Y W9Stinghouse 15.2-29 February 29,1996
!!M!!
- 15. Accident Analyaes n-1 Core Average Temperature (Deg F) 600 590 -
580 - 570 - O 560 - 550 - I I I I I I I I 540 I O 1 2 3 4 5 6 7 8 9 10 T i me ( Sec) j Figure 15.23-5 Core Average Temperature ('F) vs. Time for Turbine Trip Accident with Pressurizer Spray and Minimum Moderator Feedback
- Revision
- 5 February 29, im 15.2-30 W Westinghouse
- 15. Accklent Analyses v
i DNB Ratio 7 6 - 3 - 4 - 3 2 - l l I I I q I I I I O 10 20 30 40 50 60 70 80 01 100 Time (Sec) Figure 15.2.3-6 DNB Ratio vs. Time for Turbine Trip r*} Accident with Pressurizer Spray and Minimum Moderator Feedback " G Revision: 5 3 "v'eduiliOUS8 15.2-31 February 29,1996 l
- 15. Accident Analyses e
Nuclear Power (Fraction of Nominal) 1.4 12 - 1.0'- 4 0.8 - 0.6 - g 0.4 - 0,2 - k 0 O 10 20 30 40' 50 60 70 80 90 100 Time CSec) i Figure 15.2.3-7 Nuclear Power (Fraction of Nominal) vs. Time for Turbine Trip Accident with Pressurizer Spray and Maximum Moderator Feedback Revision: 5 February 29,1996 15.2-32 W Westilighouse l
I
.:Z..
- 15. Accident Analyses L l 1 l
l O lU Pressur izer Pressure (ps ia) l 2600 2500 - 2400~ - 2300 2200 - 0 2,00 - 2000 - 1900 - I I I I I I I 1800 I I O 10 20 30 40 50 60 70 80 90 100 ' Time CSec) Figure 15.2.3-8 l Pressurizer Pressure (psia) vs. Time for Turbine Trip Accident
- jm, with Pressurizer Spray and Maximum Moderator Feedback O
Revision: 5 Y W96tingh00S8 15.2-33 February 29,1996
n=ii
- 15. Accident Analyses en-O Pressurizer Water Vo l ume (Cub i c Feet) 1700 1600 -
1500 - 1400 - 1300 - 1200 - 1100 - hl 1000 - l 900 - I 800 - I I I I I I 700 I I I O 1 2 3 4 5 6 7 8 9 10 Time (Sec) l Figure 15.2.3-9 Pressurizer Water Volume (ft') vs. Time for Turbine Trip Accident with Pressurizer Spray and Maximum Moderator Feedback ^ Revision: 5 l February 29,1996 15.2-34 [ WBStingh0US8 l 1
h vm
- 15. Accident Analyses l r~w
' () l l l Core inlet Temperature (Oeg F) ! 600 590 580 - 570 - 560 - i 550 - Cs 540 - 530 520 - 510 - 500 0 10 20 30 40 50 60 70 80 90 100 Time (Sec) Figure 15.2.3-10 Core Inlet Temperature ('F) vs. Time for Turbine Trip Accident ' ,rm with Pressurizer Spray and Maximum Moderator Feedback (_- Revision: 5 [ W95tillgt100S8 15.2-35 February 29,1996
- 15. Accident Analyses o
l l 1 l
)
Core Average Temperature (Deg F) ' 600 590 - 580 -
)
570 -
.560 -
e 550 - I I I I I I 540 I I I O 1 2 3 4 5 6 7 8 9 10 Time CSec) l l Figure 15.2.3-11 Core Average Temperature ('F) vs. Time for Turbine Trip Accident with Pressurizer Spray and Maximum Moderator Feedback j Revision: 5 February 29,1996 15.2-36 T Westinghouse l
- 15. Accident Analyses .
,rh DNB Ratio 7
6 - 5 4 ___ A 3 2 - q l l l l l l l l l 0 10 20 30 40 50 60 70 80 90 100 T i me ( Sec) i Figure 15.23-12 DNB Ratio vs. Time for Turbine Trip Accident with Pressurizer Spray and Maximum Moderator Feedback Revision: 5 [ Westingh0Use 15.2-37 February 29,1996
- 15. Accident Analyses e
l Nuclear Power (Fract ion of Nomi na l) 1.4 1 2 - 1.0
-]
0.8 - 0.6 - g 0.4 - 0.2 - 0 O 10 20 30 40 50 60 70 80 90 100 T i me ( Sec) Figure 15.2.3-13 Nuclear Power (Fraction of Nominal) vs. Time for Turbine Trip Accident without Pressurizer Spray and Minimum Moderator Feedback Revision: 5 February 29,1996 15.2-38 3 Westinghouse
- 15. Accide~t Analyses
() w Pressur izer Pressure (ps ia) 2600 2500 - 2400 - 2300 2200 - 13 V 2100 - 2000 - 1900 - I I I I I 1800 I I I I O 10 20 30 40 50 6.0 70 80 90 100 Time CSec) l Figure 15.2.3-14 Pressurizer Pressure (psia) vs. Time for Turbine Trip Accident without Pressurizer Spray and Minimum Moderator Feedback Revision: 5 [ W86tingfl00S8 15.2-39 February 29,1996
ii=_ !!
- 15. Accide:t Analyses au --
s. O Pressur izer Water Vo l ume (Cub i c Feet) 1700 1600 - 1500 - l 1400 - 1300 - 1200 - 1100 - ~ 1000 - 900 - 800 - i I I I I I I 700 I I I O 1 2 3 4 5 6 7 8 9 10 Time (Sec) Figure 15.2.3-15 Pressurizer Water Volume (ft') vs. Time for Turbine Trip Accident without Pressurizer Spray and Minimum Moderator Feedback Revision: 5 February 29,1996 15.2 4 T Westliighouse
- 15. Accident Analyses
(
,ex, U
Core Inlet Temperature (Deg F) 600 590 - 580 - 570 - 560 - 550 - r~
- b) 540 -
530 520 - 510 - I I I I 500 I I I I O 10 20 30 40 50 60 70 80 90 100 Time (Sec) l Figure 15.23-16 Core Inlet Temperature ('F) vs. Time for Turbine Trip Accident !g] without Pressurizer Spray and Minimum Moderator Feedback
%.)
Revision: 5 [ W86tingh00S8 15.2-41 February 29,1996
- 13. Accid nt Analyses e
Core Average Temperature (Deg FJ 600 590 - 580 - 570 - 560 - O 550 - I I I I 540 I I I I I O 1 2 3 4 5 6 7 8 9 10 Time CSec) Figure 15.23-17 Core Average Temperature ('F) vs. Time for Turbine Trip Accident without Pressurizer Spray and Minimum Moderator Feedback Revision: 5 February 29,1996 15.2-42 T W85tingh00S8
- 15. Accident Analyses r
b) DNS Ratio 7 6 - 5 - 4 - o L) 3 - 2 - l l l l l I I I l q 0 10 20 30 40 50 60 70 80 90 100 T i me ( Sec) Figure 15.2.3-18 DNB Ratio vs. Time for Turbine Talp Accident Without p
.sj Pressurizer Spray and Minimum Moderator Feedback Revision: 5 T Westinghouse 15.2 43 February 29,1996
- 15. Accident Analyses e
Nuclear Power (Fraction of Nominal)
- 1. 4 1.2 -
1.0'- 0.8 - 06 -- 0.4 - 1 0.2 -
' I I I ' ' '
0.0 ' O 10 20 30 40 50 60 70 80 90 100 l Time (Sec) Figure 15.2 3-19 Nuclear Power (Fraction of Nominal) vs. Time for Turbine Trip Accident Without Pressurizer Spray and Maximum Moderator Feedback Revision: 5 February 29,1996 15.2-44 3 Westingttouse
inamm=
- 15. Accident Analyses j-l L)
Pressurizer Pressure (psia) 2600 - 2500 - 2400 - 2300 2200 - 2100 - 2000 - 1900 - 1 I I I 1800 I I I I i 0 10 20 30 40 50 60 70 80 90 100 Time (Sec) Figure 15.23-20 Pressurizer Pressure (psia) vs. Time for Turbine Trip Accident Without Pressurizer Spray and Maximum Moderator Feedback %p.) Revision: 5 Y W8Stingh00S8 15.2-45 February 29,1996
l
- 15. Accident Analyses e
1 Pressur izer Water Vo I ume CCubic Feet) ! 1700 1600 -- 1500 -- 1400 - 1300 - 1200 - 1100 - 1000 - 900 - 800 - I I I I I I 700 I I O 1 2 3 4 5 6 7 8 9 10 Time (Sec) Figure 15.2.3-21 Pressurizer Water Volume (n') vs. Time for Turbine Trip Accident Without Pressuriar Spray and Maximum Moderator Feedback Revision: 5 February 29,1M 15.2-46 [ W85tingh00S8
- 15. Accident Analyses'
[. 1 'q b Core inlet Temperature (Deg F) 600 590 - 580 - 570 - 560 - 550 - O 540 - 530 520 - 510 - I I I I I 500 I I I I O 10 20 30 40 50 60 70 80 90 100 Time (Sec) l , I l Figure 15.2.3-22 Core Inlet Temperature (*F) vs. Time for Turbine Trip Accident . g]
- . Without Pressurizer Spray and Maximum Moderator Feedback 2 k./
Revision: 5 Y WOStingh0088 15.2-47 February 29,1996. l
- 15. Accident Analyses e
I Core Average Temperature (Deg F)
\
600 I 590 - k 580 - l l 570 - O 560 - 550 - I I I I 540 I I I I I O 1 2 3 4 5 6 7 8 9 10 Time (Sec) I I I Figure 15.23-23 Core Average Temperature ('F) vs. Time for Turbine Trip Accident Without Pressurizer Spray and Maximum Moderator Feedback Revision: 5 February 29,1996 15.2-48 T Westinghouse
1
- 15. Accident Analyses I t
r - l I l l l 1 DNB Ratio 7 I I 6 5 4 - (O%J 3'- 2-l l I I I I I I I q O 10 20 30 40 50 60 70 80 90 100 Time C5ec) Figure 15.2.3-24 1 ! DNB Ratio vs. Time for Turbine Trip Accident Without lO\ Pressurizer Spray and Maximum Moderator Feedback Revision: 5 T W8stingh0US8 15.2-49 February 29,1996
- . . . - . . . ~--. . . _ . _ - . . -- . - . . - _ . ....- . . . . . . . . .
i = ii
- 15. Accid =.t Analyses e
Core Nuclear Power
-- 1 2 "E -
.2 -
E - g 1- - - o -
@ 8- -
n - E - 4: y .6- - u -
$ ~
$ .4- 2 I
=
o -
"U -
g 2- ; l o - 1 o - c.J 0
l '; -
1 2 3 4 5 10 10 10 10 10 Time (Seconds) l Figure 15.2.6-1 l Nuclear Power Transient For Loss of Nonemergency AC Power to the Plant Auxiliaries l Revision: 5 February 29,1M 15.2-50 3 W65tingh00Se
._..-4._.. ..-...m____-__-- . _ . . . _ _ . _ _ . . _ . _ _ _ . . _ _ . _ . . . _ _ . . _ . . _ _ . - . . . _ . _ _ . _ . . . _ _ _ . _ - .
3
- 15. Accident Analyses g[r-~wg1 ti. ~d
-'4 9
J 1
\
Core Heat Flux i 12
=n -
..C -
g 1 - 4 o _
- 2 _
o 8- - m- _ 4- o
.g _
o - I j k
- 6- -
M _ m
; g .4- -
a - o .
.c 2- -
es - C.) 0 '! 8 3 5
' ' ' ' ' ' i i
1 2 3 4 5 l 10 10 10 10 10 ' Time (Seconds) ! l 1 Figure 15.2.6 2 Core Heat Flux Transient For Loss of Nonemergency AC Power to the Plant Auxiliaries Revision: 5 W Westinghouse 15.2-51 Febr- 29,1996
- .. ._. _. .. . . . - - ..__ . .- = . - - . .
- M!!
- 15. Accident Analyses sm-l 9:
Pressurizer Pressure 2600
^< 2400 --
cn / L a) u 2200 - a ~
~
E as - g 2000 - - b n
*C 18 0 0 --
c - E u ~ A 1600 - -
~
1400 ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '
'l l I 2 3 4 5 10 .10 10 10 10 Time (Seconds)
Figure 15.2.6-3 Pressurizer Pressure Transient for Loss of 1 Nonemergency AC Power to the Plant Auxiliaries ; Revision: f i February 29,1996 15.2-52 [ W85tingh0US8
~
- 15. Accident Analyses -
/'
A i 1 Pressurizer Water Volume 1 _ 1600 Ye ~ o 1400 - - 3 -
.m -
1200 - - s -
.2 :
8 1000 - - u - , :r~ .S - e - L > 800 -- 6 - N ~
.g -
m 600 -- i E
=
A ' ' 400 'l
'"l I 2 3 4 5 10 .10 10 10 10 Time (Seconds) i Figure 15.2.6-4 Pressurizer Water Volume Transient For Loss of / Nonemergency AC Power to the Plant Auxiliaries Revision. 5 T Westinghouse 15.2 53 February 29,1996
,-g k i i 15. Accid =1 Analyses
{ C -- O i I Saturation
----Hot Leg
----CoId Leg 700 -
650 -_ rm
's s g
. 600 -:
N s__--, s e 550 - 1 . . . . . s.~.\\ ,. . . . - - s s y s e 500 - - o -
.s.' ss. .
i 2 - s g 5 450 - i / E b I
- \/ ."'s
~
/
400 -2 i.s/
~
350 ' ' ' ' ' ' "
'l ';
1 2 3 4 5 10 10 10 10 10 Time (Seconds)
~ _ . . _.
i Figure 15.2.6-5 t RCS Temperature Transients in Loop Containing the PRHR For Loss of Nonemergency AC Power to the Plant Auxiliaries Revision: 5 February 29,1996 15.2-54 y Westinghotise
. . . _ . . . . . - . . . . - - . - . ~ . . _ . - . - . . . . . - . . . - - _ . - - - - - - . ~ - . - - . _ . . .
1 l ry~ : a
- 15. ' Accident Analyses fI' f LO i
l' !' j
.Saturalion l ----Hot Leg
----Cold Leg l 700 -
l _ 650 - -
- e
, rm., -
#s ~
. :_ _ _ _ s gp ..
. g 600 - -
's,,-p *.,,
! v - s.
,/ \. ._. #'
- s 550 -
! , . ... . ... a i g l 8' o n.500- - E - i , F : i i 450 - -
!i ,
si 1 400 1
'l 1
'''l 3
'l' 4 5
10- 10 10 10 10 i Time (Seconds) l l: f-Figure 15.2.6-6 i RCS Temperature Transients in Loop Not Containing the PRHR For Loss of Nonemergency AC Power to the Plant Auxiliaries hO i I i Revision: 5 [ Westkigh0088 15.2-55 February 29,1996 l I
n -!;
- 15. Accident Analyses n_
O\ l i Loop WIth PRHR
----Lcop W1thout PRHR 1150 -
1100 - I m - 5 rn 1050 - - s', _____ w - e 1000 - 4 3 L 950 - h
@ 900 -
850 -
~
800 '! '"l 1 2 3 4 5 10 .10 10 10 10 Time (Seconds) f Figure 15.2.6-7 Steam Generator Pressure Transients For Lc; of Nonemergency AC Power to the Plant Auxiliaries Revision: 5 February 29,1996 15.2-56 3 Westingflouse
. _- - . . - ~ - ~ . . ~ _ . . . . - . . ~ . . . . . - ~ ~ . ~ . . - . . . . . _ . . . . . - . . . . .
r
- 15. Aceklent Ans' ' *g il F'~-;S) h .;
-a k
9 i PRHR-Flow Rote 120 e 100 - - 9 E - N - Ei 80 . S y - as 60 - - E: - A > ~ V t$ 40 -~ - i. I llI: - Ct: - cL. 20 - - W
=
i f f I f f elf I f f I I Ilit 0 . . I f f f ' I f 19 f f I I I f ff 1 2 3 4 5 10 10 10 10 10 Time (Seconds) Figure 15.2.6-8 PRHR Flow Rate Transient For Loss of Nonemergency AC Power to the Plant Auxiliaries O Revision: 5 Y WOStkigh0088 15.2-57 February 29,1996
- 15. Accidest Analyses e4 PRHR Heat Flux 25E-01
= -
g y 2 E f O _ C -
- j. .15 E -
U ~
~
r.s . E g 1 E - g 5 :
$ 5E -
= :
- n ~
Z A ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' " ' ' ' ' ' ' ' ' ' ' ' ' 0 I 2 3 4 5 10 10 10 10 10 Time (Seconds) l. i i Figure 15.2.6-9 PRHR Heat Flux Transient For Loss cf Nonemergency AC Power to the Plant Auxiliaries Revision: 5 February 29,1996 15.2-58 W Westinghouse
_ -. . . _ . . - . . _ - . . . ~ . ~ . . _ .... -. . . - . .- . _ . . - - . . . - - , . . . ~
- 15. Accide;t Analyses f
_m /% Loop WIth PRHR
----Loop Without PRHR e 1.2
~
1 z - 8- -
~ 6--
*4"
.g 1 .2 --
2! : g a_- ~-----------;___.
$ .2 ' ' ' ' ' ' ' " ' ' ' ' ' ' ' " ' ' ' ' ' ' ' ' ' ' ' '
1 2 3 4 5 10 10 10 10 10 Time (Seconds) Figure 15.2.6 10 Reactor Coolant Mass Flow Rate Transient For Loss of Q Nonemergency AC Power to the Plant Auxiliaries ,V Revision: 5 Y W8Stingh0088 15.2-59 February 29,1996
._ . . .. . . - . . -. . .- .. .. -_~ . - - . ...
- 15. AccidInt Analyses e
Loop With PRHR
----Lcop Without PRHR 140000 _
120000 -{ E 100000 - - bx 80000 - b 8 : 5as 60000 - E : g 40000 - - 20000 - _
'N 0
''l I 2 3 4 5 10 10 10 10 10 Time (Seconds)
Figure 15.2.6-11 Steam Generator Inventory Transients O Revision: 5 I February 29,1996 15.2-60 [ W85tingh00S8 l
=
nn
- 15. Accident Analyses o
V Core Nuclear Power n 12
~B -
a .
'g _
o 1- - - 1 2 _ o -
'O" " .
a - o , e -
~
I
\
r5 w .6- - ,
- s. I e
/ >
~
o - Q. .4- - 5 o - m 2- - C 8 - C) 0 1- '2 '3 4 5 10 10 10 10 10 Time (Seconds) Figure 15.2.7-1 Nuclear Power Transient For O Loss of Normal Feedwater Flow }L Revision: 5 T Westinghouse 15.2-61 February 29,1996 l l
ii- ii
- 15. Accidet Analyses m-O Loop With PRHR
-~~-Loop Without PRHR c 1.2 .
j - 1 - 8- -
.6- -
4- - 3 : A - b .2 --
$ o :
u
, , , , ,,,.. , , , , , , , - , - . . . . - - , . , , , , . - - + , , , . . .
1- 't '3 '4 5 10 10 10 10 10 Time (Seconds) Figure 15.2.7-2 RCS Mass Flow Transient For Loss of Normal Feedwater Flow l Revision: 5 I February 29,1996 15.2-62 [ W85tingh00S8
n--n
- 15. Accident Analyses l'
p Soturation
----Hot Leg
----Cold Leg 700 650 --
T on -i
.. 600 -___ t e _
s Clll3 -
*o550 - 2._. ..... .% .-~
.*.0 y -
. ,' % s \
a - ss I o e 500 - w _ i. s K 8, ;
' s. s's
~
E 450 - 'N ' ' e"
's,'
s
/.
400 - - s. ,' 350 ~ ' ' ' ' ' ' ' ' ' '
'l 1 2 3 4 5 10 10 10 10 10 Time (Seconds)
Figure 15.2.7-3 RCS Temperature Transients in Loop Containing the PRHR For O Loss of Normal Feedwater Flow Revision: 5 [ W85tinghouse 15.2-63 February 29,1996
. - . . .-.. . .- ~. . - - - . - ~ _ _ . !!ME ;;
- 15. Accident Analyses !
nam - 0 l Soturation ,
----Hot Leg l
----Cold Leg ;
700 i
. 1 , 650 - -
r - a 40 _____,,'g ,
- 600 - -
g
- w. . \
be -
.!. ~.A
~
w 550 - -
...'s ..' %. ,
R g
%. ,s c
[ 'w f%. g
\
" 500 - \
,. * "'g
{ g \ ! g (
\
- s\ f !
' ' ' ' h *y '
450 'l 1* 2 3 4 $ 10 10 10 10 10 Time (Seconds) 1 Figure 15.2.7-4 l l RCS Temperature Transients in Loop Not Containing the PRHR For Loss of Normal Feedwater Flow Revision: 5 February 29,1996 15.2-64 [ WSStingh00S8
i !- 15. Accident Analyses ia , 1 l l l Pressurizer Pressure 2600
~
1 m d!i 2 4 00 - - v3 - l L
- i 8 -
8 2200 - - E . 8 . i L ~ n i\J
.g 2000 -_
u a - B w 1800 - - L - 1600 ". ". ". i i . , - . . I 2 3 4 5 10 10 10 10 10 Time (Seconds) ! Figure 15.2.7-5 j Pressurizer Pressure Transient For f Loss of Normal Feedwater Flow Revision: 5 W a % 00$6 15.2-65 February 29,1996
--u
- 15. Accide:t Analyses e
Pressurizer Water Volume l _ 1600 T* - r ;
$m 1400 - -
8 - o E 1200 - - 2 . u -
.S a
1000 - -
>s ~
k N ~
*c 800 - -
2 . B w
" 600
.I
': 2 3
4 5 10 10 10 10 10 Time (Seconds) Figure 15.2.7-6 Pressurizer Water Volume Transient For Loss of Normal Feedwater Flow Revision: 5 February 29,1M 15.2-66 [ W8stirighouse
l
- 15. Accident Analyses -
(o> l Loop With PRHR
----Loop Without PRHR 1200 1000 - -
2 's s 95 b 800 -- ' g I I g ,e~% g - s , D. 600 - -
, s /
C -
/
~ ,
r _ 400 - - 200 'l 1 2 3 4 5 10 10 10 10 10 Time (Seconds) Figure 15.2.7-7 Steam Generator Pressure Transients For Loss of Normal Feedwater Flow Revision: 5 T Westinghouse 15.2-67 February 29,1996
i=:i
- 15. Accident Analyses e
Loop With PRHR
----Lcop Without PRHR 140000 _
120000 -{ E 100000 - - 3 g 80000 -; 3 - m
,0 c:
60000 -{
~
g g 40000 - 20000 - -
~
, , ,,,iii. , , , , , . . . , , , ,,,,,e g i , ,,. i 1 2 3 4 5 10 10 10 10 10 Time (Seconds)
Figure 15.2.7-8 Steam Generator Inventory Transient For Loss of Normal Feedwater Flow Revision: 5 February 29,1996 15.2-68 3 Westirigh00S8
. .- . . . . . _. . . . . . . . _ . . . _ . -- .. . ~ _ . _ . _ . _ .
l WL
- 15. Accideit Analyses * ~~
! / i I , v 1 1 PRHR Heat Flux
.7E-01 m -
m -
's . 6 E-01 -j o _
- ac -
's . 5 E-01 -_:
,. c o 3e 4E-01 -[ j rs. w . 3 E - A. 3 (] SC 2 E - m - G - Z : a: 1 E - l2: og : Q* - 0
, ,.i1 .,
I 2 3 4 5 10 10 10 10 10 Time (Seconds) Figure 15.2.7-9 i PRHR Heat Flux Transient For 1 - Loss of Normal Feedwater Flow
^'
- <J Revision
- 5 i-T Westinghouse 15.2-69 February 29,1996 i
p-- ei
- 15. Accid:: t Andyses m-O
~
CMT lnjection Line Flow 120 T -
- l N % 100 - -
E
- S : ,
80 - - o E _
~
g 60 - -
- a -
.m E 40 - -
h.
~
tl5 : 20 - - l C.)
~
l 0 1 2 3 4 5
)
10- - 10 10 10 10 Time (Seconds) Figure 15.2.7-10 CMT Injection Flow Transient For Loss of Normal Feedwater Flow Revision: 5 February 29, IN 15.2-70 [ W85tiligh0US8
i , I r i
- 15. AccW;t Andym F i, +
t r' Ew] r i I l l 12 o . C. 1__ E - o C _ o .a_ _ J C _ l O o
- o 6- - l w -
q - g - _ m _ 4_ _ O _ Q. o e .2- - u "3 _ z 0 '"", ' ' ' ' ' ' " ' 1 2 3 4 5 6 10 10 10 10 10 10 TIME (S) s; Figure 15.2.8-1 l l Nuclear Power Transient For ! Main Feedwater Line Rupture t i Revision: 5 T Westinghouse 15.2-71 February 29,1996
-}}