ML20198G986

Rev 1 to WCAP-13856, AP600 Implementation of Regulatory Treatment of Nonsafety-Related Systems Process

ML20198G986
Person / Time
Site:	05200003
Issue date:	01/05/1998
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20198G970	List: ML20198G986 NSD-NRC-98-5509, Forwards Rev 1 to TR WCAP-13856, AP600 Implementation of Regulatory Treatment of Nonsafety-Related Systems Process
References
WCAP-13856, WCAP-13856-R01, WCAP-13856-R1, NUDOCS 9801130166
Download: ML20198G986 (102)

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Westinghouse Non Proprietary Class 3

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AP600 Implementation of the Regulatory Treatment of Nonsafety-Related Systems Process i

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Westinghouse Energy Systems -

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AP000 DOCUMENT COVER SHEET TDC

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CO6SIRM RF6s, RFS ITEM e; AP900 DOCWEN* NQ. RCVISlON NO. ASSIGNED 1O GWGLO26 1 Page 1 of 9 ALTERNATE DOCUMENT NUMBER; WCAP 13856, Rev.1 WORK BREAKDOWN #:

DESIGN AGENT ORGANIZATION Westinghouse  !

TITLE: AP600 Implementation of the Regulatory Treatment of Nonsafety Related Systems Process

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0 WCAP 13856 Retision 1 AP600 Implementation of the Regulatory Treatment of Nonsafety-Related Systems Process I

l January 1998 This document contains information proprietary to Westinghouse Electric Company, a Division of CBS Corporation; it is submitted in confidence and is to be used solely for the purpose for which it is fumished and retumed upon request. This document and such information is not to be reproduced, transmitted, disclosed or used otherwise in whole or in part without prior written authorization of Westinghouse l Electric Company, Energy Systems Business Unit.

4 l Westinghouse Electric Company Energy System Business Unit P.O. Box 355 Pittsburgh, PA 15230-0355 l C 1998 Westinghouse Electric Company All Rights Reserved WCAP-13856 Revision: 1 eM9543 I.spf.lM10596-lh

TABLE OF CONTENTS Section Page EX EC UTI V E S UM M A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vil 1.0 I NTR ODUCTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 2.0 FOCUSED PRA SENSmVITY STUDY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 '

2.1 FOCU S ED PRA ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 2.2 R ES U LTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 l 2.3 ~ PRA UNCERTAINTY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2 3.0 INmAT1NO EVENT FREQUENCY EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 3.1 MAIN S7EAM LINE STUCK OPEN SAFETY VALVE . . . . . . . . . . . . . . . . . . . . . . . 3 2 3.2 R CS LEA l'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . 3 3 3.3 LOCAs.........................................................34 3.4 SECONDARY SIDE B REA KS . . . . . . . . . . . . . . . . . . . . . . . . ................35 3.5 TRA N S I ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -6 3.6 ANTICIPATED TRANS!ENT WmlOUT SCRAM . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8 3.7 MISCELLANEOUS SPECIAL INITIATORS . . . . . . . . ......................39 3.8 SI I UTDOWN LOCA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 10 3.9 SilUTDOWN LOSS OF 0FFSITE POWER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 10 3.10 S!!UTDOWN LOSS OF DECAY llEAT REMOVAL . . . . . . . . . . . . . . . . . . . . . . . . 311 3.11 REACTOR COOLANT SYSTEM OVERDRAIN . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 11 3.12 S U M M A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 12 4.0 10 CFR 50.62 (ATWS RULE) ..............................................41 4.1 10 CFR 50.62 EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1 4.2 10 CFR 50.62 CONCLUSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1 5.0 10 CFR 50.63 (LOSS OF ALL AC POWER RULE) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 5.1 10 CFR 50.63 EVALU ATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......51 5.2 10 CFR 50.63 CONCLUSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 16.0 POST.72 iiOuR ACriONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-i 6.1 POST 72 IlOUR ACTIONS EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1 6.2 POST.72 IlOUR ACTIONS CONCLUSION . . . . . , . . . . . ......... . . . . . . . . . 6- 2 7.0 CONTAINMENT PERFORM ANCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1 7.1 CONTAINMENT PERFORMANCE EVALUATION .........................71 7.2 CONTAINMENT PERFORMANCE CONCLUSION . . . . . . . . . , . . . . . . . . . . . . . . . . 7 1 8.0 A DVERSE SYSTEM S INTERACTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 1 8.1 ADVERSE SYSTEMS INTERACTION EVALUATION . . . . . . . . . . . . . . . . . . . . . . . 8 1 l 8.2 ADVERSE SYSTEMS INTERACTIONS CONCLUSION . . . . . . . . . . . . . . . . . . . . . . 81 9.0 SEISMIC CONSIDERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 1 9.1 SEISMIC CONSIDERATIONS EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . .. 9-1 9.2 SEISMIC CONSIDERATIONS CONCLUSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 WCAP-13856 Revision: I o msaw-13put*oloS98-120na til

TAllLE OF CONTENTS Section Page 10.0 MISSION STATEMENTS AND PROPOSED REGULATORY OVERSIGIIT RECOMM ENOA110NS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 1 10.1 IMPORTANT NONSAFETY RELATED SSCS . . . . . . . . ......... .. ....... 10-1 10.2 MISSION STA1T3 TENTS . . . . . . . . . . . . . . . . . . . . . .................... 10-3 10.3 PROPOSED REGULATORY OVERSIGliT RECOMMENDATIONS . . . . . . . . . ... 10-4 WCAP-13856 Revirion: 1 on9s4w.t.wpr.iutos9s-12w y

LIST OF TABI.13 Table Page Table i Summary List of Investment Protection Short Term Availability Contsols .............in Table 11 Nonsafety Related Systems Evaluated in the AIWK) RTNSS Process . . . . . . . . . . . . . . . 16 Table 21 Nonsafety Related Systems and Functions Removed From Baseline PRA Ana'ysis . . . . . 2-4 Table 31 initiating Event Criteria Appbcation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 13 Table 101 List of Investment Protection Short. Term Availability Controls . . . . . . . . . . . . . . . . . 1016 Table 10 2 Investment Protection Short Term Availability Controls . . . . . . . . . . . . . . . . . . . . . . 10-17 WCAP 13856 Revision: I oV984*-1214 its010598-l.20ru y

LIST OF FIGURES Figure Page l Fi ure f 11 AP600 RTNSS hocess Evolution and implementation . . . . . . ................... 14 Figme 31 Evaluation of Nonsafety related SSCs impact on Initiating Event Frequency . . . . . . . . . 314 WCAP 13856 Revision: 1 oues4..I wpr thol059s.1,20ru g.i ,

EXECUTIVE SifMMARY l %e regulatory treatment of nonsafety related systems (RTNSS) in advanced reactor passive p' ant designs has a wide ranging effect on both the design and licensing of the AP600. Unlike the current l generation of light water reactors, the AP600 uses pasive safety systems that rely exclusively on natural forces such as density differences, gravity, and stored energy to provide water for core and containment cooling. These passive systems do not include active equipment such as pumps. One-time alignment of safety related valves actuates the passive safety related systems using valve operators such as de motor operators with power provided by Class lE buteries, air operators that reposition to the safeguards position on a loss of the nonsafety related compressed air that keeps the safety related equipment in standby, or check valves that operate by the pressure differential across the valve disc. The operation of the safety related passive systems does not require a: electrical power.

For the AP600 the active systems are designated as nonsafety related syn,tems except for limited portions of the systems that provide safety related isolation functions, such as containment isolation.

In current plants, the NRC has treated many of the active systems as safety related systems. Ilowever, for AP600, the active systems are not classified as safety related systems and credit is not taken for ther,e active systems in the Chapter 15 licensing design basis accioent analyses unless their operation makes the consequences of an accident more limiting.

%c nonsafety related active systems in the AP600 provide defense in-depth functions and supplement the capability of the safety related passive systems. nus, the NRC and industry have defined a process to evaluate the importance of the nonsafety related systems and for maintaining appropriate regulatory oversight, as necessary, of these active systems in the A?600. This process of identifying tegulatory oversight on nonsafety related systems is referred to as RTNSS.

De AP600 RTNSS evaluation is consistent with the process agreed to between the industry and the NRC on May 20,1993 and documented in the draft SECY paper issued on September 7,1993.

He RTNSS process summarized in this report includes three parts:

• Identification of the signincant nonsafety related systems, structures, and components (SSCs)

• Development of specific retirLility/ availability missions for the significant nonsafety related SSCs

• Specincation of proposed regulatory treatment for each of the missions developed.

He RDISS evaluation w?s performed in the following probabilistic and deterministic areas:

• Focused probabilistic risk assessment (PRA), that is a sensitivity study removing credit for mitigation functions of all nonsafety-related systems from the baseline PRA WCAP 13856 Revision: I c V9s4w l*pr.itwol059s-I R 4 yil

4 c-

• Focused PRA initiating event frequency evaluation

• ATWS rule (10 CFR 50.62)

• Loss of all ac power rule (10 CFR 50.63)

• Post.72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> actions

• Containment performance

• Adverse interactions with the AP600 safety related systems l* Seismic considerations l The results of the RTNSS evaluation of the AP600 identined ponions of several nonsafety related l systems that are RTNSS important and should have additional regulatory oversight. These systems are l listed in Table I along with the basis for their importance.

Section 10 of this report provides the proposed regulatory oversight recommendations, including short-term availability controls where appropriate.

WCAP 13856 Revision: 1 OM984w l wpf it410598 123rw yijj

O l Table 1

SUMMARY

LIST OF INVESTMENT PROTECTION SHORT. TERM AVAILAMLITY COWROLS Systems, Structures, Components Number MODES Basis Trales (a) Operat6on (b) (c) 1.0 Instruroentation Systems 1.1 DAS ATWS Mitigation 2 I (2,4) 1.2 DAS ESF Actuation 2 1,2,3,4,5.6 (C) (2) 2.0 Plant Systems 2.1 RNS I 1.2.3 (2) 2.2 .NS . RCS Open 2 5.6 (B.C) (3)e

- 2.3 CCS . RCS Open 2 5,6 (B,C) (3) 2.4 SWS . RCS Open 2 5,6 (B,C) (3) 2.5 PCS Water Makeup . Long Term Shutdown i 1.2.3,4,54 (E) (6) 2.6 MCR Cooling . Long Term Shutdown 1 1.2.3,4.5,6 (6) 2.7 I&C Room Cooling . Long Term Shutdown 1 1.2.3.4.5,6 (6) 2.8 Hydrogen Ignitors i 1.2.5.6 (B.C) (2) 3D Electrict) Power Systems 3.1 AC Power Supplies i 1.2.3,4,5 (2) 3.2 AC Power Supplies . RCS Open (A) 5,6 (B.C) (3) 3.3 AC Power Supplies . Long Term Shutdown i 1,2,3,4,$.6 (6) 3.4 DC Power Supplies . DAS 2 1,2,3,4,5,6 (C) (2,4)

Alpha Notas:

(a) Refers to the number of trains covered by the availability controls.

(b) Refers to the MODES of plant operation where the availability controls apply.

(c) Refers to the RINSS evaluation that identified the SSC as R1NSS important dettM (A) 2 of 3 AC power supplies (2 standby diesel generators and I offsite power supply).

(B) MODE $ with RCS open, (C) MODE 6 with upper internals in place and cavity level less than full, (D) MODES 5 and 6 with the calculated core decay heat greater than 6 Mwt, Bank Notas:

(1) Focused PRA. See section 2.1, (2) PRA uncertainty, See section 2.3.

(3) . PRA initiating event frequency, See section 3.0, (4) A1WS Rule,10 CFR 50.62, See section 4.0 (5) Loss all AC power rule,10 CFR 50.63. See section 5.0 (6) Post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> actions, See section 6.0.

(7) Containment performance. See section 7.0.

(N) Adverw systems interactions. See section 8.0.

(9) Sc.5nic watiderations. See section 9.0, W ?AP-13856 - .

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1.0 INTRODUCTION

The purpose of this repon is to surnmarin the evaluation performed to detennine the significant nonsafety related systems, structures, and components (SSCs) for the AP600 and the appropriate additional regulatory oversight associated with these SSCs. His evaluation is consistent with the process agreed to twtween the industry and the NRC on May 20,1993 and documented in the draft SECY paper issued on September 7,1993 l Figure 1 1 depicts the evolution and implementation of the process documented in the draft SECY l paper to assess the importance of AP600 nonsafety related SSCs. Table 1 1 lists the nonsafety related AP600 systems evaluated in the RTNSS process.

His summary report relies on the AP600 PRA and SSAR as supponing documentation. The PRA report provides infonnation conceming the development of the PRA including methodology, ast,umptions, models, quantifications, and results. De SSAR provides information for the AP600 nonsafety related SSCs including drawings, system descriptions, and system functions.

Various AP600 nonsafety related SSCs are classified as Equipment Class D. Equipment Class D is defined by Regulatory Guide 1.26 as an intermediste, nonsafety related equipment classification, with specific AP600 criteria based on SSC functions. This equipment class is part of a comprehensive, graded-classification process used for AP600 mechanical, electrical, and I&C equipment, as described l in Section 3.2 of the SSAR.

His document summarizes the RTNSS process used to evaluate the nonufety related SSCs. He l specific detailed results, such as the focused PRA quantified results, are included in the PRA report.

De RTNSS process summarized in this report includes three pans:

• Identification of the significant nonsafety-related SSCs

• Development of specific reliability / unavailability missions for the significant nonsafety related SSCs

• Specification of proposed regu!..ory treatment for each of the missions developed.

He first step in the process is to identify the significant nonsafety related SSCs. To identify the significant nonsafety related SSCs, the AP600 nonsafety related systems were evaluated against criteria in the following probabilistic and deterministic areas:

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i PROSASILISTIC - I 1

l-* Pocused PRA sensitivity study.-  !

• Focused PRA in'tlating event frequency importance j i

DE1ERMINISTIC- j i

=* 'ATWS ru!e (10 CFR 50.62) j e- loss of all ac power mle (10 CPR 50.63) l

• Post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> actions  ;

• Containment performance  !

e Adverse interactions with the AP600 safety related systos  !

e Seismic considerations

l [-- Section 2 of this report summarises he focused PRA sensitivity study. Quantified results are included L ll < in Chapter $2 of the PRA report. Tne focused PRA calculates core damage frequency and large  :

l release frequency assuming no credit for nonsafety relatcJ SSCs to mitigate at power and shutdown events. The focused PRA results indicate that with no credit for nonsafety related SSCs in mitigating events, the AP600 meets the NRC safety goal for core damage frequency and also meets the large 4

release frequency goal. The core damage frequency is less than 1 x 10 per reactor year and the large release frequency is less than 1 x 104per reactor year. These results show that there are no i~

L l significant nonsafety related SSCs with respect to the focused PRA. However, to address uncertainties L - l in the focused PRA, portions of nor. safety related SSCs were designated as RTNSS important.

Section 3 of this report summarizes the evaluation of the importance of nonsafety related SSCs with j l respect to the focused PRA initiating event frequencies. Since the focused PRA sensitivity study was l lI performed using the bsseline PRA at power and shutdown initiating event frequencies, an evaluation [

was performed to identify those nonsafety related systems that are important to the focused PRA d l - initiating event frequencies. Section 3 concludes that during shutdown, RCS open conc.itions, based

{

on shutdown loss of offsite power and loss of decay heat removal initiating events, confirming the  ;

availability of certain functions of the following nonsafety related systems prior to entering this condition helps to reduce the potential for these two initiating events:

• : Offsite power system :  !

1 - Main ac~ power system -  ;

. Onsite standby power system- j

.* ' < Normal residual heat r-moval system  ;

e- Component cooling water tvsrem '

• - Service water system j 1lOlhe criteria for this conclusica is discussed in Section 3 and summarized in Table 3-1. i

-: 3 2 -

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Section 4 Summarires an evaluation of the functions relial upon to comply with 10 CIR 50.62 (N!WS rule). The evaluation identified the nonsafety related SSC functions that are relied upon to comply with this rule.

Section 5 summarires an evaluation of the functions relied upon to comply with 10 CTH 50.63 (las of all ac power rule). This evaluation identified no nonsafety related SSCs that are relied upon to comply with this rule.

Section 6 summarizes the evaluation of the post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> actions. Although event scenarios that result in an extended loss of both offsite and onsite ac power sources for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or longer are very unlikely, this potential is considered in the AP600 design. As part of this process, the post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> actions have been evaluated to identify installed, nonsafety related SSC functions relied upon in the l AP600 design. The evaluation summarized in Section 6 concludes that nonsafety related SSCs are relied upon to support post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> actions.

Section 7 6ummarizes the evaluation of the SSCs needed to meet the containment performance goal (SECY 93 087, issue IJ), including containment bypass, during severe accidents. The evaluation l summarired in Section 7 concludes that nonsafety related SSCs are needed.

Section 8 summarizes the evaluation of the potential for the nonsafety telated SSCs to adversely interact with the safety-related systems. As indicated in Section 8, safety related functions have been included in the design to preclude the identified interactions and no nonsafety related SSC functions are relied upon to preclude such interactions.

Section 9 summarires the evaluation of nonsafety related SSCs with respect to seismic considerations.

The evaluation summarized in Section 9 concludes that f.o nonsafety related SSCs are credited in the seismic margins analyses.

Section 10 providen concise reliability / unavailability mission statements for those nonsafety related SSC functions identified as important by the evaluations described in Sections 2 through 9. Section 10 also provides proposed regulatory oversight corresponding to each system mission. The proposed regulatory oversight includes recommended short term availability contmis. Long term availability considerations are not provided in the recommendations from this evaluation since long term availability considerations will be addressed in the plant specific implementation of the maintenance rule. -

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. r 4

l l

l l

NRC questions moweg r poeshe erseem espebenes (seCY4Hes, December 1900) 5 i,

'sen+64641s* eran noemd, leonetylne sevaret nor:_^ ., .1: eyenem leaves ter peewve pienes (Pobruary teer) >

iF Sortes of eneetnes between hdustry/NRC 4 eeveio,RTiset presses w weluene the imputenne of r:-- ^ :, ..r_ : sece ,

ir_

Ape 0098AM and mech agreement en RTN06 pmees(seer 1983) NRCleeuse ten s, n ,png RTNOS SECY sutennend to NRC paper (June 1seig (sepeoneer sees) ir

AP900 ' ;' .z._ :: . of the RTN06 process -

<r Continued Figure 11 (Sheet 1 of 2)

AP600 RTNSS Process Evolution and Implementation WCAP-13856 Revision: 1 OV984w l wpf Ib410598-l:20ru j4

i i

i RTNSS Process Implementation j V

j Cort.plete RTNSS Evaluations by Screening Nonsafety Related Systems in Table 1 1 Against RTNSS Criteria I

n Baseline PRA 10 CFR $0.62 10 CFR 50.63 Focused PRA Initiating Events ATWS Blackout Containment Adverse Systems Seismic Post 72 Ilour Performance Interaction C(r -'deratons Actions 1r Nonsafety.

Basel on RTNSS Criteria,is No m Related SSC not Nonsafety Related SSC Impoitant? "

Important to RTNSS Process Yes if r

Define R/A Missions and Propose Additional Regulatory Oversight if AP600 KrNSS Submittal

.twu hum Figure 11 (Sheet 2 of 2)

AN40 RTNSS Process Evolution and Implementation l

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s ,  ;

l Table 1 1 NONSAFETY.RELATED SYSTEMS EVALUATED IN THE AP600 RTNSS PROCESS P

Annex / Aux Building Nonradioactive Ventilation  !

Auxiliary S"vn Supply  ;

Cathodic Prt. tion Central Chilleu Water i Chemical and Volume Control r Circulating and Se vice Water Chemical Injection Circulating Water Closed Circuit TV i Communications Component Cooling Water Component and instrument Air  ;

Condensate Polishing  ;

Condensate Condenser Air Removal Condenser Tube Cleaning Containment Air Filtration  !

Containment Leak Rate Test Containment Recirculation Cooling Cooling Tower Data Display and Processing Demineralized Water Transfer and Storage Demineralized Water Treatment Diesel Generator Building Ventilation Diverse Actuation Excitation and Voltage Regulation Fire Protection Fuel llandling and Refueling Oaseous Radwaste Generator Hydrogen and C03 Oland Seal Olavity and Roof Drain Collection Grounding and Lightning Protection Health Physics and 110t Machine Shop HVAC . 1 Heater Drain Hot Water Heating Hydrogen Seal Oil incore Instrumentation Liquid Radwaste f

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i l Table 1 1 (Cont.)

NONSAFETY.RELATED SYSTEMS EVALUATED IN THE AP600 RTNSS PROCESS Main AC Power Main and Startup Feedwater Main Generation Main Steam Main Turbine and Generator Lube Oil Main Turbine Control and Diagnostics Main Turbine Mechanical llandling Mete >rological and Environrnental Monitoring Non Class IE DC and UPS Narmal Residual lleat Removal Nuclear Islar.d Nonradioactive Ventilation Onsite Standby Power Operation and Control Centers Plant Control Plant Oas Plant Lighting ,

Plant Security Potable Water Primary Sampling Pump ilouse Ventilation Radiation Monitoring Radioactive Waste Drain Radiologically Controlled Area Ventilation Radwaste Building IIVAC Raw Water Sanitary Drainage Secondary Sampling Security Lighting Seismic Monitoring Senice Water Solid Radwaste Special Monitoring Special Process lleat Tracing -

Spent Fuel Pit Cooling -

Standby Diesel and Auxiliary Boiler Fuel Oil Stator Cooling Steam Generator WCAP l??% Revision: I oW64*-tmpf;lt430598 l:20ru j.7

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l Table 1 1 (Cont.) )

l NONSAFETY.RELATED SYSTEMS EVALUATED IN THE AP600 RTNSS PROCESS Storm Drainage Transmission Switchyard and Offsite Power i Turbine Building Closed Cooling Water l Turbine Building Ventilation Turbine Island Chemical Feed Turbirie Island Vents, Drains and Relief Waste Water i

I l

i i

)

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_ . . . - _ __ . - _ _ _ _ _ _ _ _ ~ . ._

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i l2.0 FOCUSED PRA SENSITIVITY STUDY 3 i

.l De purpose of the focused PRA senshivity study is to determine if the safety related syste ns, when >

challenged, can provide sufficient capability without relianu on nonsafety related SSC mitigation  !

functions to meet the NRC safety goal guideline for core damage frequency less than ! x 10 4 events l

4

l. Nonsafety per reactor related year and tofunctions SSC mitigation meetthat a are large reliedrelease upon in the frequency focused PRA sensitivity of lessstudy than 1 x 10 evI to meet the safety goals will be assigned reliability / unavailability missions as appropriate and will be  ;

subject to additional regulatory oversight,  !

- ,' ne focused PRA sensitivity study is based on the AP600 baseline PRA. De baseline and focused _

l PRAs include an eva'.uation of both intemal and extemal events. De PRAs consider events that occur j at power, as well as during the spectrum of shutdown operations including plant cooldown, reduced - i reactor coolant system inventory (includiseg midloop), refueling, and heatup. Seismic margins are used ,

L to evaluate seismic events, -

l - For the focused PRA sensitivity study, the initiating event frequencies remain the t.ame as in the t l baseline PRA. %e mitigation functions of the nonsafety related systems are removed from the .,

l baseline PRA, and then the focused PRA sensitivity study model is quantified. If the core daman,e  ;

frequency and large release frequency calculated in the focused PRA are acceptable and no mitigation ,

credit is taken for nonsafety related SSCs, then no additional regulatory oversight is necessary for the  ;

l nonsafety related SSCs, based on the results of the focused PRA sensithity study. l l De AP600 systems classified as nonsafety related systems and that are modeled in the baseline PRA,  :

l which are not credited for mitigation purposes in the focused PRA analysis are listed in Table 21.

l Table 2-1 also contains a list of the safety.related systems that sae credited in this analysis. _i 2.1 FOCUSED PRA ANALYSIS l- he focused PRA sensitivity study is performed using the same basic methodology as the baseline >

l - PRA. he baseline event trees and fault trees are used as a starting point for the focused PRA I L sensitivity study. Since the baseline event tree and fault tree models credit nonsafety related SSCs, the l event trees and fault trees must be modified for the focused PRA sensitivity study so that only safety-l telated functions are credited. De fault tree reduction, core damage frequency quantification, accident

_l ; class quantification, and release frequency calculations are performed using the same methods as for l J the baseline PRA. De tr.ethodology for performing the focused PRA sensitivity study is described in

._l Chapter 52 of the AP600 PRA report. : -

- g1 l 722 RESULYS l L ne focused PRA' core damage frequency' and large release frequency with no credit for the nonsafety-l 2 l related mitigation functions of the nonsafety related SSCs are reported in Chapter 52 of the AP600 l

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l PRA repon. An importance analytis of the focused PRA core damage frequency was performed to l rank the initiating ennts. De results from this analysi: and a comparison with the AP600 baseline l initiating event imponance ranking can also be found in Chapter 52 of the AP600 PRA report.

De focused PRA sensitivity study assumes no credit for the nonsafety related mitigation functions of the nonsafety related SSCs. The results from it?s focused PRA analysis show that the AP600 core l damage frequency and large release frequency meet the specified safety goals with no credit for l mitigations by the nonsafety rtlated SSCs.

l 2.3 PRA UNCERTAINTY l To facilitate resolution of the RTNSS issue with the NRC, several additional nonsafety related 53Cs l are designated as RTNSS important in order to add margin to the AP600. These additional SSCs are glected using insights from the AP600 PRA to add margin, compensating for potential uncenalnties 1 identified by the NRC. Dese areas of potential unceneinty include the following:

. 'Ihennal/ hydraulic (T/II) uncertainty -- The thermal / hydraulic calculations used to justify the PRA success criteria may have uncertainty. Some low margin, high risk accident sequences have been identified for both short term and long term conditions. These low margin, high risk sequences include large LOCAs, direct vessel injection (DVI) line breaks, and long-term cooling. A d', tailed thermal / hydraulic uncenainty evaluation was performed and is documented in WCAP 14800, "AP600 PRA nermal/llydraulic Uncenainty Evaluation for Passive System Reliability." The result of the evaluation confirms that the majority of the success criteria rpecified in the AP600 PRA for passive-only accident sequences lead to successful core cooling, even wben conservatisms consistent with design basis methodology l are applied. For multiple failure accident sequences that exceed the 2200*F PCT core cooling analysis criterion using conservative assumptions, the effect on both the focused PRA sensitivity study and the baseline PRA was detennined in WCAP-14800. The effect on the PRA is small, and the focuseo PRA sensitivity study total core damage frequency and large release frequency remain witlJn the goals established in SECY-94-084.

• Equipment failure rate uncenainty - The equipment failure rates used for the IRWST check valves, squib valves, and the RCP trip breakers may have uneenainty. An uncenainty analysis was performed for equipment failure rates, human error probabilities, and initiating event frequencies. De results of the uncenainty analysis are reported in AP600 PRA Chapter St.

The result of the uncenalnty analysis confirmed that the results are within the required safety goals.

• Focused PRA sensitivity study modeling uncenainty -- The def' m ition of focused PRA may have some uncenainty, One issue is whether normally operating nonsafety related systems l such as AC power and instrument air should remain available if they make operation of WCAP !3856 Revisen: I oV984w lwpf it,010598-1.20N 22

passive safety related systems less likely. The AP600 focused PRA assumes that such systems do not continue to operate.

• Nonsafety related SSC importance in an initiating event frequency - The evaluation of the importance of nonsafety-related SSCs with respect to initiating events may have some uncertainty. One issue is what is an appropriate measure of risk imponance; is an initiating event significant if the core darnage frequency or large release frequency resulting from it is 1 percent or 10-percent of the total.

The objective of this PRA uncertainty evaluation is to determine which nonsafety related SSCs should l be identified to compensate for these PRA uncertainties. Ideally the SSC will directly compensate for the uncertainty. It is recognized that for some of these uncertainties, there are no nonsafety related SSCs that can directly compensate for these uneenainties. In such situations, margin is provided in the focused PRA by adding regulatory oversight on nonsafety related SSCs that improve the focuwd PRA results for other sequences. For example, there are no nonsafety related SSCs that can improve the focused PRA r.ensitivity study results associated with DVI line breaks. During a DVI line break, the RNS injection flow spills out the break and does not inject water into the RCS. In such a case, providing short term availability controls on a system such as the DAS for ATWS events is a way to add margin to the focused PRA sensitivity study by improving the overall focused PRA results, even though it does not add margin to DVI line break events.

The result of this PRA uncertainty evaluation is that a few additional nonsafety related SSCs are designate <l as RTNSS important in order to add margin to the AP600 design to compensate fc.r potential uncertainties. 'the nonsafety related SSCs designated by this process include the following:

l* DAS ATWS And ESF actuation (provide margin for T&li uncertainty) l l* Normal rendual heat removal system capability (provide margin for ADS / IRWST injection /

l containment recirculation valve reliability uncertainty, long term cooling T&ll uncertainty) l l* Onsite AC power supplies (provide margin for ADS /lRWST injection / containment l recirculation valve reliability certainty, long term cooling T&li uncertainty) l l* 11ydrogen ignitors (provide margin for uncertainty in hydrogen bum consequences)

Section 10 provides a list of the RTNSS important SSCs, a description of the functions they perform, and the proposed short term availability control regulatory oversight for .hese SSCs.

WCAP 13856 Revision: 1 oV964=-lipt it>01059s 120ru 23

l Table 21 l NONSAFETY.RELATED SYSTEMS AND FUNCTIONS REMOVED FROM BASELINE PRA ANALYSIS i

l + Chemical and> Volume Control System (CVS) l

• Main and Startup Feedwater System (FWS) l

• Condensate System (CDS) l +

Normal Residual Heat Removal (RNS) l + Main AC Power System (ECS)(connection from grid and diesel) l l

• Non Class IE DC and UPS System (EDS) l l

• Diverse Actuation System (DAS) I l

• Plant Control System (PLS)

-l

• Circulating _ Water System / Cooling Tower (CWS) l + Main Steam System (MSS)

• Turbine Building Closed Cooling Water l

• Central Chilled Water System (VWS) l

• Component Cooling Water System (CCS) l + Senice Water System (SWS) l + Compressed and Instrument Air Systems (CAS)

+ liydrogen Ignitors There are several nonsafety related systems that are not included in this list. To address the -

systems that are not included in the list, the following rule is applied, "Any system that relies on nonsafety related AC power for active equipment such as pumps, compressors, or fans is not credited in the analysis."

l l Safety Related Systems and Functions Credited in the Focused PRA Sensitivity Study .

l

• Passive Core Cooling System (PXSJ .

• IRWST Injection / Containment Recirculation l +

Core Makeup rank (CMT)

+ Accumulator

• Passive Residual lleat Removal

+ Automatic Depressurization l +

Passive Containment Cooling (PCS)

+

Containment isolation

-l +- Class lE DC and UPS System (IDS) l +

Protection and Safety Monitoring System (PMS)

• Steam Generator Isolation l

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3.0 ' INITIATING EVENT FREQUENCY EVALUATION An evaluation was performed to study Om importance of the nonsafety related systems to the initiating event frequencies using the calculation of shutdown and at power initiating event frequencies i:1 the l AP600 PRA.

De initiating events identified in the PRA were reviewed for the impact of nonsafety related system l unavailability. De assessment of the importance of nom.afety related systems, stnictures, an(

l components (SSCs) on initiating event frequency is based on the specific PRA methodologie, used to l calculate the initiadng event frequencies.

For the purpose of evaluating the importance of the nonsafety related SSC unavailability 19 the l calculation of initiating event frequency, the specific baseline PRA and focused PRA sentitivity study initiating events were categorized by the PRA methodology used to calculate the initiating event l frequencies. Eleven categories of initiating events were identified for at power and shutJown conditions. A brief discussion of the initiating event frequency calculational methodolt.gy for each category is pros.ded to assist in understanding the process used to determine the impo;tance of nonsafety related SSC teliability on the initiating event frequency.- The specific categories are listed below along with the section of this report that evaluates the importance of the nonstiety related SSCs:

AT POWER INITIATING EVENTS

• Section 3.1 Main Steam Line Stuck Open Safety Valve

• Section 3.2 - RCS Leak

• Section 3.3. LOCAs l* _

Section 3,4. Secondary Side Breaks

• Section 3.5 - Transients

• Section 3.6 - Anticipated Transient Without Scram

• Section 3.7 . Miscellaneous Special Initiators SHUTDOWN INITIATING EVENTS

• Section 3.8 - Shutdown LOCA

• Section 3.9. Shutdown Lot - of Offsite Power

• Section 3.10 Shutdown Loss of Decay Heat Removal l +1  ; Section 3.11_- Shutdown RCS Overdrain l De initiating events and the associated initiating event frequencies calculated in the baseline PRA and l focused PRA sensitivity study are provided in Chapter 2 of the AP600 PRA. As discussed in

l

Chapter 2 of the AP600 PRA, the initiating event frequencies were determined using several different methodologies.'

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I The evaluation of the importance of the nonsafety-related SSC unavailability to the initiating event fiequency requires identifying appropnate criteria for use in determining the importance of the nonsafety-related SSCs. The following three criteria were developed for use in the evaluation:

Criterion 1 Are nonsafety related SSCs considered in the calculation of the initiating event frequency?

Criterion 2 Does the unavailability of the nonsafety related SSCs significantly affect the calculation of the initiating event frequency?

Criterion 3 Does the initiating event significantly affect core damage frequency and large release frequency for the focused PRA7 The criteria are applied to the individual initiating events in each of the initiating event categories.

For each initiating event, if the response to any one of the three criteria is *No," then the unavailability of the nonsafety-related SSCs is not important to the calculation of initiating event frequency for the l foctased PRA. The discussions in Sections 3.1 through 3.11 include the results from applying the l screening criteria to the initiating events discussed in each section. Section 3.12 provides a summary l of the results of the evaluation and Table 31 shows the results of the criteria application. Figure 3-1 shows a diagram of the evaluation proce.ss.

The third screening criterion was developed for initiating events where nonsafety-related SSCs affect the calculation of the inLiating event frequency, but the initiating event itself is not significant to the focused PRA from the perspective of its contribution to the core damage frequency and the large l release frequency. The rationale for this criterion is tinit a change in the nonsafety-letated SSC unvailability can impact the calculated initiating event frequency, but the change does not have a significant effect on the core damage frequency and large release frequency. For the purposes of this l screening criteria, individual initiating events with a contribution to core damage frequency and large l release frequency of less than approximately 10 percent are not considered to be significant.1 AT POWER INITIATING EVENTS 3.1- MAIN STEAM LINE STUCK OPEN SAFETY VAINE A plant specific calculation was performed to determine the initiating event fa quency for spurious opening of the steam generator safety valves and power-operated relief valves. The steam generator I

Note that in Revision o. the seteemng entena definition far an tmtiator to not be a sigmficant concibutor to core dannge frequency or large release frequency was consersauvely chosen as "less than about one percent.'in the absence of any conventionally estabbshed entena. Since then, this entena was reviewed against nsk sipiricance definiuons suggested by EPRI and those that are used in maamenance rule yphcations. Moreover. It was observed that the nonsafety related system failure probabthnes are already modeled conservauvely in the AP600 PRA so that any imbanng event frequencies calculated by these models already include pessinusni it was concluded that h.aintaimng the one percent enter" tvould be unduly conservanve. From a nsk pint of view, in order to avoid unduly flagging imtiatmg ewnts for posantsal regulatory ove. sight treatmers, the screemng hmit for Cntenon 3 of 10 percent is chosen.

' )

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system,:ncluding the steam generator safety valves and the power-operated relief valve block valves-

' * ~

are safety related.t ' Dh design of the steam generator power operated relief valve is such that the valve . i

' litself and W closing function'of the actuator are safety-related, while the valve actuator opening I capability (required fofthis initiating event) and the required support functions (compressed air, plant = 1

. conuel system, de power) are nonsafety-related.ine initiating event frequency calculation implicitly _-

includes the nonsafety related component initiators that contribute to Spurious actuation of the steam j

. generator power-operated relief valve. Since this ir.itiating event considers nonsafety-related SSCs in

the calculation of the initiating event frequency, the response to Criterion 1 is "Yes.*

{

De' initiating event frequency. calculation considers the historical' data for a enous opemng of both the j steam generator safety' valves and the power opersted relief valve. De w.. al data reflect, the -

5 s

- difference in the number of safety valves per plant compared to the numier of powe -operated relief '

valves per plant. Based.on a comparison of the contributions from each type of valve, the contribudon 5 from tt.e relatively small population of power-operated relief valves is not significant when compared y 16_the contribution from the larger population of safety valves in the historical data. Since nonsafety. -

< related SSCs do not significantly affect the calculation of initiating event frequency, the response to '

Criterion 2 is *No."- An additional consideration for this event is that increased unavailability of the associated nonsafety related SSCs will reduce the reliability of the power-operated relief valves to open on demand and therefore, reduce the initiating event frequency.

l l De response to Criterion 2 is "No." Derefore, the nonsafety related SSCs associated with this event are not considered to be important with respect to their effect on this initiating event.

3.2 - RCS LEAK l For an RCS leak event, the initiating event frequency is derived from a review of historical data that' l identifies leaks that result in a RCS leak with an equivalent pipe diameter of less than 3/8-inch and the l' unavailability of the chemical and volume control system (CVS) to provide reactor coolant system l makeup. De components that are considered in the portion of the RCS leak event frequency-l calculation related to teactor coolant system leakage are safety-related. 'm chemical and volume' l control system and the associated support systems are nonsafety-related. Since this initiating event

considers nonsafety-related SSCs in the calculation of the initiating event frequency, the response to Criterion 1 is "Yes."

De effect of overall nonsafety-related system unavailability on this initiator results in a proportional

! change in the initiating event frequency, it is also possible to evaluate sensitiviry to changes in l hunavailability ofipecific equipment and r<,'nponents modeled in the chemical and volume control-system fault tree. Lhe initiating event frequency changes in proportion to the importance of the -

(specific component to the overall nonsafety-related system unavailability. Based on a comparison of

the probability.of a leak initiator with the probability for unavailability of the chemical and volume

~

control system, the nonsafety-related SSC impact on the initiating event frequency is significant. _.ince 1 WCAP 13856 = .

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the nonsafety related SSCs significantly affect the calculation of initiating event frequency, the response to Criterion 2 is "Yes."  !

l As shown in Tables 52 5 and 5212 of the PRA, this initiating event is not an important contributor to the core damage frequency and large release frequency for the focused PRA. Since this in% ting I l event does not significantly contribute to the core damage frequency and large release frequency, the response to Criterion 3 is "No." j 1

The response to Criterion 3 is "No," Therefore, the nonsafety-related chemical and volume control system SSCs and other nonsafety related SSCs (that are required for normal plant power oper? tion) associated with this event are not considered to be important with respect to their effect on this initiating event.

3.3 LOCAs l For LOCA events, the initiators are caused by piping and valve leaks and breaks end by spuriaus opening of certain sr.fety related valves such as reactor coolant system safety valves or automatic depressurization systera valves. The PRA methodology identifies the piping segments (or tube segments for tube rupture events) within the appropriate areas of the various systems and calculates the initiating event frequency based on a basic failure rate for these piping sections.

l l There are two nonsafety-related systems (normal residual heat removal system and chemical and l volume control system), identified in the LOCA events, however, all of the seem 'ns of the system that l could contain reactor coolant and potentially initiate a LOCA event use safety-related piping and l components.

In addition, these piping c.ections have redundant safety riated isolation valves that are either normally closed during plant operation or automatically closed following LOCA events. For example, the chemical and volume control system purification loop and discharge piping includes redundant letdown isolation valves and redundant containment isolation valves to isolate leaks that initiate in this piping and to prevent leakage from lines that exit containment.

The piping sections of the nonsafety-related systems included in the initiating event frequency calculation are the same sections that are designed using safety-related piping as discussed previously.

l Since these initiating events do not consider nonsafety related SSCs in the calculation of the initiating l event frequencies, the response to Criterion 1 is "No." Therefore, nonsafety-related SSCs are not l considered to be important with respect to their effect on these initiating events.

l For spurious ADS actuation events leading to a large LOCA event, the answer to the first criterion l quesuon is "Yes" since DAS is one of the means of spurious actuation. However, the answer to the l second criteria question is "No" since the contribution of DAS to spurious ADS actuation is much less l

WCAP-13856 Revision: 1 oA3984*-Ispr.itatoS98-120ru 34 i

l than that of PhtS. %us, spurious ADS actuation leading to a large LOCA is also dismissed due to tre l "No" response to Criterion 2.

l l As outlined in Chapter 2 of the PRA, the initiating event frequency for the interfacing system LOCA l event is a result of the erroneous opening of normal residual heat removal system isolation valves, due l to either hardware failure or operator error, in conjunction with the rupture of safety-related normal l residual heat removai system components due to overpressurization. Since these components are l safety related and the operator errors that contribute to this initiating event have no relationship to l nonsafety-related systems, the response to Criterion 1 is "No." Derefore, nonsafety-related SSCs are l not considered to be important with respect to their effect on the interfacing system LOCA initiating l event.

l3.4 SECONDARY SIDE BREAKS

- l For the secondary side break events, the initiators are caused by pipe leaks and breaks. Similar to the calculation for LOCA events, the PRA methodology identifies the piping segments within the appropriate areas of the various secondary systems and calculates the initiating event frequency based on a basic failure rate for these piping sections.

De initiating event frequency calculation consists of a plant specific analysis that includes pipe segments in several nonsafety-related systems that can be leak initiators for specific events.

l In the AP600 baseline PRA and focused PRA sensitivity study, two main steam line pipe break initiating events were identified. De initiating events listed below include piping segments in nonsafety related systems, as identified below:

l. Secondary Side Break - Upstream of the Main Steam Isolation Valves or Downstream of the l hiain Feedwater Isolation Valves hiain steam system hiain and startup feedwater system l* Secondary Side Breaks - Downstream of hiain Steam Isolation Valves cr Upstream of the l Main Feedwater Isolation Valves hiain steam system hiain and startup feWwater system l

l For the two nonsafety-related systems identified for the secondary side break events listed above, the initiating event frequency calculation includes both safety-related and nonsafety-related piping sections.

Since these initiating events consider nonsafety-related SSCs in the calculation of me initiating event frequencies, the response to Criterion 1 is "Yes."

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l l

l 1

l De initiating event frequencies for the main steam line break transient events are calculated only considering piping integrity for these nonsafety-related systems. Therefore, the operational unavailability of the nonsafety related systems has no impact on the initiating event frequency, assuming that piping integrity is unchanged.

De integrity of this nonsafety related piping directly affects the calculation of the initiating event frequencies. To provide conservative treatment of the nonsafety-related SSCs (piping) for these nonsafety related systems, piping integrity is assured to impact availability for the purposes of screening against this criterion. Since nonsafety-related SSCs significantly affect the calc 11ation of initiating event frequency, the response to Criterion 2 is "Yes."

l As shown in Tables 52-5 and $212 of the PRA, the main steam line break initiating events are not important contributors to the core damage frequency and large release frequency for the focused PRA l sensitivity study. Since these initiating events do not significantly contribute to the core damage l frequency or large release frequency, the response to Criterion 3 is "No." Therefore, the nonsafety-related FSCs (that are required for normal plant power operation) associated with these events are not considered to te important with respect to their effect on these initiating events.

3.5 TRANSIENTS I

The initiating event requencies for the transient events are calculated using historical failure data. The historical data for applicable initiating events is sorted into categories for calculating the initiating event frequencies for the specific initiating events. In general, the initiating event frequency is determined based upon the number of initiating events per year from the historical data. Once the historical data used in the :alculation of initiating event frequency for a specific event was identified, the historical data that is not applicable to the AP600 design for a specific initiating event is not included in the initiating event frequency calculation.

For some events, the available historical data base was used to calculate the initiating event frequency.

These events include the following:

. Core power excursion

l. Loss of RCS flow l* Loss of offsite power

l. Loss of condenser

l. Loss of main feedwater flow to both steam generators l -

Spurious safeguards actuation l

l As described in Chapter 2 of the PRA, the loss of offsite power initiating event frequency is based on l the frequency provided in Appendix A of the Advanced Light Water Reactor (ALWR) Utility l _ Requirements Document. He value provided in the ALWR Utility Requirements Document is based l nn historical data, wM is also provided.

WCAP-13856 Revision: I cA3954w-1.wpf:Ib-010598-1h 3-6

s For other initiating events,'some of the historical data was not' applicable to the AP600 design; For'. l these events, the non applicable data was removed from'the calculation of the initiating event - 'l frequencies. De events that included data that was inappropriate for the AP600 include the following:

~ -. . ,

• Transient with main feedwater flow .

-l . -

Spurious reactor trip lL '

Turbine trip

, .lI : Loss of main feedwater flow to one steam generator -

[l1 *. = Loss of main feedwater flow to both steam generators  ;

tl .-- - Total loss of main feedwater flow i

~l; ~- ' Secondary to primary power mismatch -

ll .

. For example, the loss of main feedwater event data was removed to exclude events where a plant trip :

l 9 occurred following the loss of a single main feedwater pump. His is due to the AP600 design'that [

allows for continued plant operation' following a loss of one main feedwater pump. For loss of main

'l feedwater events and secondary-to-primary power mismatch events, an adjustment was made in

. _ calculating the initiating event frequencies to account for the lower number of steam generators in l AP600. Since AP600 has two steam generators and the historical data includes data from plants that =

_ have more than two steam generators, an adjustment to the initiating event frequency is required to prevent calculation of an overly conservative initiating event frequency. ,

-l Several of the transient events have been groaped under similar transient event categories. These l transients have been grouped for quantification purposes. This is possible since the plant response is -

l identical for the group transients. For example the spurious trip and turbine trip events are grouped l under the transient with main fdlwater category, l 'Ihe initiating event frequencies for the seven initiating events listed above are impacted by the il unavailability of various nonsafety related SSCs. Chapter 2 of the PRA lists typical nonsafety related

'i SSCs and associated malfunctions, identified in the historical data, and considered in the calculation of

. -l) the AP600 initiating event frequencies for these initiators. Since the seven initiating events consider nonsafety-related SSCs in the calculation of the initiating event frequencies, the response to Criterion' I is "Yes."

The unavailability of the nonsafety related SSCs impacts the number of events documented in the historical data and therefore, contributes directly to the calculation _of the initiating event frequencies.

However, some of tt.e associated nonsafety-related SSCs are more significant than others in the

calculation of the initiating event frequencies. - Siace some nonsafety related SSCs significantly affect

~

the calculation of the initiating event frequencie*, the response to Criterion 2 is "Yes."

ll As shown in Tables 52-5 and 52-12 of the PRA, six of the seven initiating events discussed in this [

. section do not significantly contribute to the core damage frequency and large release frequency for-the focused PRA. nese events include the following:

8-

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.~r'

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a; i

0 --_

s G

n----

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tel 4 Core power excursion. i y ll% Loss of main feedwater flow to both steam generators N

!lLe i ilms of RCS flow; -

l 114ssjf main feedwater flow to one steam generator ~ L

,[lJe.l[

; Loss of condenseri

~

j

" ~

+1 Loss of offsite power; ,

I liSince these six initiating events do not significantly contribute to' the core damage frequency or large r = j l x release l frequency, .the response to. Criterion 3 is "No." "Iherefore, the nonsafery related SSCs l

l associated with theie events are not considered to be important'with respect to the effect on these six; ,

. initiating events.D l J Aishown in Table 52-5 of the PRA, the transient with main feedwater flow event does not i l / significantly contribute to the core damage frequency for the focused PRA sensitivity study. However. .

il[as shown in Table 5212, the event is within the cutoff range to the large release frequency. Since the - 4 l transient with main feedwater initiating event category contribution to the large release. frequency is .

l near the cutoff limit, the response to Criterion 3 is "Yes."

l : '!he tesponses to Criteria _1, 2, and 3 for the transient with' main feedwater initiating event are "Yes."

.i, Therefore, the nonsafety related SSCs that are required for normal at-power operation associated with j l 'this event are important with respect to the effect on this initiating event. Section 10 provides'a list of i important SSCs, the functions they perform and the proposed regulatory oversight recommendations.

[-

- 3.6 ANTICIPATED TRANSIENT WITHOUT SCRAM

- l ..- As identified'in Chapter 2 of the PRA, there are three ATWS initiating events considered i. um core l damage frequency and large release frequency calculations for the PRA. The three initiating events l are comprimi of the following:

l lle LATWS precursor without main feedwater flow  ;

l t +1 ATWS precursor with safeguards actuation ll 7 e ATWS_ precursor with main feedwater flow available

,f '

. llThe calculation of the individual ATWS initiating event frequencies is outlined in Chapter 2. As can-i lLbe seen froni the. ATWS events descriptions in Section 2.2.4, the calculation of all three of thi ATWS L l$ initiating event frequencies consider'nonsafety related SSCs. Since the ATWS initiating events '

ll* consider nonsafety related SSCs in the calculation of the initiating event fmquency,' the response to - -

'LCritenon 1 is "Yes."' ,

. = .

w l l? 'The probability of an ATWS event is the combination of the probability of the initiating events as

l Jdescribed in Section 2.2.4 of the PRA and the probability of a failure of safety-related SSCs to insert -

l3 control rodi'Ihe actual failure probability of the automatic and manual reactor trip is not included in -

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A ,

1

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?

4 - -

?

3

• llithe ATWS frequency cakulation; however, automatic and manual reactor trip are modeled as top ' -

} lI events iri the ATWS event trees, Chapter 4.1herefore, if either the automatic or manual reactor trip -

-)

~

il? are successful,' the event does not dEveldp into an ATWS event.L Since the failure'of 6fety relatedl l

~

@ SSCs determines if an initiating event develops into an ATWS event, nonsafety-related SSCs do not  !

l(significantly affect the calculation of the probability _of an ATWS; the response to Criterion 2 is "No."- , j i

Since the response to Criterion 2 is "No," the nonsafety-related i .ssocitd a e with this event are not ' j considered to be important with respect tr '7ir effect on this initiating event .

l

~

3.72 .- MISCELLANEOUS' SPECIAL INITIATORS a

. For several initiating events, plant specific fault trees were developed and evaluted for the specified [

nonsafety related systems to determine the initiating event frequencies for these events. 'Ihese

+

miscellaneous events are typically referred to as special initiators and they include the following:

lk

• 1hss of component cooling water / service water ,

i o? Im of compressed and instrument air -

l-Il Since the' plant response to the loss of service water or component cooling water is the same, the l Jinitiating event frequencies are combined under the same initiating event category.

t

^

l

- I Since these initiating events consider nonsafety related SSCs in the calculation of the mitigating event frequencies, the response to Criterion 1 is "Yes."

4

>Ihe unavailability of these nonsafety-related systems completely determines the initiating event

= frequency of the associated special initiator, For example, if the overall component cooling water system unavailability increases by a specified amount, the initiating event frequency directly increases by this same amount. Since nonsafety-related SSCs significantly affect tne calculation of these initiating event frequencies, the response to Criterion 2 is "Yes."

[ l L As shown in Tables 52 5 and 52-12 of the PRA, these initiating events are not important contributors

. to the core damage frequency and large release frequency for the focused PRA. . Since these initiating.

events do not significantly contribute to the core damage frequene or large release frequency, the response to Criterion 3 is "No."

I Since the response to Criterion 3 is'"No," the nonsafety related SSCs (the are required for normal

' l # plant power operation) associated with these events are not considered to be important with respect to

$\ their effect on these four initiating events.  ;

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SHUTDOWN INITIATING EVENTS 3.8 . SHUTDOWN LOCA For the shutdown LOCA event, as with the at power LOCA events, the initiators are caused by piping l ' leaks and breaks, with one exception. As discussed in Chapter 54, the PRA methodology identifies the pipe segments and calculates the initiating event frequency based on a basic failure rate for these pipe segments. De shutdown evaluation considers a number of additional normal residual heat removal system pipe segments in containment that are not included in the at power calculation (almost three times as many pipe segments). For the shutdown LOCA event, the normal residual heat removal-system is assumed to be the source of the LOCA. De normal residual heat removal system piping is safety related. Therefore, no nonsafety-related SSCs are considered in the calculation of the initiating event frequency from niping breaks or leaks.

An additional mechanism that contributes to the initiating event frequency is included in the calculation of the !nitiating event frequency for shutdown LOCAs. De calculation includes the potential for inadvertent operator opening of the normal residual heat removal system discharge l valve (s) to the in-containment refueling water storage tank. This mechanism reduces the reactor coolant system inventory by diverting flow from the shutdown cooling flowpath. This operator error in this initiating event mechanism has no relationship to nonsafety-related SSCs.

De piping sections for the nonsafety-related normal residual h.;at removal system included in the initiating event frequency calculation are designed using safety-related piping. The potential for the

- operator to erroneously initiate a loss of reactor coolant system inventory is not related to the unavailability of nonsafety related SSCs. Since this initiating event does not consider nonsafety-related SSCs in the calculation of the initiating event frequency, the response to Criterion 1 is "No."

l Berefore, nonsafety-rel:Md SSCs are not considered to be important with respect to their effect' on l this initiating event.

3.9 SHUTDOWN LOSS OF OFFSITE POWER A calculation was completed to determine the initiating event trequency for the loss of offsite power during shutdown conditions. The calculation uses the initiating event frequency for this event from the -

historical data with an adjustment for the length of time spent in shutdown conditions. The calculation

- considers effects from onsite nonsafety-related systems such as the transmission switchyard and the main ac power systems, as well as the offsite power system. Since this initiating event considers nonsafety related SSCs in the calculation of the initiating event frequency, the response to Criterion 1 is "Yes."

A review of the historical data used in the calculation of initiating event frequency for this event shows that the contribution from onsite nonsafety-related SSCs such as transformers, high voltage switchyard circuit breakers, or the main ac power circuit breakers are significant to the calculation of

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t N

.a i m

.~ "the'inidating ' event frequency.H Since nonsafety-related SSCs significantly affect the calculation of

~

Linitiating event frequency, the response to Criterion 2 is "Yes "

~

lfAs seen from T5ble $215 of the PRA, this event during RCS * ' 2 conditions contributes to the

~

l fcore damage frequency for the fdcused PRA.; Since this initist vent contributes to the core?

. damage frequency, the response to Criterion 3 is "Yes."' ,

The responses to CWeria 1,2,=and 3 for this initiating event are "Yes," Since.the responses to all Lthree criteria are "Yes," the _nonsafety-related SSCs (that are sequired to provide offsite power;during ;

. [! shutdown RCS drained conditions) associated with this event are important with respect to their effect -

on this initiating event.' Section 10 provides a list of important SSCs, the functions they perform, and

' the proposed regulatory; oversight recommendations.

q L 3.10 i : SHUTDOWN LOSS OF DECAY HEAT REMOVAL DA plant-specific fault tree was developed and evaluat'ed to determine the initiating event frequency for (l . two initiating events that represent a loss of decay heat removal during shutdown corditions.- De two -

l L initiating events include the loss of decay heat removal capability due to failure of the normal residual =

l heat removal system and the loss of decay heat removal capability due to failure of the component

= l cooling water or service water system. De evaluation for a loss of decay ' neat removal during I

-l: shutdown'is provided in Chapter 54 of the PRA iport. This initiating event results from the loss of a

-l? nonsafety related system normally used to provide decay heat removal during shutdown conditions.

The_nonsafety related SSCs considered in this evaluation include the normal residual heat removal l system, the component cooling water system, and the service water system. . Since this initiating event

. considers nonsafety related SSCs in the calculation of the initiating event frequency, the response to Criterion 1 is "Yes."

- The unavailability of these nonsafety-selated systems significantly affects the initiating event frequency. Since nonsafety-related SSCs significantly affect the calculation of these initiating event

. . frequencies, the response to Criterion 2 is "Yes."

l /As seen from Table 5215 of the PRA, this event is an important contributor to the core damage

' l frequency;for the focused PRA. . Since this initiating event contributes significantly to the core damage

, . l[ frequency,' the response to' Criterion 3 is "Yes." .

q De responses to Criteria 1,2, and 3 for this initiating event are "Yes." Since the responses to all

. three criteria are "Yes," the nonsafety related SSCs (that are required for decay heat removal during

i plant shutdown conditions) associated lwith this event are considered to be important with respect to L i

. their effect on this initiating event. Section 10 provides a list of important SSCs, the functions they _

1 perform, and the Nposed regulatory oversight recommendations.

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-l l/ 3.11 ; REACTOR' COOLANT SYSTEM OVERDRAIN1 ,s-q

.l 5 For the reae:or~coblant system overdrain event, two scenarios have been identified for which this 4 l f accideat could occur.' The first scenatio is a combination 'of thE failure of the hot leg level instmment -

~

- lIand operator failure to recognize that the hot leg Jevel instruments have failed.' Scenario two is the -

$ l combination of a failure of the valves CVS V045 and V047 to close on the receip* of a closure single, :

l l1 in conjunction' with an operator failure to recognize that the valves have not closed and manually

. l ? isolate % valves.l The failure mechanisms for these two scenarios are detailed in Chapter 2 of the.'

lhPRAE ,  ;

I

. l Ihe SSCs considered in the reactor coolant system overdrain event include the hot leg level -

l s instruments', valves CVS V045 and V047, and protection r.nd safety monitoring system to actuate the ~

lL componentsi Since these SSCs are safety-related and the operator errors that contribute to this

-- li initiating event have ne relationship to nonsafety-related systems, the response to C'iterion 1 is "No." q l?

~ l8 The response to Criterion 1 is "No." 1herefore, nonsafety-related SSCs are not considered to be

. l. important with respect to their effect on this initiating event.

l 3.12 -

SUMMARY

l The results of the screening criteria application for each initiating event category are provided in l Table 31. Section 10 provides a list of important SSCs, the functions they perform, and the proposed

. regulatory oversight recommendations for nonsafety related SSCs that impact the calculation of the l focused PRA sensitivity study initiating event frequencies, based on the initiating event criteria

. application,

+

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~ l,

l i

l Table 31 .l INITIATING EVENT CRITERIA APPLICATION I l

l l

Criterion ? Criterion 2 Criterion 3  ;

Are nonsafety related Does the unavailability of Does the initiating event  !

SSCs considered in the nonsafety related SSCs sig.ilficantly affect core )

the calculation of the significantly affect the damage frequency and 1 Initiating Event initiating event calculation of the initiating large release frequency Category frequency? event frequency? for the focused PRA?

l 3.1 - Main Yes No' N/A Steam Line Stuck Open Safety Valve 3.2 RCS Leak Yes Yes No 3.3 LOCAs No N/A N/A 3.4 Seco idary Yes Yes No Side Breaks 3.5 Transient Yes Yes Yes with MFW Flow Other Yes Yes No Transients 3.6 ATWS Yes No N/A 3.7 Miscellane Yes Yes No ous Special Initiators 3.8 Shutdown No N/A N/A LOCA 3.9 Shutdown Yes Yes Yes loss of offsite power.

3.10 Shutdown .Yes Yes Y4 loss of decay heat removal 3.11 Shutdown No N/A N/A RCS Overdrain g WCAP-13856 - Revision: 1 eM984w-l*p(.lM10598-1:20PM 3 13

4 Unavailatsity of Cettorion 1 the assoaisted Are non='-;i,_L^z f SSCs considered in re.1 :i related the 8=hd% of the initiating event SSCs is not 7 7__ a f? Important to the focused PRA Yes 2

U"***Ii*tdY'I Does the unevelletdlity of the nonsalsty-related esca significantly enect the -  % Seceis not

• # important to the Yes Ur c ' - ri of Cetterton 3 the m'8 Does the initiating event significantly nonealoty related atteot covo damage frequency and large SSCsis not

' ; mf er the t tooused PRA' c_; r r.: to the focused Pfu Yes Identify nw ^ ^, n' red 99Cs, proposed miselons, and proposed regulatory W Figure 3-1 Evaluation of Nonsafety related SSCs Impact on Initiating Event Frequency WCAP 13856 0* I eM984w-1 wpt;nb clo398-1.20m 3,34

3 y -

y -

.< u q- : , ,

? 4.0 ' . le CFR 50.62 (ATWS RULE)-  :

' ~

L4.1 z '.10 CFR $0.62 EVALUATION - _ q 110 CFR 50.62 sets forth the requirements for reduction of risks from anticipated transients without'- r scram (ATWS). : 1he rule requires diverse actuation of auxiliary feedwater (for decay. heat removal) l

, L and turbine trip /~Ihe AP600 design _ includes a diverse actuation system, diverse from the protection-

~

. _ and safety monitoring system, that trips the turbine and actuates passive residual heat removal, to i l1 provide decay heat rt.Jwval for AP600 'Ihe DAS also provides reactor trip.  !

~i The diverse actuation system is nonsafety-related." 1he diverse actuation system is powered by the

= non-Class IE de and UPS system. ; To support the diverse actuation system functions, a power supply -

jh must be provided! The following nonsafety related system supports the diverse actuation system functions needed to meet the requirements of 10 CFR 50.62.

-l-* . Non-class IE de and UPS system 4

Since the diverse actuation system relies upon power to actuate, the function of the non class IE dc-and UPS system to provide the diverse actuation system with power is needed to meet the L requirements of 10 CFR 50.62.' The diverse actuation system is designed to function for at least one

. . :hour following loss of HVAC. 'Ihe diverse actuation system can perform its required functions

~ l (reactor trip, turbine trip and passive residual hsat removal actuation) prior to the point where environmental conditions degrade the diverse actuation system capabilities,-_'Iherefore, environmental

- l - control ~ systems are not necessary to meet the requirements of 10 CFR 50.62, t

4.2 10 CFR $0.62 CONCLUSION -

The following nonsafety-related system functions are needed to meet the requirements of 110 CFR 50.62:

I

~ lI

• Diverse actuation system functions of reactor trip, turbine trip and passive residual heat

. l- removal actuation functions during power operation -

l7 d- Non-class IE de 'and UPS system support of the' diverse actuation' system and required l1 - component actuation associated with reactor trip, turbine trip and passive residual heat removal-

- l4 actuation functions during power operation' Ihe missions for these systems along with the corresponding proposed regulatory oversight -

recommendations are included in Section 10 of this documerit.'

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l 5.0 , 10 CFR 50.63 (LOSS OF ALL AC POWER RULE) ,

5,1 10 CFR 50.63 EVALUATION 10 CFR 50.63 sets forth~the requirements for addressing the capabilities to safely shutdown a reactor-

~

following a loss of all ac power.

l ' De'AP600 design minimizes the potential tisk contribution of station blackout by not requiring ac _  ;

power sources for design basis events. Safety related systems do not need nonsafety-related ac power sounes to perform safety related functions.

De AP600 safety related systems automatically establish and maintain safe shutdown conditions for-the plant following design basis events, including an extended loss of ac power sources, The safety-related systems can mentain these safe shutdown conditions after design basis events, without operator 1 action, following a loss of both onsite and offsite ac power sources. Derefore, no nonsafety-related

-SSCs are relied upon to establish and maintain safe shutdown conditions following a loss of all ac power for a period of up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. For an extended loss of ac power beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, Section 6 provides an evaluation of the post-72 hour actions.

5.2 10 CFR 50.63 CONCLUSION No installed nonsafety-related SSCs are relied upon to meet the requirements of 10'CFR 50.63.

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Y  ; '} .

j ll/6.0 POST.72 HOUR ACTIONSL -

,6,1 POST 72 HOUR ACTIONS EVALUATION- ,

> l)1he AP600 includes safety-related passive systems and equipment that are sufficient to automatically--

3 l 0 wablish and maintain safe shutdown conditions for the plar.t following design basis events, assuming  ;

l5 tigthe most limiting single failure occurs.- The safety related passive systems maintain safe shutdown -- 0 l? conditions after an event - without operator action, without onsite and offsite ac power sourcesc

'l; 4

li'Ihe AP600 includes nonsafety related active systems 'and equipment designed to provide multiple .

~

l11evels of defense for a wide range of events. For the more probable events, these nonsafety-related-- j

\i systems automatically actuate to provide'a first level of defense to reduce the likelihood of y ll unnecessary actuation and ' operation of the safety-related passive systems. 'Ihese nonsafety-related -

!lj systems establish and maintain safe shutdown conditions for the plant following design basis events, ll$ provided that at least one of the standby nonsafety related ac power sources is available. ,

l:  ;

l .. Although event scenarios that result in an extended loss of the nonsafety related systems or both l

'll offsite and onsite ac' power sources for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are very unlikely, this potential is 1l: considered in the AP600 design.

.l  :

-l' 'Ihe safety' functions required following an extended loss of these nonsafety-related systems include the l: following: -

l

l. Core cooling.' inventory, and reactivity control

'.- l :

• Containment cooling and ultimate heat sink l:.-

• Main control room habitability ~

l f *~ - Post accident monitoring il? * - Spent fuel pool cooling g, I?

lL In ords to support extended operation of the passive safety-related systems, the AP600 includes both l l j nonsafety-related onsite' equipment and safety-related connections for use with transportable equipment zl dand' supplies to provide the following extended support actions:

-- l c Provide electrical power to supply the post-accident and spent' fuel pool monitoring instrumen-

~

l1 tation, using the ancillary diesel generators or transportable, engine-driven ac generators that ,

l1 Lconnects to safety-related electncal connections, j l:

l Provide makeup water to the passive containment cooling water storage tank to maintain (l) *'  : external containmer.t cooling water flow, ustug a PCS recirculation pump powered by an -

, ;l( ancillary diesel generator or a transportable, engine-driven pump that connects to a safety-l~~  : related makeup connection.

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Ventilation' and cooling of the main control room, the instrumentation and control rooms, and.

l. _ the de equipment rooms is provided by open doors and ancillary fans.

l --

. ll . Provide makeup water to the spent fuel pool from the passive containment cooling water ;  ;

' lI - storage tank and from the long term makeup connection. ,

lt .

l- Dese actions are accomplished by the site' support personnel, in coordination with the main control ,

l . room operators. Dese actions are performed separate from, but in parallel with, other actions taken lj by the plant operators to directly mitigate the consequences of an event.

I

'l-6.2 Poss72 Hour Actions Conclusion i l- I li No nonsafety related SSCs are required to operate to support post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> actions.- However, in order lL- to provide margin for events that may challenge the ability to secure offsite transportable equipment N l - within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the following nonsafety-related onsite equipment should be available:

. l .l

-l' * - Ancillary diesel generator and ancillary diesel generator fuel oil storage tank l<* PCS recirculation pump and ancillary PCS water storage tank .

l..- Main control room ancillary fan .

Instrumentation room ancillary fan

ll.

l *Ihe missions for this equipment along with the contsponding proposed regulatory oversight l recommendations are included in Section 10 of this document.

1 l

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7.0' CONTAINMENT PERFORMANCE 7.1 : CONTAINMENT PERFORMANCE EVALUATION The SSCs relied upon to support containment performance assumptions in the baseline PRA were l evaluated using the MAAP code as described in Chapter 44 of the AP600 PRA report, ne following l containment performance criteria is identified in SECY 93-087 and used in the AP600 PRA:

De containment should maintain its role as a reliable, leak-tight barrier by ensuring that containment stresses do not exceed ASME ser ice level C limits for a minimum period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the onset of core damage, and that following this 24-hour period the containment should continue to provide a barrier against the uncontrolled release of fission products, ne containment performance evaluation includes consideration for the following functions:

• Reactor coolant system depressurization-

• - Passive core cooling system injection

• Containment isolation

• ' Passive containment cooling

• Ex vessel coolable geometry l The only nonsafety-related SSCs included in the evaluation of the containment integrity are the l hydrogen igniters and the reactor vessel insulation.

l ' He bounding evaluation performed in Chapter 41 of PRA for containment performance following l hydrogen combustion shows that the AP600 containment design is sufficient to meet the containment l performance criteria without credit for the nonsafety-related hydrogen igniters, in addition, the AP600 l meets the safety margin basis which examines the containment % ability to satisfy the structural l requirements in 10 CFR 50.34(f) when subjected to the pressure and temperature loads associated with l a LOCA, combired with the burning of hydrogen produced by the oxidation of 75 percent of the

_l active cladding and without credit for the hydrogen igniters.

l l In order to support in vessel retention of a damaged core, water in the containment must be able to l flow to the reactor vessel and steam (generated on the reactor vessel outside surface) must be vented l away from the reactor vessel. The AP600 reactor vessel insulation is designed to allow this to occur.

l, 7.2 ' CONTAINMENT PERFORMANCE CONCLUSION l The reactor vessel insulation design is required to support in vessel retention.

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l[~Altngh not required to meet the containment performance functions, at least one hydrogen ignitor -

~ l ;- group should'be available. -

_l lh The missions for this equipment along with the conesponding proposed regulatory oversight =

-l'. recommendations are included in Section 10 of this document.1 H

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! F.0 - . ADVERSE SYSTEMS INTERACTION  ;

'8.1- ADVERSE SYSTEMS INTERACTION EVALUATION

'l . De potential adverse systems interactions considered here are those where nonsafety-related systems l : may adversely interact with the, safety related systems. 'Ihe following three types of interactions have been addressed:

• - Functional intenctions Spatial interactions l* Human intervention interactions -

l The AP600 Adverse System Interactions Evaluation Report'(WCAP-14477) summarizes the systematic llf and through approach used to evaluate the AP600 for potential adverse system interactions. Potential

lc- sdverse systems interactions that reduce the capability or degrade the performance of the safety-related l systems to perform their safety related missions are identified in this report.

8.2 Adverse Systems Interactions Conclusion l De AP600 Adverse System Interactions Evaluation Report documents that the AP600 SSAR and the l AP600 PRA have properly considered potential adverse system interactions. As a result, no nonsafety-l related SSC's are captured by this evaluation,

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3 9.0 - SEISMIC CONSIDERATIONS 9.1 SEISMIC CONSIDERATIONS EVALUATION The seismic margins analysis used to perform the AP600 seismic evaluation does not credit nonsafety.

related SSCs. SSCs relied upon to address design basis events are designed in accordance with the AP600 seismic design criteria provided in Section 3.7 of the SSAR. -

9.2 SEISMIC CONSIDERATIONS CONCLUSION No nonsafety related SSCs are reliec upon to support the AP600 seismic margins evaluation.

1 r

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.i 10.0' MISSION STATEMENTS AND PROPOSED REGULATORY OVERSIGHT RECOMMENDATIONS 10.1 IMPORTANT NONSAFETY RELATED SSCS '

- Following are the nonsafety related SSCs that are identified as important by the evaluations summarized in Sections 2 through 9 of this report:

l '10.1.1 Focued PRA

-l- No nonsafety-related SSCs are identified as important.

l ' 10.1.2 Initiating Event Frequency l Three initiating events are identified (in' Section 3) as having nonsafety-related SSCs that are l- important.' nese events include transients with main feedwater available, shutdown ioss of offsite

~

l power and shutdown loss of decay heat removal.

l ' -- Transients with Main Feedwater Available l De evaluation of the at-power turbine trip / spurious reactor trip and loss of main feedwater l transient events identifies several nonsafety-related secondary plant systems whose continuous l operation during power production prevents plant trips. These nonsafety related systems.

l therefore, impact the at-power turbine trip / spurious reactor ip and loss of main feedwater l transient initiating event frequencies for the focused PRA. Rese nonsafety related systems l ' include the following:

l l + Main steam system i

• Maii feedwater system l

• Condensate system l -= Main turbine ,

= Main turbine control and diagnostics system l

-l' -* Plant control system portions which control main steam, main feedwater (including

-l steam generator water level control subsys. 7), condensate and main turbine whose l malfunction can cause a reactor trip l

-l: Section 10.2 provides the missions for these nonsafety-related systems. Section 10.3 provides

~

l an evaluation of the need for additional regulatory oversight based on the impact on the l - focused PRA initiating event frequencies.

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-l-- . Shutdown Losi of Offsiti Power / Loss of Decay Heat Removal LThe evaluation of the shutdown loss of offsite power and loss of decay heat removal events ildentifies several nonsafety-related SSCs whose continuous operation during shutdown, RCS > _

l i

open conditions prevents'a loss of shutdown decay heat removal.' lhese nonsafety-related -

systems, therefore, impact the shutdown loss of offsite power and loss'of decay heat removal -  ;

~ . initiating event frequencies for the. focused PRA.

l

l. J ihe following nonsafety related SSCs are identified as imponant for these two shutdown, RCS l li open initiating events:

• L Offsite power system (only portions of the system needed to provide electrical power ll -

to onsite equipment required to support decay heat removal operation during _RCS ' l lw- open conoitions) 1 l

• Onsite standby power system (only need diesel generators as a. backup source of  ;

'l. electrical power to support decay heat removal operation during RCS open conditions) -

• : Normal residual' heat removal system (only portions of the system needed to provide  ;

l- shutdown decay heat removal during RCS open condition.s)

• . Component cooling water system (only portions of the system needed to support

- normal residual heat removal system shutdown decay heat removal operation during l RCS open conditions)

- Senice water system (only ponions of the system needed to suppon component l cooling water system shutdown decay heat removal operation during RCS open i

conditions)-

Section 10.2 provides the missions for these nonsafety related systems. Section 10.3 provides an evaluation of the need for additional regulatory oversight based on the impact on the focused PRA initiating event frequencies.

- l L10.li PRA Uncertainty .

ll .

l 1he' following nonsafety-related SSCs are identified as important.

e

.l' _

Tl: --

DAS ATWS and ESF actuation (provide margin for T&H uncenainty) li ,

l Normal residual heat removal system injection (provide margin for ADS / IRWST z

l~ injection / containment recirculation valve reliability uncenainty, long term cooling

--l i T&H uncertainty)

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Onsite AC power supplies (provide margin for ADS / IRWST injection / containment

-l recirculation valve reliability uncertainty, long term cooling T&H pncertainty)_

Hydrogen !gnitors (provide margin for uncertainty in hydrogen bum consequences)

Section 10.2 provides the missions for these nonsafety related systems. Section 10.3 provides the proposed regulatory oversight r'ecommendations for these systems.

l 10.1.4 10 CFR 50.62 (ATWS Rule) l The following nonsafety related SSCs are identified as important.

l- Diverse actuation system (only portions of the system needed to provide reactor trip, turbine trip and passive residual het.t removal actuation functions during power operation) l-- -

Non class lE DC and UPS system (only portions of the system needed to support the diverse l- actuation system and required *:tuation components to provide reactor trip, turbine trip and passive residual heat removal actuation functions during power operation)

Section 10.2 provides the missions for these nonsafety-related systems. Section 10.3 provides the proposed regulatory oversight recommendations for these systems.

l 10.1.5 10 CFR 50.63 (Loss of all AC Power Rule) l No nonsafety related SSCs are identified as important.

l 10.1.6 Post 72 Hour Actions

-l l The following nonsafety related SSCs are identified as important.

l- PCS ancillary water makeup l- Main control room ancillary cooling l_- Instrumentation room ancillary cooling l- Onsite AC ancillary power supply (to supply post accident monitoring and above functions)

Section 10.2 provides the missions for these nonsafety-related systems. Section 10.3 provides the proposed regulatory oversight recommendations for these systems.

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4-l 10.1.7 - Containment Performance l De following nonsafety.related SSCs are identified as important.

l- Reactor vessel insulation (to support in. vessel' retention) l Section 10.2 provides the missions for these nonsafety related s? 8tems. Section 10.3 provides the

. l . proposed regulatory oversight recommendations for these systems.

l - 10.1.8 ' Adverse Systems Interaction l No nonsafety related SSCs are identified as important l 10.13 Seismic Considerations ~

l No nonsafety related SSCs are identified as important 10.2 MISSION STATEMENTS This section provides the mission statements for the nonsafety-related SSCs identified as important in the l evaluations summarized in Sections 2 through 9. The mission statements are grouped by type of system l (instrumentation, plant, electrical systems).

l 10.2.1 Instrumentation Systems l- Diverse Actuation System (ATWS) l The diverse actuation system provides the capability to automatically actuate reactor and turbine trip and initiate passive residual heat removal under conditions indicative of an ATWS during power operation.

l- Diverse Actuation System (ESP)

I l De diverse actuation system provides the capability to automatically actuate passive safety related features l during ' ta power and shutdown MODES.

I l 10.2.2 Plant Systems

- Miscellaneous Secondary Plant Systems Continuous operation of the following nonsafety-related secondary plant systems during power production prevents plant trips:

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k Main steam system

'* Main feedwater system

• - Condensate system

-* Main turbine Main turbine control and diagnostics system

- l--

• Plant control system ponions which control main steam, main feedwater (including steam l- generator water level control subsystem), condensate and main turbine whose malfunction l can cause'a reactor ti;p

-l- Normal Residual Heat Removal System (RCS Open)

-l De normal residual heat removal system provides shutdown decay. heat removal during RCS open

- l shutdown conditions.-

l= Component Cooling Water System (RCS Open)

The component cooling water system provides cooling to support normal residual heat removal systern l shutdown decay heat removal operation during RCS open shutdown conditions. ,

l- Service Water System (RCS Open)

The service water system providea cooling to support component cooling water system shutdown decay l ' heat removal operation during RCS open shutdown conditions.

-l - Passive Containment Cooling Water Makeup (Long Term Shutdown)

' l De PCS recirculation pumps provide the capability to transfer water from the PCS ancillary water storage l tank to the PCS water storage tank to support post 72 hou, operation of passive safety-related SSCs. Bis l capability is required when the decay heat of the core is sufficient to require PCS water evaporative l cooling.

l- Main Control R' om Cooling (Long Term Shutdown) l he MCR ancillary room fans provide cooling of the MCR to suppe.. post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> MCR habitability l during all modes of plant operation.

-l--- InstruCGM En. Cooling (Long Term Shutdown) l ne Inrtrumentation Room Fans provide cooling of the 1E instrumentation rooms to support post-72 hour l ' post accident moriitoring during all modes of plant operation.

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l- Hydrogen Ignitors l De hydrogen ignitors prevent combustion of hydrogen that may cause failure of the containment l . following a core melt.

l- Reactor Vessel Insulation

'l The reactor vessel insulation supports in-vessel retention of a moltant core during a severe accident by l allowing water from the containment remove heat fron the reactor vessel outer surface from plant l eperating or shutdown conditions.

l 10.2.3 Elect ical Systems l

l l- Onsite AC Power Sulply l

l The onsite standby power system provides a backup source of electrical power to onsite equipment needed l to provide PMS actuation and to support normal residual heat removal operation during at power and l shutdown conditions following a loss of offsite power.

l l- AC Power Supplies (RCS Open) l l he offsite power system provides electrical power to onsite equipment needed to sups. . decay heat l removal operation during RCS open shutdown conditions.

I l- AC Power Supply (Long Term Shutdown) l l The ancillary diesel generators provide power to support post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> nperation following at power and l shutdown events.

l- Non-class IE DC and UPS System (DAS) l Re non-class IE DC and UPS system provides electrical power to the diverse actuation system and l actuation components to actuate reactor and turbine trip and initiate passive residual heat removal under conditions indicative of an ATWS during power operation.

10.3 PROPOSED REGULATORY OVERSIGHT RECOMMENDATIONS l De proposed regulatory oversight recommendations are grouped in the same manor as the mission l statements (by system type). Table 101 lists the nonsafety-related SSCs that have short-term availability l controls. Table 10-2 contains the short term availability controls WCAP-13856 Revision: 1 oM984=-1.wpr;1b.oloS9s-tco 10-6

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_l[ 10.3.I' lastrt spentation Systems i

.lE* . ' Divene Actuation System (ATWS) -l l A description of the diverse actuation system is incloded in SSAR Subsection 7.7.1.11. De  ;

4l~ - AP600 D-RAP includes the DAS in SSAR Table 17.4 l.' ITAACs are pravided in Section 2.5.1.

De quality l assurance guidance provided in Generic Letter 8546 is applicable to the diverse' actuation system.

r

- l -- - Table 10 2. (item 1.1) provides recommendations for diverse actuation system short-term l- availability controls covering the ATWS function.

5 l~e: Diverse Actuation System (EFS) .

l l A~ description of the diverse actuation system is included in SS.^A Subsection 7.7.1.11. De - i l AP600 D RAP includes the DAS in SSAR Table 17.4-1. = ITAACs are provided in Section 2.5.1.

l l- Table 10-2 (item 1.2)- provides recommendations for diverse actuation system short term l availability controls covering the ESF actuation function, iL l 10.3.2 Plant Systems 7

l+ Reactor Trip Initiating Event Systems l For these initiating events, the impact of nonsafety-related SSCs on the focused PRA initiating event frequencies is identified as important, There are several factors that must be considered in evaluating the potential benefit of additional regulatory oversight to the initiating event frequencies calculated in the focused PRA. Derefore, these factors must be considered in identifying missions for these nonsafety related systems and developing-proposed additional regulatory

oversight based on these missions. De four initiating events are the following:

• _ Turbine trip / spurious reactor trip.

+ Loss of main feedwater -

For purposes of this discussion.-the turbine trip / spurious reactor trip and loss of main feedwater transients can be= grouped together as they, were in the trensient initiating event frequency- -

evaluation in Section 3.5.

-.De responses to the three.yriteria for these two initiating events indicate that some of the associated nonsafety related SSCs are significant to -the calculation- of the-initiating event .

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O frequencies and these initiating events affect the focused PRA results 11( ever, there are other considerations related to the nonsafety related SSCs that contnbute to the calculation of these initiating events that must be considered in deciding upon the benefit of additional regulatory oversight for these SSCs.

The rionsafety related SSCs contributing to the historical data for these two events in curicit J pla.its are also nonsafety-related systems and perfonn essentially the same functions for the j AP600. These nonsafety related systems include the following: l

• Main s'eam system

. Main feedwater system

• Condensate system

• Main turbine

. Main turbine control and diagnostics system

-l + Plant control system portions which control main steam, main feedwater (including steam i

l generator water level centrol subsystem), condensate and main turbine whose malfunction l can cause a reactor trip The AP600 nonsafety related systems include varices design features nnd improvements that help to increase system reliability ami availability. There is currently regulatory oversight in the design for these nonsafety related syittms and the associated design improvements, based on the descriptions of the various systems provided in the AP600 SSAR. Examples of these design improvements and the SSAR subxction that describes the trecific AP600 design features are p.uvided below:

• Digital steam generator water level system (7.7)

+ Variable speed, motor-drivsi mairt feedwater pumps (10.4.7.2.2)

• Improved main feedwater r:gulating valve throttling control features (10.4.7.2.2)

+ Elimination of main feedwater bypass control valves (10.4.7.2.2)

+ No plant trip following the loss of one main feedwater pump (10.4.7.1.2)

+ Full load irjection capability (10.4.4)

• Digital turbine electrohydraul;c convol system C0.2.2.3)

Although the nonsafety selated SSCs affect the inidating event fiequencies for these two transientt, there are several considerations in evaluating the need for and benefit from additional regula'. cry oversight for the associated nonsafety-related SSCs.

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As discussed above, the AP600 ayste.ns, including the nonsafety related systems listed above, contain numerous design improvements that incorporate current regulatory oversight provided by the NRC to address various plant safety issues. De nonsafety related systems that impact these two initiating events are required to continuously operate to support normal plant power operation.

%erefore, there is strong incentive to establish and maintain reliable system performance. By providing rnore fault tolerarit system designs that increase plant relishility and availability, the design improvements also directly increase plant safety by reducing the potential for plant transierits or trips that could present challenges to the plant.

De PRA benefit frorn these design improvements is not fully reflected from the perspective of the initiating event frequency calculation. As discuired in Section 3.5, the accepted PRA methodology for calculating the initiating event frequencies for these two transient events is to use the available historical data for these eveats, as the histoneal data is applicable to die AP600. In calculating the initiating event frequency, it is possible to adjust the calculated initiating event frequency for some design improvements. For example, t. ; calculation can ignete historical events where the loss of one main fe 4ater pump taused a plant tr4. Ilowever, there is no attempt to adjust the historical data to compentate for improvernents such as increased main feedwater pump or main feedwater control valve reliability Either the events in the historical data for a component failure are included or they are not included, without consideration of whether the AP600 component is more reliable. De calculation includes events where the loss of both main feedwater pumps causes a plant trip and these evente are not adjusted to account for increased AP600 main fetdwater pump reliability.

Based on this PRA methodology, the AP600 design improvements are not fully credited in the calculation of the initiating event frequencies for these two events, he increased reliability of the AP600 systems, when compared to the same systems in current plants, results in a conservative calculation of initiating event frequency for the ArQ0 PRA. His historical data, which conservatively bounds the calculadon of initiating event frequency for these events,is based upon the cuneat level of regulatory oversight for similar nonsafety related SSCs in cunent plants.

Derefore, additional regulatory oversight for the AP600 nonsafety related SSCs that impact these two initiating events, beyond that provided via the SSAR design details and via existing operational controls on current plants, will not provide significant benefit in reducing either the initiating event frequency, core damage frequency, or large release frequency. In addition, it is not meaningful to consider additional regulatory oversight, that is intended to increase the reliability of nonsafety related systems that are nonnally in standby operation, to nonsafety related syttems that are required to operate during powe' production.

De cunent level of regulatory oversight, including the SSAR design oversight, is sufficient to assure that the changes in unavailabilities of the nonsafety related SSCs that impact these two specific initiating events are conservatively bounded from the perspective of calculating the WCAP 13856 Revision: I ou984w.I wpr tulos98-120m 10 9

focused PRA initiating event frequency, and the resulting core damage frequency and large release frequency.

Derefore, from the perspective of initiating event frequencies for the focused PRA, the evaluation for these two initiators identify no nonsafety related SSCs where additional regulatory oversight would significantly reduce the associated init..iting event frequencies, the core damage frequency, and the large release frequency. No proposed regulatory oversight recommendations have been identified for these nonsafety related SSCs.

l* Shutdown loss of Offsite Power De shutdown loss of offsite power is only significant from the perspective of core damege l frequency and large relear.c frequency during shutdowns with the RCS open conditions. Als initiating event is not important duting any of the other shutdov.n conditions considered in the l shutdown PRA and discussed in Appendix F.4.3 of the PRA report, j herefore, the missions developed for these nt ,tafety related SSCs as a resuh of this initiating event, and any associated additional regulatory overright rec, mendations, are only applicable l during RCS open shutdown conditions, which represent a small percentage of the overall time spent in plant shutdown. j

%e responses to the three criteria for this initiating event indicate that the associated nonsafety- f related SSCs are significant to the calculation of the initiating event frequency and this initiating event affects the focused PRA results.

Dis initiating event is impacted by the loss of electrical power from the offsite grid sources, which is independent of onsite nonsafety related SSCs. De initiating esent is also impacted by nonsafety related SSCs such as the transmission switchyard system, which is site-specific and not part of the AP600 de.ign as described in the SSAR or modeled in the PRA report. However, the onsite nonsafety related SSCs contributing to the historical data for this event in current plants are ,

also nonsafety related systems and perform essentially the sann functions for the AP600. Dese nonsafety telated systems include the following:

• Offsite power system equipment including the main step-up, unit auxiliary, and reserve auxlary transformers 1

+

Main ac power system equipment including the offsite power supply circuit breakers to the onsite switchgear buses De nonsafety related SSCs identified w ve required to be in continuous operation to support shutdown plant operations during reduced reactor coolant system inventory conditions. - These

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nonsafety related SSCs prov'<de electrical power to onsite ac power system. The desigr. and operation of these nonsafety-related SSCs are described in Chapter 8 of the AP600 SSAR.  ;

l Additicnally, the operation of these nonsafety related SSCs during RCS open shutdown operations follow industry guidelines and practices, which minimize the potential for, and enhance mitigation of, this event.

Even though not specifically required for the evaluation of initiating event frequency impact, adMinnel neplatory oversight recommendations are also proposed for the nonafety-relsted diesel-gm1itors of the onsite standby power system. This initiating event frequency evaluation corisiders only nonsafety related SSCs that impact the probability of an event occuning. From this perspective, where the failure of a site transformer can cause a loss of offsite power event, failure of a nonsafety related diesel Eenerator does not initiate a loss of offsite power and availability of the diesel generators does not prevent a loss of offsite power. Diesel-generator unavailability does impact the plant response and mitigation of this event. 11owever, mitigation impact is not the intent of this evaluation.

In oc, eloping the proposed regulatory oversight recommendations for this event, the need for shon tenn availability control of the offsite power sources can benefit from credit for availability of the diesel generators. The proposed oversight recommend &ns are not onerous, considering l the normal plant actions expected in preparing for RCS opest a dtdowr operations, in addition, the proposed oversight enhances plant safety during these conditions and provides flexibility by l allowing for preventive or corrective maintenance that may need to be performed while in RCS l open shutdown conditions, such as during special plant evolutions that are required over the lifetime of the plant.

• Shutdown Loss of Decay lleat RemovJ As with the shutdown loss of offsite power, the shutdown loss of decay heat removal is only significant from the perspective of core damage frequency and large release frequency during l l shutdowns with RCS open conditions. This initiating event is not imponant during any of the i other shutdown conditions considered in the shutdown PRA and discussed in Appendix F.4.3 of the PRA report.

l Therefore, the missions developed for these nonsafety-related SSCs as a result of this initiating event, and any associated additional regulatory oversight recommendations, are only applicable l during RCS open shutdown conditions, which represent a small percentage of the overall time spent in plant shutdown.

The responses to three criteria for this initiating event indicate that the associated nonsafety related SSCs are significant to the calcula' On of the initiating event frequencies and these initiating events affect the focused PRA results.

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His initiating c: vent is impacted by the loss of nonsafety related SSCs that are used to provide core decay heat removt.! - the normal residual heat removal system and its support systems.

He AP600 includes a number of various design features and improvements that help to improve plant safety by increasing thutdown decay heat removal reliability and availability. Dere is currently regulatory oversight in the design for these nonsafety related systems and the associated design improvements, based on the descriptions of the various systems provided in the AP600 SSAR. A discussion of these design features and specific design improvements to support shutdown decay heat removal is included in Subsection 5.4.7 of the AP600 SSAR.

Dere is also regulatory oversight in the operation of the shutdown decay heat temoval systems through industry documents such as Generic Letters 8712 and 8817. Subsection 1.9.5.1 of the SSAR provides the AP600 response to the SECY 90-016 issue of midloop operation, which references the guidance provided in the generic letters.

Examples of these design improvements, which are discussed in Subsection 5.4.7 and l Appendix 1.9.5.1 of the SSAR, include the following:

+ Loop piping offset

+ llot leg level instrumentation

• Self venting pump suction line

• No residual heat removal system throttling required during midloop operation

• Capability for full normal residual heat removal system flow with saturated fluid conditions ne nonsafety related systems identified above are required to be in continuous operation to support shutdown core decay heat removal during reduced reactor coolant system inventory conditions. De nonsafety related normal residual heat removal system and its nonsafety related support systems are normally available and fully operational to support the plant cooldown prior l to entering RCS open shutdown maintenance conditions. In addition, these nonsafety-related systems are required to be available prior to initiating reduced textor coolant system inventory operations during a plant shutdown. Planned maintenance for these systems will not be scheduled l during RCS open shutdown conditions.

l Additionally, the operation of these nonsafety related SSCs during RCS open shutdown operations follow industry guidelines and practices, which minimize the potential for, and enhance mitigation of, this event.

WCAP-13856 Revision: 1 0V954=-lmrt Ib410595-I 20ru .

g().12

Normal Residual lleat Removal System A description of the normal residual heat removal system is included in SSAR Subsection 5.4.7.

l The AP600 D-RAP includes the normal residual heat removal system in SSAR Table 17.41.

l ITA.ACs are provided in Section 2.3.6.

l Table 10 2 (item 2.2) provides recommendations for normal residual heat removal system shon.

term availability controls.

+

Component Cooling Water System l A description of the component cooling water system is included in SSAR Subsection 9.2.2. De l AP600 D RAP includes the component cooling water system in SSAR Table 17.41. ITAACs are l provided in Section 2.3.1.

l Table 10 2 (item 2.3) provides recomrnendations fo* component cooling water system short-term availability controls.

+

Senice Water System l A description of the service water system is included in SSAR Subsection 9.2.1. De AP600 D.

l RAP includes the service water system in SSAR Table 17.41. ITAACs are provided in l Section 2.3.8.

l Table 10-2 (item 2.4) provides recommendations for service water system short term availability controls.

l* PCS Water Makeup (Long Term Shutdown) l Makeup to the PCS water supply post 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is provided by the PCS recirculation pumps taking

\ suction from the PCS ancillary water storage tank. A description of this arrangement is provided l in SSAR Section 6.2.2. ITAACs are provided in Section 2.2.2.

l l Dis equipment should be available following seismic and high vind events that may make l procurement of offsite equipment more difficult. As a result, this equipment is Seismic 11 as l 2hown in SSAR Table 3.2 3. Tabic 10-2 (item 2.5) provides recommendations for PCS water makeup short term availability controls.

WCAP-13856 Revision: 1 eM9s4w l wr(It@l059s 1;20ru 10 33

1 l

l* MCR Cooling (Long Term Shutdown) l MCR cooling post 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is provided by opening doors and using the MCR ancillary fans. A l description of this cooling capability is provided in SSAR Section 9.4.1. ITAACs are provided l in Section 2.7.1.

l l This equipment should be available following seismic and high wind events that may make l procurement of offsite equipment more difficult. As a result, this equipment is Seismic !! as l shown in SSAR Table 3.2 3. Table 10-2 (item 2.6) provides recommendations for MCR fan l short term availability controls.

l* Instrumentation Room Cooling (Long Term Shutdown) l Instrumentation room cooling post 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is provided by opening doors and using the l instrumentation room ancillary fans. A description of this cooling capability is provided in SSAR l Section 9.4.1. ITAACs are provided in Section 2.7.1.

l l This equipment should be available following seismic and high wind events that rnay make l procurement of offsite equipment more difficult. As a result, this equipment is Seismic !! as l shown in SS AR Table 3.2 3. Table 10 2 (item 2.7) provides recommendations for instrumentation l room fan short-term availability controls.

l* lydrogen Ignitors A description of the hydrogen ignitors is provided in SSAR Section 6.2.4. 'Ihe AP600 D-RAP l includes the hydrogen ignitors in SSAR Table 17.41. ITAACs are provided in Section 2.3.9.

Table 10-2 (item 2.8) provides recommendations for PCS water makeup short term availability controls.

l* Reactor Vessel Insulation l

l A description of the features of the reactor vessel insulation that provide in vessel retention of a l molten core are described in SSAR Section 5.3.5. ITAACs are provided in Section 2.2;3. Short-l ' term availability controls are unnecessary for this passive component.

WCAP 13856 Revision: I oV0s4w 14pr It>01059s-lZ8u l0 34

l l 10.3J Doctrical Systeens j

• AC Power Supply System [

l_ _ A descrii nion of the onsite power system is included in SSAR Subsection 8.3. De AP600 D-l RAP includes the onsite standby power system in SSAR Table 17.41. ITAACs are provided in l l Section 2.6.4.

j i

I l Table 10 2 (item 3.1) provides recommendations for onsite power system short term availability

- controls. -

I .

l*.

AC Power Supplies (F.CS Open) f i

l A description of the offsite power system is included in SSAR Subsection 8.2. A description of lf the main AC power system is included in SSAR Subsection 8.3.1. i i

l Table > 2 (item 3.2) provides recommendations for AC power supply short term availability f

' controls. l l*- AC Power Supply (Long Term Shutdown)-

I '

l De ancillary diesel generators provide power for post accident monitoring, PCS water makeup l -(recirculation pumps), MCR cooling (MCR ancillary fans), and Instrumentation room cooling t i

l (Instrumentation room ancillary fans). A description of the ancillary diesel generators is included l in SSAR Section 8.3.1. The AP600 D RAP includes the ancillary diesel generators in SSAR ,

l Table 17.41. ITAACs are provided in Section 2.6.1. ,

I  !

I l Table 10 2 (item 3.3) provides recommendations for the ancillary diesel generator short term

l availability controls.

l* AC Power Supply (DAS) l De non class IE de and UPS system provides power to the DAS. A description of the non class f l lE de and UPS system is included in SSAR Subsection 8.3.2. ITAACs are provided in l Section 2.6.2.  !

l> - Table 10 2 (item 3.4) provides recommendations for non class IE de and UPS system short term ,

availability controls. l 1

i i

WCAP 13856 ' Revision: I oues4w t wpr Ibet059s 1.2w 30 15  ;

_ . . _ . . . _ _ . . , _ . . . . - . . _ . . - - _ . . _ . ,, a . . _ _ _ n_ _ . , . . _ _ . . _ . . _ . , _ . _ _ , , _ - . c.;

,._,.a,___,~.,.

l Table 101 LIST OF INVESTMENT PRO'IT.CTION SilORT. TERM AVAILABILITY CONTROLS Systems, Stevctures, Components Number MODES Trains (a) Operation (n) 1.0 instrumentation Systems 1.1 DAS ATWS Mitigation 2 1 1.2 DAS ESF Actuation 2 1,2,3,4,5,6 (3) 2.0 Plant Systems 2.1 RNS 1 1,2,3 2.2 RNS RCS Open 2 5,6 (2,3) 2.3 CCS RCS Open 2 5,6 (2,3) 2.4 SWS RCS Open 2 5.6 (2,3) 2.5 PCS Water Mateup . Long Tenn Shutdown i 1,2,3,4,5.6 (4) 2.6 MCR Cooling Long Term Shutdown i 1,2,3,4,5,6 2.7 I&C Room Coohng . Long Term Shutdown 1 1,2,3,4,5,6 2.8 Ilydrogen Ignitors 1 1,2,5,6 (2,3) 3.0 Electrical Power Systems 3.1 AC Power Supplies 1 1.2.3,4 J 3.2 AC Power Supplies RCS Open (1) 5.6 (2.3) 3.3 AC Power Supplies . Long Term Shutdown i 1.2.3,4,5,6 3.4 DC Power Supplies DAS 2 1,2,3,4,5,6 (3)

Alpha Notest (a) Refers to the nutWr of trains covered by the availability controls.

(b) Refers to the MODES of plant operation where the availability controls apply, Nett11 (1) 2 of 3 AC power supplies (2 standby diesel generators and I offsite power supply).

(2) MODE 5 with RCS open.

(3) MODE 6 with upper internals in place and cavity level less than full.

(4) MODES 5 and 6 with the calculated core decay heat greater than 6 Mwt.

WCAP 13556 Revision: 1 om&4* t+pf it clos 98-tmu 10 16

_~

c - . .__ . _

l Table 10 2 INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS l

1.0 instrumentation Systems 1.1 Diverse Actuation System (DAS) ATWS Mitigation OPERABILITY: DAS ATWS mitigation function listed in Table 1.1-1 should be operable

< 4 APPLICABILITY: MODE 1  !

I r

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DAS ATWS Function A.I Notify (chief r.uclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

/ith one or mote required [on call attemate).

channels inoperable.

AND A.2 Restore required channels to 14 days operable status. -

B. Required Action and B.1 Submit report to (chief nuclear i day i, associated Completion officer) or [on-call alternate)

Time of Condition A not detailing interim compensatory .

met, measures, cause for inoperability,  !

and schedule for restoration to OPERABLE.

AND B.2 Document in plant records the i month justification for the actions taken .

to restort the function to OPERABLE. ,

l 1

i

- WCAP.I3856 Revision: 1

. oves 4w.1wg IHimmt:20* ~- 10-17

0 l Table 10 2 (cont)

INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS 1.0 Instrumentation Systems .

1.1_ DAS ATWS Mitigation ,

l SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY j SR 1.1.1 Perform CHANNEL CliECK on each required channel. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> SR 1.1.2 Perfonn CHANNEL OPERATIONAL TEST on each required 92 days chenel.

SR I.l.3 Perform CHANNEL CALIBRATION on each required 24 months channel.

i -_

Table 1.11, DAS ATWS Functions DAS Initiating Number Channels Setpoint Function Signal Installed Required Rod Ddve MO SO Wide 2 per SG 1 per 50 > [25,000 lb]

Set Tdp, Turbine Range Level Trip and PRHR HX Actuation WCAP 13856 Revision: I r

atws4 -tar ib-otos98 idoru 103 g

l . Table 10 2 ' cont.)

INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS -

i 1.0 Instmmentation Systems 1.1 DAS A*IWS Mitigation BASES:

De DAS ATWS mitigation function of reactor trip, turbine trip and passive residual heat removal heat exchanger (PRHR HX) actuation should be available to provide ATWS mitigation capability. This ,

function is important based on 10 CFR 50.62 (ATWS Rule) and because it provides margin in the

~ PRA sensitivity perfonned assuming no credit for nonsafety rela'ed SSCs to mitigate at power and shutdown events. De margin provided in the PRA study acumts a minimum availability of 90% for this function during the MODES of applicability, considering both maintenance unavailability and  ;

failures to actuate.

De DAS uses a 2 out of 2 logic to actuate automatic functions. When a required channelis unavailable the automatic DAS function is unavailable. SSAR section 7.7.1.11 provides additional information. De DAS channels listed in Table 1.1 1 should be available.

Automated operator mids may be used to facilitate performance of the CHANNEL CHECK. An automated tester may be used to facilitate performance of the CllANNEL OPERATIONAL TEST.

De DAS ATWC mitigation function should be available during MODE I when ATWS is a limiting event. Planned maintenance affecting this DAS function should be performed MODES 3,4,5,6; these MODES are selected because the reactor is tripped in these MODES and ATWS can not occur.

. WCAP-13856 .

Revision: I 99:4 v.it4tos9s-th 30 39

C l Table 10 2 (cont) l INVESTMENT PROTECTION SHORT-TERM AVAILABILITY CONTROLS 1.0 Instrumentation Systems 1.2 DAS Engineering Safeguards Futures Actuation (ESFA)

OPERABILITY: DAS ESFA functions listed in Table 1.21 should be operable APPLICABILITY: MODE 1,2, 3,4,5, MODE 6 with upper internals in place and cavity level less than full l

ACTIONS

] i CONDITION REQUIRED ACTION COMPLETION TIME {

A. DAS ESrA Functions A.1 Notify [ chief nuclear officer] or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with one or more required [on-call altemate]. '

channels inoperable. AND A.2 Restore required channels to i operable status. 14 days B. Required Action and B.I Submit report to [ chief nuclear I day associated Completion officer] or [on-call altemate]

Time of Condition A not detailing interim compensatory met. . measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND B.2 Document in plant records the I month justification for the actions taken to restore the function to OPERABLE.

4 i

i

.W CAP-13856 Revision: 1 i om .impt.ib410sw 1:20* '10 20

l Table 10 2 (cont.)

INYFJ!iTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS 1.0 instrumentation Systems 1.2 DAS ESFA SURVElu.ANCE REQUIREMENTS SURVEILLANCE FREQUENCY S R ~1.2.1 Perform CilANNEL CilECK on each required 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> CHANNEL.

SR 1.2.2 - Perform CHANNEL OPERATIONAL TEST on each 92 days  ;

required CllANNEL.

SR 1.2.3 Perform CHANNEL CALIBRATION on each required 24 months CHANNEL.

Table 1.2-1 DAS ESFA FUNCTIONS .

DAS Initiating Number Channels Setpoint Function Signal Installed Required PRilR HX SG Wide 2 per SG 1 per SG > [25,000 lb]

Actuation Level or '

llL Temp 1 per HL 1 per HL < [620]F CMT Actrion Pzr Level 2 2 > [7]%

and RCP t'.p Passive Cont.. Cont. Temp 2 2 < [200]F Cooling and -

Selected Cont.

-Isolation Actuation-WCAP.13856 Revision: I wo9s4.-t*pt.ibol0598-120ru 10 21

1. . . . . _ _ _ _ _ _ _._ _. ._. _ __ _ .

.lc >

Table 10 2 (cost.)

INVESTMENT PRO 1ECTION SHORT. TERM AVAILABILITY CONTROLS '

- 1.0 lastrumentation Systems l 1.2 DAS ESFA-  !

l BASES:

• !he DAS ESFA functions listed in Table 1.21 should be available to provide accident mitifation capability.; This function is important because it provides margin in the PRA sensitivity performed .

assuming no credit for nonsafety related SSCs to mitigate at power and shutdown events. The raargin-provided in the PRA swdy assumes a minimum availability of 90% for this function during the; 1 MODES of applicability, considering both maintenance unavailability and failures to actuate. -l

'Ihe DAS uses a 2 out of 2 logic to actuate automatic functions. When a required channel is' .

unavailable the automatic DAS function is unavailable. SSAR section 7.7.1.11 provides additional--

information.D1he DAS channels listed in Table 1,21 should be available.  ;

Automated operator aids may be used to facilitate performance of the CHANNEL CHECK. An I automated tester may be used to facilitate performance of the CHANNEL OPERATIONAL TEST. .

l 1he DAS ESFA mitigation functions should be available during MODES 1, 2, 3, 4, 5, 6 when  !

- accident mitigation is beneficial to the PRA results. 'The DAS ESFA should be available in MODE 6 l

. with upper intemals in place and cavity level less than full. Planned maintenance affecting dese DAS functions should be performed in MODE 6 when the refueling cavity is full; this MODE is selected because requiring DAS F.SFA are not anticipated in this MODE. j i

i k

?

+

. a A

4

I

[

- WCSP 13856 - ,

'e e 'wpr m io m-12 w - g Revision: :.1 -

] o i

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l Table 10 2 (coat.)

INVESTMENT PROTECTION SHORT TERM AVAH. ABILITY CONTROLS 2.0 Plant Systems 2.1 Nonnal Residual Heat Removal System (RNS)

OPERABILITY: One train of RNS injection should be operable APPLICABILITY: MODE 1,2,3 ACTIONS CONDITION REQIflRED ACTION COMPLETION TUdE

' A. - One required train not A.1 Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> operable. [on-call alternate).

AND A.2 Restore one train to operable status 14 days B. Required Action and B.! Submit report to [ chief nuclear 1 day ast.ociated Completion officer) or [on-call alternate)

Time not met. detailing interim compensatory measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND B.2 Document in plant records the I month justification for the actions taken to restore the function to OPERABLE.

WCAP 13856 : Revision: 1 omIwpr italoses 120* 10-23

O l Table 10 2 (cont.)

INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS 2.0 Plant Systems 2.1 RNS l SURVEILLANCE REQUIREMENTS _

SURVEILLANCE FREQUENCY SR 2.1.1 Verify that one RNS purnp develops a differential head of 92 days ,

[330] feet on recirculation flow 1

SR 2.1.2 - Verify that the following valves stroke open 92 days j RNS V0ll RNS Discharge Cont. Isolation RNS V022 RNS Suction Header Cont. Isclation RNS V023 RNS Suction from IRWST Isolation  ;

t t WCAP 13856 =

Revision: I nws4w I wpr;tbalows-im - 10-24

l Table 10 2 (cont.)

INVESTMENT PROTECTION SIIORT TERM AVAILAlllLITY CONTROLS 2.0 Plant Systems 2.1 RNS BASES:

%e RNS injection function provides a nonsafety related means of injecting IRWST water into the RCS following ADS actuations. De RNS injection function is important because it provides margin in the PRA sensitivity perfonned assuming no credit for nonsafety related SSCs to mitigate at power and shutdown events. De margin provided in the PRA study assumes a minimum availability of 90%

for this function during the MODES of applicability, considering both maintenance unavailability and failures to opera'e.

One train of RNS injection includes one RNS pump and the line from the IRWST to the RCS. Eree of the valves in the line between the IRWST and the RCS are normally closed and need to be opened to allow injection. His equl[ ment does not normally operate during MODES 1,2,3. Refer to SSAR section $.4.7 for additional information on the RNS.

The RNS injection function should te available during MODES 1. 2,3 because decay heat is higher and the need for ADS is greater.

Planned maintenance on redundant RNS SSCs should be performed during MODES 1,2,3. Such maintenance should be perfonned on an RNS SSC no', required to be available. The bases for this recommendation is that the RNS is more risk important during shutdown MODES when it is normally operatirig than during other MODES when it only provides a backup to PXS injection.

Planned maintenance on non redundant RNS valves (such as V0ll, V022. V023) should be performed to minimize the impact on their RNS injection and their containment isolation capability. Non-pressure boundary maintenance should be performed during MODE 5 with a visible pressurizer level or MODE 6 with the refueling cavi'y full. In these MODES, these valves need ta be open but they do not need to be able to close, Containment closure which is required in these MODES can be satisfied by one nonnally open operable valve. Pressure boundary maintenance can not be performed during MODES when the RNS is uwd to cool the core, therefore such maintenance should be performed during MODES 1,2,3. Since these whes tre also containment isolation valves, maintenance that rtnders the valves inoperable requires that the containment isolation valve located in series with the inoperable valve has to closed and de-activated. Be baies for this recommendation is that the RNS is more risk important during shutdown MODES when it is nonnally operating than during other MODES when it only provides a backup to PXS injection. In addition, it is not possible to perform pressure boundary maintenance of these valves during RNS operation, WCAP 15&56 Revision: 1 eM964.13pr.it410598-1.20ne 10 25

4 a

l Tame 16 2 (eemt.)

INVES1 MENT PROTECTION BHORT.1ERM AVAILABILITY CONTROLS

' 2.0 ~ Mant Systems

~ 2.2 Normal Residual Heat Removal System (RNS) . RCS Open _ _

1 OPERABILITY: . Both RNS pumps should be eperable for RCS cooling

. AvPLICABILITY: MODE $ with RCS pressure boundary open,.

MODE 6 with upper intemals in place and cavity level less than full 5 ACTIONS-

- CONDITION: REQUIRED AC110N COMPLETION TIME _ ,

-4

- A. One pump not operable. A.1 Remove plant from applicable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> MODES B. Required Action and B.1 Submit report to [ chief nuclear i day

. associated Completion officer)'or [on. call allemate)

Time not met. detailing interim compensatory measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND o B.2 Document in plant records the 1 month justification for the actions taken to restore the function to OPERABLE.

i

' k S

iWCAP-13856 . .bvision: I s- t o9ete=,t.wyr.lbelopes-1:3cm : 10 26

I l Table 10 2 (cont.)

INVENTMENT PRO 1ECTION SHORT. TERM AVAILABILITY CONTROLS ,

2.0 Plant Systems l 2.2 RNS - RCS Open i

SURVEILLANCE PEQUIREMENTS i

SURVEILLANCE FREQUENCY i SR 2.2.1 Verify that one RNS pump is in operation and that each Within I day prior tv ,

RNS pump operating individually circulates reactor entering the MODES l coolant at a flow > [900] gpm of applicability OR Verify that both RNS pumps are in operation and ,

circulating reactor coolant at a flow > [1800] spm F

i

?

.t i

.WCAP-13856 Revision: 1 wovm.. .wyr.tdatoses.ty

~ # y,,,r,- , - , - , - . . ..r . . -- . - , --

l  : Table 10 2 (cont.)

INVESTMF.NT PROTECTION SHORT TERM AVAILABILITY CONTROLS 2.0 Plant Systems i

2.2 RNS RCS Open BASES:

]

ne RNS cooling function provides a nonsafety related means to normally cool the RCS during shutdown operations (MODES 4,5,6). His RNS cooling function is important during conditions when the RCS pressure boundary is open and the refueling cavity is not flooded because it reduces '

the probability of an initiating event due to loss of RNS cooling and because it provides margin in the PRA sensitivity performed assuming no credit for nonsafety related SSCs to mitigate at power and shutdown events. %e RCS is considered open when its pressure boundary is not capable of being re-i established from the control room. %e RCS is also considered open if there is no visible level in the pressurizer. De margin provided in the PRA study assumes a minimum availability of 90% for this function during the MODES of applicability, considering both maintenance unavailability and failures to operate.

%e RNS cooling of the RCS involves the R. :3 suction line from the RCS HL, the two RNS pumps and the RNS discharge line returning to the RCS through the DVI lines. De valves located in these lines should be open prior to the plant entering these conditlons. One of the RNS pumps has to be operating: the other pump may be operating or may be in standby. Standby includes the capability of being able to be plactd into operation from the main control room. Refer to SSAR section 5.4.7 for 1 additionalinformation on the RNS.

Both RNS pumps should be available during the MODES of applicability when the loss of RNS cooling is risk important. If both RNS pumps are not available, the plant should not enter these conditions. If the plant has entered these conditions, then the plant should take action to restore system operation or leave the MODES of applicability.

Planned maintenance affecting this RNS cooling function should be performed in MODES 1,2,3 when the RNS is not normally operating. The bases for this recommendation is that the RNS is more -

risk important during shutdown MODES, especially during the MODES of applicability conditions than during other MODES when it only provides a backup to PXS injection.

WCAP 13856 Revision: I

. o*.39s4w t.wpf.It41059s t:20N 30 2g

s. r .. _ . _ _ . _ . _ - . _ _ _ - , ._ _ . _ _ _ . _ . _ .

o

'l Table 10 2 (cont.) -  !

-INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTitOLS i 2.0 Plant Systems t

2.3 Component Cooling Water System (CCS) . RCS Open OPERABILITY: Both CCS pumps should be operable for RNS cooling l APPLICABILITY: - MODE $ with RCS pressure boundary open, MODE 6 with upper intemals in place and cavity level less than full .;

i i

ACTIONS  !

CONDITION - REQUIRED ACTION COMPLETION TIME i

-A. One pump not operable. A ! Remove plant from applicable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />  ;

MODES l B. Required Action and .

B.1 Submit report to [ chief nuclear i day ,

associated Completion officer) or [on-call attemate)

Timo not met. detailing interim compensatory  ;

measures, cause for inoperability, and schedule for restoration to ,

OPERABLE. j AND ,

B.2 Document in plant records the 1 month i justification for the actions taken  ;

to restore the function to OPERABLE.

I b

F

, iWCAP 13856-- Revision: I

. o9964w.t.wyr tb-oloses 120m -

.10-29

, - ~ .,, . . , _ _ _ .. . - _. - _ . -

l Table 16 2 (cont.)

INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS 2.0' Plant Systems 2.3 CCS . RCS Open 4

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY -

S R 2.3.1 Verify that one CCS pump is in operation and each CCS Within 1 day prior to pump operating individually provides a CCS flow through entering the MODES ~ ,

one RNS heat exchanger > [2$20) gpm of applicability I OR I Verify that both CCS pumps are in operation and the CCS  !'

flow through each RNS heat exchanger is > [2$20] gpm -

i l

I b

l 4

WCAP 13856 "I I

• WW1.wgit4tosw.t m, 10 30

+

- . 4 , - - = , , , . . . , ----

3 I TaWe 16 2 (cov.)

. INVESTMENT PROTECTION #HORT.'IERM AVAILABILITY CONTROLS -  !

i

+2.0 Plant Systems' l 2.3 CCS + RCS Open j

~

BASES: ' i ne CCS cooling of the RNS HXs provides a nonsafety related :neans to normally cool the RCS  :

- dudng shutdown operations (MODES 4,5,6). , This RNS cooling function is important because it - l reduces the probability _of an initiating event due to loss of RNS cooling and because it provides _

- margin la the PRA sensitivity performed assuming no credit for nonsafety related SSCs to mitigate at.  !

power and shutdown events; he RCS is considered open when its pressure boundary is not capable  ;

of being re established from the control room. %e RCS is also considered open if there is no visible- j level in the pressuriser. %e margin provided in the PRA study assumes a minimum availability of j 90% for this function during the MODES of applicability, considering both maintenance unavailability -  !

' and failures to operate, j

- De CCS cooling of the RNS involves two CCS pumps and HXs and the CCS line to the RNS HXs.

he valves around the CCS pumps and HXs and in the lines to the RNS HXs should be open prior to the plant entedng these conditions. One of the CCS pumps and its HX has to be operating. One of i the lines to a RNS HX also has to be open. De other CCS pump and H.X may be operating or may - =;

be in standby. Standby includes the capability of being able to be placed into operation from the main  :

control room. Refer to SSAR section 9.2.2 for additional information on the CCS.  !

Both CCS pumps shot.id be available during the MODES of applicability when the loss of RNS  !

cooling is risk importar.t. . If both CCS pumps are not available, the plant should not enter these j conditions. If the plant has entered these conditions, then the plant should take action to restore both l

. CCS pumps or to leave these conditions.  !

Planne.d maintenance affecting this CCS cooling function should be performed in MODES 1,2,3 when the CCS is not supporting RNS operation.- De bases for this re:ommendatice is that the CCS is  ;

more risk important during shutdown V' DES, especially during the MODES of applicability {

conditions than during other MODES.  ;

i

l 7

i 4

I e

I 4 . WCAP 13856 .

Revision: 1 8

, 7 eussewimpr.ib4toses im , 10 31- ,

d 1

• 4 - g y - e e.-, ghe w- , . . . .m y ,p , g -g.,g-e eh , _%r - w e-r ow.- g,

l Table 10 2 (cont.)

INVESThtENT PROTECTION SilORT TERht AVAILABILITY CONTROLS 2.0 Plant Systems 2.4 Service Water System (SWS) RCS Open OPERABILITY: Both SWS pumps and cooling tower fans should be operable for CCS cooling APPLICAlllLITY: hiODE 5 with RCS pressure boundary open, hiODE 6 with upper intemals in place and cavity level less than full ACTIONS CONDITION REQUIRED ACTION COhfPLETION TIhtE A. One pump or fan not A.I Remove plant from applicable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> operable. htODES T1. Required Action and D.1 Submit report to [ chief nuclear i day associated Completion officer] or (on-call alternate)

Time not met. detailing interim compensatory measures, cause for inoperability, and schedule for restoration to OPERABl.E.

AND D.2 Document in plant records the I month justification for the actions taken to restoit the function to OPERABLE.

WCAP-13856 Revision: I e 0964= 1 mpr.lbalos9s 1;20ru lg32

a l Table 16 2 (cont.)

INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS 2.0 Plant Systems .

2.4 SWS RCS Open

,. SURVEILLANCE REQU'EMENTS SURVEILLANCE FREQUENCY SR 2.4.1 Vedfy that one SWS pump is operating and that each Within I day prior to SWS pump operating individually provides a SWS flow entedng the MODES

> [6200) spm of applicability SR 2.4.2 Operate each cooling tower fan for > 15 min Within 1 day pdor to entedng the MODES of applicability

- WCAP-13856 . Revision: I e09 4w 1.wyt.iboloses tJona 10 33

l Table 10 2 (cont.)

INVESThtENT PROTECTION SilORT.TERht AVAILABILITY CONTROLS 2.0 Plant Systems 2.4 SWS . RCS Open BASES:

%e SWS cooling of the CCS IlXs provides a nonsafety related means to normally cool the RNS IlX which cool the RCS during shutdown operations (hiODES 4,5,6). This RNS cooling function is important because it reduces the probability of an initiating event due to loss of RNS cooling and because it provides margin in the PRA sensitivity performed assaming no credit for nonsafety related SSCs to mitigate at power and shutdown events. He RCS is cr+sidered open when its pressure boundary is not capable of being re-established from the control room. The RCS is also considered open ;f there is no visible level in the pressurizer. De margin provided in the PRA study assumes a minimum availability of 90% for this function during the htODES of applicability, considering both maintenance unavailability and failures to operate.

The SWS cooling of the CCS llXs involves two SWS pumps and cooling tower fans and the SWS line to the RNS IlXs. De valves in the SWS lines should be open prior to the plant entering these conditions. One of the SWS pumps and its cooling tower fan has to be operating. The other SWS pump and cooling tower fan may be operating or may be in standby. Standby includes the capability of being able to be placed into operation from the main control room. Refer to SSAR t,ection 9.2.1 for additional information on the CCS.

Both SWS pumps and cooling tower fans should be available during the h10 DES of applicability when the loss of RNS cooling is risk important. If both SWS pumps and cooling tower fans are not available, the plant should not enter these conditions, if the plant has entered these conditions, then the plant should take action to restore both SWS pumps / fans or to leave these conditions.

Planned maintenance affecting this SWS cooling function should be performed in MODES when the SWS is not suplerting RNS operation, i.e., dunng h10 DES 1,2,3. De bases for this recommendation is that the SWS is more risk important during shutdown h10 DES, especially during the htODES of applicability conditions than during other h10 DES.

l WCAP 13856 Pevision: 1 au9:4. I wg it,-olos9s.im io.34

l Table 10 2 (cont.)

INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS 2.0 Plant Systems 2.5 Passive Containment System Water Storage Tank (PCCWST) Makeup . Long Term Shutdown OPERABILITY: Long term makeup to the PCCWST should be operable APPLICABILITY: MODES 1, 2, 3, 4 MODES 5 and 6 with calculates cora decay heat > 6 MWt.

ACTIONS CONDITION REQUIRED ACTION CCMPLETION TIME

- A. Water volume in PCS A.! Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ancillary tank less than [on-call alternate).

limit.

AND A.2 Restore volume to within limits 14 days B. One required PCS B.I Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> recirculation pump not [on-call alternate).

operable, AND B.2 Restore pump to operable status 14 days C. Required Action and C,1 Submit report to (chief nuclear i day

- associated Completion officer) or [on call attemate)

Time of Condition A, B - detailing interim compensatory not met. measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND C.2 Document in plant records the 1 month justification for the actions taken to restore the function to

. OPERABLE.

WCAP 13856 .

Revision: 1

. oV964w iwpr.lb010594120PM 10 ._-_ __-___-- __ _-__-_--_ _

. ,,-. . - . . . . _ . _~ .

'~'

a ll  ! Table 10 2 (cont)-

INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS 2.0 Plant Systems

~ 2.5 PCCWST Makeup - Long Term Shutdown -

. 1 i

1 SURVEILLANCE REQUIREMENTS-I SURVEILLANCE - FREQUENCY S R -2.5.1- Verify water volume in the PCS ancillary tank is 31 days

-> [362,000] gal. ,

SR'2.5.2 Record that the required PCS recirculation pump provides 92 days recirculation of the PCCWST at > [62) gpm.

SR 2.5.3 Verify that each PCS recirculation pump transfers 10 years

> [62] gpm from the PCS ancillary tank to the PCCWST, During this test, each PCS recirculation pump will be '

powered from a ancillary diesel.

l i

k

-z 4 WCAP 13856 - .

Revision: 1 0:09tuw l.wpf.It>010596. .20PM 1 'l0-36

{r ,

!l~ Tab 8e 10 2 (cont.)

INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS 2.0 ' Pla.7t Systems 2.5' PCCWST Makeup Long Term Shutdown BASES:

The PCS recirculation pumps provide long term shutdown support by transferring water from the PCS ancillary tank to the PCCWST, his water is used to maintr'n PCS cooling during the 3 to 7 day time .

. per od fellowing an accident. After 7 days water brought in from offsite allows the PCCWST to i

continue to provide PCS tooling and also to provide makeup to the spent fuel pit. This PCCWST makeup function is important because it supports lor.g term shutdown operation. A minimum svailability of H)% is assumed for this function during the MODES of applicability, considering both l maintenance unavailability and failures to operate.

' De PCCWST makeup function involves the use of one PCS recirculation pump, the PCS ancillary tank and the line connecting the PCS ancillary tank with the PCCWST, One PCS recirculation pump

- normally operates to recirculate the PCCWST. Refer to SSAR section 6.2.2 for additional information on the PCCWST makeup function.

The PCCWST makeup function should be available during MODES of operation when PCS cooling is required; one PCS recirculation pump and PCS ancillary tank should be available during MODES

- 1,2,3,4, and MODES S and 6 with calculated core decay heat > 6 MWt. De PCS is required to be available during these MODES; it is not required to be available in MODE 5 or 6 with calculated core decay heat < 6 MWt.

Planned maintenance affecting the required PCCWST recirculation pump should not be performed during required MODES; planned maintenance should be performed on tac redundant pump (i.e., the pump not required to be available).' Planned maintenance affecting the PCS ancillary tank that requires )ess than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to perform can be performed in any MODE of operation. Planned maintenance requiring more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> should be performed in MODES 5 or 6 when the calculated core decay heat is < 6 MWt. De bases for this recommendation is that the long-term PCS makeup is not required in this condition.

.s

~

~

.WCAP 13856 - Revision: I

_ oM984 4wpr:150tos9s-th 30 37

.=

l Table 10-2 (cont.)

. !NVESTMENT PROTECTIGX SHORT. TERM AVAILABILITY CONTROLS -

2.0 - Plant Systems -

2.6. Main Control Room (MCR) Cooling - Long Term Shutdown OPERABILITY: Long term cooling of the MCR should be operable -

APPLICABILITY: MODES 1, 2, 3, 4, 5, 6 ACTIONS _

CONDITION - REQUIRED ACTION COMPLETION TIME 1

I A. One tegttired MCR A.! Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

- ancillary fans not . [on-call altemate).-

operable. ^

l AND A.2 Restore one fan to operable status 14 days B. Required Action and B.1 Sub@ report to [ chief nuclear i day associated Completion offiv ; or [on<all attemate]

Time not met, detailing interim compensator, measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND

. B.' Document in plant records the 1 month justification for the actions taken to restore the function to OPERABLE.

i

= WCAP-13856 -

Revision: I a.u96:w.13pt: belosw120'M 10-38

,n

. . .l! Table 10 2 (cont.)-

INVESTMENT PROTECTION SHORT-TERM AVAILABILITY CONTROLS

~

2.O Plant Systems-2.6 MCR Cooling Long Term Shutdown SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 2.6.1 . Operate required MCR ancillary fan for > 15 min 92 days Sk 2.6.2 Verify that each MCR ancillary fan can provide a flow of 10 years . .

air into the MCR for >l5 min During this test, the MCR ancillary fans will be powered from the ancillary diesels.

I

(-

t WCAP-13856 Revision: 1

. o:u984 i*rf',1 bates 9s-1:20ru 10-39

I l- Table 10 2 (cont.)

' INVESTMENT PROTECTION SHORT-TERM AVAILABILITY CONTROLS- l 2.0 Plant Syster.is

' 2.6 MCR Cooling - Ing Term Shutdown -

-l BASES:

The MCR ancillary fans provide long term shutdown support by cooling the main control room. For the first three days after an accident the emergency HVAC system (VES) together with the passive ,

. heat sinks in the MCR provide cooling of the MCR. After 3 days, the MCR ancillary fans can be used to circulate ambient air through the MCR to provide cooling. The long term MCR cooling

. function should be available during all MODES of operation. This long term MCR cooling function is important because .t supports long-term shutdown operation. A minimum availability of 90% is assumed for this function during the MODES of applicability, considering both maintenance unavailability and failures to operate.

'Ihe long term MCR cooling function involves the use of a MCR ancillary fan, During SR 2.6.1 the fan will be run to verify that it operates without providing flow to the MCR. During SR 2.6.2 each fan will be connected to the MCR and operated such that they provide flow to the MCR. Refer to SSAR section 9.4.1 for additional information on the long term MCR cooling function.

One MCR ancillary fan should be available during all MODES of plant operation.- Planned maintenance should not be performed on the required MCR ancillary fan during a required MODE of operation; planned maintenance should be performed on the redundant MCR ancillary fan (i.e., the fan not required to be available) during MODES 3 or 4, MODE 5 with a visible pressurizer level or MODE 6 with the refueling cavity full; these MODES are selected because the reactor is tripped in these MODES and the risk of core damage / hydrogen generation is low.

i 1

. WCAP-13856 Revision: 1 oM984w.lapr:lt>010398-l:20m 10-40

- - . . . . . . . . . ~

_. u .~ ,_ . .

- - _ _ . - , . ~

'lh.i: p? '

T

= Table 10 2 (cont.) .. _

1

'N

INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS c ,

'i

~

2.0 Plant Systems -

[

2.7. I&C Room Cooling - Long Term Shutdown

=

q OPERABILITY: = Long term cooling of I&C rooms B_& C should be ~ operable  ;

- APPLICABILITY:1 MODES 1, 2, 3, 4, 5, 6 :

i ACTIONS CONDITION'- REQUIRED ACTION: COMPLETION TIME -

" A. - One required I&C room L ' A.1 Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> - ,

ancillary fan not operable, ' [on-call altemate]. 4 AND A.2 Restore one fan' to operable status 14 days -

B. Required ' Action and B.I. Submit report to [ chief nuclear i day [

associated Completion officer) or [on-call.altemate]

Time not met . detailing interim compensatory -

measures, cause for inoperability, <

ane schedule for restoration to OPERABLE.

AND:

B.2 Document in plant records the 1 month

.: justification for the actions taken to restore the function to OPERABLE.

9 a

i

. h.. .

WCAP.13856 . . . . .

Revision:' I s,  ; oA)es4w l#:st4tomt:20rw l . t(y l '

9

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.g - , y-

+

.y y j

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1 . : Table 16 2 (cont.) - .. q n

/ C INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS e  :

.r q

t

2.0 Plant Systems -

. 2.7. l&C Room Cooling Long Term Shutdown SURVEILLANCE REQUIREMENTS _

4 SURVEILLANCE FREQUENCY.-)_ -

.i SR ' 2.7.1 '  ; Operate required I&C room ancillary fan for > 15 min 92 days f

- SR :.2.7.2. LVerify tiist each I&C room ancillary fan can provide a: 10 years flow of air into an I&C room for >15 min' ' During this test, the I&C room ancillary fans will be powered from an -  :;

. ancillary diesel, e

k t

..: I r

. .i l

=

l

WCAP-13856 . ..-

Revision: 1

- tuossw.I,wyt.l u tM w tao m . 3().42

.'. . , . - . . . . - . . . . .:. - - , ,  ; a - .. - ,.

e ,

z.: -

k .

LTable 10 2 (cont.)" ,

= INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS- ,

' 2.0: Plani Systems.

2.7 il&C Room Cooling Long Term Shutdown - l i

BASES: -

I The I&C room' ancillary fans provide long term shutdown support by cooling'I&C rooms B & C ;

I which contain post accident instrument processing equipment. For the first three days after an accident the passive heat sinks in the I&C rooms provide cooling. After 3 days, the I&C room -  :

.  ; ancillary fans can be used to circutaw ambient air through the I&C room to provide cooling. The long :

term IAC room cooling function.should.be available during all MODES of operation.' 'Ihis long term ~

IAC room cooling fanction is important because it supports long term shutdown operation. A a l minimum availability of 90% is assumed for this function during the MODES of applicability,--

?considering both maintenance unavailability and failures to operate.

l'Ihe long term 1&C room cooling function involves the use of two I&C room ancillary fans; each fan is associated with one 1&C room (B or C). During SR 2.6.1 the required fan will be run to verify that fit operates without providing flow to the I&C room. During SR 2.6.2 each fan will be connected to

~

~ its associated 1&C room and operated such that flow is provided to the I&C room. Refer to SSAR section 9.4.1 for additional information on the long term I&C room cooling function.

- One I&C room ancillary fan should be available during all MODES of plant operation. Planned maintenance should not be performed on the required I&C room ancillary fan during a required --

• MODE of operation; planned maintenance should be performed on the redundant I&C room ancillary fan.

5 A 4 7

li l

~

4 4

> !_WCaP 13856 ~ - . Revision: ~ -

*09:4w,1.wyr.lt>01059s th c 10 # '

- ,,-o.e, .-, e e - , , vn- mn e w ~~, e- +g en

. , . . - . . - . - ~. , - - . . .~

)N Table 10 2 (cont.)i

^ ~

INVESTMF.NT PROTECTION SHORT. TERM AVAILABILITY CONTROLS ~

m.

/2.0: Plant Systemst 2.8 Hydrogen Ignitorsi

.z

- OPERABILITY: '

'Ihe hydrogen ignitors listed in Table 2.8-1 should be operable APPLICABILITY: - MODE 1, 2, ~ ._ .!

' - MODE 5 with RCS pressure boundary open,;

iMODE 6 with upper intemals in place and cavity' level less than full 4

-c ACTIONS - -l CONDITION REQUIRED ACTION COMPLETION TIME l

. . . i A. One or more required A.I Notify (chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> -

hydrogen ignitor .. (on-call altemate).-

inoperable.

AND-  !

A.2 Restore required ignitors to - 14 days 1 operable status.-  ;

B.- Required Action and B.1 Submit report to [ chief nuclear i day ,

associated Completion -officer) or (on-call altemate)

Time of Condition A not detailing interim compensatory met. measures, cause for inoperability, and schedule for restoration to OPERABLE.-

AND B.2 : Document in plant accords the I month justification for the actions teken to restore the function to

' OPERABLE.

~

WCAP-13856 Revision:- l~

- orws4..t.wyt.ibelos96-120m - Ig.g ;

?L

. 3 7...

- g 5l Table 10 2 (coat.) -

'1 INVESTMENT PROTECTION SHORT - TERM AVAILAL- TY CONTROLS

~2.0 Plant Systems  ;

2.8 Hydrogen Ignitors -  !

s ,

- SURVEILLANCE REQUIREMENTS

- SURVEILLANCE ' FREQUENCY - .

SR 2.8.1 Energize each required hydrogen ignitor and verify the surface Each refueling outage..

tempera *ure is > [1700] . .

I s

4 4

I t-4 4

WCAP-13856 . Revision: I n:0964 l.wyt:lb-01059s-t;.w 10 45 ,

O l- Table 10 2 (cont.)

INVESTMENT I'ROTECTION SHORT. TERM AVAILABILITY CONTROLS Table 2.81, hydrogen Ignitors (1)-

Location Hydrogen Ignitors Group 1 Group 2 Reactor Cavity note 2 note 2 Loop Compartment #1 12,13 11,14.

Loop Compartment #2 5,8 6,7 Pressurizer Compartment 49,60 50,59 Tunnel connecting Loop Compartments 1,3,31 2,4,30 Southeast Valve Room & Southeast 21 20 I Accumulator Room ,

East Valve Room. Northeast Accumulator 18 17,19 Room, & Northeast Valve Room North CVS Equipment Room 34 33 Lower Compartment Area 22,27,28,29,31,32 23,24,25,26,30 (CMT and Valve Area)

IRWST 9,35,37 10,36,38 IRWST inlet 16 15 Refueling Cavity 55,58 56,57 Upper Compartment

- Lower Region 39,42,43,44,47 40,41,45,46,47

- Mid Region -51,54 52,53

- - Upper Region 61,63 62,64 Notes:

1) 'Ihe tatic lists the hydrogen ignitors. In each location, all of the ignitors in Group 1 or Group 2 should be u wilable. It is not nec,:ssary for all of the available ignitors to be in one group.

2) Ignitors in this lu:ation are shared with other locations.

~ WCAP-13856 Revision: 1 oM984w.l.wpf.lb410$98-120m 10 46

+ l lI - Table 10 2 (cont.)

INVESTMENT PROTECTION SHOPT. TERM AVAILABILITY CONTROLS 2 0 Plant Systems 2.8 Hydrogen Ignitors BASES: ,

The hydrogen ignitors should be available to provide the capability of burning hydrogen generated during severe accidents in order to prevent failure of the containment due to hydrogen detonation.

- These hydrogen ignitors are required by 10 CFR 50.34 to limit the buildup of hydrogen to less than 10% assuming that 100% of the active zircaloy fuel cladding is oxidized.

~

This function is also important because it prodoes margin in the PRA sensitivity performed assuming no credit for nonsafety related SSCs to mitigate at-power and shutdown events. The margin provided in the PRA study assumes a minimum availability of 90% for this function during the MOD 2J of applicability, considering both maintenance unavailability and failures to operate.

The ignitors are distributed in the containment to limit the buildup of hydrogen in local areas. Two groaps of ignitors are provided in each area; one of which is sufficient to limit the buildup of hydrogen. When an ignhor is energized, the ign; tor surface heats up to 2 [1700] F. This temperature is sufficient to ignite hydrogen in the vicinity of the ignitor when the lower flammability limit is reached. SSAR section 6.2.4 provides additional information.

The hydrogen ignitor function should be available during MODES 1 and 2 when core decay heat is high and during MODES 5 when the RCS pressure boundary is open and in MODE 6 when the refueling cavity is not full. - Planned maintenance should be performeJ on hydrogen ignitors when they are not required to meet this availability control.

s WCAP 13856 ' Revision: 1 o \3984*-1.wpf.lM10596-1:20PM 10-47 i

A l Table 10-2 (cont.) '

INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS 3.0 Electrical Power Systems i 3.1 AC Power Supplies -

' OPERABILITY: - One standby diesel generator should be operable

' APPLICABILITY: . MODES 1, 2, 3, 4, 5 -

ACrlONS CONDITION ' REQUIRED ACTION COMPLETION TIME A. Fuel volume in one A.1 Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> required standA 4 1iesel - [on-call alternate).

fuel tank less than limit. -

I AND l

A.2 Restore volume to within limits 14 days l l

B. One recuired fuel transfer B.1 Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> pump or standby diesel - [on-call attemate).

generator not operable.

AND l B.2 Restore pump and diesel generator 14 days to operable status C. Required Action and C.1 Submit report to [ chief nuclear I day associated Completion officer) or [on-call attemee]

Time not met detailing interim compensatory measures, cause for inoperability and schedule for restoration to OPERABLE.

AND C.2 Document in plant records the 1 month ,

justification for the actions taken to restore the function to OPERABLE.

WCAP 13356 Revision: 1 eM9s4w l. pf;nb.ot0598-1:_w - 3 o.4g

.. - . . . - ~ -

'I

~

. Table 10 2 (cont.)

' INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS -

7 3.0 : Electrical Power Systems 3.l AC Power Supplies SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENC's SR 3.1.1 Verify that the fuel oil volume in the required standby diesel 31 days-generator fuel tank is > [50,000] gal.

SR - 3.1.2 Record' that the required fuel oil transfer pump provides a ' 92 days recirculation flow of > [8] gpm.

SR 3.1.3 Verify that the required standby diesel generator starts and- 92 days operates at > [3800) kw for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This test may utilize diesel engine prelube prior to starting and a warmup period  ;

prior to loading.

SR 3.1.4 Verify that each standby diesel generator starts and operates at 10 years

> [3800] kw for > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This test may utilize diesel-engine prelube prior to starting and a warmup period prior to loading. Both diesel generators will be operated at the same 7 tirne during this test.

WCAP-13856 - - ,. Revision: I eM964w.1*pr ibolos98-icom . 10 49

l Table 10 2 (cont.)

INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS 3.0 Electrical Power Systems :

3.1 AC Power Supplies BASES:

AC power is required to power the RNS and to provide a nonsafety-related means of supplying power to the safety related PMS for actuation and post accident monitoring. De RNS provides a nonsafety-related means to inject water into the RCS following ADS actuations in MODES 1,2,3,4 (when steam generators cool the RCS). His AC power supply function is important because it adds margin to the ,

i

PRA sensitivity performed assuming no credit for nonsafety-related SSCs to mitigate at-power and-shutdown events. De margin provided in the PRA study assumes a minimum availability of 90% for this function during the MODES of applicability, considering both maintenance unavailability and failures to operate.

. Two standby diesel generators are provided. Each standby diesel generator has its own fuel oil transfer pump and fuel oil tank. De volume of fuel oil required is that volume that is above the connection to the fuel oil transfer pump. Refer to SSAR section 8.3.1 for additional information.

This AC power supply function should be available during MODES 1,2,3,4,5 when RNS injection end' PMS actuation are more risk imponant. Planned maintenance should not be performed on required AC power supply SSCs during a required MODE of operation; planned maintenance should be performed on redundant AC power supply SSCs during MODES 1,2,3 when the RNS is not normally in operation. The bases for this recommendation is that the AC power is more risk important during shutdown MODES, especially when the RCS is open as defined in availability control 2.2, than during other MODES.

- WCAP 13856 Revision: 1 o:u964 twpt tt4tos98- :20PM 10 50

. -. y  ;

. _ _ _ . . ._ .m.____ ,

< = ,

g I~ ' l5

- Tame 10 2 (coat.)  :

i i INVESTMENT PROTECTION SHORT TERM ' AVAILABILITY _ CONTROLS ..

j

,j 13.0 ' Elostrical Power Synems ; _;

L -

2

?3.O AC Power Supplies RCS Open j e

OPERABILITY:: . Two AC power supplies should be operable to support RN5 operation :

APPLICABILITY: MODE $ with RCS pressure boundary open, I MODE 6 with pyytt intemals in place 'and cavity level less than full . , _

ACTIONS l CONDITION.  ; REQUIRED ACTION COMPLETION TIME

'A. . One required AC power A.1 Remove plant from applicable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

= supply not operable.E . MODES B, . Required Action and B.! Sr* ' woort to [c5ief nuclear i day.

associated Completion-  : o' s pn-call altemate]

Time not met, det. . interim compensatory measures, cause for inoperability.

and schedule for restoration to OPSRABLE. -

e AND- .

B.2 Document in plant records the . I month justification for the actioas taken

' to testore the function to

-OPERABLE. 'l s Y

{

s 1

4 c

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INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS :

3.b: Electrical Power Sptems

~

3.2' AC Power Supplies RCS Open. ,

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! SURVEILLANCE REQUIREMENTS

-SURVEILLANCE - FREQUENCY.-

SR : 3.2.1 Verify that the required number of AC power supplies are Within 1. day prior to- 3

operable entering the MODES'  : j of applicability i 4

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, Table 10 2 (cont.)

INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS 3.0 Electrical Power Systems 3.2 AC Power Supplies - RCS Open

. BASES:

AC power is required to power the RNS and its required support systems (CCS & SWS); the RNS provides a nonsafety related means to normally cool the RCS during shutdown operations. This RNS cooling fun: tion is important when the RCS pressure boundary is open and the refueling cavity is not flooded because it reduces the probability of an initiating event due to loss of RNS cooling during these conditions and because it provides margin in the PRA sensitivity performed assuming no credit for nonsafety related SSCs to mitigate at power and shutdown events. Th: RCS is considered open when its pressure boundary is not capable of being re-established from the control room. The RCS is also considered open if there is no visible level in the pressurizer. De n.argin provided in the PRA-

study assumes a minimum availability of 90% for this function during the MODES of applicability, considering both maintenance unavailability and failures to operate.

Two AC power supplies, one offsite and one onsite supply, should be available as follows:

a) Offsite power through the transmission switchyard and either the main step-up transformer

/ unit auxiliary transformer or the reserve auxiliary transformer supply from the

- transmission switchyard, and b) Onsite power from one of the two standby diesel generators.

Refer to SSAR section 8.3.1 for additional infonnation on the standby diesel generators. Refer to SSAR section 8.2 for information on the offsite AC power supply.

One offsite and one onsite AC power supply should be available during the MODES of applicability when the loss of RNS cooling is important. If both of these AC power supplies are not available, the plant should not enter these conditions. If the plant has already entered these conditions, then the plant should take action to re> tone this AC power supply function or to leave these conditions.

Planned maintenance should not be performed on required AC power supply SSCs, Planned maintenance affecting the standby diesel generators should be performed in MODES 1,2,3 when the RNS is not normally in operation. Planned maintenance of the other AC power supply should be performed in MODES 2,3, or MODE 6 with the refueling cavity full. He bases for this recommendation is that the AC power is more risk important during shutdown MODES, especially

'during the MODES of applicability conditions than during other MODES.

WCAP-13856 Revision

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Table 10 2 (cont.)

INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS '

3'.01 Electrical Power Systems -- a 3.31 AC Power Supplici- Long Term Shutdown OPERABILITY:) One' ancillary diesel generator should be operable APPLICABILITY: MODES 1, 2, 3, 4, 5, 6 '

- ACTIONS" )

-CONDITION 1 REQUIRED ACTION- COMPLETION TIME .

A. Fuel' volume.in ancillary A.1 Notify [ chief nuclear officer) or 72 he its diesel fuel tank less than [on call alternate].

? limit. .

AND .;

A.2 Restore volume to within limits - '14 days {

. B. One required ancillary - -B.! Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> diesel generator not [on call attemate). .

. operable. .

AND-B.2. Restore one diesel generator to 14 days ,

, operable status

. C.. Required Action and C ! Submit report to [ chief nuclear l' day associated Completion officer) or [on-call alternate] .

- Time not met, detailing interim compensatory measures, cause for inoperability, t and schedule for restoration to OPERABLE.

AND C.2 Document in plant records the 1 month justification for the actions taken to restore the function to

-OPERABLE.

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INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS 3.0' Electrical Power Systems 3.3 AC Power Supplies - kng Term Shutdown ,

SliRVEILLANCE REQUIREMENTS

' SURVEILLANCE FREQUENCY f

l lSR _ 3.3.l! Verify fuel volume in the ancillary fuel tank is >[450] gal 31 days SR 3.3.2 Verify that the required diesel generator starts and operates for 92 days l >l hour connected to a test load > [24] kw.- Ris test may utilize diesel engine warmup period prior to loading.

SR 3.3.3 Verify that each diesel generator starts and operates for 4 10 years hours while providing power to the regulating transformer, an ancillary control room fan, an ancillary I&C room fan and a passive containment cooling water storage tank recirculation pump that it will power in a long term post accident condition.

Test loads will be applied to the output of the regulating transformers that represent the loads required for post accident monitoring and control room lighting, his test may utilize .~

diesel engine warmup prior to loading. Both diesel generators will be operated at the same time during this. test.

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[ Table 10 2 (cont.)

INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS

'3.0 Electrical Power Systems 3.3 ' AC Power Supplies - 1xmg Term Shutdown BASES:

l%e ancillary diesel generators provide long term power supplies for post accident monitoring, MCR

]

and 1&C room cooling, PCS and spent fuel water makeup. For the first three dsys after an accident

]

the 'E batteries provide power for post accident monitoring. Passive heat sinks provide cooling of the '

MCR and the I&C rooms. The initial water supply in the PCCWST provides for at least 3 days of PCS cooling. = The initial water volume in the spent fuel pit normally provides for 7 ' days of rpent fuel -

cooling; in some shutdown events the PCCWST is used to suppiement the spent fuel pit. A minimum

- availability of 90% is assumed for this function during the MODES of applicability, considering both maintenance unavailability and failures to operate.

After 3 days, ancillary diesel generators can be used to power the MCR and I&C room ancillary fans, the PCS recirculation pumps and MCR lighting. In this time frame, the PCCWST provides water makeup to both the PCS and the spent fuel pit. An ancillary generator should be available during all MODES of operation. This long term AC power supply function is important because it supports long-term shutdown operation.

The long term AC power supply function involves the use of two ancillary diesel generators and an ancillary diesel generator fuel oil storage tank. Refer to SSAR section 8.3.1 for additional information on the long term AC prever supply function.

One ancillary diesel generator and the ancillary diesel generator fuel oil storage tank should be available during all MODES of plant operation. Planned maintenance should not be performed on the required ancillary diesel generator during a required MODE of operation.: planned maintenance should be performed on .he redundant ancillary diesel generator. Planned maintenance affecting the ancillary diesel fuel tank that requires less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to perform can be performed in any MODE of operr. ion. Planned maintenance requiring more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> should be performed in MODE 6 with the refueling cavity full. De basis for this recommendation is that core decay heat is low and the risk of core damage is low in these MODES he inventory of the refueling cavity results in slow response

. of the plant to accidents.

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l1 4 Table 10 2 (c at.)

-INVESTMENT PR'OTECTION SHORT TERM AVAILABILITY CONTROLS :l 3.0 = Electrical Power Sjstems : Ei 3.4 6DC Power Supplies - i LOPERABILITY: - Power for DA3 ' automatic actuation functions listed in 1.1 and 1.fiould be operable'; .,

P APPLICABILITYi MODES 1, 2, 3, 4, 5,-

-- MODE 6 with upper internals in place and cavity level less than full j ,

2 ACTIONS -

CONDITION REQUIRED ACTION - COMPLETION TIME - ,

i TA. Power to DAS Function - - A.1 Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />  !

inoperable. . [on-call altemate).

i ANDx A.2 Restore power supply to DAS to 14 days ,

operable status B.- Required Action and - B.1 Submit report to [ chief nuclear i day associated Completion officer) or [on-call altemate]

2 -Time of Condition A not detailing interim compensatory met, measures, cause for inopere. 3ty, and schedule for restoration to OPERABLE

-AND B.2 Document in plant records the 1 month justification for the actions taken

_ to restore the function to -

OPERABLE.

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,  ? WCAP-13856 : .

. Revision: 1

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4 l - Table 10 2 (cont.)-

INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS 3.0 Electrical Power Systems 3.'4 ' Non Class IE DC and UPS System (EDS)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR~ ~3.4.1 Verify power supply voltage at each DAS cabinet is 120 92 days -

volts i 5% 1 1

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,W CAP 13856 Revision: 1 a:\3984*-l.wpf.lM10598-lh 10-58

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Tame' 10 2 (cont.)

INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS i

3,0_ Electrical Power Systems 3.41Non Class IE DC and UPS System (EDS)

BASES:

The EDS function of providing power to DAS to support ATWS mitigation is importrant based 'on 10 CFR 50.62 (ATWS Rule) and to support ESFA is important based on providing margin in the PRA sensitivity performed assuming no credit for nonsafety related SSCs to mitigate at-power and shutdown events. The margin provided in the PRA rtudy assumes a minimum availability of 90% for

. this function during the MODES of applicability, considering both maintenance unavailability and ,

failures to operate, i De DAS uses a 2 out of 2 logic to actuate automatic functions. EDS power must be available to the DAS sensors, DAS actuation, and the devices which control the actuated components. Power may be provided by EDS to DAS by non.1E batteries through non lE inverters. Other means of providing -

powe' to DAS include the spare battery through a non-1E inverter or non lE regulating transformers.

The EDS support of the DAS ATWS mitigation function is required during MODE 1 when ATWS is a limiting event and during MODES 1,2,3,4,5,6 when ESFA is important. The DAS ESFA is required in MODE 6 with upper internals in place and cavity level less than full. Planned maintenance should not be pe formed on a required EDS SSC during a required MODE of operation; planned maintenance should be performed on redundant supplies of EDS power

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