ML17296A440

Forwards TMI-2 Lessons Learned Implementation Rept.

ML17296A440
Person / Time
Site:	Palo Verde
Issue date:	09/28/1979
From:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17296A441	List: ML17296A440 ML19249E367
References
ANPP-13905-JMA1, NUDOCS 7910010222
Download: ML17296A440 (526)

Text

REGULATORY INFORMATION DISTRIBUTION SYb ZM "'('RIDS) r ACCESSION NBR:7910010222 DOC ~ DATE: 79/09/28 NOTARIZED: YES DOCKET FACIl.;STN 50-528 Pal o Verde Nuclear Station~ Uni t l i Arizona Publ i 05000528 STN 50-529 Palo Verde Nuclear Stations Unit 2~ Arizona Publ i 05000529 STN-50-530 Palo Yerde Nuclear Station~ Unit 3i Arizona Publi 05000530 AUTH, NAME AUTHOR AFFILIATION VAN BRUNT,E,E, Arizona Public Service Co.

RECIP ~ NAME, RECIPiENl AFFILIATION Office of Nuclear Reactor Regulation

SUBJECT:

Forwards "TMI-2 Lessons Learned Implementation Rept,"

DISTRIBUTION CODEX'001B COPIES RECEIVED:LTR g ENCL ~ SIZE:

~ ~9 TITLE:

6 PSAR/FSAR Amdts, 8, Related Correspondence 79/Q/3/C>W~Z4 ES + reeyesw~eeeeyepreetw~mmew~~mte~mn/ar RECIPIENT COPIES yae7~meemyeee~mmwe~ee~wee~mmeo~wmw~reoewerwm~aeeww>>woomp RECIPIENT COPIES ID CODE/NAME CTTR ENCL ID LODE/NAME LTTR ENCL ACTION; 05 PM K. 'L/C/ 7 ER" AD 1 0 BC C~4 4S 1 0 LA 1 0 INTERNAL: 0 RE

' II E 1 1 02 NRC PDR 1 06 2 2 08 OPERA LIC uR 1 1 09 GEOSCIEN BR 10 QAB 1 1 11 MECH ENG BR 1 1 12 STRUC ENG UR 13 MATL ENG BR 2 15 REAC SYS BR 1 1 16 ANALYSIS BR 1 1 17 CORE PERP HR 1 1 18 AUX SYS BR 1 1 19 CONTAIN SYS 1 20 I L C SYS BH 1 1 21 PONER SYS BR 1 1 22 AD SITE TECFl 1 0 26 ACCDNT ANLYS 1 1 27 EFFl.=. TRT SYb i 1 28 RAD ASMT Bh 1 1 29 KIRKiOOD 1 1 AD FOR ENG 1 0 AD PLANT SYS 1 0 AD REAC SAFETY 1 0 AD SITE ANLYSI8 1 0 DIRECTOR NRR 1 0 HYDRO METEOR BA 2  ? MPA 0 GELD 1 0 EXTERNAL: 03 LPDR 1 1 04 NSIC 1 1 30 ACRS 10 10 Plt'S L6-

/WC Wod WE-z ~g)g e/.s ~~ 8'c 7P' R. G/c8sEr E~e ~~ ~A' In'CT 2 1979 NUMBER OF COPIES REQUIRED: LTTR W ENCL y5'OTAL

a p P. 0 BOX 2I666 PHOENIX'RIZONA 85036 September 28, 1979 ANPP-13905 - JMA/DBK Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Coranission Washington, D.C. 20555 Re: Palo Verde Nuclear Generating Station Units 1, 2 and 3 Docket Nos. STN 50-528/529/530

Dear Sir:

As a result of the Three Mile Island accident, Arizona Public Service Company (APS) is in the process of assuring that the lessons learned from TMI-2 are reflected in the Palo Verde Nuclear Generating Station (PVNGS) design and operation.

To accomplish this objective, the APS President and Chief Executive Officer, by a directive issued April 12, 1979, established a Safety Evaluation Task Force "to conduct a complete review and analysis of the Three Mile Island accident and its implications for the Palo Verde Nuclear Generating Station project." This task force is com-prised of fourteen (14) members from knowledgeable, high levels of management of the utilities participating in the Palo Verde project, the engineer-constructor for the project, the nuclear steam supply system supplier, and turbine generator supplier to the project and the academic technical consultants to the Arizona State Senate and House legislative committees organized to review and consider the implications of the TMI accident.

The task force is exercising oversight of'he on-going independent in-depth reviews, initiated since the TMI accident, of the Palo Verde design, planned administrative and operational procedures, training and qualification programs, and emergency response plans. In this con-nection, it is closely following the work and recommendations of the .

many industry and governmental organizations currently conducting similar, but more generic investigations related to the TMI accident, including particularly the NRC's Lessons Learned Task Force.

79yo 010 zz

n I

Director of Nuclear Reactor Regulation September 28, 1979 ANPP-13905 - JMA/DBK Page 2 Upon receipt of the TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations (NUREG-0578), modified as set forth in the memorandum dated August 20, 1979 fr om Harold R. Denton, Director, Office of Nuclear Reactor Regulation to the NRC Commissioners, the recommendations related to the design and operation of the Palo Verde project were evaluated. The results of this evaluation are incorporated in the report entitled TMI-2 Lessons Learned Implementation Report submitted herewith. The report identifies those NUREG-0578 recommendations which are incorporated into the present Palo Verde design, identifies recommendations that are not applicable to the Palo Verde design and sets forth APS'ommitments to develop the analyses, design modifications, operational procedures and other measures necessary to meet such recommendations in accordance with the schedule specified in NUREG-0578. The report will be amended as appropriate to incorporate the manner in which such commitments are implemented.

Five (5) copies of this report are enclosed for your review. It is hoped that this report, in conjunction with the PVNGS Final Safety Analysis Report to be submitted in October, 1979, will assist in your timely review of the implications of TMI on PVNGS.

Respectfully submitted, ARIZONA PUBLIC SERVICE COM By:

~

~~ n~

E wean E. Van Brunt, Jr.

Vice President STATE OF ARIZONA )

) ss. On its own behalf and as agent for County of Maricopa all other joint participants.

s s

Subscribed and sworn to before me this~/day of , 1979.

My Commission expires:

otary Pu sc Nil immission Expims /an. gs, esse s7

"' ll \0 I 1

~ ~

~ n

~ I (

f 'k I r I 1 .Cp ',-"

II

~ II

~ rI P ~ ~ ~

'g ~a 0

1 I

HE( ULA ruH v iwFvR(ih io>> DISTR Idu ( IO'u SvSTEi( (BIDS) 1 ACCF6S I ()~;>>de - 791 O010222 UUC irATE: 79/09/28

~ NOTARIZED: YBS  !)()CKE FACII; S I Palo verae nuclear Stations unit Jf arizona Puoli l'500')528 rv 50-52 i alo verae iuuclear Station, unit 2r Arizona Puoli 05000529 0 a V Palo veroe nuclear Station~ Unit 3~ Arizona Puoli OS')00530 A() TH,"(4 "E A()Tc(vH At FILI ATION<

VA') ')HvivT g E ~ E ~ Arizona Puolic Service Co ~ tat f EC IP. uAHE RECI H I Er~ AF F ILIAT TU 1 1 Office of i'iuc ear Reactor 1 ~e~gui at ion SuH J EC T: FOr>ar uS "T'> I "2 LeSSOnS Lear nea .Imp emen 1 ) On Rept ~"

nI 8 I'RIUUTIUiu CODE: d00lb ('Ol'IES RECEI VEO: L TITLE: PSALM(/FS,(R Amdts 5 Pe a ted Corresponaence

~ 1 rg E')CL ~ SIZE: Q+'~++

i'u UTES:

REC IP I E uT';0F'IES REC IP I E.ar COPIES It) cuOE/<" Ai"iE (.1 ri< E<lcL I D CUD'./lA'lt, LTIH EilCL ACTIU<': 0g P(" E'. HC-/ 7 RR" ZCcfg I 0 LC Q~g AF 0 1 0 C.~g 4 8 I 0 Ii')TEHiiAL: 01 I I ' 02 hHC vugg I I Ob a ~ e. V<3 Oi hHA LIC tsar 1 1 09 GEOSCIc,'1 UR 4 1(1 AAd 1 I i)EC4 EivG 1 1 12 ST<vC Eric i)H 1 I I 3:NATL F(uG 15 HEAc,svs a>< 1 I 1 o A>uALYS IS I 17 COHc. PEHF 4H 1 l8 AUX SYS 6H I I 19 CU)~i I hli'i SYS 1 20 I. C C SYS dry I 21 PU>uER SYS t.'K I I 22 4() S I TE TEC:l I 26 4CCD") 1 A "~Lv S 1 27 Ef FL r((T s vs 1 1 HAU AS.'t r I 1 29 w I i(K.)U()0 1 1 A() r OH (riG 1 0

>>i) PLA "i l SYS 0 A>> REAC SAFt.rv I 0 AD S I I E (i>uLYS I S 1 V UI~EC'I UH ~Hu 1 0 1>>v()H(.'-~ETEUH d< 2 i~ PA 1 0 UEL() I 0 Ex TE~(uAL: 03 LPDR I 1 0< fuS IC 1 1 30 ACl;S IJ 10

)c'/C. ES

/1/ WC wad HE.

cap R 7-6W 8 6'/c, 8<E 7 EPB wM oct 2 $ 79 T() r AL r<< "idE>~ ()F CUP IES HEGV lijD: Lr ri ~ E uCL

o EFEPlHM(Q OWKMMW P. O. BOX 2I666 PHOENIX'RIZONA BSO36 September 28, 1979 ANPP-13905 - JMA/DBK Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Re: Palo Verde Nuclear Generating Station Units 1, 2 and 3 Docket Nos. STN 50-528/529/530

Dear Sir:

As a result of the Three Mile Island accident, Arizona Public Service Company (APS) is in the process of assuring that the lessons learned from TMI-2 are reflected in the Palo Verde Nuclear Generating Station (PVNGS) design and operation.

To accomplish this objective, the APS President and Chief Executive Officer, by a directive issued April 12, 1979, established a Safety Evaluation Task Force "to conduct a complete review and analysis of the Three Nile Island accident and its implications for the Palo Verde Nuclear Generating Station project." This task force is com-prised of fourteen (14) members from knowledgeable, high levels of management of the utilities participating in the Palo Verde project, the engineer-constructor for the project, the nuclear steam supply system supplier, and turbine generator supplier to the project and the academic technical consultants to the Arizona State Senate and House legislative committees organized to review and consider the implications of the TMI accident.

The task force is exercising oversight of the on-going independent in-depth reviews, initiated since the TNI accident, of the Palo Verde design, planned administrative and operational procedures, training and qualification programs', and emergency response plans. In this con-nection, it is closely following the work and recommendations of the many industry and governmental organizations currently conducting similar, but more generic investigations related to the TMI accident, including particularly the NRC's Lessons Learned Task Force.

2.

7910 010 z2

Director gf Nuclear Reactor Regulation September 28, 1979 ANPP-13905 - JMA/DBK Page 2 Upon receipt of the TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations (NUREG-0578), modified as set forth in the memorandum dated August 20, 1979 from Harold R. Denton, Director, Office of Nuclear Reactor Regulation to the NRC Commissioners, the recommendations related to the design and operation of the Palo Verde project were evaluated. The results of this evaluation are incorporated in the report entitled TMI-2 Lessons Learned Implementation Report submitted herewith. The report identifies those NUREG-0578 recommendations which are incorporated into the present Palo Verde design, identifies recommendations that are not applicable to the Palo Verde design and sets forth APS'ommitments to develop the analyses, design modifications, operational procedures and other measures necessary to meet such recomnendations in accordance with the schedule specified in NUREG-0578. The r'eport will be amended as appropriate to incorporate the manner in which such commitments are implemented.

Five (5) copies of this report are enclosed for your review. It is hoped that this report, in conjunction with the PVNGS Final Safety Analysis Report to be submitted in October, 1979, will assist in your timely review of the implic'ations of TMI on PVNGS.

Respectfully submitted, ARIZONA PUBLIC SERVICE COM By:

Edunn E. Van Brunt, Jr.

Vice President STATE OF ARIZONA ss. On its own behalf and as agent for County of Maricopa all other joint participants.

Svbscribed and sworn to before me this~day of , 1979.

My Commission expires:

otary Pu sc XIII ibmlllissign fsgimg iso. SS, Igss

l LOS)

• HE GUi A r UH Y IN F URMA r li)~> I) I S TH I HU I I OC4 S YS TE ~~ (R ACGESSL()'u hr3H -'9 1001'0222 OUC ~ L)ATE '9/09/2)) i'4O l'AHLZEO! YES  !)s)cwE r FACIL: S I tu 50 "52d Pal " o Verce uuc ear Station 1 Uni t 1 ~r i zona Pvo I i

~ t)5000528 Si'~- , lo verae (uuclear Stations'U~it 2i arizona Puo)i A!~T,, '"

S .-50-530 l'o veroe iuuc)ear Station< unit 31 Arizona Puo>i AUTr)uH Ai FILIATION 05000529 05')00530

'VAN ~)Huul',E,E ~ Af iZOna PuOI iC Ser ViCe CO.

REC IP. uAHE REC I P IE u l AF F I L I AT IU')

Oft ice of r.uc ear Heac tnr He~gu I at i on 1

SUHJECT: Forwards "l<<l-2 Lessons Learnea Lmpleeentatlon Rept."

nISTR f JT LO>u

~3 (OUE'001b TITLE: PSALM(/FS4R L0/ LES HECE I VEO: LTH Amdts.

g E)')CL

'elated'.'orresponaence

~ SIZE'+'~++

N'OTES'M'WQOWO\W WWMWP1MM'WW'W W 0 'WOOt%M WMMWOOW Me MMP\ W %0 WM'WWWWMW REC I P IE uT CUP IES RECLPLEur CO)' ES I ') CUOE/f'ARE w I l H E)uCL 10 CUUE/')A:1E. LTrR E~CL Ac T I ~)i;; 05 P)" 2'. LID./7 RR AO 6 CcJC '

/Cud 4Q

C 1 0 in LmR <e 1 1 0 IVI TER)i A i: 1 1 ~2 t)HC Pva L 1 06 x a E 08 OPERA LLC UH 1 l 09 GEOSCIc,') UH 4 10 qAd 1 1

>>EC~ Eius 1 1 12 ST~UC E:)L'<H 1 L 3

'<<'rL E')G ')H l5 HEAC SYS ak 1 1 Lo Ai')ALYSLS 1 17 CURE PEH)'H 1 1 18 AUX SYS 6H 1 1 19 CU~"> I A Ii'i SYS 1 20 I A C SYS dH 1 21 PU>u4H SYS 1 1 22 AO SITE TEC.< 1 0 26 ACCQ~>l ANLVS 1 1 27 EFi-L rHT sYs 1 1 RAV 1 1 29 afar,iUOO 1 1 Ao F0H t;iver) 1 0 A0 PLAht'YS 0 Ao REAc sAFErv 1 0 S I r E AiuLYS IS 1 0 UIHRC rUK luHH 1 0

)< Y i)Hl!->E,TEOH dH h(P A 0 OELO 1 0 EATS,Rl'~AL: 03 L~UR 1 1 V4 iuSLC 30 ACHS 10 LO P/c, E's

/L/WC WddRE.

76< 8 R. G/c.ds/7 E~e ~~ ~8 ggy R 578 TL) l Ai i'~)'i:~@EH<<UF C".P I ES iEUV LHE,O: L r TH W Ei~ci

C~

I<

P a IISI lul P. O. BOX

~ 0 CIOHIMIHW 2I666 'HOENIXi ARIZONA 85036 September ANPP-13905 28, 1979

- JMA/DSK Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Coranission Washington, D.C. 20555 Re: Palo Verde Nuclear Generating Station Units 1, 2 and 3 Docket Nos. STN 50-528/529/530

Dear Sir:

As a result of the Three Mile Island accident, Arizona Public Service Company (APS) is in the process of assuring that the lessons learned from TMI-2 are reflected in the Palo Verde Nuclear Generating Station (PVNGS) design and operation.

To accomplish this objecti've, the APS President and Chief Executive Officer, by a directive issued April 12, 1979, established a Safety Evaluation Task Force "to conduct a complete review arid analysis of the Three Mile Island accident and its implications for the Palo Verde Nuclear Generating Station project." This task force is com-prised of fourteen (14) members from knowledgeable, high levels of management of the utilities participating in the Palo Verde project, the engineer-constructor for the project, the nuclear steam supply system supplier, and turbine generator supplier to the project and the academic technical consultants to the Arizona State Senate and House legislative committees organized to review and consider the implications of the TMI accident.

The task force is exercising oversight of the on-going independent in-depth reviews, initiated since the TMI accident, of the Palo Verde design, planned administrative and operational procedures, training and qualification programs, and emergency response plans. In this con-nection, it is closely following the work and recommendations of the ,

many industry and governmental organizations currently conducting similar, but more generic investigations related to the TMI accident, including particularly the NRC's Lessons Learned Task Force.

gIIp~'r 2.

yoq 001 p Z2

0 Director. of Nuclear Reactor Regulation September 28, 1979 ANPP-13905 - JMA/DBK Page 2 Upon receipt of the TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations (NUREG-0578), modified as set forth in the memorandum dated August 20, 1979 from Harold R. Denton, Director, Office of Nuclear Reactor Regulation to the NRC Commissioners, the recommendations related to the design and operation of the Palo Verde project were evaluated. The results of this evaluation are incorporated in the report entitled TMI-2 Lessons Learned Implementation Report submitted herewith. The report identifies those NUREG-0578 recommendations which are incorporated into the present Palo Verde design, identifies recommendations that are not applicable to the Palo Verde design and sets forth APS'oamitments to develop the analyses, design modifications, operational procedures and other measures necessary to meet such recormendations in accordance with the schedule specified in NUREG-0578. The report will be amended as appropriate to incorporate the manner in which such commitments are implemented.

Five (5) copies of this report are enclosed for your review. It is hoped that this report, in conjunction with the PVNGS Final Safety Analysis Report to be submitted in October, 1979, will assist in your timely review of the implications of TMI on PVNGS.

Respectfully submitted, ARIZONA PUBLIC SERVICE COMP By:

Edwin E. Van Brunt, Jr.

Vice President STATE OF ARIZONA ss. On its own behalf and as agent for County of Maricopa ) all other joint participants.

Subscribed and sworn to before me this~day of , 1979.

My Commission expires:

otary u >c bip Csmmiggiog Espimg mg, 33 Seas

V~ I 1'l l'

I

7 II-~2 ILESSGRS ILEMN=D IIMPIL'MEN7A7ilGN BEPGH7 no (<

cnatrc M~~/OppZ-

~~8/a~/~~

prie /<

~ of Oocttmeab REGULARY BOCKEf QL

~l~n gpss

~70 gggog ggKT

0' N

I~

,~ ~

p v P a.,

~ 1 ~~

msE3M(Q P. 0 BOX 2I666 'HOENIX'RIZONA 65036 September 28, 1979 ANPP-13905 JMA/DBK Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Re: Palo Verde Nuclear Generating Station Units 1, 2 and 3 ,

Docket Nos. STN 50-528/529/530

Dear Sir:

As a result of the Three Nile Island accident, Arizona Public Service Company (APS) is in the process of assuring that the lessons learned from TMI-2 are reflected in the Palo Verde Nuclear Generating Station (PVNGS) design and operation.

To accomplish this objective, the APS President and Chief Executive Officer, by a directive issued April 12, 1979, established a Safety Evaluation Task Force "to conduct a complete review and analysis of the Three Mile Island accident and its implications for the Palo Verde Nuclear Generating Station project." This task force is com-prised of fourteen (14) members from knowledgeable, high levels of management of the utilities participating in the Palo Verde project, the engineer-constructor for the project, the nuclear steam supply

,system supplier, and turbine generator supplier to the project and the academic technical consultants to the Arizona State Senate and House legislative committees organized to review and consider the implications of the THI accident.

The task force is exercising oversight of the on-going independent in-depth reviews, initiated since the THI accident, of the Palo Verde design, planned administrative and operational procedures, training and qualification programs, and emergency response plans. In this con- .

nection, it is closely following the work and recommendations of the many industry and governmental organizations currently conducting similar, but more generic investigations related to the THI accident, including particularly the NRC's Lessons Learned Task Force.

~,

I I

J i

Director of Nuclear Reactor Regulation September 28, 1979 ANPP-13905 - JMA/DBK Page 2 Upon receipt of the TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations (NUREG-0578), modified as set forth in the memorandum dated August 20, 1979 from Harold R. Denton, Director, Office of Nuclear Reactor Regulation to the NRC Commissioners, the recommendations related to the design and operation of the Palo Verde project were evaluated. The results of this evaluation are incorporated in the report entitled TMI-2 Lessons Learned Implementation Report submitted herewith. The report identifies those NUREG-0578 recommendations which are incorporated into the present Palo Verde design, identifies recommendations that are not applicable to the Palo Verde design and sets forth APS'ommitments to develop the analyses, design modifications, operational procedures and other measures necessary to meet such recommendations in accordance with the schedule specified in NUREG-0578. The report will be amended as appropriate to incorporate the manner in which such commitments are implemented.

Five (5) copies of this report are enclosed for your review. It is hoped that this report, in conjunction with the PVNGS Final Safety Analysis Report to be submitted in October, 1979, will assist in your timely review of the implications of TMI 'on PVNGS.

Respectfully submitted, ARIZONA PUBLIC SERVICE COMP n~'y:

Edwin E. Van Brunt, Jr.

Vice President STATE OF ARIZONA )

) ss.. On its own behalf and as agent for County of Maricopa sslsiiiiiii(,,

all other joint participants.

,:",l',>',I...'.I;Snbscribed and sworn to before me this~/day of , 1979.

s i

My Commission expires:

otary Pu ic ssjl Csmmissios Rrpims isn. 23, 3333

'>>l l'iiillll00

J

'NTRODUCTION The Palo Verde Nuclear Generating Station NUREG-0578 Evaluation addresses the "TMI-2 Lessons Learned Task Force Status Report And Short Term Recommendations" (NUREG-0578), July 1979, as supplemented by letter dated August 20, 1979 from Harold R. Denton, Director, Office of Nuclear Reactor Regulation to the NRC Commissioners.

Each recommendation is addressed, as it applies to the Palo Verde Nuclear Generating Station (PVNGS) design and operation, in the following manner:

(1) Descriptions of the existing PVNGS design correspond with information presented in the PVNGS Final Safety Analysis Report (FSAR).

(2) Proposed modifications, evaluations and analyses of the PVNGS design, as discussed herein, will be summarized in revisions to this report.

(3) Recommendation numbering corresponds to that used in NUREG-0578.

This report will be referenced by subsequent PVNGS FSAR amendments, as appropriate, to address in the FSAR, required modifications and analyses as discussed herein.

f.

'TABLE'OF'CONTENTS INTRODUCTION DESIGN AND ANALYSIS 'PAGE NO.

2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs. ~ 2.1.1-1 2.1.l.a Pressurizer Heater'ower Supply . . 2.1.1-1 2.1.l.b Power Supply for Pressurizer Relief and, Block Valves and Pressurizer Level Indocators. . 2.1.1-5 2.1.2 Performance Testing for BWR and PWR Relief and Safety Valves. . 2.1.2-1 2.1.3 Information to Aid Operators in Accident Diagnosis and Control. 2.1.3-1 2.1.3.a Direct Indication of Power Operated Relief Valve and Safety Valve Position for PWRs and BWRs . . 2.1.3-1 2.1.3.b Instrumentation for Detection

. of Inadequate Core Cooling in PWRs and BWRs . . . . . . . . . . . 2.1.3-3 2.1.4 Containment Isolation Provisions for PWRs and BWRs .. 2.1.4-1 2.1.5 Post-Accident Hydrogen Control Systems for PWR and BWR Containments. 2.1.5-1 2.1.5.a Dedicated Penetrations for External Recombiner or Post-Accident External Purge System. . . . . . . . . . . . . 2.1.5-1

I

'((

l fl',

DESIGN AND ANALYSIS (continued) 2.1.5.b Inerting BWR Containments.

2.1.5.c Capability to Install 8ydrogen Recombiner at Each Light Water Nuclear Power Plant. .

2.1.6 Post Accident Control of Radiation in Systems Outside Containment of PWRs and BWRs . . . . . . . . . . . . . . . .

2.1.6.a Integrity of Systems Outside Containment Likely to Contain Radioactive Mater ial s (Engineered Safety Systems and Auxiliary Systems) for PHRs and BWRs . .

2.1 .6.b Design Review of Plant Shielding and Environmental gualification of Equipment for Spaces/Systems Which May Be Used in Operations 2.1.7 Improved Auxiliary Feedwater System Reliability for PWRs .

2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWRs.

2.1.7.b . Auxiliary Feedwater Flow Indication to Steam Generator for PWRs. . . . .

2.1.8 Instrumentation to Follow the Course of an Accident .

'.1.8.a Improved Post-Accident Sampling Capabili'ty .

2.1.8.b Increased Range of Radiation Monitors.

2.1.8.c Improved In-Plant Iodine Instru-mentation ..

2.1.9 Analysis of Design and Off-Normal Transients and Accidents . . . . . . . . .

DESIGN AND ANALYSIS (continued) PAGE NO.

2.1.10 Instrumentation to Monitor Contain-ment Conditions During the Course of an Accident. . . . . . . . . . . . . . . . . . 2.1.10-1 2.1.11 Installation of Remotely Operated High Point Vetits in the Reactor Coolant System. . . . . . . . . . . . . . . . . . 2.1.11-1 OPERATIONS 2.2.1 Improved Reactor Operations Command Function. 2.2.1-1 2.2.l.a Shift Supervisors 2.2.1-1 Responsibilities'.2.l.b Shift Technical Advisor . 2.2.1-4 2.2.l.c Shift Relief and Turnover Procedure . 2.2.1-6 2.2.2 Improved In-Plant Emergency Procedures and Preparations. 2.2.2-1 2.2.2.a Control Room Access . 2.2.2-1 2.2.2.b Onsite Technical Support Center . 2.2.2-3 2.2.2.c Onsite Operational Support Center . 2.2.2-5 111

2.1.1 EMERGENCY POWER'SUPPLY RE UIREMENTS FOR THE PRESSURIZER HEATERS, POWER OPERATED RELIEF VALVES AND BLOCK VALVES, AND PRESSURIZER LEVEL INDICATORS IN PWRs 2.l.l.a PRESSURIZER HEATER POWER SUPPLY NUREG-0578 Position

1. The pressurizer heater power supply design shall provide the capability to supply, from either the offsite power source or the emergency power source (when offsite power is not available), a predetermined number of pressurizer

,heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions.

The required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supply capability.

2. Procedures and training shall be established to make the operator. aware of when and how the required pressurizer heaters shall be connected to the emergency buses. If required, the procedures shall identify under what con-ditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters.

3. The time required to accomplish the connection of the pre-selected pressurizer heater to the emergency buses shall be consistent with the timely ini tiation and maintenance of natural circulation conditions.

2.1.1-1

4. Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with safety-grade requirements.

PVNGS Evaluation The PVNGS pressurizer heaters are powered as follows:

1. The pressurizer heaters (36 elements rated at 50 KW each, connected in 12 groups of 3 delta-connected elements each) have a total capacity of 1800 KW. The heaters are supplied 480 volt power from the non-Class IE and Class IE distribution systems as follows:

Number of Heaters Ca acit KW 480V Bus IE Bus A. 5-3 element groups 750 NGN-Lll No B. 5-3 element groups 750 NGN-L12 No C. 1-3 element group 150 PGA-L33 Yes D. 1-3 element group 150 PGB-L32 Yes Maintenance of natural circulation at hot standby con-ditions requires 150 KW of heaters. These heaters are immediately available from either of two redundant Class IE buses, fed independently by Safety Trains A and B. Controls for these heaters are likewise inde-pendent. As the Class IE buses are a part of the ESF distribution system, they are fed either from the off-site power system or the diesel generators when offsite power is not available.

2.1.1-2

Heaters supplied from the non-Class IE power system are fed",from the offsite power system and are available for manual interconnection to the diesel generators when offsite power is not available.

2. Heaters fed from the Class IE power system (150KW) are automatically shed upon receipt of a Loss of Offsite Power (LOP) or a Safety Injection Acti-vation Signal (SIAS). They may subsequently be manually reconnected to the ESF buses without shedding of any loads from the Class IE buses.

Heaters fed from non-Class IE buses can also be manually connected to the Class IE buses for powering by either the offsite power system or the diesel generators through the ESF transformers. These heaters are available in 150 KW blocks (0-150 KW in the case of the two proportional groups). Sufficient margin exists in the diesel generators for at least an additional 300 KW of heaters to be backfed from each Class IE bus without the shedding of selected emergency loads.

2.1.1-3

e

3. If an SIAS or LOP signal are received during an accident or severe transient (shedding the heaters connected to the Class IE buses), the heaters are available for immediate reconnection to the Class IE bus from the control room.

4. The redundant pressurizer heaters required for maintenance of, natural circulation at hot standby are fed from Class IE buses via Class IE load side breakers qualified in accordance with safety-grade requirements.

The PVNGS pressurizer heaters, as presently designed, comply with the recommendations of NUREG-0578, July, 1979.

2.1.1-4

2.1.1 EMERGENCY POWER SUPPLY RE UIREMENTS FOR THE PRESSURIZER HEATERS, POWER OPERATED RELIEF VALVES AND BLOCK VALVES, AND PRESSURIZER LEVEL INDICATORS IN PWRs e

2.1.1.b POWER SUPPLY FOR PRESSURIZER RELIEF AND BLOCK VALVES AND PRESSURIZER LEVEL INDICATORS NUREG-0578 Position

1. Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.

2. Motive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.

3. Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be through devices that have been qualified in accordance with safety-grade requirements.

4. The pressurizer level indication instrument channels shall be powered from the vital instrument buses. These buses shall have the capability of being supplied from either the offsite power source or the emergency power source when offsite power is not available.

2.1.1-5

~ '

I

PVNGS Evaluation

1. PVNGS does not use PORVs and block valves.

2. PVNGS does not use PORVs and block valves.

3. PVNGS does not use PORVs and block valves.

4. Two channels of Class IE level instrumentation are provided for PVNGS. Pressurizer level channels L-llOX and L-llOY (refer to Figure 2.1.11-2) are indicated in the control'oom. Channel L-110X is also recorded in the control room.

The pressurizer level instrumentation is powered from 120V AC Class IE instrument buses E-PNA-D25 and E-PNB-D26 (refer to Figure 2.1.1-1). These buses are normally powered through inver ters from Class IE

'batteries. The Class IE battery chargers are powered from offsite power or from the diesel generators when offsite power is not available.

The pressurizer level indicators, as presently designed, comply with the recommendations of NUREG-0578, July 1979.

2.1.1-6

t 480V MCC 4807 MCC ~ SOV MCC 480V MCC 480V MCC 480V MCC ~ Soy MCC ~ SOV MCC 4SOV MCC E PHA M)S E-PHA M)S E-PHA-M)3 E PHA M)I E PHA M)l 2;PHS M)6 E PHS M)6 EIPHS M)4 E-PH8 M)t E.PHS.M32 DWG I)-E PHA 005 DWC I) E PHA<0$ OWC l3 E PHAGO) owG l) E PHA ool Dwc I) E PHA Ool OWG I3-E-PHA-006 OWG I) ~ E PHA 006 DWG I)-E-PHA-004 OWC I)-E-PHA not DWG I) E PHA 002

~ 480420V ZSKVA I VCLT ACE Q 2SKYA 480 l20V I ZSKVA 480 Itov I Z5KVA I 480 I20V VOLTAC E VOI.TACE REGULATOR i IOEZV I RKCVLATOR I VOI.TACE RKCVLATOR RECVLATOR E PNA-Vts l600 AH 880 AH I l

lo,t 27 lo 2 27 l 600 AH 790 AH lo.t 236 i E PNC V27 E bNS VZS E PND VZS DwC lt)v CC J I) E-PNAOOI It5VDC SATTERT A i2 PKA Fll l25V OC OATTEltY C K'rC fi)

I ~ IOWC I).EWNA OOI I DWC l3.2 PNA4NZ CATTERY 8 f PKS FI2 IZSVOC CATTERY O' PrD-FI4

~I I)DWG E PNA-OOZ 400A ~ OOA 400A E SAT CHCR It)yoc 8AT CHCR I Syoo 8AT CHCR SAT CHGR SAT C<<CR IZSVDC SAT CHCR D'

A E-PKA.HII Ac E PKA HIS 0

K PKC HI) f~ Hlt ltSVOC 2

I 8D PK8-Hl6 I'KD.HI4 300A IZ3 V DC I- r-q I)-'-( )OOA lt5VDC I

I NOTE 3 NOTE 3 I 1 1

1 I I NO) NC) NC) I I25V DC ct' NC)

CONTROL CENTElt 2-trA-M4t lt)VOC COHTltOL CENTER 1E.PKC.M43 I

I ")

I25V DC CONTROL J)

CENTER E PKS M42 I

I I

I lZSV DC NC)

CONTROL CENTER E PKD M44 I

f IZSv OC l I I IZSV DC I I

I I I 5$ v oc I I

I I

I I

ltsv oc I olsTR PNL I I DISTR PNL I DISTR PNL I I I I I

I I I

L DWC IS E PKA 003 DWC l) E PKA 002 L DWC I) K tKA 004 J, 2)ryl L DWC I)-E-PKA-006 DWC I) E PKAWS J

I I

DWC I)-K-trA-007 AJ...

48OV.)O.-

ZSKVA INVERTER INVKRTER IZSV OC TO IZSV Dc To nl E PKC nl f-PKD-N44 REACTOR Tlt[P REACTOR TRlt IZ5V DC To IZSV OC To SWITCHCEAR A ~

DC MOVS SWITCHCKAlt C REACTOR I RIP REACTOlt TRIP

'0' J-58A Co) ITYPICAL FOR J-SSC-CO) LOCAL 480Y )O. SWITCHCEAlt 8 SWITCHCEAlt VP.OWC I) l0407 A.c VP.DWCI) l0407 STAItTEI\ J $ 88 Co) SSD CO) LOCAL 480Y, )II.

NOOI I).0) 50 CHANNELS AND D)

NOOI-I)~ 03%0 E PKC S43

)-l 7 P. OWC I 0407 Vt. DWC I) l0407 $ TARTER STAltTER 2'KD 844 NOOI I) 0) 30 NCOI I) ~ 03 30 3C SHUTDOWN C SHUTDOWN 2.6 CLC ISOL.VLV

/ii J-SIC UV 653 2.6 CLC ISOL.VLY J SIO UV-CS4 DWC l) E SIS OI4 DWC I) E SIS.OI4 ltoy, lo+

I INVERTER INVERTE R INVERTER I I A C 'D I K PNA NII E.PNC-NI) I E PNS.NIZ/ E PND NI4 I I I OFf NC TRANSfER Sw MANUAL NC MANUAL I

MANUAL NC 0 0

f TRAN5FKR SW No I TRANSFER sw 'TRANSFER Sw I NO 0 Ho I NOTE 2 1

) I )

I I

) 1 I

) ]

I I I I I I I I I I JZOVAC VNCROUNOfo I I l20YAC VHCROUNOED l20V AC UNGROUNDED l20V AC VNCROVNDEO I IC,VITAL INST CISTR PNL i il4.vlTAL Irsf a)TR PNL i ( I4.VITAL INST DISTR PNLI l4. VITAL INST DISTR PNL I

I DWG CHANNEL A I) E PNA OOI OWG I) E-PNA OOI J I CHANNEL OWG l3 E PNA 8

002 i CHANNEL '0 Dwc l)-E-PNA.OOZ I

LOAD CROUP I LOAD CROUP 2 13-E-PKA-001 REV 1 Palo Verde Nuclear Generating Station NOTES I. FOR LEGEND 8 CfrfRAL NOTKS REFER TO OWCS l)-E ZZS.OOI SINGLE LINE DIAGRAM THROVGH 007 THIS INDICATES MANVAL TWO POLE THRKf POSITION TRANsffR DC POWER SYSTEM SWITCH OUTPVT SWITCHES IN SATTEIIY CHARCElt) AC'd'80'RE 3.

MECHANICALLY INTERLOCKED To OFFER THE to)5ISILIIT Of EITHER SOTH OPEN OR ONE CLOSED AND ONE OPEN AT Figure 2.1.1-1 AN1 TIME

f.l

\

PERFORMANCE TESTING FOR BWR AND PWR RELIEF AND SAFETY VALVES NUREG-0578 Position Pressurized water reactor and boiling water reactor licensees h

and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected oper-ating conditions for design basis transients and accidents.

The licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2. The single failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized. Test pressures shall be the highest predicted by conventional safety analysis procedures. Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry piping and supports, as well as the valves themselves.

PVNGS Evaluation PVNGS does not utilize pressure operated relief valves.

An analysis will be performed to determine the expected code safety valve operating conditions based upon the accidents and anticipated operational occurrences described in the PVNGS FSAR.

2.1. 2-1

I l'

When this analysis is completed, generic code safety valve qualification tests will be performed, as necessary, to qualify the valve for expected operating conditions following an accident or anticipated operational occurrence.

In addition, the discharge piping and supports will be analyzed for their expected operating conditions. Per-formance of qualification tests is dependent on the availability of test facilities capable of simulating the expected operating conditions.

The results of these analyses and tests will be provided prior to PYNGS Unit 1 Operating License.

2.1.2-2

II Direct indication of Power-Operated Relief Valve and Safety Valve Position for PVfR's and BWR's

2.1.3 'INFORMATION'TO'AID OPERATORS IN ACCIDENT DIAGNOSIS'AND'CONTROL 2.1.3.a DIRECT INDICATION OF POWER-OPERATED RELIEF VALVE AND SAFETY.

VALVE POSITION FOR PWRs AND BWRs NUREG-0578 Position Reactor system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detection device or a reliable indication, of flow in the discharge pipe.

PVNGS Evaluation PVNGS does not utilize 'power operated relief valves.

The PVNGS pressurizer code safety valves are headered into the reactor drain tank (RDT), inside containment. Upstream of the common header each code safety valve is monitored for seat leakage by an in-line RTD (refer to Figure 2.1.11-2). Indirect indication of code safety valve operation is provided by RDT pressure, pressurizer pressure and pressurizer level instrumentation.

An evaluation will be performed to determine if the multiple indirect indications of code safety valve status provided in the PVNGS design provide unambiguous indication that a small break area LOCA is in progress.

2.1. 3-1

i 0

If the evaluation determines that direct valve status indication in the control room is required (e.g. by "acoustic techniques, flow measurement or valve position indication), it will be incorporated in the PVNGS design prior to Unit 1 Operating License.

Results of this evaluation will be provided by January 1980.

2.1.3-2

J j

Instrumentation for Detection of Inadequate Core Cooling in PWR's and BV/R's

2.1.3 INFORMATION TO AID OPERATORS IN DIAGNOSIS AND CONTROL 2.1.3.b INSTRUMENTATION FOR DETECTION OF INADE(UATE CORE COOLING IN PWRs AND BWRs.

NUREG-0578 Position

1. Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available instrumentation. The licensee shall provide a description of the existing instru-mentation for the operators to use to recognize these conditions. A detailed description of the analysis needed to form the basis for operator training and procedure development shall be provided pursuant to another short-term requirement, "Analysis of Off-Normal Conditions, Including Natural Circulation".

In addition, each PWR shall install a primary coolant saturation meter to provide on-line indication of coolant saturation condition. Operator instruction as to use of this meter shall include consideration that is not to be used exclusive of other related plant parameters.

2.1.3-3

0 I.

f t

~

2. Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-interpret indication of inadequate core cooling. A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.

PVNGS Evaluation Information to diagnose inadequate core cooling is as follows:

1. Prior to PVNGS Unit 1 Operating License, procedures will be developed to aid the operator in recognition of inade-quate core cooling. These procedures will utilize existing redundant safety related instrumentation such as:

A. Pressurizer Pressure 1500-2500 psia B. Pressurizer Pressure 0-3000 psia C. Pressurizer Level 0-100K D. Coolant Temperature

- Hot Leg 375-675F

- Cold Leg 465-615F 2.1.3-4

These procedunes will be based upon analyses performed as discussed in Section 2.1.9. Redundant subcooling meters will be installed prior to PVNGS Unit 1 Operating License to provide additional indication of primary coolant saturation condition. The procedures discussed in Item 1 will incorporate the use of primary coolant saturation indication.

2. Upon completion of- the analyses described in Section 2.1.9, an evaluation will be performed to determine the need, if any, for additional .instrumentation, such as reactor vessel level indication. Any instrumentation determined to be necessary will be installed prior to PVNGS Unit 1 Operating License.

2.1.3-5

CONTAINMENT ISOLATION PROVISIONS FOR PWRs AND BWRs NUREG-0578 Position

l. All containment isolation system designs shall comply with the recommendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the init-ation of containment isolation.

2. All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essential, shall identify each system determined to be non-essential, shall describe the basis for selection of each essential system, shall modify their containment isolation'esigns accordingly, and shall report the results of the re-evaluation to the NRC.

3. All non-essential systems shall be automatically isolated by the containment isolation signal.

4. The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic re-opening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action.

2.1.4-1

0 l

fl

PVNGS'Evaluation The PVNGS containment isolation system is designed as follows:

l. As required by SRP 6.2.4, a containment'solation signal is diversely gener ated by either a high containment pressure signal (5 psig) or a low pressurizer pressure signal (1600 psig). These parameters additionally generate a safety injection actuation signal. The power access purge is additionally isolated by high containment purge radioactivity.

2. A review of the PVNGS containment isolation system will be performed to determine that no essential systems are auto-matically isolated by a containment isolation actuation signal(CIAS). Essential systems will be selectively iso-lated by the operator after their use is no longer required.

The results of this review will be provided to the NRC staff by January 1980. Plant modifications, if any, will be per-formed prior to PVNGS Unit 1 Operating License.

3. All non-essential systems penetrating the PVNGS contain-ment are isolated by CIAS.

4. Resetting of a CIAS does not result in the automatic opening of containment isolation valves. Re-opening requires operator action for each valve.

2.1.4-2

0 Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems

e

'2.1.5 POST'ACCIDENT HYDROGEN'CONTROL'YSTEMS FOR'PWR'AND

'BWR'CONTAINMENTS 2.1.5.a DEDICATED PENETRATIONS FOR EXTERNAL RECOMBINER OR POST-ACCIDENT EXTERNAL PURGE SYSTEM NUREG-0578 Position Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere should provide containment isolation systems for external recombiner or purge systems that are dedi-cated to that service only, that meet the redundancy and single failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR Part 50, and that are sized to satisfy the flow requirements of the recombiner or purge system.

PVNGS Evaluation Combustible gas concentration in the PVNGS containment building is maintained by the Containment Hydrogen Control System.

The PVNGS Containment Hydrogen Control System (refer to Fig-ure .2.1.5-1) shares no equipment or penetrations with the normal containment purge system and includes:

2.1. 5-1

0 II t 0

1. Two redundant and independent hydrogen gas analyzers with separate piping, valves, controls and instrumentations (refer to Section 2.1.10).

2. Redundant and independent mobile, hydrogen recom-biners (two for the PVNGS site) with separate supply and return lines, utilizing separate penetrations for each line.

3. Redundant and independent supply and return line isolation valving (one valve inside containment and one outside containment) that meets the requirements of 10 CFR 50, Appendix A, Criterion 54 and 56.

4. Dedicated containment penetrations sized for maximum recombiner flow requirements.

The PVNGS Containment Hydrogen Control System, as currently designed, provides dedicated penetrations and isolation systems that meet redundancy and single failure requirements as recommended by NUREG-0578, July 1979.

2.l.. 5-2

fl Il

2.1.5.b lnerting 8%R Containments

l f

I I

2.1.5 POST ACCIDENT HYDROGEN CONTROL SYSTEMS FOR PWR AND BWR CONTAINMENTS 2.1.5.b INERTING BWR CONTAINMENTS NUREG-0578 Position It shall be required that the Vermont Yankee and Hatch 2 Mark I BWR containments be inerted in a manner similar to other operating BWR plants. Inerting shall also be required for near term OL licensing of Mark I and Mark II BWRs.

PVNGS Evaluation PVNGS utilizes Combustion Engineering,'-Inc. System 80' PWRs with a containment with 2.6 xl0 ft3 net free volume.

Inerting of the PVNGS containment is not required.

r 2.1.5-3

i 1

l j

Capability to Install Hydrogen Recombiner at each Light Water Nuclear Power Plant

0 l

2.1.5. POST ACCIDENT HYDROGEN CONTROL SYSTEMS FOR PWR AND BWR CONTAINMENTS 2.1.5.c CAPABILITY TO INSTALL HYDROGEN RECOMBINER AT EACH LIGHT WATER NUCLEAR POWER PLANT NUREG-0578 Position (Minority View)

l. All licensees of light water reactor plants shall have the capability to obtain and install recombiners in their plants within a few days following an accident if containment access is impaired and if such a system is needed for long-term post-accident combustible gas control.

2. The procedures and bases upon which the recombiners would be used on all plants should be the subject of a"review'by, the licensees:in considering 'shielding requirements and personnel exposure limitations as demonstrated to be necessary in the case of TMI-2.

PVNGS Evaluation The PVNGS hydrogen recombiners are designed as follows:

1. PVNGS will have available, onsite, two mobile hydrogen recombiners designed to limit the concentration of hydrogen in containment. These hydrogen recombiners are capable of being installed in the affected unit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an accident.

2.1.5-4

i fr I

2. Installation procedures for the hydrogen recombiners will be developed and shielding requirements re-analyzed prior to PVNGS Unit 1 Operating License.-

2.1.5-5

0 i f

~ 4M

,4

~

CR w CR 4 CR

~

g 4 StCT" SELKC'I

~

1 TK5T at Sk A Sc A 50 A A ~

CCNTAI )CENT HZ POST LOCA A I HTDltoCEN I NCHITOR 1 /4 HV 5

k

~ /4

~ ~ ~ ~ 1 VOX5 SCK NOTE 2 CR 0 HS o ~

5 A LJ ~

CIAS ~ Q TtST CLOSES Vol C ll

'Q5

~

TEST X

' ~ O A 005 HCDA 27

3. ~ /I C ~

~ A 005 HCCA 2 1 SEC No'lC ~ I VOOZ o NV ZSNL 5 TA A

Tt ST

~ ~ ~ \R ~ ~ ~ ~ ~ ~ RC ~ ~ ~

SCE HOTE 2 4

4 I HFA Ik)1 4

~ To *TNDSPHERC X HPS.II)1'1 Hs CLOSES>+ 0 EII)1OLoscs ~

55 0 Nl I

4~

~ ~

il

~ ~

UV ZSNI VV ZSNL ~ ~ 4 CONTkl NIENT ~ ~

u NT CROCK N ~ ~ tDI tDSL A A 001 HCCA 2 ~

~ ~

5$ 5$

I STSTEH

~

~ ~

~

4

,SCE SECHTEL LOC

)set Holt ~ ) A 001 NCCA I Vol 0 I No, NSTS 44 V015 ~

~ ~

0 0 0 A OOT NCCA 2 I

~

~

~

~

~

tl 2$

4 PO I 55 PO I 54 LJ LJ 44 ~

~

~

~

~

~ 8 TEST TEST ~EKJlQL ~

~

~

~

IZZE N 015 NCOk 2 To PLANT 4

~

~

~

~ Ctt 001 VENT

~

4

~

~ ~

ZS IF I )

0 0 N 00$ NCDA 2 ~

~ ~

2

~ ~

42 1

~

~

~

~ 55

~ ~

~ N 011 HCDA I

~ ~

II 4 I

Q

~ ~

~

'4

~

4 I'u CO---'I 1

~

'o

~

X

+

~

~ ~

Cl ccoscs1 y+I 0 OOS HCSA 2 4~ 4

~

~

~

~

~

~

4 4 0 ST

)TEST I ' I

~

~

~

~

SERT I Ct UV ZSHL UV ZSNL Vol '5 CCNTAINICNT

~

~

~

~

Ik 044 HCDA I U 2 N ~ NTOIIOCEN ~ ~ IAP 002 4 gg I/IC 8 D 00$ HCSA 2 I RECCNDINER ~

~

~

~ ~

0 VOOT lc TI

,Stt STSTEN DECNTCL LOC.

~ ~

u Z ~IA NCOA HCOA ISCE NOTE I )

..J O R o ~

Vol I No 4555 44 Vol 2

~

~

~

~

Z 004 HCCA

~ ~

~ ~ ~ ~ ~ k RK ~ ~ ~ ~ ~ ~ ~ ~ ~

I I ~

~

~

~

LJ Tcs'I L J Nt* NOIR

~

~

~

~

Z NOTES' AILJI ~

~

~

ZSNL HV INSTALL VALVES Wl'TH FLOW DIRECTION

~ I ARROW IN OIRECtloN NOTED CA To CLOSE ACAINST CCNTAIIAENT OESICN tltKSSVltt THE $ 1$ TEN OESICNATOR Ht ls To PRECEDE ALL 2 ~ PIPE SHALL CE TNREADED TO ACCEPT LINE VALVE~ AND INSTRVLENT Iklloclts SNOXN ON THIS DRAWINO VNLLSSO'tHCRWIst INDICATED A 2 2 I CkttED TEST RCDVCER TEST IJ AS SHOWN CtLOW.

I Fl 4 4 UV ZSHL C

~

4 N

0 I /I 4 4

Hs 4 ~ ELECT))ICAL NEA'I TltkCINO 5NALL FRON IE SOURCE OF PONER.

St towCRCD TEST C~/ 2 OVRI NO NatHAL OPERATION WNEN C ~

C 004 NCCA 2 END ol'ltE Is OttN, PROVIDE SCREN COVPLINCS Folt PROTECTION

)SEE NOTE I) V004 OF IHE 'tldtEAO 0 004 HCSA Zw/ Clt akDI HEI\ T AIR HS F II.THAT)OH VNI 't 3 TEST 01 CR w CR 4 CR HV STST TES'I SELECT 5 Cs CON'tAIIAENT 0

~1 10 4 I Z 4 13-M-HPP&01 REV 3 0 tosT LOCA 0 NTOROCKN HORIZON I I

SCE NOTE 2 o 4 W J I/4 A Palo Verde Nuclear Generating Station NZ Vo)1 lo

~ k V052 J Nts COZ 01 C COMBUSTIBLE GAS CONTROL SYSTEM Figure 2.1.5-1 4

Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PV/R's and BNR's

~,

)

~

2.1.6 POST ACCIDENT CONTROL OF RADIATION'lN SYSTEMS:OUTSIDE C NTAINM N F PWRs AND'B JRs 2.1.6.a INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE MATERIALS (ENGINEERED SAFETY SYSTEMS AND AUXILIARY SYSTEMS) FOR PWRs AND BWRs NUREG-0578 Position Applicants and licensees shall iranediately implement a program to reduce leakage from systems outside containment that would or could 'contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels.

This program shall include the following:

l. Immediate Leak Reduction

a. Implement all practical leak reduction measures for all systems that could carry radioactive fluids outside of containment.

b. Measure actual leakage rates with system in operation and report them to the NRC.

2. Continuing Leak Reduction Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels in systems outside containment 1'ikely to contain 'highly radioactive fluid during a serious transient or accident.

2.1.6-1

i

'PYNGS Evaluation The PVNGS program of preventive maintenance will reduce leakage to as low as practical. The program will include:

l. In-service inspection of ASME Class 2 and 3 systems in accordance with AMSE Boiler and Pressure Vessel Code Section XI, including pressure tests.

2. Alarming of abnormal airborne radioactivity levels by the radiation monitoring system for gaseous rad-waste system leakage.

3. Routine monitoring of plant components.

Undesirable leakage identified during tests or inspections will be reduced to a level as low as is practical.

2.1.6-2

0 Design Review of Plant Shielding 5 Environmental Qualification of Equipment for Spaces/Systems Nhich May be Used in Post-Accident Operations .

I 2.1.6 POST ACCIDENT CONTROL OF RADIATION IN SYSTEMS OUTSIDE CONTAINMENT F PHRs ND B Rs 2.1.6.b DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR SPACES/SYSTEMS WHICH MAY BE USED IN OPERATIONS NUREG-0578 Position With the assumption of a post-accident release of radio-activity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50/ of the core radioiodine and 100% of the core noble gas inventory are contained in the primary coolant), each licensee shall perform a radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.

2.1.6-3

0 II

'PVNGS'Evaluation As part of the PVNGS design process, shielding in areas requiring personnel access for accident mitigation is designed such that access is not unduly limited.

In addition, safety related equipment is qualified for the maximum expected equipment life dose.

An analysis of the PVNGS shielding design criteria, now based on ALARA, will be performed to determine if the results of TMI invalidate a y of the PVNGS shielding criteria. If, as a result of this analysis, the shielding design crit'eria in a plant area requires modification, the shielding in that area will be redesigned based on revised criteria. Any increase in permanent or temporary shielding, redesign or procedural control determined to be necessary will be completed prior to PVNGS Unit 1 Operating License. Refer to Section 2.1.8.a for a discussion of sampling system shielding.

2.1.6-4

Automatic Initiation of the Auxiliary Feedwaier System for PNR's

~

H

2.1.7 IMPROVED AUXILIARY FEEDWATER SYSTEM RELIABILITY FOR PWRs 2.1.7.a AUTOMATIC INITIATION OF THE AUXILIARY FEEDWATER SYSTEM FOR PWRs NUREG-0578 Position Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10 CFR Part 50 with respect to the timely initiation of the auxiliary feedwater system, the following requirements shall be implemented in the short term:

1. The design shall provide for the automatic initiation of the auxiliary feedwater system.

2. The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function.

3. Testability of the initiating signals and circuits shall be a feature of the design.

4. The initiating signals and circuits shall be powered from the emergency buses.

5. Manual capability to initiate the axuiliary feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function.

2.1.7-1

J

,I I

6. The AC motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses.

7. The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFWS from the control room.

8. In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety-grade requirements.

PVNGS Evaluation

1. The Auxiliary Feedwater System shown in Figure 2.1.7-1 consists of two Seismic Category I pumps and their associated valves and one non-Seismic Category I pump and its associated valves.

The Seismic Category I. portion of the auxiliary feedwater system is provided to automatically initiate residual heat removal capability during emergency conditions, such as a steam line rupture, loss of normal feedwater, or loss of offsite and normal onsite power.

2.1.7-2

i The Seismic Category I portion of the auxiliary feedwater system is automatically actuated by an Auxiliary Feed-water Actuation Signal (AFAS) from the Engineered Safety Features Actuation System (ESFAS). The AFAS is initiated for each steam generator by a low steam generator level coin-cident with a "not ruptured" calculated signal for that steam generator'(refer to Figure,2.1.7-2).

The AFAS logic determines whether a steam generator is not intact in the event of a secondary system break by sensing:

- The steam generator has initiated a low water level trip.

- The steam generator pressure is less than the other by a predetermined value.

- The other steam generator has been calculated as not being ruptured.

The non-Seismic N

Category I portion of the auxiliary feedwater system is provided for normal non-emergency operation during star tup, cooldown, and hot standby.

2. The Seismic Category I portion of the auxiliary feedwater system is composed of components in two separate and diverse load groups (i.e., load group 1 and load group 2. Each of the four auxiliary feedwater valves associated with each steam generator is automatically actuated in such a manner that no single failure can prevent either the supply of auxiliary feedwater to an intact steam generator or the 2.1.7-3

0 I

isolation of auxiliary feedwater from a ruptured steam generator. 'Load group 2 powers the Seismic Category I motor-driven auxiliary feedwater pump and its associated valves .and controls. Load group 1 supplies dc power to the steam-driven turbine controls and the valves associated with the turbine-driven auxiliary feedwater pump. No ac power is required for support of the turbine-driven auxiliary feedwater train. The instrumentation and controls of the components and equipment in load group 1 are physically and electrically separate and independent of the instru-mentation and controls of the components and equipment in load group 2. This separation is maintained such that both trains are not terminated on common logic circuits.

3. Provisions are made to permit periodic testing of the auxiliary feedwater initiation signals and circuitry.

These tests cover the trip actions from sensor input through the protection system and actuation devices.

The system test does not interfere with the system protective function. The testing system meets the criteria of IEEE-338-1971, "IEEE Trial-Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protective Systems," and Regulatory Guide 1.22, 2.1.7-4

"Periodic Testing of Protection System Actuator Functions." Testing is performed in accordance with the surveillance test requirements of the PVNGS Technical Specifications.

4. AFAS circuits are a part of the Engineered Safety Features Systems. The initiating signals and circuits are powered from Class IE buses in separate load groups as discussed in Item 2. The initiating sensors are powered from separate and redundant Class IE nuclear instrumentation and control panels, each of which is supplied by either offsite power or the diesel generators when offsite power is not available and is backed up by Class IE batteries.

/

5. Manual .initiation capability for each auxiliary feedwater train exists in the control room. Control of individual auxiliary feedwater system components is also available in the control room. No single failure in the manual initiation portion of the circuit can result in the loss of auxiliary feedwater, system function.

6. The Seismic Category I AC motor-driven pump and power operated valves in its train are automatically and sequentially loaded on the diesel generator bus upon loss of offsite power.

2.1.7-5

~ i

7. Failure of the AFAS will not result in the loss of manual capability to control auxiliary feedwater system components from the control room.

8. AFAS circuits are Class IE as presently designed.

The PVNGS auxiliary feedwater system, as presently designed, complies with the recommendations of NUREG-0578, July 1979.

2.1.7-6

l I

l

Auxiliarjj Feedwater Flow Indication to Steam Generators for PNR's

4, IMPROVED AUXILIARY FEEDWATER SYSTEM RELIABILITY FOR PWRs AUXILIARY FEEDWATER FLOW INDICATION TO STEAM GENERATOR FOR PWRs NUREG-0578 Position Consistent with satisfying the requirements set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function, the following requirements shall be implemented:

1. Safety-grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.

2. The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary .feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.

PVNGS Evaluation

1. Monitoring of both trains of auxiliary feedwater flow is pro-vided at PVNGS. These flow indicator channels are made from highly reliable non-Class IE components and are displayed on the 2.1.7-7

l main control boards. Class IE (safety grade) pressure indicators located upstream of the manual block valves are also provided. Pressure indicator PI-18A and'low indicator FI-40 monitor flow from the train A turbine-driven auxiliary feedwater pump (refer to Figure 2.1.7-1). Pressure indicator PI-17A and flow indicator FI-.41 monitor the flow from the train B motor-driven auxiliary feedwater pump.

2. The safety grade pressure indication channels are powered from their associated trains redundant Class IE buses., The instrument bus powering the auxiliary feedwater flow instrumentation is normally powered from a non-Class IE 480 V motor control center (NCC).

Upon loss of this NCC the instrument bus is auto-matically transferred to a Class IE NCC via a Class IE isolation transformer . This, Class IE NCC is powered from offsite power or from its diesel generator when offsite power is not available.

2.1.7-8

sc-N-0>o-ccoe-So

~oa saoet NYDCAtlaf 5CP 001 N.otl.<<CDA ~

ADD>1>OI~ P(p<<5 If Sl $ C YOCF oil PI I V

It ~>$

fAON A<<NON>A aooll>DN PUIIPS LCP 001 If 2I sc-N4Q CCC4 N 020'ec(OA I I

'a I

ca N 014 ocoa-4 5CP If Oot 141 10 Ftfoi(Alta

<<CA(44 SIC AII cca(aa'loa acf ~ CDIA

$ C-VOCO Yo>2 I Q~ ga O I I I c>>

ccoa Ncoa ce

~

ce O o I s NON C T~ 00 I Ct ~ Ie Olt NCOA 10 ee otl eccoa lo 4 N 025 DCoa C LCP 00t 10 FMDICATfa NCAOIN STfAN COIDCIC>atf TANK ID 2> TOO I AFN SOI v012 vOlt T IC ICI ccafaatoa acc foi~

O 4 C ce

~ \ svztszatz ~SO V o I~ e

'10 CONDCN5ATC LI (t ATN Ct ~ Dol Ie 001 OCDA 2 0 IC 5$ 0 TANK I (5 '21 a voos lOCO Nooe LP fNON NAIN 5(/'ol sc 4 002 toce c lo D

~-.4R 5'6'2 ~ OI4 Nece I STLAN CfsetaaTDC 1$ 2 il 12I Toot ~ VOCF

'~ 520

~

4 Oll DO(0 ~ L 4 Icsts Po el 5 Ialp I 2SNL 0

aio OCC oe I I

Faoa AUX>LTAAT DN~NOO(0 0 ZSN L 411 Aa IC-4>>

ASP OOI I 4\

8544 I 2sl tC OCCI(iti Qfo VZ NA at N 022 CSOO ~

QSO A 'TSZ~~/I 5$

I 'g Vl NA Et>LZZ(>Z 0 I C 510 0 IIV (~~OOI ~ 5 0 PUN/ ~ Y4 IIA Coco DOD0 ooco Oooo O ee C 54 IC Itl afa SD2 OIC L.o. Coo(fa 4 051 04(0 I vs IIA VC TOST 154C ZZP DO>I QOO C ~

oI<< tla sot l 0 fol Afa Asee 0>c D ftltfa N-o>S-st.C'Y.

N Oll OSOC I O 01 CNML vfNOON V INC V'2 VOSC SUOSTtTCII

~.

514 II02> lo isM Mp-ooi>

1'1 NA va NA vs-s a VC V04C 4-CSI-(OCO-I 250-Ceto-I-W + T I L>

Y ovfaaiof Sia AFN FOR L ~$ 2( A 0

N 01 ~

D nV NODS o on" Vl NA V4 Na 15 NA VC VIIO Vl Na Va NA VS IIA VC sta VIIt

, s 9,, ~ ovf aa I of I f I 12A ca 6 1/I C

9%44@5

, II I.~j r--Q s ~$ 5 a ca C

ca

'~

$ 14 ~ AFN soa AFN F05 AFAt fata NZACTDN NAKCUP Nasfa Ce<<ool I r 4 OtC IICCA 0 O Ie 4

oo(D

/ $ 20 I / I (ODI LJSS Iaotf 21 rOSe o 4 o n voos~ o nv N<<*F4t Ir I

I A T lie A 6-'-'-'"Pic./

CYJ I 420 OIC <<QX I I

~ sea 10 fffovatfa O'::=:Jv oe O O NT 11 Nf aof 4 O o $2 $2 Stf AN p 4 Cf ICLNATON SICOA ICCC 4 SZC 4X Saoel COND(4541$

ctp oo Ct 4 Otl Ncca ~

e 045 ee(ca 0 4 LO 4 00$ OCCA 4 INOTF II A

4 Ool OCCA 5 4 00 ~ OCOA C J acc KDIA KBl STOAACC TlNK Volt VO(C SCP 002 ID-21 VOOC TOOF AFA-SOI OI~COit vote" IP IOI C.OOC-OCOA.C.~

4 ee 4 c OCCA OCOA o(I

~

ee ce o ovfaaiof 05 D Og ovf atof V Z Is o 51C O O OIO 5$ 0 I Oct~A DCOA Pt I lo I cet I I

~

5ZL ZUI( s s I rr---~

~

~STA I 0 V LO 02$ OCCa A c 4 TOC I I s as VOI 1 I s I

6 I- I AFAS I 4 o re~

//

asas ~510

$ 50 I oa C as I 4

c 110 t9Ãlicf r s I

Nv sea zt x Sl 5$ $1 C

~ \ (Notf I I 4 OIC DCCA 4 (NOTS 'I I A

4 OIC OCOA-5 10 CONOC USA TC 5104ACf TANK ctp.oos 4 Ool OCCA C 0(~COI 0 (alai SLON ANO 4 44 << DCCa OCOa elf ac>sf > Ic ZI VOI 0 0

Djo V

O 0

~ aoa afacTOA X

Ca ca ILL>cc'eea 0

--- Ca gI ca 9-- -,

CIS 0

CI

~fo No l.

lit Yaivt NA5 its YDc NOToa acTvlvoa.

NAKCUP NAtfa TANK CN N 4$ > NCOA 4 0 020 NCCA 4' Pt I I as Qfo 2~ Lift 15 O'Ic TO SICST Fit(0 (4 SI

~F ~>ta ASAS Ij.

ICCLO OU15I DC' IA IN SMAN 0 Tool ASA$ Q I lit0 SUPPOAI STAUCTUaf O

~pnv :6!

O I Cl

&-- 4$

tNMK/

// I I I-- 1st I as

%$ 5lK./

// I

~

I

~1St ce P

V 5

AFAS STAAIS o $ >o I ecv tl SI\ zt

'l I 150 0 Uv $ea <

~COit C OOC OCta 4'4 l SCP 001 I ~ IOI Ail llaf Vaivf allo INSTAUNCNT NUlloflvt 5<<oe<<e (Ne lee>5 DAANINC UN(fls OTNCAIC>sf 4 OCCA OCOA l(co(CaTTO ~ O D V jfft>XAZZS~ ZCZZZ~ZZZZ(> <<'f~ B-0 0 O D ~ QLC 0 4 SO ca 0 ca CYJaaitf f I los 6 Ca ytgaaiÃ~ '~ TOC LO 0 OTS DCCA 5 ~PS( Pa 0 I I 0 I 0 I TORt I 0$ afae y-Qi 13-M-AFP-001 REV 5 Co // ~l10 / ~TOO 6- -/ I 4 as kS I ~ I loo 8'-"- $ 40 as CLYJNCM./ I UV SCIL O O 0 50 0 '14 0 Palo Verde Nuclear Generating Station 0 OIF OCCA C OI ~ OCOA 4 10 CO<<Of Nlalt SIOIIACC 1ANK 0 Oil OCCA 4 OIC 510 0(0 51D I(IIXI SL(al If 1> v021 Dcc La~as OCOAAADCOA ecsoa fsf AC>SCI 30 30 V014 toes Votl VO44 AUZZLZARY FEEDWATER SYSTEM Figure 2 1 7 1 PRESS LOW SG.2 PRESS LOW SG 1 LEVEL SG 2)SG I LEVEL SG.1 )SG 2 A 8 C 0 A 8 C 0 A 8 C D A 8 C D AND AND AND AND AND AND AND AND A 8 C D AND AND AND AND AND AND AND AND 2/4 LOGIC 2/4 LOGIC EF AS-1 EFAS 1 EFAS 2 EFAS.2 REMOTE REMOTE REMOTE REMOTE MANUAL MANUA MANUA ~ MANUAL OR OR OR OR OR OR OR T 1A 18 2A 28 3A 38 4A 48 IA 18 2A 28 3A 38 4A 48 PPS AUX 1A A 3A 4A 18 28 38 48 RELAY lA 2A 3A 4A 18 28 38 48 CABINET SE LECTI VE SELECTIVE SELECTIVE SELECTIVE 2OUTOF 4 2OUTOF 4 2 OUT OF 4 2OUTOF 4 LOGIC LOGIC LOGIC LOGIC MANUAL MANUAL OR OR OR OR VALVES VALVES VALVES VALVES TRAIN A TRAIN 8 TRAIN A TRAIN 8 EF AS.1 EFAS.2 EFAS EMERGENCY FEEDWATER ACTUATION SIGNAL H~ NOT C-E I Figure ESFAS SIGNAL LOGIC (EFAS 1, EFAS 2) $ 8tSPS / i 2.1.7-2 @i,l (t ) 0 l improved Post-Accident Sampling Capability I If I I e 2.1.8 INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2.1.8.a IMPROYED POST-ACCIDENT SAMPLING CAPABILITY NUREG-0578 Position A design and operational review of the reactor coolant and containment atmosphere sampling systems shall be performed to determine the capability of personnel to promptly obtain (less than 1 hour) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18 3/4 Rems to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria. A design and operational'eview of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly quantify (less that 2 hours) certain radioisotopes that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums ( which indicate high fuel temperatures), and non-volatile isotopes (which indicate fuel melting). The -initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release. 2.1.8-1 0 The review should also consider the effects of direct radi-ation from piping and components in the auxiliary building and possible contamination and dir ect radiation from air-borne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria. In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly; i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift. PVNGS Evaluation An analysis of the design and operation of the reactor coolant and containment atmosphere sampling systems will be performed to ensure that samples can be obtained promptly (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) during accident conditions without exposure to any individual to greater than'3 and 18-3/4 Rems to the whole body or extremities, respectively. Any design or procedural changes will be imple-mented prior to PVNGS Unit 1 Operating License. 2.1.8-2 0 An analysis of PVNGS spectrum analysis facilities will be performed to ensure that the design allows prompt quantification (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) of previously prepared radioisotope samples that may be indicative of core damage. The assumed reactor coolant spectrum will correspond to a Regulatory Guide 1.3 or 1.4 release. PYNGS chemical analysis facilities will be designed to allow prompt analysis of boron concentration (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and chloride concentration (less than 8 hours) assuming a radioactive sample corresponding to a Regulatory Guide 1.3 or 1.4 release. 2.1.8-3 0 0 increased Range of Radiation Monitors j 2.1.8 INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2.1.8.b INCREASED RANGE OF RADIATION MONITORS NUREG-0578 Position The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a revision to Regulatory Guide 1.97, "Instrumentation to Follow the Course of an Accident," which has already been initiated, and in other Regulatory Guides, which will be promulgated in the near-term. 1., Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions; multiple monitors are considered to be necessary to cover the ranges of interest.

a. Noble gas effluent monitors with an upper range capacity of 10 pCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.

b. Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal condition (ALARA) concentrations to a maximum of 10 pCi/cc (Xe-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors should overlap by a factor of ten.

2.1.8-4 0

2. Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for .

the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.

3. In-containment radiation level monitors with a maximum range of 10 rad/hr shall be installed. A minimum of two such monitors that are physically separated shall be provided. Monitors shall be designed and qualified to function in an accident environment.

PVNGS Evaluation The PVNGS radiation monitoring system is a follows;

1. Noble gas effluent monitoring of the plant vent is presently designed to have an upper limit of 10 yCi/cc'(Xe-133).

Additional monitors will be added with an upper range of '04 pCi/cc (Xe-133). The range of the plant vent monitors shall overlap by at least a factor of ten.

2. Post-accident effluent monitoring of radioiodines is provided for by sampling of plant vent iodine monitor adsorption media followed by onsite laboratory analysis.

Procedures for this sampling will be developed prior to Unit 1 Operating License. 2.1.8-5 0 t

3. The PVNGS power access purge monitors (two redundant, ESF monitors) are located around the power access purge duct adjacent to the containment. These monitors are capable of monitoring the purge duct during normal operation and the containment (through the containment wall) subsequent to an accident. When monitoring the containment radiation level these monitors will have 6

a maximum range of at least 10 rad/hr. The monitors will be qualified to post-accident condition. The design of the monitors will be evaluated further to determine if a maximum range of 10 8 rad/hr is achievable. This evaluation will be completed by March, T981. 2.1.8-6 Improved In-Plant Iodine Instrumentation I INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT IMPROVED IN-PLANT IODINE INSTRUMENTATION NUREG-0578 Position Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where personnel may be present during an accident. \ PVNGS Evaluation Prior to PVNGS Unit 1 Operating License, procedures will be implemented for determination of airborne iodine concen-tration by sampling iodine monitor cartridges for spectrum analysis. These procedures will be used when the plant radiation monitoring system indicates iodine levels requiring the use of respiratory equipment. Spectral analysis equipment is available at each PVNGS unit. 2.1.8-7 ANALYSIS OF DESIGN AND OFF-NORMAL TRANSIENTS AND ACCIDENTS NUREG-0578 Position Analyses, procedures, and training addressing the following are required:

1. Small break loss-of-coolant accidents;

2. Inadequate core cooling; and

3. Transients and accidents.

Some analysis requirements for small breaks have already been specified by the Bulletins and Orders Task Force. These should be completed. In addition, pretest calculations of some of the Loss of Fluid Test (LOFT) small break tests (scheduled to start in September 1979) shall be performed as means to verify the analyses performed in support of the small break emergency procedures and in support of an eventual long term verification of compliance with Appendix K of 10 CFR Part 50. In the analysis of inadequate core cooling, the following conditions shall be analyzed using, real istic (best-estimate) methods:

1. Low reactor coolant system inventory (two examples will be required - LOCA with forced flow, LOCA without forced flow).

2. Loss of natural circulation (due to loss of heat sink).

These calculations shall include the period of time during which inadequate core cooling is approached as well as the 2.1.9-1 0 I If j J ej period of time during which inadequate core cooling exists. The calculations shall be carried out in real time far enough that all important phenomena and instrument indications are included. Each case should then be repeated .taking credit for correct operator action. These additional cases will provide the basis for developing appropriate emergency procedures. These calculations should also provide the analytical basis for the design of any additional instru-mentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy. The analyses of transients and accidents shall. include the design basis events specified in Section 15 of each FSAR. The analyses shall include a single active failure for each system called upon to function for a particular event. Consequential failures shall also be considered. Failures of the: operators to perform required control manipulations shall be given consideration for permutations of the analyses. Operator actions that could cause the complete loss of function of a safety system shall also be considered. At present, these analyses need not address passive failures or multiple system failures in the short term. In the recent analysis of small break LOCAs, complete loss of auxiliary feedwater was considered. The complete loss of auxiliary feedwater may be added to the failures being considered in the analysis of transients and accidents if it is concluded 2.1.9-2 rj tr If II ( that more is needed in operator training beyond the short-term actions to upgrade auxiliary feedwater system relia-bility. Similarly, in the long term, multiple failures and passive failures may be consider ed depending in part on staff review of the results of the short-term analyses. The transient and accident analyses shall include event tree analyses, which are supplemented by computer calculations for those cases in which the system response to operator actions is unclear of these calculations could be used to provide important quantitative .information not available from an event tree. For example, failure to initiate high-pressure injection could lead to core uncovery for some transients, and a computer calculation could provide information on the amount of time available for corrective action. Reactor simulators may provide some information in defining the event trees and would be useful in studying the information available to the operators. 'The transient and accident analyses are to be performed for the purpose of identifying appropriate and inappropriate operator actions relating to important safety considerations such as natural circulation, prevention of core uncovery, and prevention of more serious accidents. 2.1.9-3 The information derived from the preceding analyses shall be included in the plant emergency procedures and operator training. It is expected that analyses performed by the NSSS vendors will be. put in the form of emergency procedure guidelines and that the changes in the procedures will be implemented by each licensee of applicant. In addition to the analyses performed by the reactor vendors, I analyses of selected transients should be performed by the NRC Office of Research, using the best available computer codes, to provide the basis for comparisons with the analytical methods being used by the reactor vendors. These comparisons together with comparisons to data, r including LOFT small break test data, will constitute the short-term verification effort to assure the adequacy of the analytical methods being used to generate emergency procedures. PVNGS Evaluation Arizona Public Service Company will work with Combustion Engineering to determine the generic analyses necessary for the System 80 design. The analyses will be performed on a schedule consistent with development of procedures, training of operator's installation of hardware (refer to Section 2.1.3) prior to PVNGS Unit 1 Operating License. 2.1.9-4 2.1.10 INSTRUMENTATION TO MONITOR CONTAINMENT CONDITIONS DURING THE COURSE OF AN ACCIDENT NUREG-0578 Position Consistent with satisfying the requirements set forth in General Design Criterion 13 to provide the capability in the control room to ascertain containment conditions during the course of an accident, the following requirements shall be implemented:

l. A continuous indication of containment pressure shall be provided in the control room. Measurement and indication capability shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and

~ minum five psig for all containments.

2. A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room. Measurement capability shall be provided over the range of 0 to 101 hydrogen concentration under both positive and negative ambient pressure.

3. A continuous indication of containment water level shall be provided in the control room for all plants.

A narrow range instrument shall be provided for PWRs and cover the range from the bottom to the top of the containment sump. Also for PWRs, a wide range instrument shall be provided and cover the range 2.1.10-1 0 from the bottom of the containment to the elevation equivalent to a 500,000 gallon capacity. For BWRs, a wide range instrument shall be provided and cover the range from the bottom to 5 feet above the normal water level of the suppression pool. The containment pressure, hydrogen concentration and wide range containment water level measurements shall meet the design and qualification provisions of Regulatory Guide 1.97, including qualification, redundancy, and testibility. The narrow range containment water level measurement instrumentations shall be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being periodically tested. PVNGS Evaluation PVNGS containment monitoring instrumentation is as follows:

1. Eight, Class IE containment pressure indicators are provided in the control room. Four indicators have a -4 to 20 psig rahge. The other four indicators have a -4 to 85 psig range.

An analysis performed for the containment maximum nega-tive pressure (refer to FSAR Section 6.2.1.1.3.6) demon-strates a maximum credible negative pressure of -2.6 psig. The existing pressure indicator range (to -4 psig) is, therefore, considered adequate to measure containment negative pressure under any postulated conditions. 2.1.10-2 l'g I,' The containment is designed for 60 psig internal pressure and tested to 69 psig. The maximum calcu-lated containment pressure under accident conditions \ is 49.2 psig. The existing pressure indicator range (to 85 psig) provides a 40% margin above the contain-ment design pressure and a 70K margin above the con-servatively calculated peak pressure. Therefore, a containment pressure range greater than the existing 85 psig maximum is not considered to be necessary. The addition of monitors with a range to 180 psig would add unnecessary confusion to the available containment I pressure instrumentation by:

a. Displaying at only 255 of span during the worst postulated accident.

b. Displaying containment pressure with an accuracy of only +3.6 psi (+2%).

2. A continuous indication of hydrogen concentration is available in the control room. Two Class IE redundant monitors are presently provided with a range of 0-10Ã.

The hydrogen monitors are manually initiated following a LPGA, Once initjated, they provide a continuous measure-ment of hydrogen concentration (refer to Figure 2,1.5-1). 2.1.10-3 i'

3. Continuous indicatjon of containment water ]eyel will be provided for PVNGS, An evaluation will be performed to determine the design requirements for this instrumentation. The results of this evaluation will be presented to the NRC staff by March;"1'981.

2.1.10-4 2.1.11 'NSTALL'ATION'OF'REMOTELY'OPERATED'HIGH POINT VENTS IN THE REACTOR COOLANT'SYSTEM NUREG-0578 Position Each applicant and licensee shal.l install reactor coolant system and reactor vessel head high point vents remotely operated from the control room. Since these vents form a part of the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A to 10 CFR Part 50 General Design Criteria. In particular, these vents shall be safety grade, and shall satisfy the single failure criterion and the requirements of IEEE-279 in order to ensure a low probability of. inadvertent actuation. Each applicant and licensee shall provide the following information concerning the design and operation of these high point vents:

l. A description of the construction, location, size, and power supply'for the vents, along with results of analyses of loss-of-coolant accidents initiated by a break in the vent pipe. The results of the analyses should be demonstrated to be acceptable in accordance with the acceptance criteria of 10 CFR 50.46.

2. Analyses demonstrating that the direct venting of non-condensable gases with perhaps high hydrogen concentrations does not result in violation of com-bustible gas concentration limits in containment as described in 10 CFR Part 50.44, Regulatory Guide 1.7 (Rev. 1), and Standard Review Plan Section 6.2.5.

3. Procedural guidelines for the operators'se of the vents. The information available to the operator for initiating or terminat'ing vent usage shall be discussed.

PVNGS Evaluation PVNGS will utilize existing penetrations on the pressurizer and reactor vessel head to provide high point venting of the Reactor 7 Coolant System (RCS). Refer to Figure 2,1.11-1 for a schematic of the RCS. The pressurizer vent (refer to Figure 2.1.11-2) penetrates the top of the pressurizer and is currently utilized for steam space sampling. The reactor vessel head'ent (refer to Figure 2.1.11-2) penetrates the top of the reactor vessel head and is currently piped to the Reactor Drain Tank (RDT), within containment, via a manual valve. Piping and valves will be provided to allow remote operation of the pressurizer and reactor vessel vents from the control room. In order to ensure a low probability of inadvertent action, valves and controls will satisfy the single failure criterion and IEEE-279-1971. 2.1.11-2 El The following information will be supplied as the RCS Vent System design is finalized (prior to Unit 1 Operating License):

l. A description of .the vents, piping, valves and controls utilized for the RCS Vent System. An .

analysis demonstrating the acceptability of a vent. piping'break in accordance with 10 CFR 50.46.

2. An analysis demonstrating that the venting of RCS

'non-condensable gases via the RCS Venting System does not result in containment hydrogen concentrations greater than those allowed by 10 CFR 50.44, Regulatory Guide 1.7, Revision 0 and Standard Review Plan 6.2.5.

3. Operating procedures for use of the RCS Vent System.

2.1.11-3 i 4l PRESSURIZER STEAM STEAM GENERATOR GENERATOR No. 2 No. 1 PUMP No. 1B'UMP No. 2B PUMP No. 2A PUMP No. 1A REACTOR VESSEL C-E I Figure SK5Hi~ I i ISOMETRIC VIEW OF THE REACTOR COOLANT SYSTEM 2.1.11.1 i'i 0 Shift Supervisor's Responsibilities 2.2.1 'MPROVED REACTOR 'OPERATIONS 'OMMAND FUNCTION 2.2.1. a SHIFT SUPERVISORS RESPONSIBILITIES NUREG-0578 Position

1. The highest level of corporate management of each licensee shall issue and periodically reissue a management directive that emphasizes the primary management responsibility of the shift supervisor for safe operation of the plant under all conditions on his shift and that clearly establishes his command duties.

2. Plant procedures shall be reviewed to assure that the duties, responsibilities, and authority of the shift E

supervisor and control room operators are properly defined to effect the establishment of a definite line of coranand and clear delineation of the command decision authority of the shift supervisor in the control room relative to other, plant management personnel. Particular emphasis shall be placed on the following:

a. The responsibility and authority of the shift supervisor shall be to maintain the broadest perspective of operational conditions affecting 2.2.1-1

t1, t l the safety of the plant as a matter of highest priority at all times when on duty in the control room. The idea shall be reinforced that the shi ft supervisor should not become totally involved in any single operation in times of emergency when multiple operations are required in the control room.

b. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators. Persons authorized to relieve the shift supervisor shall be specified.

c. If the shift supervisor is temporarily absent from the control room during routine operations, a lead control room operator shall be designated to assume the control room command function.

These temporary duties, responsibilities, and authority shall be clearly specified.

3. Training programs for shift supervisors shall emphasize 1

and reinforce the responsibility for safe operation and the management function the shift supervisor is to pro-vide for assuring safety. 2.2.1-2 , I PVNGS'Evaluation The responsibilities and authority of the Shift Supervisor is delineated in FSAR Chapter 13. 1. In addition, the Station Manual and the Operating Department Program also prescribe responsibilities and authority of the Shift Supervisor. The shift organization at PVNGS is such that the Shift Foreman will perform many of the administrative functions that are normally completed during a shift. 2.2.1-3 I 2.2.).b Shift Technical Advisor IMPROVED REACTOR OPERATIONS COMMAND FUNCTION SHIFT TECHNICAL ADVISOR NUREG-0578 Position Each licensee shall provide an on-shift technical advisor to the shift supervisor. The shift technical advisor may serve more than one unit at a multi-unit site if qualified to perform the advisor function for the various units. The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received spycific training in the response and analysis of the plant for transients and accidents. The shift tech-nical advisor shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign normal duties to the shift technical advisors that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience. PVNGS Evaluation There is sufficient time prior to PVNGS Unit 1 Operating License to assess fully the alternatives for providing 2.2.1-4 fl f' I additional technical and analytical capability for diagnosis of off-normal events. Two alternatives considered by the Lessons Learned Task Force have merit for PVNGS:

l. Upgrade the requirements for operators to include more engineering and plant response training.

2. Improve plant response diagnosis capabilities by backfit of computer and plant status display inno-vations.

The PVNGS Operator Training Program will be revised and upgraded to include more engineering and plant response training. The PVNGS Training Simulator will be used exten-sively to accomplish this objective. 'he PVNGS design already has incorporated some advanced informational displays. These include the Safety Equipment Status Panels, computer-generated graphics displays and the various information displays which can be generated by the Core Monitoring Computer and the Plant Monitoring Computer. 2.2.1-5 I Shift and Relief Turnover Procedures l ~ I l l 2.2.1 IMPROVED 'EACTOR 'OPERATIONS COMMAND 'UNCTION 2.2.1.c SHIFT RELIEF AND TURNOYER PROCEDURE NUREG-0578 Position The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:

l. A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete and sign. The following items, as a minimum, shall be included in the check-list:

a. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist).

b. Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the control console (what to check and criteria for acceptable status shall be included on the checklist);

2.2.1-6

c. Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications.

For such systems and components, the length of time in the degraded mode shall be compared with the Technical Specifications action statement (this shall be recorded as a separate entry on the checklist).

2. Checklists or logs shall be provided for completion by the offgoing and oncoming auxiliary operators and technicians. Such checklists or logs shall include any equipment under maintenance of test that by them-selves could degrade a system critical to the pre-vention and mitigation of operational transients and accidents or initiate an operational transient (what to check and criteria for acceptable status shall be included on the checklist); and

3. A system shall be established to evaluate the effect-iveness of the shift and relief turnover procedure (for example, periodic independent verification of system alignments).

PVNGS Evaluation PVNGS will have a procedure for shift relief which will include the use of checklists and the review of operating 2.2.1-7 il jl ,f logs. The relieving operator or supervisor will be required to sign that he has read the appropriate logs and completed the watch turnover as specified in the procedure. Provisions will be made for periodic verification of the relief and turnover procedure. The shift relief and turnover procedure and the provisions for its periodic verification will be prepared and imple-mented prior to PVWGS.:Unit 1 Operating License. 2.2.1-8 I I 2.2.2.a Control Room Access l 2.2.2 IMPROVED IN-PL'ANT'EMERGENCY'PROCEDURES 'AND'PREPARATIONS 2.2.2.a CONTROL ROOM ACCESS NUREG-0578 Position The licensee shall make provisions for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations supervisor, shift supervisor, and control room operators), to technical advisors who may be requested or required to suppor t the operation, and to predesignated NRC personnel. Provisions shall include the following:

1. Develop and implement an administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access.

2. Develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event of an emergency. The line of succession for the person in charge of the control room shall be established and limited to persons possessing a current senior reactor operator's license.

The plan shall clearly define the lines of communication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room. 2.2.2-1 fr PVNGS Evaluation Prior to PVNGS Unit 1 Operating Ljcepse, admjnistratjye procedures will be developed to limit personnel access to the unit control room during operational transient and accident conditions. These procedures will delineate:

1. That the shift super visor has the. authority to restrict control room access to personnel responsible for plant operations and technical advisory support to operations.

2. Clear lines of communications, authority, and responsibilities in the event of an emergency. Only senior licensed operators will direct the licensed operators.

2.2.2-2 0 Onsite Technical Support Center ~ f 2.2.2 IMPROVED 'IN-'PLANT'EMERGENCY 'ROCEDURES 'AND 'PREPARATIONS 2.2.2.b ONSITE TECHNICAL SUPPORT CENTER NUREG-0578 Position Each operating nuclear power plant shall maintain an onsite technical support center separate from and in close proximity to the control room that has the capability to display and transmit plant status to those individuals who are knowledge-able of and responsible I for engineering and management support of. reactor operations in the event of an accident. The center shall be habitable to the same degree as the control room for postulated accident conditions. The licensee shall revise his emergency plans as necessary to incorporate the role and location of the technical support center. Records that pertain to the as-built conditions and layout of structures, systems and components shall be stored and filed at the site and accessible to the technical support center under emergency conditions. Examples of such records include a system description, general arrangement drawings, piping and instrumentation diagrams, piping system isometrics, electrical schematics, wire and cable lists, and single line electrical diagrams. It is not intended that all records described in ANSI 45.2.9-1974 be stored and filed at the site 2.2.2-3 and accessible to the technical support center under emergency conditions; however, as stated in that standard, storage systems shall provide for accurate retrieval of all information without undue delay. PVNGS Evaluation An Onsite Technical Support Center for Palo Verde Nuclear Generating Station will be established prior to PVNGS Unit 1 Operating License. This center will be located within the site boundary and will provide a work area for supervisory and technical personnel from Arizona Public Service Company and the NRC. It will have communications with the control room and the other, response centers. An analysis will be performed to determine equipment to be incorporated into the center to monitor and display the status of the affected unit. The center will be habitable to permit occupancy following a loss of coolant accident. As-built drawings and other appropriate records shall be available in readily retrievable form at the site and shall be accessible to the Onsite Technical Support Center. The Palo Verde Nuclear Generating Station Emergency Plan will be revised to describe the Onsite Technical Support Center and its function. 2.2.2-4 S' Onsite Operational Support Center IMPROVED" IN-"PLANT EMERGENCY PROCEDURES AND PREPARATIONS ONSITE OPERATIONAL SUPPORT CENTER NUREG-0578 Position An area to be designated as the Onsite Operational Support Center shall be established. It shall be separate from the control room and shall be the place to which the operations support personnel will report in an emergency situation. Communications with the control room shall be provided. The Emergency Plan shall be revised to reflect the existence of the center and to establish the methods and lines of communi-cation and management. PVNGS Evaluation Prior to PVNGS Unit 1 Operating License, an area separate from the unit control rooms will be designated as the PVNGS Onsite Operational Support Center prior to receipt of operating license. This will be an assembly area where plant operations support personnel will report in an emergency situation for further orders or assignment. The Operational Support Center will have communication with the control room and the Emergency Control Center (refer to the PVNGS Emergency Plan). The Emergency'lan will reflect the existence of the Operational Support Center and will establish the methods and lines of communication and management. 2.2.2-5 0 8 'REVISED'LIMITING CONDITIONS "FOR'OPERATION OF NUCL'EAR 'P W R'PLANTS B SED UPON SAFETY'SYSTEM AVAILABILITY NUREG-0578 Position All NRC nuclear power plant licensees shall provide infor-mation to define a limiting operational condition based on a threshold of complete loss of safety function. Identifi-cation of a human or operational error that prevents or could prevent the accomplishment of a safety function required by NRC regulations and analyzed in the license application shall require placement of the plant in a hot shutdown condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The loss of operability of a safety function shall include consideration of the necessary instrumentation, controls, emergency electrical power sources, cooling or seal water, lubrication, operating procedures, maintenance procedures, test procedures and operator interface with the system, which must also be capable of performing their auxiliary or supporting functions. The limiting conditions for operation shall define the minimum safety functions for modes 1, 2, 3, 4, and 5 of operation. The limiting conditions of operation shall require the following: 2.2.3-1 8 1 ~

l. If the plant is critical, restore the safety function (if possible) and place the plant in a hot shutdown condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

2. 'ithin 24 hours, bring the plant to cold shutdown.

3. Determine the cause of the loss of operability of the safety function. Organizational accountability for the loss of operability of the safety system shall be established.

4. Determine corrective actions and measures to 'prevent recurrence of the specific loss of operability for the particular safety function and generally for any safety function.

5. Report the event within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirm by telegraph, mailgram, or facsimile trans-mission to the Director of the Regional Office, or his designee.

6. Prepare and deliver a Special Report to the NRC's Director of Nuclear 'Reactor Regulation and to the Director of the appropriate regional office of the Office of Inspection and Enforcement. The report shall contain the results of Steps 3 and 4 above, 2.2.3-2

S, along with a basis for allowing the plant to return to power operation. The senior corporate executive of the licensee responsible and accountable for safe plant operation shall deliver and discuss the contents of the report in a public meeting with the Office of Reactor Regulation and the Office of Inspection 'uclear and Enforcement at a location to be chosen by the Director of Nuclear Reactor Regulation.

7. A finding of adequacy of the licensee's Special Report by the Director of Nuclear Reactor Regulation will be required before the licensee returns the plant to power.

PVNGS'Evaluation In his August 20, 1979 memorandum to the NRC Coranissioners, Harold R. Denton, Director, Office of Nuclear Reactor Regulatory, announced his intent to initiate a rulemaking proposal on limiting conditions of operations, which would include con-sideration of the recommendations of the Lessons Learned Task Force and alternative approaches. APS will follow this proposal with great interest and will participate in any rulemaking proceedings that may be instituted. 2.2.3-3 l pygmy yElggE NIUCILEAIR GEMERA7IINIG SMfIIGH UNlTS 1,2@8 TMII-2 LESSGNS 1EABNEG . m~~a~<NTAAGM BEFOG 8~00 6 280 s~n~~<<< .>97.3 p 4 /j'WMMa P. 0 BOX ~Ba aCOZQPL5M't 2I666 'HOENIX, ARIZONA 85036 September 28, 1979 ANPP-13905 - JMA/DBK t Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission klashington, D.C. 20555 Re: Palo Verde Nuclear Generating Station Units 1, 2 and 3 Docket Nos. STN 50-528/529/530

Dear Sir:

As a result of the Three Mile Island accident, Arizona Public Service Company (APS) is in the process of assuring that the lessons learned from TMI-2 are reflected in the Palo Verde Nuclear Generating Station (PVNGS) design and operation.

To accomplish this objective, the APS President and Chief Executive Officer, by a directive issued April 12, 1979, established a Safety

'Evaluation Task Force "to conduct a complete review and analysis of the Three Mile Island accident and its implications for the Palo Verde Nuclear Generating Station project." This task force is com-prised of fourteen (14) members from knowledgeable, high levels of management of the utilities participating in the Palo Verde project, the engineer-constructor for the project, the nuclear steam supply

.system supplier, and turbine generator supplier to the project and the academic technical consultants to the Arizona State Senate and House legislative committees organized to review and consider the implications of the TMI accident.

The task force is exercising oversight of the on-going independent in-depth reviews, initiated since the TMI accident, of the Palo Verde design, planned administrative and operational procedures, training and qualification programs, and emergency response plans. In this con-nection, it is closely following the work and recommendations of the ,

many industry and governmental organizations currently conducting similar, but more generic investigations related to the TMI accident, including particularly the NRC's Lessons Learned Task Force.

4 Director of Nuclear Reactor Regulation gO September 28, 1979 ANPP-13905 - JMA/DBK Page 2 Upon receipt of the THI-2 Lessons Learned Task Force Status Report and Short Term Recommendations (NUREG-0578), modified as set forth in the memorandum dated August 20, 1979 from Harold R. Denton, Director, Office of Nuclear Reactor Regulation to the NRC Commissioners, the recommendations related to the design and operation of the Palo Verde project were evaluated. The results of this evaluation are incorporated in the report entitled THI-2 Lessons Learned Implementation Report submitted herewith. The report identifies those NUREG-0578 recommendations which are incorporated into the present Palo Verde design, identifies recommendations that are not applicable to the Palo Verde design and sets forth APS'ommitments to develop the analyses, design modifications, operational procedures and other measures necessary to meet such recommendations in accordance with the schedule specified in NUREG-0578. The report will be amende'd as appropriate to incorporate the manner in which such commitments are implemented.

Five (5) copies of this report are enclosed for your review. It is hoped that this report, in conjunction with the PVNGS Final Safety Analysis Report to be submitted in October, 1979, will assist in your timely review of the implications of THI on PVNGS.

Respectfully submitted, ARIZONA PUBLIC SERVICE'COMP Edwin E. Van Brunt, Jr.

Vice President STATE OF ARIZONA )

) ss. On its own behalf and as agent for

>,s ~ ~

County of Haricopa ) all other joint participants.

"..;Subscribed and sworn to before me tbis~gday of , 1979.

i>' I~.

I e

) Hy Commission expires:

~

lc iMy.tommississ Expires Jsn. 23, sess

~ ~

s

~

s,) ~

s ps 6 e

s s l ~ ~

s s s

~

TABLE OF'CONTENTS INTRODUCTION DESIGN AND ANALYSIS PAGE NO.

2.1.1 Emergency Power Supply Requirements for the Pressurizer Heaters, Power Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs. ~ ~ 2.1.~1-1 2.1.1.a Pressurizer Heater Power Supply . 2.1.1-1 2.l.l.b Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indocators. . 2.1.1-5 2.1.2 Performance Testing for BWR and PWR Relief and Safety Valves. 2.1. 2-1 2.1.3 Information to Aid Operators in Accident Diagnosis and Control. 2.1.3-1 2.1.3.a Direct Indication of Power Operated Relief Valve and Safety Valve Position for PWRs and BWRs . 2.1.3-1 2.1.3.b Instrumentation for Detection of Inadequate Core Cooling in PWRs and BWRs . 2.1.3-3 2.1.4 Containment Isolation Provisions for PlJRs and BWRs . 2.1.4-1 2.1.5 Post-Accident Hydrogen Control Systems for PWR and BlJR Containments. 2.1.5-1 2.1.5.a Dedicated Penetrations for External Recombiner or Post-Accident External Purge System. 2.1.5-1

i li Jr'!i 1

DESIGN AND ANALYSIS (continued) PAGE NO.

2.1.5.b Inerting BWR Containments. . . . . . . . . 2.1.5-3 2.1.5.c Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant. . . . . . . . . . . . 2.1.5-4 2.1.6 Post Accident Control of Radiation in Systems Outside Containment of PHRs and BWRs . . . . . . . . . . . . . . . . . . . . 2.1.6-1 F 1.6.a Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs . . 2.1.6-1 2.1.6.b Design Review of Plant Shielding and Environmental gualification of Equipment for Spaces/Systems Which May Be Used in Operations 2.1.6-3 2.1.7 Improved Auxiliary Feedwater Reliability for PWRs..... ~........

System 2.1.7-1 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWRs.... 2.1.7-1 2.1.7.b Auxiliary Feedwater Flow Indication to Steam Generator for PHRs........ 2.1.7-7 2.1.8 Instrumentation to Follow the Course of an Accident . . . . . . . . . . . . . . . . . 2.1.8-1 2.1.8.a Improved Post-Accident Sampling Capabili'ty . . . . . . . . . . .- . . . . . 2.1.8-1 2.1.8.b Increased Range of Radiation Monitors... 2.1.8-4 2.1.8.c Improved In-Plant Iodine Instru-mentation. 2.1.8-7 2.1.9 Analysis of Design and Off-Normal Transients and Accidents . 2.1.9-1

0 DESIGN AND ANALYSIS (continued) PAGE NO.

2.1.10 Instrumentation to Monitor Contain-ment Conditions During the Course of an Accident. . . . . . . . . . . . . . . . . . 2.1.10-1 2.1.11 Installation of Remotely Operated High Point Vents in the Reactor Coolant System. . . . . . . . . . . . . . . . . . 2.1.11-1 OPERATIONS 2.2.1 Improved Reactor Operations Command Function. 2.2.1-1 2.2.l.a Shift Supervisors Responsibilities. 2.2.1-1 2.2.l.b Shift Technical Advisor . 2.2.1-4 2.2.1.c Shift Relief and Turnover Procedure 2.2.1-6 2.2.2 Improved In-Plant Emergency Procedures and Preparations. 2.2.2-1 2.2.2.a Control Room Access . 2.2.2-1 2.2.2.b Onsite Technical Support Center . 2.2.2-3 2.2.2.c Onsite Operational Support Center . 2.2.2-5

j INTRODUCTION The Palo Verde Nuclear Generating Station NUREG-0578 Evaluation addresses the "TMI-2 Lessons Learned Task Force Status Report And Short Term Recommendations" (NUREG-0578), July 1979, as supplemented by letter dated August 20, 1979 from Harold R. Denton, Director, Office of Nuclear Reactor Regulation to the NRC Commissioners.

Each recommendation is addressed, as it applies to the Palo Verde Nuclear Generating Station (PVNGS) design and operation, in the following manner:

(1) Descriptions of the existing PVNGS design correspond with information presented in the PVNGS Final Safety Analysis Report (FSAR).

(2) Proposed modifications, evaluations and analyses of the PVNGS design, as discussed herein, will be summarized in revisions to this report.

(3) Recommendation numbering corresponds to that used in NUREG-0578.

This report will be referenced by subsequent PVNGS FSAR amendments, as appropriate, to address in the FSAR, required modifications and analyses as discussed herein.

I I

2.1.1.a Pressunzer Heater Power Supply

2.1.1 EMERGENCY POWER SUPPLY RE UIREHENTS FOR THE PRESSURIZER HEATERS, POWER OPERATED RELIEF VALVES AND BLOCK VALVES, AND PRESSURIZER LEVEL INDICATORS IN PHRs .

2.1.l.a PRESSURIZER HEATER POWER SUPPLY NUREG-0578 Position

1. The pressurizer heater power supply design shall provide the capability to supply, from either the offsite power source or the emergency power source (when offsite power is not available), a predetermined number of pressurizer heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions.

The required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supply capability.

2. Procedures and training shall be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency buses. If required, the procedures shall identify under what con-ditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters.

3. The time required to accomplish the connection of the pre-selected pressurizer heater to the emergency buses shall be consistent with the timely initiation and maintenance of natural circulation conditions.

2.1.1-.1

4. Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with safety-grade requirements.

PVNGS Evaluation The PVNGS pressurizer heaters are powered as follows:

1. The pressurizer heaters (36 elements rated at 50 KW each, connected in 12 groups of 3 delta-connected elements each) have a total capacity of 1800 KW. The heaters are supplied 480 volt power from the non-Class IE and Class IE distribution systems as follows:

Number of Heaters Ca acit KW 480V Bus IE'Bus A. 5-3 element groups 750 NGN-Lll No B, 5-3 element, groups 750 NGN-L12 No C. 1-3 element group 150 PGA-L33 Yes D. 1-3 element group 150 PGB-L32 Yes Maintenance of natural circulation at hot standby con-ditions requires 150 KW of heaters. These heaters are immediately available from either of two redundant Class IE buses, fed independently by Safety Trains A and B. Controls for these heaters are likewise inde-pendent. As the Class IE buses are a part of the ESF distribution system, they are fed either from the off-site power system or the diesel generators when offsite power is not available.

2.1.1-2.

S Heaters supplied from the non-Class IE power system are fed from the offsite power system and are available for manual interconnection to the diesel generators when offsite power is not available.

2. Heaters fed from the Class IE power system (150KW) are automatically connected to the ESF buses, following either a loss of offsite power or an engineered safety features actuation system signal, by their respective load sequencer.

Heaters fed from non-Class IE buses can be manually connected to the Class IE buses for powering by either the offsite power system or the diesel generators through the ESF transformers. These heaters are available in 150 KW blocks (0-150 KW in the case of the two proportional groups). Sufficient margin exists in the diesel generators for at. least an additional 300 KW of heaters to be backfed from each Class IE bus without the shedding of selected emergency loads.

2.1.1-3

9 I

II If t

.3.'o manual connection of pressurizer heaters to the Class IE power system is required to maintain natural circulation at hot standby.

4. The redundant pressurizer heater required for maintenance of natural circulation at hot standby are fed from Class IE buses via Class IE load side breakers qualified in accordance with safety-grade requirements.

The PYNGS pressurizer heaters, as presently design, comply with the recommendations of NURE6-0578, July 1979.

2.1.1-4

0 Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indicators

2.1.1 EMERGENCY POWER SUPPLY RE UIREMENTS FOR THE PRESSURIZER HEATERS, POWER OPERATED RELIEF VALVES AND BLOCK VALVES, AND PRESSURIZER LEVEL INDICATORS IN PWRs 2.1.l.b POWER SUPPLY FOR PRESSURIZER RELIEF AND BLOCK VALVES AND PRESSURIZER LEVEL INDICATORS NUREG-0578 Position

l. Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.

2. Motive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.

3. Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be through devices that have been qualified in accordance with safety-grade requirements.

4. The pressur izer level indication instrument channels shall be powered from the vital instrument buses. These buses shall have the capability of being supplied from either the offsite power source or .the emerqency power source when offsite power is not available.

2.1.1-5 .

PVNGS Evaluation 1; PVNGS does not use PORVs and block valves.

2. PVNGS does not use PORYs and block valves.

3. PYNGS does not use PORYs and block valves.

4. Two channels of Class IE level instrumentation are provided for PVNGS. Pressurizer level channels L-110X and L-llOY (refer to Figure 2.1.11-2) are indicated in the control room. Channel L-llOX is also recorded in the control room.

The pressurizer level instrumentation is powered from 120V AC Class IE instrument buses E-PNA-D25 and E-PNB-D26 (refer to Figure 2.1.1-1). These buses are normally powered through inverters from Class IE batteries. The Class IE battery chargers are powered from offsite power or from the diesel generators when offsite power is not available.

The pressurizer level indicators, as presently designed, comply with the recommendations of NUREG-0578, July 1979.

2.1.1-6.

0 440v uCC 4eov Mcc ~ eov MCC 440V IICC ~ eov MCC 440V IICC ~ eov MCC 4eov ucc 440V IICC PHATIC 440V UCC E PHA 11)S E PHA M)S E PHA u)) C PHA M)I E PHA M)I C PHS hi)4 E PHS Il)6 C PHS M)4 E PHS M)2 E. PH 8. M ) 2 OWC I) C PHA 005 Owo 1)-E PHA<05 Owo I) C PHAH)o) DWC I) E PHA&ol O'WC IS f PHA 001 OWC 1) ~ K PHA 004 DwC 1) C-PHA 004 OWC l)-E-PHA.C04 Owo I) E DWC I) C PHA 002

~ 44042ov 25KVA Q 25KVA 440 Itov

+ RSKVA 440 ltOV Z5KVA 440-120V i VOLTAGE VOL'TACE RCCtlLATOR VOLTACE I VOLTACE RECVLATolt i 1822X I RCCULATOR 'tECULATOR i 2'X E-PNA VRS 1400 AH 440AM I 118.2 2x io.t RX E PNS V24 1400 AH 750 AH

'8,2 E PND V2 ~

I 0'HC 1 E PNC VRY its.oc 0' I

~ I)

J I) E PNIOOI IRSVDC dlTTERY A 8ATTERY C fl)

~ I) 1 DWC CWNAH)OI I DWC I)-E-PNAK)02 IRSV DC SATTEIIY 8 f PKS Fit SATTCRY fll I DWC E PNA-002 E-PKA fli 4 00 A 400A E. PKC SAT CHCR 4AT CHCR 4 00 A SAT OICR E PKD 8AT CHCR IRSVDC 8AT CHCR 1 5VDC SAT CHCR

' 400 A 8D' Its OC D

A Ac C itevOC C-PKa.uil E-PKAHMS PKC Hl) f PKS HIR PKS.Hie f PKD HI4

)Col !OOA 12$ VOC it)VDC NOTE ) NOTE )

1 I I I No) I NC)

I I

12SV DC CONTROL CENTER f-PKl-M41 IZSV DC CONTROL CENTER E PKC-M45 I 12)V DC CONTROL CENTClt f PKS M42 125V DC CONTROL CENTER E. PKO-U44 I I I; I I 7) ct NC)

I I

+) ~)

I I

I I ") I>

I

' I I

I I

I r--

I I 2SV DC I

I I

I I

I Mtfv oc I I I

I I 0!STR PNL I

Olsl tNL D2) I I

I I

I DISTR ~ NL I I. I I

015111 tNL I

I I I

owc t)-c-PKA-00)

I L L E PKC J I J eov !8. L owc C PKS 022 t) c-PKA-004 J J I

I L E PKD Dte

~J DWC 1)-C PKA-002 DwC 1)-C-1 Ka-004 ~

ZSKVl owc t)-c-PKA-Oos DWC 0 E PKA 00'7 4eov.)e.

2)KVA 1NVERTfit lNVOITCR ItSV OC To IRSV DC To C'-PKC-N4) nl C.PKD.N ~ 4 12)v Dc To 12)V DC TO REACTOR Tait SHITCHCEllt J SSA-Co)

A'C ucvs I TYPTCAL FOR REACTOR TRIt SWlTCHCEAR C J-58C Co) LOCAL 440V.'Sll.

REACTOR SWITCHCEAlt Sed Co)

Tltlt'd' REACTOlt TRIP SNITCHCEAR D LOCAL 440V. )11.

J 580 Co)

VP.DHC1) 10407 CHAHNKLS A.c ANO 0) VP. DWC 1) 10407 STARTfR STARTKR V P. 0 wc 1).10407 V P. DWC 1) 10407 STARTER STARTER NOOI I) 0) )0 Nool-1) 0) )0 f-PKC-S4) NOOI I) 0) )0 Wool I) ~ 0) )0 C-PKO-844 3C SHVTDOWN SHVToowN CLC lsoL.VLV 2 4 CLC 1$ 0L VLV J sic Uv es) Ht J-510-UV SS4 OwC I) C-Sle 014 IZOV. I 0+ )tov,tgo im.io 1 f V.gtd RSKVA ZS KVA tsKVA 2)KVA I INVERTER lNVERTER I lNVERTER )NVERTER A I C 8' J I D I E PNA-Nil E.PliC-NI) I PNS NIRJ E.PIIO.N14 I I I Off NC MANUAL I OfF NC MANUAL UauvaL fflt SW TRAN5ffa O'W TRANSFER sw 0 I NO TRAN$ Ntl I I 1RANSFElt SW ND I L r=== =,==i I

) I )

I I

I I I I I I I I ltovac UNCROUNofo 1tovac UNCRoUNDE o 120VAC UNCROVNDED 120V AC UNCROVNOCO i 14,VI'TAL INST I CISTR PNL J 14,VTTAL )NST Cl)'TR 14 vlTaL lusT otsld PNJ '4.vlraL lust otsTR PNL i 1 OWC CHANNEL A I)-E-PNA 001 OWC I)-f PNA 001 J cuaNNKL 8 DWC 't) C PNA-002 I

CHANNEL'0'wC t)-E-PNA OOR LOAD CROUP I LOAD CROVP t 13-E-PKA-001 REV 1 Palo Verde Nuclear Generating Station

.@@i?

NOTE)1 I ~ FOR LCCflK> 4 CCNEitAL NOTES REfCR To OWCS I) E ZZS OOI SINGLE LINE DIAGRAM

'THROUCH 007 2 THIS )ND)CATE) MANVALTWO POLE 1HRfE t05I'llou TRANSffit DC POWER SYSTEM SWITCH OUTPUT SWITCHES IN SllTCRY CHARCERS MECHlNlCALLY lutERl.oCKED '10 Off Eit THC lC 8'80 ARC tos)lelLIT 1 OF Figure 2.1.1-1 ETTHER dOTH OPEN OR ONC CLOSED ANO ONC Otfu AT ANY TINE

f

-0

PERFORMANCE TESTING FOR BWR AND PWR RELIEF AND SAFETY VALVES NUREG-0578 Position Pressurized water reactor and boiling water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected oper-ating conditions for design basis transients and accidents.

The licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2. The single failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized. Test pressures shall be the highest predicted by conventional safety analysis procedures. Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry piping and supports, as well as the valves themselves.

PVNGS Evaluation PVNGS does not utilize pressure operated relief valves.

An analysis will be performed to determine the expected code safety valve operating conditions based upon the accidents and anticipated operational occurrences described in the PVNGS FSAR.

2.1.2-1

0 When this analysis is completed, generic code safety valve qualification tests will be performed, as necessary, to qualify the valve for expected operating conditions following- an accident or anticipated operational occurrence.

In addition, the discharge piping and supports will be analyzed for their expected operating conditions. Per-formance of qualification tests is dependent on the availability of test facilities capable of simulating the expected operating conditions.

The results of these analyses and tests will be provided prior to PVNGS Unit 1 Operating License.

2.1.2-2

2.1.3.a Oirect Indication of Power Operated Relief Valve and Safety Valve Position for PWRs and BWRs

2.1.3 'INFORMATION'TO'AID"OPERATORS IN ACCIDENT DIAGNOSIS'AND CONTROL 2.1.3.a DIRECT INDICATION OF POWER-OPERATED RELIEF VALVE AND SAFETY VALVE POSITION FOR PWRs'- AND BWRs NUREG-0578 Position Reactor system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe.

PVNGS Evaluation PVNGS does not utilize -'power' operated relief valves.

The PVNGS pressurizer code safety valves are headered into the reactor drain tank (RDT), inside containment. Upstream of the common header each code safety valve is monitored for seat leakage by an in-line RTD (refer to Figure 2.1.11-2). Indirect indication of code safety valve operation is provided by RDT pressure, pressurizer pressure and pressurize'r level instrumentation.

An evaluation will be performed to determine if the multiple indirect indications of code safety valve status provided in I

the PVNGS design provide unambiguous indication that a small break area LOCA is in progress.

2.1.3-.1

I I

f

If the evaluation determines that direct valve status indication in the control room is required (e.g. by acoustic techniques, flow measurement or valve position indication), it will be incorporated in the PVNGS design prior to Unit 1 Operating License.

Results of this evaluation will be provided by January 1980.

2.1.3-2

0 2.1.3.b Instrumentation for Detection of Inadequate Core Cooling in PWRs and BWRs

2.1.3 INFORMATION TO AID OPERATORS IN DIAGNOSIS AND CONTROL 2.1.3.b INSTRUMENTATION FOR DETECTION OF INADE(UATE CORE COOLING IN PWRs AND BWRs.

NUREG-0578 Position

1. Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available instrumentation. The licensee shall provide a description of the existing instru-mentation for the operators to use to recognize these conditions . A detailed description of the analysis needed to form the basis for operator training and procedure development shall be provided pursuant to another short-term requirement, "Analysis of Off-Normal Conditions, Including Natural Circulation".

In addition, each PWR shall install a primary coolant saturation meter to provide on-line indication of coolant saturation condition. Operator instruction as to use of this meter shall include consideration that is not to be used exclusive of other related plant parameters.

2.1.3-3

0

2. Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-interpret indication of inadequate core cooling. A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.

PVNGS Evaluation Information to diagnose inadequate core cooling is as follows:

1. Prior to PVNGS Unit 1 Operating License, procedures will be developed' to aid the operator in recognition of inade-quate core cooling. These procedures will utilize existing redundant safety related instrumentation such as:

A. Pressurizer Pressure 1500-2500 psia B. Pressurizer Pressure 0-3000 psia C. Pressurizer Level 0-100K D. Coolant Temperature

- Hot Leg 375-675F

- Cold Leg 465-615F 2.1.3-4

These procedures will be based upon analyses performed as discussed in Section 2.1.9. Redundant subcooling meters will be installed prior to PVNGS Unit 1 Operating License to provide additional indication of primary coolant saturation condition. The procedures discussed in Item 1 will incorporate the use of primary coolant saturation indication.

2. Upon completion of the analyses described in Section 2.1.9, an evaluation will be performed to determine the need, if any, for additional instrumentation, such as reactor vessel level indication. Any instrumentation determined to be necessary will be installed prior to PVNGS Unit 1 Operating License.

2.1.3-5

t CONTAINMENT ISOLATION PROVISIONS FOR PMRs AND BllRs NUREG-0578 Position

l. All containment isolation system designs shall comply with the recommendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the init-ation of containment isolation.

2~ All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essential, shall identify each system determined to be non-essential, shall describe the basis for selection of each essential system, shall modify their containment isolation designs accordingly, and shall report the results of the re-evaluation to the NRC.

3. All non-essential systems shall be automatically isolated by the containment isolation signal.

4. The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic re-opening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action.

2.1.4-]

PVNGS'Evaluation The PVNGS containment isolation system is designed as follows:

l. As required by SRP 6.2.4, a containment isolation signal is diversely generated by either'a high containment pressure signal (5 psig) or a low pressurizer pressure signal (1600 psig). These parameters additionally generate a safety injection actuation signal. The power access purge is additionally isolated by high containment purge radioactivity.

2. A review of the PVNGS containment i,solation system will be performed to determine that no essential systems are auto-matically isolated by a containment isolation actuation signal(CIAS). Essential systems will be selectively iso-lated by the operator after their use is no longer required.

The results of this review will be provided to the NRC staff by January 1980. Plant modifications, if any, will be per-formed prior to PVNGS Unit 1 Operating License.

3. All non-essential systems penetrating the PVNGS contain-ment are isolated by CIAS.

4. Resetting of a CIAS does not result in the automatic opening

'I of containment isolation val ves. Re-opening requires operator action for each valve.

2.1.4-2

~i t

2.1.5.a Dedicated Penetrations for External Recombiner or Post-Accident External Purge System

2.1.5 POST "ACCIDENT HYDROGEN'CONTROL SYSTEMS FOR 'PWR'AND BWR'CONTAINMENTS 2.1.5.a DEDICATED PENETRATIONS FOR EXTERNAL RECOMBINER OR POST-ACCIDENT EXTERNAL PURGE SYSTEM NUREG-0578 Position Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere should provide containment isolation systems for external recombiner or purge systems that are dedi-cated to that service only, that meet the redundancy and single failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR Part 50, and that are sized to satisfy the flow requirements of the recombiner or purge system.

PVNGS Evaluation Combustible gas concentration in the PVNGS containment building is maintained by the Containment Hydrogen Control System.

The PVNGS Containment Hydrogen Control System (refer to Fig-ure.,2.1.5-1) shares no equipment or penetrations with the normal containment purge system and includes:

2.1.5-1

1. Two redundant and independent hydrogen gas analyzers with separate piping, valves, controls and instrumentations (refer to Section 2.1.10).

2. Redundant and independent mobile, hydrogen recom-biners (two for the PVNGS site) with separate supply and return lines, utilizing separate penetrations for each line.

3.'edundant and independent supply and return line isolation valving (one valve inside containment and one outside containment) that meets the requirements of 10 CFR 50, Appendix A, Criterion 54 and 56.

4. Dedicated containment penetrations sized for maximum recombiner flow requirements.

The PVNGS Containment Hydrogen Control System, as currently designed, provides dedicated penetrations and isolation systems that meet redundancy and single failure requirements as recommended by NUREG-0578, July 1979.

2.1. 5-'2

2.1.5.b lnerting BWR Containments

2.1. 5 POST ACCIDENT HYDROGEN CONTROL SYSTEMS FOR PWR AND BWR CONTAINHENTS 2.1.5.b INERTING BWR CONTAINHENTS NUREG-0578 Position It shall be required that the Vermont Yankee and Hatch 2 Mark I BWR containments be inerted in a manner similar to I

other operating BWR plants. Inerting shall also be required for near term OL licensing of Mark I and Hark II BWRs.

PVNGS Evaluation PVNGS uti lizes Combustion Engineering,:-Inc. System 80' PWRs with a containment with 2.6x106 ft3 net free volume.

Inerting of the PVNGS containment is not required.

2.1.5-3

I, 2.1.6.c Capability to Install Hydrogen Recornbiner at Each Light Water Nuclear Power Plant

0 2.1.5. POST ACCIDENT HYDROGEN CONTROL SYSTEMS FOR PWR AND BWR CONTAINMENTS 2.1.5.c CAPABILITY TO INSTALL HYDROGEN RECOMBINER AT EACH LIGHT WATER NUCLEAR POWER PLANT NUREG-0578 Pos i ti on (Mi nori ty Vi ew)

l. All licensees of light water reactor plants shall have the capability to obtain and install recombiners in their plants within a few days following an accident if containment access is impaired and if such a system is needed for long-term post-accident combustible gas control.

2. The procedures and bases upon which the recombiners would be used on all plants should be the subject of a"review'by. the licensees:in considerihg shielding requirements and personnel exposure limitations as demonstrated to be necessary in the case of TMI-2.

PVNGS Evaluation The PVNGS hydrogen recombiners are designed as follows:

1. PVNGS will have available, onsite, two mobile hydrogen recombiners designed to limit the concentration of hydrogen in containment. These hydrogen recombiners are capable of being installed in the affected unit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an accident.

2.1.5-4

o

2. Installation procedures for the hydrogen recombiners

'ill be developed and shielding requirements re-analyzed prior to PVNGS Unit 1 Operating License.

2.1.5-5

8 l

~OQ~

'V 4

~

~

~

~

~

V 0p g

C CR w

SA CR t~i 0 A

I v

~~ ~w ts CR A

w

~ M ic CR RCPOtC A

I CONTAI laCNT POST LOCA t * ~

NTDIIOCKN 4

~ ION I ION HV ZSHL 5 VDZ4 ~

A

~ RC ~ v ~ ~ I~ ~ ~ /IXIT VDZS SCC NOTC 2 ee Ve

~ I ~

~

0 NS O ~

5

• IJ ~

CI AS ~ g rcsr CLOSCS VOI 5 II ~

I/ST I X W UV ZSHL 0 4 p 's 0 A Ool HCOA o ~ ~

I /IC 27 A

~ A Ool HCOA 2 15CC NOTC II VOOZ

~ ~ ~ ~ ~

Vl 0

~ ~

'g5 ~ ~ ~ ~ ~ ~ ~ r--+-----4 D,~

5CC NOTC 2 6

CR ~ Cll TO ATIIDSPICRE CO---'I N NPO Hole NS

-;*<+ VCIAS CLOSCS ge He

~ I 55

'g5

~ ~

'I UY ZSHL

~ ~

CONTA(eaCNT

~

~

~

~

~

~

~

V tD I V~

ltSL

~

HIDROCCN ~

p ~

~ ~ 54 I RCCOIDI NCR I A 001 NCOA 2 SISIKH

,SCC OCCHTtL LOC,

~

~

~

~

15CC NOTE ~ I A 001 HCOA 2+ Yol 0 I

~

Vol 5 ~

~

~ 0 PI PD I to I 0 0

~ ~

A 001 NCOA 2 I I ~ ~

ZS 55 54

~ ~

LJ L J ~

~

~

~

0 ~ I TEST TEST kllJCRUZQ1 ~

~

~

~ N 015 NCDA 2 ~ ro FLANt QZA ~ ~ CPP 001 VEN1

~

~

~

~

~

~

• s IF I)

VI I 0

II N 005 HCDA 2 ~

~

~

~

~

I WX

~ ~

~ ~

I g

0 ~ ~

~

N Ol ~ NCDA

~ ~

~ ~

~

~

~

~ 40

~ ~ Pl II ep~ X ls Wo ~

~

'o

~

CLOSES>+ aoscs>+ 0 004 HCOA 2 gl r ee tCV I

'g5 'g5

~

~ ~ ST gTCSI "1 ~ ~ 0 0

5CRVI CC VOI 5 C ~

~

~

~ IA 04 ~ HOOR I *IN CCIITAIIACNT ~ ~ IAP 002

~ ~

HTOROCEN lc gg I/ls 0 005 NCOA 2 ~ ~ Voor 11 I RCCOICI NEO 515'ltp C OtCNtE(. LOC.

~

~

~

~ 8 ~A NCCIA IIODA

,.J

~ ~

(SCC NOTC ~ I O 4 Yol I No Nits ~ 5 Vol 2 ~ ~ 0 0 ~ 0-004-HCOA-5/5 ~ I I

~

~

~

~ 0

~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

LJ ~ ~ 2 NOTCS ~

TEST HPA IXII ~ ~

\ INSTALL VALVES WITH FLOY DIRECTION

~

6

~

ARROW IN DIRECTION NOTED ee TO CLOSC ACAIN5T CONTAllaCNT DESICN PRCSSURE al Hts Pol

l. LINC tNE SISICP OCSICNATOR Ht IS TO PRECEDE ALL VALVC AND INSIRUICN'I IAIIDCRS SHOWN tltE 5HALL OE THIICAOCO 10 ACCEPT A r S Ueacss OTHERWISE IHOICATEO A 2 X I CAttto ICSI RCDUCCR ON THIS DRAWINC AS SHOWN Otl.OW, L

TEST 0P I

UY ZSNL I

~

Ttst n

0 ve VI I

4 ~ ELECTRICAL HCAT IRACINC SNALL OC POWCRCD SRoal IC 50URCC OF PDICR. TESI L~ 2 DURINC NORPAL OPERATION VCH p

0 ~

0 004 NCOA 2 Ins NS 0 ENO OF Saett tltt 1$ OttN, tROVIDC COUtLINCS FOR tRDTCC'IION OF IHC THRCAD

~ (SCC HOTC 11 V004 CR INCR SUPPLY AIR 0 004 ICOA Zw/ NS FILTRATION UNIT g rest 01 roti 4

~l

~ C I aa w CR CR

'Q5

~ ~ ~

o 5101 TCST SKLCCT 13-M-HPP-001 REV 3 CONTAINaCNT POST LOCA 0 HraloCCN ~ ~

~ ION I IDR SCC NOTC 2 CR Palo Verde Nuclear Generating Station V4 LJ AI NZ 10 I/4 Vole 4 0 VOlt J ~ IIPO COZ Olc COMBUSTIBLE GAS CONTROL SYSTEM 2

C Figure 2 . 1 . 5-1

0 0 2.1.6.a Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs

~ a ~g I

2.1.6 POST ACCIDENT CONTROL OF RADIATION IN SYSTEMS OUTSIDE CONT INMEN OF PWRs AND BWRs 2.1.6.a INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE MATERIALS (ENGINEERED SAFETY SYSTEMS AND AUXILIARY SYSTEMS) FOR PWRs AND BWRs NUREG-0578 Position Applicants and licensees shall imoediately implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels.

This program shall include the following:

1. Immediate Leak Reduction

a. Implement all practical leak reduction measures for all systems that could carry radioactive fluids outside of containment.

b. Measure actual leakage rates with system in operation and report them to the NRC.

2. Continuing Leak Reduction Establish and'mplement a program of preventive maintenance to reduce leakage to as-low-as-practical levels in systems outside containment 1'ikely to contain highly radioactive fluid during a serious transient or accident.

2.1.6-1

8

PVNGS 'valuation The PVNGS program of preventive maintenance will reduce leakage to as low as practical. The program will include:

l. In-service inspection of ASHE Class 2 and 3 systems in accordance with ANSE Boiler and Pressure Vessel Code Section XI, including pressure tests.

2. Alarming of abnormal airborne radioactivity levels by the radiation monitoring system for gaseous rad-waste system leakage.

3. Routine monitoring of plant components.

Undesirable leakage identified during tests or inspections will be reduced to a level as low as is practical.

2.1.6-2

8 2.'l.6.b Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces/Systems Which May Be Used in Operations

2.1.6 POST ACCIDENT CONTROL OF RADIATION IN SYSTEMS OUTSIDE CONT INMENT OF PWRs ND BWRs 2.1.6.b DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR SPACES/SYSTEMS WHICH MAY BE USED IN OPERATIONS NUREG-0578 Position With the assumption of a post-accident release of radio-activity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 501 of the core radioiodine and 1005 of the core noble gas inventory are contained in the primary coolant), each licensee shall perform a radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.

2.1.6-3

0 i

'PVNGS'Evaluation As part of the PVNGS design process, shielding in areas requiring personnel access for accident mitigation is designed such that access is not unduly limited.

In addition, safety related equipment is qualified for the maximum expected equipment life dose.

An analysis of the PVNGS shielding design criteria, now based on ALARA, will be performed to determine if the results of THI invalidate a.y of'he PVNGS shielding criteria. If, as a result of this analysis, the shielding design criteria in a plant area requires modification, the shielding in that area will be redesigned based on revised criteria. Any increase in permanent or temporary shielding, redesign or procedural control determined to be necessary will be completed prior to PVNGS Unit 1 Operating License. Refer to Section 2.1.8.a for a discussion of sampling system shielding.

2.1.6-4

0 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWRs

C, I I

2.1.7 IMPROVED AUXILIARY FEEDWATER SYSTEM RELIABILITY FOR PWRs 2.1.7.a AUTOMATIC INITIATION OF THE AUXILIARY FEEDWATER SYSTEM FOR PWRs NUREG-0578 Position Consistent with satisfying the requ.irements of General Design Criterion 20 of Appendix A to 10 CFR Part 50 with respect to the timely initiation of the auxiliary feedwater system, the following requirements shall be implemented'in the short term:

1. The design shall provide for the automatic initiation of the auxiliary feedwater system.

2. The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function.

3. Testability of the initiating signals and circuits shall be a feature of the design.

4. The initiating signals and circuits shall be powered from the emergency buses.

5. Manual capability to initiate the axuiliary feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not. result in the loss of system function.

2.1.7-1

0

6. The AC motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses.

7. The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFMS from the control room.

8. In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety-grade requirements.

PVNGS Evaluation

1. The Auxiliary Feedwater System shown in Figure 2.1.7-1 consists of two Seismic Category I pumps and their associated valves and one non-Seismic Category I pump and its associated valves.

The Seismic Category I portion of the auxiliary feedwater system is provided to automatically initiate residual heat removal capability during emergency conditions, such as a steam line rupture, loss of normal feedwater, or loss of offsite and normal onsite power .

2.1.7-2

The Seismic Category I portion of the auxiliary,feedwater system is automatically actuated by an Auxiliary Feed-water Actuation Signal (AFAS) from the Engineered Safety Features Actuation System (ESFAS). The AFAS is initiated for each steam generator by a low steam generator level coin-cident with a "not ruptured" calculated signal for that steam generator'refer to Figure,2.1.7-2).

The AFAS logic determines whether a steam generator is not intact in the event of a secondary system break by sensing:

- The steam generator has initiated a low water level trip.

- The steam generator pressure is less than the other by a predetermined value.

- The other steam generator has been calculated as not being ruptured.

The non-Seismic Category I portion of the auxiliary feedwater system is provided for normal non-emergency operation during startup, cooldown, and hot standby.

2. The Seismic Category I portion of the auxiliary feedwater system is composed of components in two separate and diverse load groups (i.e., load group 1 and load group 2. Each of '.,:

the four auxiliary feedwater valves associated with each steam generator is automatically actuated in such a manner that no single failure can prevent either the supply of auxiliary feedwater to an intact steam generator or the 2.1. 7-'3

isolation of auxiliary feedwater from a ruptured steam generator. 'L'oad group 2 powers the Seismic Category I motor-driven auxiliary feedwater pump and its associated valves and controls. Load group 1 supplies dc power to the steam-driven turbine controls V

and the valves associated with the turbine-driven auxiliary feedwater pump. No ac power is required for support of the turbine-driven auxiliary feedwater I train. The instrumentation and controls of the components and equipment in load group 1 are physically and electrically separate and independent of the instru-mentation and controls of the components and equipment in load group 2. This separation is maintained such that both trains are not terminated on common logic circuits.

3. Provisions are made to permit periodic testing of the auxiliary feedwater initiation signals and circuitry.

These tests cover the trip actions from sensor input through the protection system and actuation devices.

The system test does not interfere with the system protective functions The testing system meets the criteria of IEEE-338-1971, "IEEE Trial-Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protective Systems," and Regulatory Guide 1.22, 2.1.7-4

"Periodic Testing of Protection System Actuator

~

Functions. " Testing i s performed in accordance with the surveillance test requirements of the PVNGS Technical Specifications.

4. AFAS circuits are a part of the Engineered Safety Features Systems. The initiating signals and circuits are powered from Class IE buses in separate load groups as discussed in Item 2. The initiating sensors are powered from separate and redundant Class IE nuclear instrumentation and control panels, each of which is supplied by either offsite power or the diesel generators when offsite power is not available and is backed up by Class IE batteries.

5. Manual initiation capability for each auxiliary feedwater train exists in the control room. Control of individual auxiliary feedwater system components is also available in the control room. No single failure in the manual initiation portion of the circuit can result in the loss of auxiliary feedwater system function.

6. The Seismic Category I AC motor-driven pump and power operated valves in its train are automatically and sequentially loaded on the diesel generator bus upon loss of offsite power.

2.1.7-5

0

7. Failure of the AFAS will not result in the loss of manual capability to control auxiliary feedwater system components from the control room.

8. AFAS circuits are Class IE as presently designed.

The PVNGS auxiliary feedwater system, as presently designed, complies with the recommendations of NUREG-0578, July 1979.

2.1.7-5

Ql 2.1.7.b Auxiliary Feedwater Flow Indication to Steam Generator for PWRs

Ii IMPROVED AUXILIARY FEEDWATER SYSTEM RELIABILITY FOR PWRs AUXILIARY FEED'WATER FLOW INDICATION TO STEAM GENERATOR FOR PWRs NUREG-0578 Position Consistent with satisfying the requirements set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function, the following requirements shall be implemented:

1. Safety-grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.

2. The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.

PVNGS Evaluation

1. Monitoring of both trains of auxiliary feedwater flow is pro-vided at PVNGS. These flow indicator channels are made from highly reliable non-Class IE corn'ponents and are displayed on the 2.1.7-2

~

J

main control boards. Class IE (safety grade) pressure indicators located upstream of the manual block valves are also provided. Pressure indicator PI-18A and flow indicator FI-40 monitor flow from the train A turbine-driven auxiliary feedwater pump (refer to Figure 2.1.7-1). Pressure indicator PI-17A and flow indicator FI-41 monitor the flow from the train B motor-driven auxiliary feedwater pump.

2. The safety grade pressure indication channels are powered from their associated trains redundant Class IE buses.. The instrument bus powering the auxiliary feedwater flow instrumentation is normally powered from a non-Class IE 480 Y motor control center (HCC).

Upon loss of this HCC the instrument bus is auto-matically transferred to a Class IE HCC via a Class IE isolation transformer. This Class IE HCC is powered from offsite power or from its diesel generator when offsite power is not available.

2.1.7-8

<<010

~oa faom >svgaallsf SC CCDO Se s 02$ NCOI ie,

~

5CP 00>

aool lit<<psaos CA ~>

I ~ 5> 5C vOC1 vo'll Pg ~

SAON lo>O<<la SCP 001 Sc-<<Rat ecol 10 s 025 s>col ~

I <<siss I

~le 1/I N 01 ~ OCOI 5 SCP If 002 ISI To Ffgo<<ATCA Nflola 'SICII~

ctsgaaTOA ACC Coll aool glom Pv<<FS IF 1> $ c ~ 'voce ro)t ~ SL g I

1 I

CCOA Ncoa << '

o <<

o o I

~ CO IAO<< CT ~ S OI1 NIOA~ >0 Oll NCOa lo I s 01$ DCOA 5 le 1 1 TCNAICa CIF-OO S SCP 001 NLNZA Sita<<

Co<<of<<SAIC TISC F012 VOI ~ IC Isl Cgslaavoa ACf.COIO Io 21 roe I af<<FOs o C So o o

&Vlf&2&St I~ Ila LJ Sgtft &

10 Co>DC<<SAIC Lo <<.001,DCoa.t D 0 IC 510 IA<<g CIP OOI I

~ 5 11 voog a CA s >AC+0>OOO 9.~5 IP AS I Oi ~ -><<co I SIC la flem sais Cf<<falgoa sat SCP 001 IO ~ 21 5c a oe2 Doge 4 Lo 1002 0

@-.9 ~. ~- I VDIF I 0$ 1 Deco ~ '~

<<$ $ 5 ~ I>

~ ~

0 C I rag&2&$2

~ 1O o7

~

Sacs AVIILIIAT L 0<<~aOACO o L 'vf >loom Of C>I KL o asp

&2JCSJLL p lc.a>

' ~> QSo 4\ 1SL g&LC

'Ttf~TT//4 I IC Qfo 12 <<A M

I5 <<%22 Ceoe ~ o TS o m 0 PV>O o g << vl ma &fJL242.

0 o C NV 54 IC III

~ L afa >>02 OIC VS ml ISIS ftf Doll C

~ L 0 CgsmfA Afa COI OIC '-o ~

~

VS V0$ 4 Sla ASN 101

<<011 OOLO I OCC<<lfi 5 0 ~ Iigfa 14 COOL INC Svesrtlgm 40 -co *Y.

llgf tlf Doll r2.VOSC VS <<I V4 NA vc <<A ve reac aw-eeoc.l'-oso.eeoc-a-~

<<o SIIC ggf-001>

T i Lg vcOC LJ

. Y Ortaa1 of Sla-*fs.fet ~

O A 0<<1 o

s 014 <<000 I on" ISCC 11>> 00>S osnV Vt VIOS Vl Sl V2 V>ll v1 <<A 4<<>a<<Nfl I 1

I I

~$ I I 2

CA CA TCC CA Pvg MIPP I,

r---@

~

CIS C

0'1 o Y 2 C

dR

':~

CN N SOS.NCOA I <<5 I I N 014 >ooe I I afag IF Ag.

fsagtvp Aom AglCTOA I OIC NCCI 4

~

o o +no I At

//

ILS>le I

<<IKA c<<F 001 IOO> rose <<nv v00$ 0 0 nv

~ ~

t 120 I 6'=- K'PM'./

~

I ato OIC I<<014 21 I>

I I >2$ $ $ 1$ $ML I

afac y 0 0 lo ftfo<<alga Nf logs s 11 12 CKI<< cfagallol ~

<<CD I <<Ccl lggC Lo I 001 I<<fig II I 001 ac4 toga saos ce>CN<<sist 510<<lct Ta<<g CIP lo OOS 11 CI I 022.<<CCA 0 rooC T001 I 045 NCCI ~ <<

Ifa FOI Vei 5 VOI 4 oo OCCI C

~ I OCCA 4

os~cole OCCA OCOI I 004 Dgea 4 C 00 ~ .OCOI 4 ~ I vers 55>> 002 I'

I ~ le>

I> o o 5 o

Oj as 4 1 OI ~ 510 3-:- =. 0 LJ 0

';5 e ca ppc~&f~

CA I <<0 o

I DCS~IOC 0 I V

'~

M I

0 I I 011 OCCI 1 A VOCI I alaS t/

6 VOI 1 I ~ss 0 I~ A C AS / I I

I I NV 5>A 11 C 6.'-"-

110 prpaaM./ s 11 11 11 C

o I<<OTC 11 I OIS OCCI 5 I so 'I 1 ~ I a-esa-Ogea-e 10 Co<<of <<SAIC I 00$

510Aact Ta<<c IIII<<l SLOV A<<0 tslaCliCI

~ 00>

IC 11 OCCa C'I vol ~

o <<<<o std CA <<0 K 5 ~

8-SIC CA

-"-~I gvp am [of 1 I 1I I CA PrtaaJÃ I I

~I I Flirt Nag 12$ voc NDTDA IcTVATDA I 0 I +fo LI<<t IS 4IC lo FIAST fltle 1'A SAOI AtaCIOA ~ ~ Ofo SAITVP CN ~ 41 ~ <<Cel ~ ~ 020 NCCI 0 ~5 Nfle OOTSIDf mals Slfla

'~

aifa ra<<c CNP 001

-9 s SVFFCSI SIAVCTVAC~

Ia 5> 5 LMS Q I<<elf tl o 40 VDOS g

~SI ~ / ~%$ 0

/

<<Col NCCI 0l C 0r O I

&-- grfaaM./

'S I

ie 5- pp.apM./

-'-'-' I >>

<<p fgCD<<llta C'QK ISIS 7 'I I 0 I 1 $0 10 o NV 5>IL 2 Vr 5>A Nf ADLA

$ 1&A IS 0 m 11 SI 1$ slfla m ctsfaITDA Io ~ OOC DCCI ~ .OOL.OCCA 4 0 004 Ogel 4' ~ .COI ~

fAom CONK<<gaTc I~ CT 0 02> NCC ~ 0 Lo ~ 021 NCCI 0 >'fe 5C>'01 1st Svsltm DC$ ICsATDA IS 1510 PIITCfot STDAICC Ia<<g C IO-11 00 I 1021 rett fos v02$

4'024 o <<<<o OIC OIO f OOC OCOA ~ ~MN<< . l vteo I~ I 0 I Iii LI<<c VAlvc Iso I<<ssavmf<<I Nvmegas 5<<0<

~

I<<IS Daa<<lsc sa<<TSS 01<<LA<<lsf DCCA OCOA laoICIKO ~

o o 1

'C ggggg ff&OS&I fJLtsggt,, of K

<<50

""u LO o fo lt 0.0>S.OCC ~

0 So QFI 6 K5--' ig I

ca 0

Cs 0 l<<A e-- CA p faalpp I Q CA 0

A5

'~

vO25 afat g/

<<5 afat t 13-M-AFP-001 R EV ft soe CO /

QFO I. ~5 J~FO '7 6--

$ 00 gvtapfpf /

/

~

10 tt 10 t S'p-' LS./

AS 0

'Kvpp!

I I

Vv

'll 5>A Palo Verde Nuclear Generating Station 7 0 011 OCCI OIO OCOI I '&,:

10 Co<<OC<<gilt 510AACC Tlsc 0 Oil OCCa 0 IC $ 10 0 IC OI ~ OI ~ 510 (SI<<g FLDN Oos 4'IF If tl ~ Otf OC C~aOC 0 ~ NC 0a OCC4 OCta oceaaoacoa <<coa fsfAC154> $ 0 jo T01 ~ rOIS v011 vota AUXILIARYFEEDWATER SYSTEM Figure 2 . 1 . 7-1

f 1

'll

,,i I

I

PRESS LOW SG 2 PRESS LOW SG-I LEVEL SG.2 )SG-1 LEVEL SG 1 )SG 2 A 8 C 0 A 8 C 0 A 8 C 0 A 8 C D A A 8 8 C C D 0 ANO AND AND AND AND AND AND AND A A 8 8 C C 0 D AND AND AND AND AND AND AND AND 2/4 LOGIC 2/4 LOGIC EFAS-1 EFAS I EFAS-2 EFAS-2 REMOTE REMOTE R EiVIOTE REMOTE MANUAL ,MANUA MANUA MANUA OR OR OR OR OR OR OR V V 1A 18 2A 28 3A 38 4A 48 1A 18 2A 28 3A 38 4A 48 PPS AUX 1A A 3A 4A IB 28 38 48 RFLAY IA 2A 3A 4A 18 28 38 48 CABINET SELECTIVE SE LECTIVE SE LECTI VE SELECTIVE 2OUTOF 4 2OUTOF 4 2 OUT OF 4 2 OUT OF 4 LOGIC LOGIC LOGIC LOGIC MANUAL MANUAL OR OR OR OR VALVES VALVES VALVES VALVES TRAIN A TRAIN 8 TRAIN A TRAIN 8 FFAS I EF AS.2 EFAS ~ EMERGENCY FEEDWATER ACTUATION SIGI'IAL g~ NOT C-E j Figure SK5H ESFAS SIGNAL LOGIC.(EFAS 1, EFAS 2) 2.1.7-2

2.1.8.a Improved Post-Accident Sampling Capability

t lJ 2.1.8 INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2.1.8. a IMPROVED POST-ACCIDENT SAMPLING CAPABILITY NUREG-0578 Position A design and operational review of the reactor coolant and containment atmosphere sampling systems shall be performed to determine the capability of personnel to promptly obtain (less than 1 hour) a sample under accident conditions without incurring a radiation exposure to any, individual in excess of 3 and 18 3/4 Rems to the whole body or extremities, respectively. Accident'onditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.

A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly quantify (less that 2 hours) certain radioisotopes that, are indicators of the degree of core damage.

Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums ( which indicate high fuel temperatures), and non-volatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release.

2.1.8.-1

The review should also consider the effects of direct radi-ation from piping and components in the auxiliary building and possible contamination and direct radiation from air-borne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.

In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.

Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly; i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift.

PVNGS Evaluation An analysis of the design and operation of the reactor coolant and containment atmosphere sampling systems will be performed to ensure that samples can be obtained promptly (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) during accident conditions without exposure to any individual to greater than 3 and 18-3/4 Rems to the whole body or extremities, respectively. Any design or procedural changes will be imple-mented prior to PVNGS Unit 1 Operating License.

2.1.8-2

An analysis of PVNGS spectrum analysis facilities will be performed to ensure that the design allows prompt quantification (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) of previously prepared radioisotope samples that may be indicative of core damage. The assumed reactor coolant spectrum will correspond to a Regulatory Guide 1.3 or 1.4 release.

PVNGS chemical analysis facilities will be designed to allow prompt analysis of boron concentration (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and chloride concentration (less than 8 hours) assuming a radioactive sample .corresponding to a Regulatory Guide 1.3 or 1.4 release.

2.1.8-3

Qi 2.'I.s.b Increased Range of Radiation Monitors

P

~V 0

2.1.8 INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2.1.8.b INCREASED RANGE OF RADIATION MONITORS NUREG-0578 Position The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a revision to Regulatory Guide 1.97, "Instrumentation to Follow the Course of an Accident," which has already been initiated, and in other Regulatory Guides, which will be promulgated in the near-term.

1. Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions; multiple monitors are considered to be necessary to cover the ranges of interest.

a. Noble gas effluent monitors with an upper range capacity of 10 yCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.

b. Noble gas, effluent monitoring shall be provided for the total range of concentration extending from normal condition (ALARA) concentrations to a maximum of 10 pCi/cc (Xe-133). Multiple monitors are 5'onsidered to be necessary to cover the ranges of interest. The range capacity of individual monitors should overlap by a factor of ten.

~ ~

2.1.8-4

I 0

2. Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other'edia, i

followed by onsite laboratory'analysis.

3. In-containment radiation level monitors with a maximum range of 10 rad/hr shall be installed. A minimum of two such monitors that are physically separated shall be provided. Monitors shall be designed'and qualified r 'I*

~

to function in an accident environment.

PVNGS Evaluation The PVNGS radiation monitoring system is a follows;

1. Noble gas effluent monitoring of the plant vent is presently designed to have an upper limit of 10 yCi/cc (Xe-133).

Additional monitors will be added with an upper range of 10" pCi/cc (Xe-133). The range of the plant vent monitors shall overlap by at least a factor of ten.

I

2. Post-accident effluent monitoring of radioiodines is provided for by sampling of plant vent iodine monitor adsorption media followed by onsite laboratory analysis.

Procedures for this sampling. will be developed prior to Unit, 1 Operating License.,

2.1.8-'5

I

3. The PVNGS power access purge monitors (two redundant, ESF monitors) are located around the power access purge duct adjacent to the containment. These monitors are capable of monitoring the purge duct during normal operation and the containment (through the containment wall) subsequent to an accident. When monitoring the containment radiation level these monitors will have 6

a maximum range of at least 10 rad/hr. The monitors will be qualified to post-accident condition.

The design of the monitors will be evaluated further to determine if a Iaximum range of 10 8

rad/hr is achievable. This evaluation will be completed by March, 1'981.

'.1.8-6.

0 2.1.8.c Improved In Plant Iodine Instrumentation

0 t

INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT IMPROVED IN-PLANT IODINE INSTRUMENTATION NUREG-0578 Position Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where personnel may be present during an accident.

PVNGS Evaluation Prior to PVNGS Unit 1 Operating License, procedures will be implemented for determination of airborne iodine concen-tration by sampling iodine monitor cartridges for spectrum analysis. These procedures will be used when the plant radiation monitoring system indicates iodine levels requiring the use of respiratory equipment.

Spectral analysis equipment is available at each PVNGS unit.

2.1.8-7

0 ANALYSIS OF DESIGN AND OFF-NORMAL TRANSIENTS 'AND ACCIDENTS NUREG-0578 Position Analyses, procedures, and training addressing the following are required:

1. Small break loss-of-coolant accidents;

2. Inadequate core cooling; and

3. Transients and accidents.

Some analysis 'requirements for small breaks have already been specified by the Bulletins and Orders Task Force. These should be completed. In addition, pretest calculations of some of the Loss of Fluid Test (LOFT) small break tests (scheduled to start in September 1979) shall be performed as means to verify the analyses performed in support of the small break emergency procedures and in support of an eventual long term verification of compliance with Appendix K of 10 CFR Part 50.

In the analysis of inadequate core cooling, the following conditions shall be analyzed using realistic (best-estimate) methods:

1.. Low reactor coolant system inventory (two examples will be required - LOCA with forced flow, LOCA without forced flow).

2. Loss of natural circulation.(due to loss of heat sink).

These calculations shall include the period of time during which inadequate core cooling is approached as well, as the 2.1.9-1

period of time during which inadequate core cooling exists.

The calculations shall be carried,out in real time far enough that all important phenomena and instrument indications are included. Each case should then be repeated taking credit for correct operator action. These additional cases h

will provide the basis for developing appropriate emergency procedures.'hese calculations should also provide the analytical basis for the design of any additional instru-mentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy.

The analyses of transients and accidents shall include the design basis events specified in Section 15 of each FSAR.

The analyses shall include a single active failure for each system called upon to function for a particular event.

Consequential failures shall also be considered. Failures of the operators to perform required control manipulations shall be given consideration for permutations of the analyses.

Operator actions that could cause 'the complete loss of function of a safety system shall also be considered. At present, these analyses need not address passive failures or multiple system failures in the short term. In the recent analysis of small break LOCAs, complete loss of auxiliary feedwater was considered. The complete loss of auxiliary feedwater may be added to the failures being considered in the analysis of transients and accidents if it is concluded 2.1.9.-2

fl that more is needed in operator training beyond the short-term acti ons to upgrade aux i 1 i ary 'feedwater sys tern re 1 i a-bility. Similarly, in the long term, multiple failures and passive failures may be considered depending in part on staff review of the results of the short-term analyses.

The transient and accident analyses shall include event tree analyses, which are supplemented by computer calculations for those cases in which the system response to operator actions is unclear of these calculations could be used to provide important quantitative information not available from an event tree. For example, failure to initiate high-pressure injection could lead to core uncovery for some transients, and a computer calculation could provide information on the amount of time available for corrective action. Reactor simulators may provide some information in defining the event trees and would be useful in studying the information available to the operators. The transient and accident analyses are to be performed for the purpose of identifying appropriate and inappropriate operator actions relating to important safety .considerations such as natural circulation, prevention of core uncovery, and prevention of more serious accidents.

2.1.9-3

The information derived from the preceding analyses'shall be included in the plant emergency procedures and operator training.. It is expected that analyses performed by the NSSS vendors will be put in the form of emergency procedure guidelines and that the changes in the procedures will be implemented by each licensee of applicant.

In addition to the analyses performed by the reactor vendors, analyses of selected transients should be performed by the NRC Office of Research, using the best available computer codes, to provide the basis for comparisons with the analytical methods being used by the reactor vendors.

These comparisons together with comparisons to data, including LOFT small break test data, will constitute the short-term verification effort to assure the adequacy of the analytical methods being used to generate emergency procedures.

PVNGS Evaluation Arizona Public Service Company will work with Combustion Engineering to determine the generic analyses necessary for the System 80 design. The analyses will be performed on a schedule consistent with development of procedures, training of operator's installation of:hardware (refer to Section 2.1.3) prior to PVNGS Unit 1 Operating License.

2.1.9-4

2.1.10 INSTRUMENTATION TO MONITOR CONTAINMENT CONDITIONS DURING TH COURSE OF AN ACCIDENT, NUREG-0578 Position Consistent with satisfying the requirements set forth in General Design Criterion 13 to provide the capability in the control room to ascertain containment conditions during the course of an accident, the following requirements shall be implemented:

l. A continuous indication of containment pressure shall be provided in the control room. Measurement and indication capability shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and minus five psig for all containments.

2. A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room. Measurement capability shall be provided over the range of 0 to lOX hydrogen concentration under both positive and negative ambient pressure.

3. A continuous indication of containment water level shall be provided in the control room f'r all plants.

A narrow range instrument shall be provided for PWRs and cover the range from the bottom to the top of the normal containment sump. Also for PWRs, a wide range instrument shall be provided and cover the range 2.1.10-1

0 It

from the bottom of the containment to the elevation equivalent to a 500,000 gallon capacity. For BWRs, a wide range instrument shall be provided and cover the range from the bottom to 5 feet above the normal water level of the suppression pool.

The containment pressure, hydrogen concentration and wide range containment water level measurements shall meet the design and qualification provisions of Regulatory Guide 1.97, including qualification, redundancy, and testibility. The narrow range containment water level measurement instrumentations shall be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being periodically tested.

PVNGS Evaluation PVNGS containment monitoring instrumentation is as follows:

1. Eight, Class IE containment pressure indicators are provided in the control room. Four indicators have=

a -4 to 20 psig range. The other four indicators have a -4 to 85 psig range.

An analysis performed for the containment maximum nega-tive pressure (refer to FSAR Section 6.2.1.1.3.6) demon-strates a maximum credible negative pressure of -2.6 psig.

The existing pressure indicator range (to -4 psig) is, therefore, considered adequate to measure containment negative pressure under any postulated conditions.

2.1.10-2

The containment is designed for 60 psig internal pressure and tested to 69 psig. The maximum calcu-lated containment pressure under accident conditions is 49.2 psig. The existing pressure indicator range (to 85 psig) provides a 40Ã margin above the contain-ment design pressure and a 70/ margin above the con-servatively calculated peak pressure. Therefore, a containment pressure range greater than the existing 85 psig maximum is not considered to be necessary.

The addition of monitors with a range to 180 psig would add unnecessary confusion to the available containment pressure instrumentation by:

a. Displaying at only 25/ of'pan during the worst postulated accident.

b. Displaying containment pressure with an accuracy of only +3.6 psi (+2/).

2. A continuous indication of hydrogen concentration is available in the control room. Two Class IE redundant monitors are presently provided with a range of 0-10Ã.

The hydrogen monitors are manually initiated following a LOCA, Once initiated, they proVide a continuous measure-ment of hydrogen concentration (refer to Figure 2.1.5-1).

2.1.10-3

3. Continuous indication of containment water Ieyel will be provided for PVNGS, An evaluation will be performed to determine the design requirements for this instrumentation. The results of this evaluation will be presented to the NRC staff by Harch,'981.

2.1.10-4

Q, 2.1.11 INSTALLATION'OF 'REMOTELY 'OPERATED HIGH POINT 'ENTS THE REACTOR COOLANT'SYSTEM NUREG-0578 Position Each applicant and licensee shall install reactor coolant system and reactor vessel head high point vents remotely operated from the control room. Since these vents form a part of the. reactor coolant pressure boundary the design of the vents shall conform to the requirements of Appendix A to 10 CFR Part 50 General Design Criteria. In particular, these vents shall be safety grade, and shall satisfy the single failure criterion and the requirements of IEEE-279 in order to ensure a low probability of inadvertent actuation.

Each applicant and licensee shall provide the following information concerning the design and operation of these high point vents:

l. A description of the construction, location, size, and power supply'for the vents, along with results of analyses of loss-of-coolant accidents initiated by a break in the vent pipe. The results of the analyses should be demonstrated to be acceptable in accordance with the acceptance criteria of 10 CFR 50.46.

0

2. Analyses demonstrating that the direct venting of non-condensable gases with perhaps high hydrogen concentrations does not result in violation of com-bustible gas concentration limits in containment E

as described in 10 CFR Part 50.44, Regulatory Guide 1.7 (Rev. 1), and Standard Review Plan Section 6.2.5.

3. Procedural guidelines for the operators'se of the vents. The information available to the operator for initiating or terminating vent usage shall be discussed.

PYNGS Evaluation PVNGS will utilize existing penetrations on the pressurizer and reactor vessel head to provide high point venting of the Reactor Coolant System (RCS). Refer to Figure 2.1.11-. 1 for a schematic of the RCS. The pressurizer vent (refer to Figure 2.1.11-2) penetrates the top of the pressurizer and is currently utilized for steam space sampling. The reactor vessel head vent (refer to Figure 2.1.11-2) penetrates the top of the reactor vessel head and is currently piped to the Reactor Drain Tank (RDT), v'iithin containment, via a manual valve. Piping and valves will be provided to allow remote operation of the pressurizer and reactor vessel vents from the control room. In order to ensure a low probability of inadvertent action, valves and controls will sati'sfy the single failure criterion and IEEE-279-1971.

2.1.11-2

0 The following, information will be supplied as the RCS Vent System design is finalized (prior to Unit 1 Operating License):

l. A description of the vents, piping, valves and utilized for the RCS Vent System. 4'ontrols An analysis demonstrating the acceptability of a vent piping break in accordance with 10 CFR 50.46.

I

2. An analysis demonstrating that the venting of RCS

'non-condensable gases via the RCS Venting System does not result in containment hydrogen concentrations

'greater than those allowed by 10 CFR 50.44, Regulatory Guide 1.7, Revision 0 and Standard Review Plan 6.2.5.

3. Operating procedures for use of the RCS Vent System.

2.1.11-3

PRESSURIZER STEAM .

STEAM 'GENERATOR GENERATOR No. 2 No. 1 PUMP No. 1B'UMP No. 2B PUMP No. 2A REACTOR VESSEL Figure ISOMETRIC VIP)'/ OF THE, REACTOR COOLANT S'(STEM

0

2.2.1.a Shift Supervisors Responsibilities

0 2.2.1 'HPROYED REACTOR "OPERATIONS 'OMMAND FUNCTION 2.2.1. a SHIFT SUPERYISORS RESPONSIBILITIES NUREG-0578 Position

l. The highest level of corporate management of each licensee shall issue and periodically reissue a management directive that emphasizes the primary management responsibility of the shift supervisor for safe operation, of the plant under all conditions on his shift and that clearly establishes his command duties.

2. Plant procedures shall be reviewed to assure that the duties, responsibilities, and authority of the shift supervisor and control room operators are properly defined to effect the establishment of a definite line of command and clear delineation of the command decision authority of the shift supervisor i n the control room relative to other plant management personnel. Particular emphasis shall be placed on the following:

a. The responsibility and authority of the shift supervisor shall be to maintain the broadest Perspective of operational conditions affecting 2.2.1-1'

0 the safety of the plant as a matter of highest priority at all times when on duty in the control room. The idea shall be reinforced that the shift supervisor should not become totally involved in any single operation in times of emergency when multiple operations are required in the control room.

b. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators. Persons authorized to relieve the shift supervisor shall be specified.

c. If the shift supervisor is temporarily absent from the control room during routine operations, a lead control room operator shall be designated to assume the control room command function.

These temporary duties, responsibilities, and authority shall be clearly specified.

3. Training programs for shift supervisors shall emphasize and reinforce the responsibility for safe operation and the management function the shift supervisor is to pro-vide for assuring safety.

2.2.1-2

8 PVNGS Evaluation The responsibilities and authority of the Shift Supervisor is delineated in FSAR Chapter 13.1. In addition, the Station Manual and the Operating Department Program also prescribe responsibi'lities and authority of the Shift Supervisor.

The shift organization at PVNGS is such that the Shift I

Foreman will perform many of the administrative functions that are normally completed during a shift.

2.2.1-3

2.2.1.b Shiest Technical Advisor

IMPROVED REACTOR OPERATIONS COMMAND FUNCTION SHIFT TECHNICAL ADVISOR NUREG-0578 Position Each licensee shall provide an on-shift technical advisor to the shift supervisor. The shift technical advisor may serve more than one unit at a multi-unit site if,qualified to perform the advisor function for the various units.

The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received spycific training in the response and analysis of the plant for transients and 'accidents. The shift tech-nical advisor shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign normal duties to the shift technical advisors that pertain to the

/

engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience.

PVNGS Evaluation There is sufficient time prior to PVNGS Unit 1 Operating License to assess fully the alternatives for providing 2.2.1-4

i l

0'

additional technical and analytical capability for diagnosis of off-normal events. Two alternatives considered by the Lessons Learned Task Force have merit for PVNGS:

l. Upgrade the requirements for operators to include more engineering and plant response training.

2. Improve plant response diagnosis capabilities by backfit of computer and plant status display inno-vations.

The PVNGS Operator Training Program will be revised and upgraded to include more engineering and plant response training. The PVNGS Training Simulator will be used exten-sively to accomplish this objective.

The PVNGS design already has incorporated some advanced informational displays. These include the Safety Equipment Status Panels, computer-generated graphics displays and the various information displays which can be generated by the Core Monitoring Computer and the Plant Monitoring Computer.

2.2.1-5

0 2.2.1.c Shift Relief and Turnover Procedure

0, 0 \g

2.2.1 IMPROVED REACTOR*OPERATIONS'COMMAND'FUNCTION 2.2.1. c SHIFT RELIEF AND TURNOVER PROCEDURE NUREG-0578 Position The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:

l. A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete and sign. The following items, as a minimum, shall be included in the check-list:

a ~ Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist).

b. Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the control console (what to check and criteria for acceptable status shall be included on the checklist);

2.2.1-6

0 r,lr

'l

c. Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications.

For such systems and components, the length of time in the degraded mode shall be compared with the Technical Specifications action statement (this shall be recorded as a

'separate entry on the checklist).

2. Checklists or logs shall be provided for completion by the offgoing and oncoming auxiliary operators and technicians. Such checklists or logs shall include any equipment under maintenance of test that by them-selves could degrade a system critical to the pre-vention and mitigation of operational transients and accidents or initiate an operational transient (what to check and criteria for acceptable status shall be included on the checklist); and

3. A system shall be established to evaluate the effect-iveness, of. the shift and relief turnover procedure (for example, periodic independent verification of system alignments).

PVNGS Evaluation PVNGS will have a procedure for shift relief which will include the use of checklists and the review of operating 2.2.1-7

I 0

logs. The relieving operator or supervisor will be required to sign that he has read the appropriate logs and completed the watch turnover as specified in the procedure.

Provisions will be made for periodic verification of the relief and turnover procedure.

The shift relief and turnover procedure and the provisions for its periodic verification will be prepared and imple-mented prior to PVNGS Unit 1 Operating License.

2.2.1-8

2.2.2.a Control Room Access

2.2.2 'l]PROVED IN-'PL'ANT 'MERGENCY 'PROCEDURES AND 'PREPARATIONS 2.2.2.a CONTROL ROON ACCESS NUREG-0578 Position The licensee shall make provisions for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations supervisor, shift supervisor, and control room operators),

to technical advisors who may be requested or required to support the operation, and to predesignated NRC personnel.

Provisions shall include the following:

l. Develop and implement an administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access.

2. Develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event of an emergency. The line of succession for the person in charge of the control room shall be established and limited to persons possessing a current senior reactor operator's license.

The plan shall clearly define the lines of coranunication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room.

2.2.2-1

0 0

PVNGS Evaluation Prior-to PVNGS Unit 1 Operating License, admjnjstratjve procedures will be developed to limit personnel access to the unit control room during operational transient and accident conditions. These procedures will delineate:

1. That the shift supervisor has the authority to restrict control room access to personnel responsible for plant operations and technical advisory support to operations.

2. Clear lines of communications, authority, and responsibilities in the event of an emergency. Only senior licensed operators will direct the licensed operators.

2.2,2-2

8 2.2.2.b Onsite Technical Support Center

~

~

t

'I 0

2.2.2 IMPROVED 'N-PLANT'MERGENCY 'PROCEDURES AND PREPARATIONS 2.2.2.b ONSITE TECHNICAL SUPPORT CENTER NUREG-0578 Position Each operating nuclear power plant shall maintain an onsite technical support center separate from and in close proximity to the control room that has the capability to display and transmit plant status to those individuals who are knowledge-able of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be habitable to the same degree as the control room for postulated accident conditions. The licensee shall revise his emergency plans as necessary to incorporate the role and location of the technical support center.

Records that pertain to the as-built conditions and layout of structures, systems and components shall be stored and filed at the site and accessible to the technical support center under emergency conditions. Examples of such records include a system .description, general arrangement drawings, piping and instrumentation diagrams, piping system isometrics, electrical schematics, wire and cable lists, and single line electrical diagrams. It is not intended that all records described in ANSI 45.2.9-1974 be stored and filed at the site 2.2.2-3

0 0

and accessible 'to the'technical support center under emergency conditions; however, as stated in that standard, storage systems shall provide for accurate retrieval of all information without undue delay.

PVNGS Evaluation An Onsite Technical Support Center for Palo Verde Nuclear Station will be established prior to PVNGS Unit 'enerating 1

Operating License. This center will be 'located within the site boundary and will provide a work area for supervisory and technical personnel from A'rizona Public Service Company and the NRC. It will have communications with the control room and the other response centers. An analysis will be performed to determine equipment to be incorporated into the center to monitor and display the status of the affected unit. The center will be habitable to permit occupancy following a loss of coolant accident. As-built drawings and other appropriate records shall be available in readily retrievable form at the site and shall be accessible to the Onsite Technical Support Center.

The Palo Verde Nuclear Generating Station Emergency Plan will be revised to describe the Onsite Technical Support Center and its function.

2.2.2-4

2.2.2.c Onsite Operational Support Center

0 2.2.2 IMPROVED IN-PLANT EMERGENCY PROCEDURES AND PREPARATIONS 2.2.2.c ONSITE OPERATIONAL SUPPORT CENTER NUREG-0578 Posi tion An area to be designated as the Onsite Operational Support Center shall be established. It shall be separate from the control room and'shall be the place to which the operations support personnel will report in an emergency sit'uation.

Communications with the control room shall be provided. The Emergency Plan shall be revised to reflect the existence of the center and to establish the methods and lines of communi-cation and management.

PVNGS Evaluation Prior to PVNGS Unit 1 Operating License, an area separate from the unit control rooms will be designated as the PVNGS Onsite Operational Support Center prior to receipt of operating license.

This will be an assembly area where plant operations support personnel will repor't in an emergency situation for further orders or assignment. The Operational Support Center will have communication with the control room and the Emergency Control Center (refer to the PVNGS Emergency Plan).

The Emergency Plan will reflect the existence of the Operational Support Center and will establish the methods and lines of communication and management.

2.2.2-5

'~AS << <<L REVISED LIMITING'CONDITIONS FOR OPERATION OF 'NUCLEAR 0 0 NUREG-0578 Position All NRC nuclear power plant licensees shall provide infor-mation to define a limiting operational condition based on a threshold of complete loss of safety function. Identifi-cation of a human or operational error that prevents or could prevent the accomplishment of a safety function required by NRC regulations and analyzed in the license application shall require placement of the plant in a hot shutdown condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in a cold shutdown condition within 24 hours.

The loss of operability of a safety function shall include consideration of the necessary instrumentation, controls, emergency electrical power sources, cooling or seal water, lubrication, operating procedures, maintenance procedures, test procedures and operator interface with the system, which must also be capable of performing their auxiliary or supporting functions. The limiting conditions for operation shall define the minimum safety functions for modes 1, 2, 3, 4, and 5 of operation.

The limiting condi tions of operation shall require the 4r 1 1 ~>, 4~r I VI IV 6 I I I'JJ

~ ~

2.2.3-1 .

1. If the plant is critical, .restore the safety function (if possible) and place the plant in a hot shutdown condi ti on wi thin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

2. Within 24 hours, bring the plant to cold shutdown.

3. Determine the cause of the loss of operability of the safety function. Organizational accountability for the loss of operability of the safety system shall be established.

4. Determine corrective actions and measures to prevent recurrence of the specific loss of operability for the particular safety function and generally for any safety function.

5. Report the event within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirm by telegraph, mailgram, or facsimile trans-mission to the Director of the Regional Office, or his designee.

6. Prepare and deliver a Special Report to the NRC's Director of Nuclear Reactor Regulation and to the Director of the appropriate regional office of the Office of Inspection and Enforcement. The report shall contain the results of Steps 3 and 4 above, 2.2.3-2

-jS along with a basis for allowing the plant to return to power operation. The senior corporate executive of the licensee responsible and accountable for safe plant operation shall deliver and discuss the contents of the report in a public meeting with the Office of Nuclear Reactor Regulation and the Office of Inspection and Enforcement at a location to be chosen by the Director of Nuclear Reactor Regulation.

7. A finding of adequacy of the licensee's Special Report by the Director"'of Nuclear Reactor Regulation will be requirec before the licensee returns the plant to power.

PVNGS'Evaluation In his August 20, 1979 memorandum to the NRC Coranissioners, Harold R. Denton, Director, Office of Nuclear Reactor Regulatory',

announced his intent to ini tiate a rulemaking proposal'n limiti,ng conditions of operations, which would include con-sideration'of th recommendations of the Lessops Learned Task Force and-alter r.. tive approaches. APS will follow this proposal with great 'int'er st and will participate in any rulemaking I

proceedings I that may be instituted.

2.2.3-3'

u' I

J

PVNGS, LLIR I.C OP TING PROCEDURES I.C.1 GUID CE: FOR THE, EVALUATION AND DEVELOPMENT OF PROCED S FOR TRANSIENTS AND ACCIDENTS Position In letters of- Septembe 13. and'27; October 10 and 30', and November 9, 1979,- the. Of 'ce of Nuclear Reactor'Regulation required licensees. of opera 'ng-plants, applicants 'for operat-ing,licenses and,licensees of, lants under construction to per-form, analyses of, transients and ccidents, prepare emergency.

procedure guidelines,,upgrade emer ncy 'procedur'es, including" procedures for, operating with .natural circulation conditions,"

and to"conduct operator retraining (see also item I:.A.2.1).

Emergency procedures-are required to be c nsistent withthe actions necessary to cope with the transien and accidents analyzed. Analyses of transients and accident ,were to be com-in early 1980 and implementation of proce ures and retrain-l'leted ing were to be completed 3 months after emergency ocedure guidelines were established; however, some difficult in com-pleting these requirements has been experienced. Clara ication of the scope of the task and appropriate schedule revisio s are K

being developed. In the course of review of these matters n IP and Wilcox (BSW)-designed plants, the staff will foll w i'abcock F

up on the bulletin and orders matters relating to analysis r

methods and results, as listed in NUREG-0660, Appendix C (see

PVNGS'LLIR Table C.l, Items 3, 4, 16, 18, 24, 25, 26, 27; Table'.2, Items 4, 12, 17, 18, 19, 20; and Table C."3, Items 6, 35, 37, 38, 39, 41, 47, 55, 57).

PVNGS Evaluation Arizona Public Service Company (APS) has participated in C-E Owners Group activities conducted since the Three Mile Island accident to develop improved emergency procedure guide-lines and associated supporting analyses. The C-E Owners Group has completed numerous documents which have been submitted to the NRC for review. The C-E Owners Group is currently sponsor-ing further activities which will be completed in the first half of 1981 and documented at that time. A summary of results obtained to date and current activities is provided below.

The initial C-E Owners Group analysis of Inadequate Core Cool'ing (ICC) is documented in Report CEN-117, "Inadequate Core Cooling A Response to NRC IE Bulletin 79-06C, Item 6 for Combustion Engineering Nuclear Steam Supply Systems." This report was sub-mitted to the NRC staff for review on October 31, 1979, by the C-E Owners Group.

"Operational Guidance for Inadequate Core Cooling" was prepared by the C-E Owners Group based on the analyses in Report CEN-117.

This operational guidance was distributed to members of the C-E Owners Group for their use in review and possible revision of plant emergency procedures in December, 1979. A copy of I.C.1-2

PVNGS LLIR emergency procedure guidelines are documented in Report.,CEN-128.

This report was submitted to the NRC staff. fqr review, on, April 1, 1980.

The emergency, procedure guidelines contained in Report CEN-128 were prepared based on extensive reviews of existing emergency procedures, past safety and design analyses, the plant simula-tion and sequence of events. analyses in CEN-128, and interviews with, operations personnel at plants with operating C-E reactors.

These emergency, procedure guidelines were prepared to be used I

as a basis for reviewing 'and revising, if necessary, 'existing plant emergency procedures. APS is .using these guidelines to develop.PVNGS emergency. procedures where appropriate.

The NRC,staff-,.in a letter dated July 17, 1980, sent questions-

to the C-E Owners Group concerning the emergency procedure guide-lines documented in Report CEN-128. A meeting was held with the NRC staff in Bethesda,. Maryland on September 11-, 1980, to discuss these questions and answers to"them. The C-E Owners Group is presently, preparing answers to these questions and to the emergency procedure guidelines in Report CEN-128 'evisions

,as, are appropriate. A preliminary response to these questions was submitted 'by the C-E- Owners Group to the NRC staff in 'a letter dated December 10, 1980. The remaining responses were submitted to the NRC staff by theC-E Owners Group'n a meeting

'.on January 30, 1981. ~

I.C.1-5

PVNGS ILIR Since early 1980, the C-E Owners Group has conducted an exten-sive evaluation of specific technical characteristics of emergency procedure guidelines. These include (1) the diagnos-tic guidance to be provided in emergency procedure guidelines, (2) the need for a separate guideline for inadequate core cool-ing., and (3) the format for presentation of emergency guidance.

This evaluation is currently scheduled to be completed in the first quarter of 1981. The results of this evaluation will.

serve as one basis for possible revision of emergency procedure guidelines contained in Report CEN-128.

The C-E Owners Group agreed on December 3, 1980, to conduct a series of workshops concerning emergency procedure guidelines in early 1981. The first such work shops were conducted in January, February, and March 1981. Another workshop is planned for April 1981. These workshops are intended to provide a formal process by which the emergency procedure guidelines documented in Report CEN-128 will be revised to account for multiple failure considerations. Input to these workshops will be provided by the analysis and emergency procedure guide-,

lines studies which have been conducted by the C-E 'Owners Group since early 1980. The workshops are to be attended by staff personnel from C-E and from utilities which own C-E reactors.

These workshops will also provide the opportunity to explore multiple-failure scenarios beyond those which have been currently identified in the C-E Owners Group analyses of transients and accidents. LOCA will also be considered in these workshops.

I.C.1-6

PVNGS LLIR Following the completion of the studies currently being con-ducted by the C-E Owners Group and the emergency procedure guide-lines workshops, a revised set of emergency procedure guidelines will be submitted for review to the NRC staff by the C-E Owners Group. The C-E Owners Group held a meeting with the NRC Divisions of Systems Integration and Human Factors Safety on January 30, 1981, to discuss the process being used for revision of emergency procedure guidelines. As discussed in that meeting, the revised emergency procedure guidelines are currently scheduled to be submitted for review to the NRC staff by the C-E Owners Group on June 1, 1981.

The C-E Owners Group has initiated an effort to define the process by which plant emergency procedures should be developed or modified using emergency procedure guidelines and supporting analyses. This C-E Owners Group activity is scheduled to be completed by May 1, 1981.

Following completion of the NRC review of the revised emergency procedure guidelines to be submitted by the C-E Owners Group on June 1, 1981, and completion of, the C-E Owners Group activities to define the process for development or revision of plant, emergency procedures, APS will evaluate the need for revision of its plant emergen0y procedures. The schedule for such revision, if it is necessary, will be developed at that time.

I.C.1-7

f r lp S

t

PVNGS LLIR I.D. CONTROL ROOM DESIGN I.D.1 CONTROL-ROOM DESIGN REVIEWS Position In accordance with Task Action Plan I.D.1, Control Room Design Reviews (NUREG-0660), all licensees and applicants for operat-ing licenses will be require'd to conduct a detailed control-room design review to identify and correct design deficiencies.

This detailed control-room design review is expected to take about a year. Therefore, the 0'ffice of Nuclear Reactor Regula-

' tion (NRR) requires that those applicants for operating licenses who are unable to complete this review prior to issuance of a license make preliminary assessments of their control rooms to identify significant human factors and instru-mentation problems and establish .a schedule approved by NRC for correcting deficiencies. These applicants will be required to If complete the more detailed control room reviews on the same schedule as licensees with operating plants.

PVNGS Evaluation Arizona Public Service (APS) formed a Control Room Design Review (CRDR) Management Team (Team) and is performing 'a pre-"

liminary assessment, of the PVNGS control room. The Team is comprised of APS engineering, APS operations, Bechtel Power Corporation (BPC) engineering, and BPC consultant, Torrey Pines Technology (TPT) of San Diego, California.

PVNGSLLIR The early part of this effort was divided into three phases.

Phase I of the study developed.,the guidelines to be used -while conducting the CRDR from NRC control room audit reports of other utility facilities and NUREG/CR-1580.

Phase II of the study consisted of the detailed data taking effort and the identification of human factors deficiencies.

The three task areas addressed were human factors, systems factors, and operator preparedness factors. The deficiencies identified were analyzed for'proper resolution and assigned priorities to assist in determining a schedule for implementation.

Phase III, which, is currently in progress, includes preparation and publication of a preliminary report.

Because of the stage of development of emergency operating procedures (EOPs), that portion of .the CRDR involving the walk-through and videotaping of EOPs will start in the third quarter of 1981. When this later part of the CRDR is completed, a final report for submittal to the NRC will be prepared.

The submittal date is targeted for December, 1981.

The review has resulted in APS initiating implementation of the following to date:

o Color demarcation o Instrument relocation o Alarm prioritization o Additional instrumentation P I.D.1-2

PVNGS LLIR I.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE Position In accordance with Task Action Plan I.D.2, Plant Safety Parameter Display Console (NUREG-0660), each applicant and licensee shall install a safety parameter display system (SPDS) that will dis-play to operating personnel a minimum set of parameters which define the safety status of the plant. This can be attained through continuous indication of direct and derived variables as necessary to assess plant safety status.

PVNGS Evaluation A plant safety parameter display system (SPDS) is being developed to display to operating personnel a minimum set of parameters which define the safety status of the plant. The SPDS will pro-vide continuous indication of direct, and derived variables. The requirements of NUREG-0696 will be utilized in development. of the SPDS. The SPDS will be provided and installed using the

'I guidance of NUREG-0696. A schedule will be developed consistent with the operating license review process.

I.D.2-1

PVNGS, LLIR

.I I,... SITING,AND DESIGN II.B CONSIDERATION OF DEGRADED OR MELTED CORES IN SAFETY REVIEW II.B.l ,

REACTOR COOLANT SYSTEM VENTS Position Each applicant and licensee shall install reactor'"coolant sys-'em (RCS) and reactor vessel head high point vents remotely operated from the control room. Although the -purpose of'the system is tovent noncondensible, gases from the RCS which may inhibit core cooling during natural circulation, the vents: must not lead to an unacceptable increase in the probability of a loss-of-coolant accident, (LOCA) or. a challenge to containment integrity. Since these vents, form a part of the reactor, coolant pressure boundary, the design of the events shall. con-form to the requirements of Appendix A to 10 CFR Part 50, "General Design Criteria." The vent system shall be designed.

I with sufficient, redundancy I that assures a low probability.,of

">F inadvertent or irreversible actuation.

Each licensee, shall provide the following information concern-ing.the design and operation of the high point vent system:

(1) .- Submit a'escription of the design, location, 'size, and

...,-- pow'er supply 'for'he vent system along with results of

." analyses -for. lo'ss-of-coolant accidents initiated by a break, in the vent pipe. The results of the analyses 4

PVNGS LLIR should demonstrate '.compliance with the acceptance criteria of 10 CFR 50.46.

(2)'ubmit procedures and supporting analysis foi operator use T

of the vents that also include the information available to the operator for initiating or terminating vent usage.

PVNGS Evaluation Refer to CESSAR Appendix B, Item II.B.l. In addition, Arizona Public Service Company has actively pursued a joint effort through the C-E Owners Group to develop a system design for a reactor coolant gas vent system for C-E plants, including design basis, system description, component requirements, operating guidelines, failure mode and effects analysis, and interface requirements. From this generic effort which has been completed, a system design specific to PVNGS is being developed..

The PVNGS system will utilize existing nozzles and piping on the pressurizer and reactor vessel head to provide high point venting of the reactor coolant system (RCS). Refer to FSAR Figure 5.1-1 for a PAID of the RCS. The pressurizer vent pene-trates 'the top of the pres'surizer and is currently utilized for steam space sampling. The reactor vessel head vent penetrates the top of the reactor vessel and is currently isolated by a manual valve. Piping and valves will be provided to allow remote operation of the pressurizer and reactor vessel vents from the control room. The system will have venting capability to. the reactor drain tank (RDT) and containment atmosphere. In order II.B.1-2

PVNGS I LIR to ensure availability and a low probability of inadvertent actuation, valves and controls will satisfy the single failure criterion for system operation. Valve position indication "will be provided in the control room. The system will be'seismically and environmentally qualified.

The following information will be provided as the RCS vent system design is finalized prior to Unit 1 fuel load:

1. A description of the design, location, size and power supply for the RCS vent system. The description will address compliance to 10 CFR 50.46.

2. Operating procedures for use of the reactor coolant vent system will be available onsite for NRC review.

3. Operating procedures to remove non-condensible gas from steam generator tubes will be available onsite for NRC review.

II.B.1-3

j,i PVNGS LLIR II.B.2 DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR SPACES/SYSTEMS WHICH MAY BE USED IN POSTACCIDENT OPERATIONS Position With the assumption of a postaccident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50% of the core radioiodine, 100% of the core noble gas inventory, and 1% of the core solids are contained in the primary coolant), each licensee shall perform E

a radiation and shielding-design review of the spaces around systems that may, as a result of an accident, contain highly radioactive 'materials. The design 'review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control center's, and instrument 'areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during postaccident operations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or postaccident procedural The design review shall determine which types of I'ontrols.

corrective actions are needed for vital areas throughout the facility.

II.B.2-1

PVNGS LLIR PVNGS Evaluation As part of the PVNGS design process before the incident at TMI-2, the shielding of areas-that require personnel access for accident mitigation (such as the control room) was reviewed to determine that access would not be unduly limited. For mainte-nance actions, this review principally considered shielding separation between redundant ESF components to ensure that repairs to a failed component would not be unduly restricted by radiation from an operating component. In addition, as noted in FSAR Section 3.11, safety related equipment is qualified for the maximum expected radiation dose.

An analysis of the PVNGS shielding design is being performed to determine if TMI level source strengths would inhibit main-tenance access or violate 10 CFR 50, Appendix A, General Design Criterion 19 (GDC 19). The analysis will also review equipment qualification dose limits in accordance with Commission Order and Memorandum CLI-80-21 and NUREG 0588. Specific shielding criteria have been established for systems outside containment that may function during a serious transient or accident, and which may contain highly -radioactive materials. These criteria are applicable to two categories of systems:

1. Systems required for plant shutdown or mitigation of an accident whose consequences may cause the system to contain highly radioactive materials.

II.B.2-2

PVNGS" LLIR

2. Other systems directly connected to the reactor coolant system, containment sump or containment atmosphere which, while not expected to contain highly radioactive material, are postulated to contain such material.

Initial core releases used in the, analyses are equivalent to those recommended in Regulatory Guides 1.4 and 1.7-- and Standard Review Plan 15.6.5. =-Source term concentrations will be calcu-lated for each release by considering conservative dilution volumes. As an example, spray system piping will-be analyzed for a dilution volume that includes RCS I h volume, safety injec-tion tank volume, refueling water tank volume, and spray system piping volume. Safety injection system or sampling system .

piping that could potentially circulate undiluted coolant from a degraded core inside of an intact primary system will be analyzed for RCS volume dilution only. Sump recirculation portions of the safety injection system would, however, be reviewed for full dilution (as with spray system).

The design will be evaluated based upon GDC 19. The following radiation limit guidelines will be used to evaluate occupancy and accessibility of plant vital areas:

1. Vital areas requiring continuous occupancy:

The review shall verify that for vital areas such as control room and the onsite Technical Support Center, the direct dose rate shall not be more than 15 mr/h at all times from all sources.

II.B.2-3

PVNGS LLIR

2. Vital areas requiiing infrequent access or passage-ways to these vital areas:

The review will verify that for these" areas the dose rate is less than 5 R/h. For dose rates greater than 100 mr/h a man-rem calculation, including time.

and motion analysis, will be performed to ensure that U

the integrated exposure for an operator action does not exceed 5 rem as given in GDC 19. For dose rates less than 100 mr/h, a man-rem calculation will not be performed..

Any increase in permanent or tempoiary shielding or procedural control determined to be necessary by the analyses will be completed prior to PVNGS Unit 1 fuel load.

II.B.2-4

PVNGS L'LIR II.B.3-'-'POSTACCIDENT:SAMPL'ING'APAB'ILITY Position I,

J A design and operational review of the reactor coolant and '4

~ k )

containment atmosphere sampling line systems shall be performed to determine the capability of personnel to promptly obtain

'I ~

(less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a radiation exposure 'to any individual in excess of

-~ '

3 and 18-'3/4 rem to the whole body or extremities, respectively.

Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates I

c that personnel could not promptly and safely obtain the t samples, additional design features or shielding should be provided to meet the criteria.

A design and operational review of the radiological spectrum analysis 'facilities shall'e performed to determine the 'capa-bility to promptly- quantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radionucl'ides that are indicators of" the degree of core damage.

Such'adionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel tem-peratures'), and nonvolatile isotopes (which indicate fuel The initial reactor coolant-spectrum should corre-'" 'elting).

spond to a Regulatory Guide 1.3 or 1.4 release. The .review should also consider the, effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents.

II.B.3-1

PVNGS LLIR If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be under-taken to meet the criteria.

In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.

Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable .of being completed promptly (i.e.,

the boron sample analysis within an hour and the chloride sample analysis within a shift).

PVNGS Evaluation An analysis of the design and operation of the PVNGS sampling system is being performed to determine design changes that will reduce exposure during normal and post accident (highly radioactive sample) conditions. This analysis is considering the use of in-line chemical and radiological analyses of samples, as well as the use of syringe type sampling equip-ment for obtaining grab samples.

Required modifications will be completed prior to fuel load.

Procedures for the operation of PVNGS sampling systems (includ-ing post accident conditions) will be in place prior to PVNGS Unit 1 fuel load.

II.B.3-2

PVNGS LLIR II.D.3 DIRECT INDICATION OF REIIEF AND SAFETY-VALVE POSITION Position Reactor coolant system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe.

PVNGS Evaluation PVNGS does not utilize power operated relief valves. The PVNGS primary code safety valves, located at the top of the pressurizer, are headered into the reactor drain tank (RDT) inside contain-ment. Upstream of the common header each code safety valve is monitored for seal leakage by an in-line resistive-temperature device (RTD) (refer to FSAR Figure 5.1-1).

Indirect indication of code safety valve leakage is provided by an increase of RDT pressure and a decrease of pressurizer pressure and pressurizer level, monitored by safety-grade instrumentation.

Positive indication of safety valve position will be provided in the control room. The instrumentation will be environ-mentally c{ualified in compliance with Regulatory Guide 1.89. A plant annunciator alarm will be provided to alarm valve opening.

Installation of positive pressurizer safety valve 'position indication and development of emergency procedures will be completed prior to fuel loading of PVNGS Unit 1.

II.D.3-1

Table II.E.3.1-1 PVNGS PRESSURIZER HEATERS Capacity IE IE 'Reset from, Number of Heaters (kW) 480V Bus Power Controls SIAS Trip Control Room 5-3 element groups 750 NGN-Lll No No No N/A I

'o 5-3 element groups 750 NGN-L12 No No N/A .

1-3 element groups 150 PGA-I 33 Train A Train A Yes No 1-3 element groups 150 PGB-L32 Train B Train B Yes No

PVNGS LLIR diesel generators are sized to accommodate 'this heater capacity concurrent with a forced shutdown or LOCA (refer to FSAR Table 8.3-3). In the event that heater capacity beyond that powered from a Class IE source (150 KW on each train) is required, heaters can be supplied from the non-Class IE power system that is fed from the offsite power system.

Heaters fed from the Class IE power system (300 kW)'re automatically shed upon receipt of a loss of off-site power (LOP) or a safety injection actuation signal (SIAS). They may subsequently be manually reconnected to the ESF buses without shedding of any loads from the Class IE buses.

Sufficient margin exists in each diesel generator "for at least an additional 150 kW heater.

The redundant, pressurizer heaters required for main-tenance of natural circulation at hot standby are fed from Class IE buses via Class IE load side breakers qualified in accordance with safety-grade requirements.

The PVNGS pressurizer heaters comply'with the recommendations of NUREG-0737, November 1980.

II.E.3.1-4

PVNGS "L'QIR-II.E.4.1.,DEDICATED HYDROGEN.PENETRATIONS.

Position ""

using external recombiners or purge systems for post-

'lants r ~ ~

accident combustible gas control of the containment atmosphere',$

~

provide containment penetration systems for external I'hould recombiner 'or'ur'ge'='systems that are dedicated to'that service U '

~

o'nly, that meet the redundancy and'single-'failu're r'equirements of General -Design, Criteria,54- and'6 of Appendix A'o 10 CFR- 50, and that are sized, to satisfy the flow;-requirements of the recombiner, or, purge system.

The procedures for the use of combustible'gas control systems an.accident that =results in a degraded core and 'ollowing, release of radioactivity to 'the containment must be reviewed and revised, if necessary.

PVNGS Evaluation Redundant independent combustible gas control systems are pro-vided. Each system has* dedicated containment penetrations, external hydrogen monitors, and connection points for an external hydrogen recombiner. Two portable hydrogen recom-biners will be onsite and available for connection to the affected unit. Either recombiner is capable of reducing hydrogen levels as noted in FSAR Section 6.2.5. The two systems are completely independent and meet single failure criteria.

II.E.4.1-1

PVNGS LI IR'.

An additional hydrogen"reductio'n"capabi'lity is provided by a" charcoal filtered purge exhaust unit. This non safety grade unit can be connected to either set of gas control containment penetrations. This capability would only be utilized in the l

event of separate failures in both recombiner units.

The arrangement of these combustible gas control methods is shown in- FSAR Figure 6.2.5-1.

Procedures for use of the PVNGS hydrogen control system following an accident, resulting in a-degraded core and release of radioactivity in the containment will be reviewed and revised if necessary.

Therefore, the PVNGS containment'ydrogen control system corn<<

plies.,with the recommendations .of NUREG-0737.

k I 1 II.E.4.1-2

PVNGS LLIR Table II.E.4.2-1 ESSENTIAL SYSTEMS 'PENETRATING THE.PVNGS CONTAINMENT Normal Post CIAS System, Position Position Notes HPS I "Closed Closed ZPSI Closed Closed Containment Spray Closed Closed Recirculation Sump Suction Closed Closed 3 Long Term Recirculation Closed Closed Auxiliary Feedwater Closed .Closed 5 H Control Closed Closed 4,7 Containment Pressure Sensor Open Open 6

1. Opens on Safety Injection Actuation Signal

2. Opens on Containment Spray Actuation Signal

3. Opens on Recirculation Actuation Signal

4. Manually opened from control room

5. Opens on Auxiliary Feedwater Actuation Signal

6. Function requires it to remain open

7. Isolates on CIAS II.E.4.2-7

PVNGS' LIR Table II.E.4.2-2 NON-ESSENTIAL SYSTEMS PENETRATING CONTAINMENT (Sheet 1 of 2)

THE'VNGS Normal Post CIAS System Position Position Notes Demineralized Water C C Locked Fire Protection C C' Locked Shutdown Cooling C 'ocked t

Pool Cooling C C Locked Fuel Transfer C C Flanged

~

Containment Test C C Flanged Service Air C' C Locked Integrated Leak Rate Test C Flanged Personnel Lock C C Equipment Hatch C C Emergency Lock C C Pressurizer Sample - Water C C Pressurizer Sample - Steam C C Hot Leg Sample C C 1 High Pressure Nitrogen C C 1 Containment Purge (Refueling) C C -1 2 Radiation Monitor 0 C 1 Low Pressure Nitrogen 0 C 1 Instrument Air 0 C 1 Nuclear Cooling Water 0 C 1 CVCS Letdown 0 C 1 Reactor Drain Tank (RDT) Vent 0 C 1 CVCS RDT Drain/Fill 0 C 1

,Chilled Water 0 C 1 Power Access Purge 0 C 1,2 Containment Normal Sump 0 C 1 Main Steam 0 0 ~

3 Main Feedwater 0 0 Steam Generator Blowdown 0 0 Steam Generator, Blowdown Sample 1 XI.E.4.2-8

PVNGS.LLIR II,F. INSTRUMENTATION. AND CONTROLS II.F.1 ADDITIONAL ACCIDENT-MONITORING INSTRUMENTATION ATTACHMENT 1,,NOBLE GAS EFFLUENT'ONITOR Position Noble gas effluent monitors, shall be installed with an extended range designed to function during accident conditions as well, as during normal operating conditions. Multiple

'moni~-'ors.,are considered, necessary to, cover the ranges of interest'.

(1) Noble gas effluent monitors with an upper range capacity of 10,.pCi/cc (Xe'<<133.,) are considered to be practical and should be installed. in all operating plants.'2)'oble gas effluent'onitoring shall be provided for the total range of concentration extending from normal condi-tion (as low as"reasonably achievable (ALARA)) concentra'-

tions to a maximum of 10 pCi/cc (Xe-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors should overlap by a factor of ten.

ATTACHMENT 2 ~ SAMPL ING AND ANALYSI S OF PLANT EFFLUENTS Position Because iodine gaseous effluent monitors for the accident con-dition are not considered'to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted

PVNGS LLIR by adsorption on charcoal or other, media, followed'by'onsite laboratory analysis.

ATTACHMENT 3, CONTAINMENT HIGH-RANGE RADIATION MONITOR Position In containment radiation-level monitors with a maximum range of 10 8 rad/hr shall be installed. A minimum of two such monitors that are physically separated shall be provided.

Monitors shall be developed and qualified to function in an I

accident environment.

This requirement was revised in the October 30, 1979 letter from H.R. Denton to All Operating Nuclear Power Plants to provide for a photon-only measurement with an upper range of 10 R/hr.

ATTACHMENT 4, CONTAINMENT PRESSURE MONITOR Po'sition A continuous indication of containment pressure shall be provided in the control room of each operating reactor.

Measurement and indication capability shall include three, a

times the design pressure of the containment for concrete, four times the design pressure for steel, and -5 psig for all containments.

II.F.1-2

PVNGS. LLIR ATTACHMENT,5, CONTAINMENT WATER LEVEL MONITOR.

Position A continuous indication of containment water level shall be provided in the control room for all plants.'A narrow range instrument shall be provided for PWRs and covei the'ange from the bottom to the top of the containment sump. A 'wide range instrument shall also be provided for PWRs and shall cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon capacity., For.BWRs, a wide range instrument shall be provided and cover- the, range from the bottom to 5 feet. above the normal water level of the suppres-sion pool.

L ATTACHMENT 6, CONTAINMENT HYDROGEN MONITOR Position' continuous indication of hydrogen concentration in the con-tainment atmosphere shall be provided in the control room.

Measurement capability shall be provided over the range of 0 to 10% hydrogen concentration under both positive and negative ambient pressure.

PVNGS.Evaluation Procedures for estimating'noble gas, radioiodine,'and particu-late .release rates are being developed in ac'cordance with Revision 2 to Regulatory Guide 1.97.

II.F.1-3

PVNGS LLIR Continuous indication in the control room is provided or will be provided prior to fuel load for the following parameters in accordance with Revision 2 to Regulatory Guide 1.97:

a. ,

Containment. pressure from minus 5 psig to three times the design pressure of the concrete containment at PVNGS

b. Containment, water level from (1) the bottom to the top of each of two containment normal sumps, and (2) redundant instrumentation from the bottom of the containment to a level 'equivalent to 819,700 gallons, which corresponds to the maximum flood level co Containment atmosphere hydrogen concentration from 0 to 10 volume percent (will be available within 30 minutes of the initiation of safety injection)

d. Containment radiation up to 10 R/hr

e. Noble gas effluent from each potential release point from normal concentrations up to the limits specified in Revision 2 to Regulatory Guide 1.97 for each source.

Capability to continuously sample and perform onsite analysis of the radionuclide and particulate effluent samples will be provided in accordance with Revision 2 to Regulatory Guide 1.97.

PVNGS LLIR A human factor analysis will be performed to ensure that the displays and controls added to accomplish this monitoring do not increase the potential for operator error (see sec-tion I.D.l).

II.F;1-5

PVNGS- LLIR II.F.2 INSTRUMENTATION FOR DETECTION"OF INADEQUATE CORE COOLING Position Licensees shall provide a description of any additional'nstru-mentation or controls (primary or backup) proposed for the plant to supplement existing instrumentation (including primary coolant, saturation monitors) -in order= to provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC);

A description of the functional design requirements for the system shall also be included. -A description of the procedures to be used with the.. proposed equipment, the analysis used in .

developing these procedures, and a schedule for installing the equipment shall be provided.

PVNGS Evaluation 4

Since early 1980, the C-E Owners Group has conducted an evalua-tion of response characteristics of instrumentation under conditions of inadequate core cooling. An outline of this evaluation was discussed with the NRC staff at a meeting in Bethesda, Maryland on May 28, 1980. The instruments whose response characteristics have been evaluated are the subcooled margin monitor, the heated junction (vessel level) thermocouple system, core exit (in-core) thermocouples, self-powered neutron detectors, hot leg resistance temperature detectors, and ex-core neutron detectors. This evaluation was completed in December 1980, and the results have recently been distributed to I I . F,. 2.-1

PVNGS LLIR members of the C-E Owners Group. Arizona Public Service Company (APS) is currently evaluating the results of the C-E Owners Group study to determine which instruments would be included in an instrumentation system for monitoring inadequate core cooling.

APS has participated over the past year in the C-E Owners Group development of a technique for measurement of water level in the reactor vessel. The technique which has been selected by the C-E Owners Group is use of heated junction thermocouples (HJTC) distributed at various radial and axial locations in the reactor vessel above the fuel align-ment plate. The design objective of this system is to provide a measurement of the water inventory in the reactor vessel above the fuel alignment plate. The details of this design activity were discussed with the NRC staff at a meeting in Bethesda, Maryland on May 28, 1980. APS is evaluating the HJTC system as a possible component of an instrumentation sys-tem for monitoring inadequate core cooling.

APS is also currently investigating the feasibility of the differential pressure method of reactor vessel level measurement. This system is currently under development by Westinghouse Electric Corporation.

II.F.2<<2

PVNGS LLIR II.K.3.25 EFFECT OF LOSS OF ALTERNATING-CURRENT POWER ON PUMP SEALS Position The licensees should determine, on a plant-specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers.

The pump seals should be designed to withstand a complete loss of alternating-current (ac) power for at least 2,hours.

Adequacy of the seal design should be demonstrated.

PVNGS Evaluation t

Refer to CESSAR Appendix B, Item II.K.3.2.5. In addition, the reactor coolant pump normal cooling water system (nuclear cooling water system) is backed up by the essential cooling water system to supply cooling water to the seals during loss of ac power. On loss of ac power the nuclear cooling water pumps will stop, but the redundant essential cooling water system, powered from the Class IE redundant power supplies, is available to supply the reactor coolant pump seals with cooling water. The essential cooling water system is described in the PVNGS FSAR Section 9.2.2.

II.K.3.25-1

PVNGS LLIR (d) 3 Remote interrogation of systems producing meteoro-J logical data and effluent transport and,diffusion elements.

PVNGS Evaluation I

(1) Comply withf 10CFR50 Appendix E The PVNGS-Emergency Plan submitted with the FSAR complies with. Appendix E, ".Emergency Facilities" to 10 CFR-Part 50 and to Regulation. Guide 1.101 as discussed>in- FSAR Sec-tion 1.8.

The Arizona Division of Emergency Services is currently preparing a comprehensive emergency response plan, for fixed nuclear facilities which will incorporate actions to be taken by levels of government within the State of Arizona. This plan is expected to meet the essential elements of NUREG-75/111 and is expected to be forwarded to FENA for review prior to fuel loading of Unit 1.

(2) Comply with NUREG-0654 The PVNGS Emergency Plan is currently being revised and updated to address the emergency planning criteria contained in NUREG-0654, including the means for providing prompt notification to the public, staffing for emergencies, and an upgraded meteorological program. The comprehensive emergency response plan for fixed nuclear facilities cur-rently being developed by the Arizona Division of Emergency III.A.1.1-3

PVNGS LLIR Services is expected to address the emergency planning criteria contained in NUREG-0654.

(3) Condu'ct Exercise 1

Prior to fuel loading of Unit 1, an emergency response exercise will be performed. The Arizona Division of Emergency Services will be invited to participate and major elements of both the PVNGS Emergency Plan and the state emergency response plan for fixed nuclear facilities-will be tested.

(4) Meteorological Data Response to this item is being evaluated.

III.A.1.1-4

PVNGS;.LLIR III.A.1.2 UPGRADE EMERGENCY SUPPORT FACILITIES Position (NRC letter, D.G. Eisenhut to All Licensees of r

Operating Plants, and,Holders of Construction Permits, dated February 18,,1981)

Each operating nuclear power plant shall maintain'=an -onsite technical support. center (TSC)-separate from and in close proximity to the, control room -that has the capability to"display and-transmit plant.'status.,to those'ndividuals who -are knowledge-able of- and responsible 'for engineering and man'agement .support'f reactor. operations in the-event of an- accident. The center' shall be habitable to the same degree's the control room for accident conditions. The licensee shall revise his 'ostulated emergency plans as necessary to incorporate the role and,location of the TSC. Records, that pertain to the as-built conditions 1

and layout of structures, systems, and components shall be.

readily available to personnel in the TSC.

An operational support center (OSC) shall be established separate from the control, room and other emergency response facilities as a place. where operations support personnel can assemble and.

report in .an emergency situation.to receive instructions from-m the operating staff. ,Communications shall be provided=between the OSC,,TSC, EOF, .and control room.

.An emergency. operating facility (EOF) will be operated by the

'for 'continued evaluation and coordination of all 'icensee licensee activities related to an emergency having or:potentially

~ ~ ~

• .having:environmental .consequences.

~ ~

III.A.1.2<<1

PVNGS" LLIR PVNGS Evaluation Prior to PVNGS Unit 1 Operating License, an area separate from the unit control rooms will be'esignated as the PVNGS onsite Operational Support Center (OSC); This will be an assembly area where, plant operations support personnel will report.'in an emergency situation for further orders or assignment. The OSC will have communication with the control room and the onsite Technical Support Center (TSC). Satellite TSCs will be esta-blished in each control room to facilitate face to face com-munications between control room personnel and support personnel.

The satellite TSCs will have the same display capabilities as the main TSC.

The Emergency Plan will establish will reflect the existence of the OSC and 4 the methods and lines of communication and management.

An onsite TSC and the satellite TSCs for Palo Verde Nuclear Generating Station will be established prior to fuel load.

This Center will be located within the site protected area and will provide a work area 'for supervisory and technical personnel from Arizona Public Service Company and the NRC. It will have communications with the -control room and the other response centers. An analysis will be performed to determine equipment to be incorporated into the Center to monitor and display the status of the affected unit. The Center will be habitable to permit occupancy following a loss of coolant accident.- Monitor-ing eguipment will be provided for direct and airborne radioactive 0 III .A. 1. 2-2

Q ~

PVNGS LAIR; contaminants to provide warning;if,radiation'evels"in the TSC are reaching dangerous levels.. As-built. drawings and'other

~

appropriate records shall be available.in readily, retrievable, a

form at the site and shall be accessible to the TSC.

The Palo Verde Nuclear Generating Station EmergencyI Plan. will,,

be revised to describe the TSC and its function.,

An Emergency Operations Facility (EOF) for PVNGS will be established prior to PVNGS Unit 1 operating license. This facility will be located outside of the protected area and will provide space for operations personnel in support of the TSC.

Plant personnel will be able to evaluate the magnitude and effect of radioactive releases from the plant and recommend appropriate protective measures. The facility will be habitable following a loss of coolant accident. A study will be con-ducted to determine analysis and monitoring equipment required.

Communications with the TSC will be provided.

The EOF will provide space for recovery operations following an accident as well as training facilities for the site.

The PVNGS Emergency Plan will be revised to describe the EOF and it's functions.

The TSC will have permanently installed radiation monitoring and filtered ventilation systems. The nonredundant systems in conjunction with building shielding will ensure that the com-bined doses from airborne radioactivity and direct radiation meet GDC 19.

I-II.A.1.2>>3

PVNGS LLIR The 'EOF will have ventilation isolation capability and be shielded against direct radiation.

Display of data at the TSC and EOF will be in accordance with NUREG 0696.

Implementation of these items will be complete before PVNGS Unit 1 fuel load.

III.A.1.2-4

PVNGSLLIR ~

III.D RADIATION PROTECTION III.D.l.l INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN, RADIOACTIVE MATERIAL FOR PRESSURIZED>>WATER REACTORS AND BOILING-.WATER REACTORS Position Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids, during a:serious transient or accident to as-.low-as-practical, levels. This program shall. include the following:,

'(1) Imme'diate'eak'.reduction

'y y (a) Implement all practical leak reduction measures y

for all systems that could carry radioactive fluid outside of containment.

l (b) 4 Measure ll actual leakage rates with system in operation and report them to the NRC.

(2) Continuing Leak Reduction -- Establish and implement a program of preventive maintenance to reduce leak-E age to as-low-as-practical levels. This program shall include periodic integrated leak tests at l

intervals not to exceed each refueling cycle.

h ~

y l

PVNGS LLIR PVNGS Evaluation Refer to CESSAR Appendix B, 'Item III.D.1.1. In addition, a PVNGS design review was performed on the system below. to assure that potential radioactive release paths following a serious transient or accident is reduced to as-low-as-reasonably achievable (ALARA) levels.

o Shutdown Cooling System (SCS)

The existing design" incorporates all-welded piping.

Vent and drain lines throughout the system are capped when not in use. Relief valves on the system relieve to the equipment drain .tank (a tank designed to accept radioactive fluids). The leakage from the LPSI pump seals and system valve stems is ALARA.

Potential leakage from -the SCS into the essential

'cooling water system (through the shutdown heat-exchanger) can be detected during normal operation by installed radiation monitoring.

o Containment Spray Recirculation System (CS)

The existing design incorporates all-welded piping.

Vent and drain lines throughout the system are h

capped when not, in use. Relief valves on the system (external to the containment) relieve to the equip-ment drain tank. The leakage from the CS pump seals and system valve stems is ALARA. Potential leakage during normal operation from the CS into the essential III.D.1.1-2

PVNGS LLIR cooling water system (through the shutdown heat exchanger) can be detected by installed radiation monitoring.

~ CVCS Charging and Letdown System

'The existing'design incorporates all-welded piping.

The letdown 'system is isolated upon CIAS and SIAS.

Relief valves on the system relieve to the equipment dr'ain tank.

The leakage from the CVCS charging pumps (positive displacement pumps) and other system equipment is ALARA as they are hard-piped to drains. The nuclear cooling water system is monitored for potential

.leakage from the CVCS through the letdown heat exchanger during normal operation.

~ Sampling System The existing design of the normal sampling system incorporates "Swagelok" connections, however, the design will be upgraded to all-welded piping for sections which would come into contact with highly radioactive fluids. The system is isolated upon CIAS and SIAS. Relief valves" relieve to 'the equipment drain tank. Leakage from the system is also mini-mized by the small size of the lines.

The post-accident sampling system will also be constructed of all-welded piping.

III.D.1.1-3

PVNGS LLIR High-Pressure Injection Recirculation (HPSI)

The design incorporates all-welded piping. Relief valves on the system (external to the containment) relieve to the equipment drain tank. The vent and drain lines throughout the system are capped when not in use. Leakage from the HPSI pump seals and system valve stems is as-low-as reasonably achievable.

Miniflow connections to the refueling water tank (RWT) are isolated upon Recirculation Actuation Signal (RAS).

Manual cross over valves to the CVCS are normally locked shut.

Waste Gas System The waste gas system is isolated from the contain-ment upon CIAS. (The normal vent path from the reactor drain tank (RDT) and the reactor head vent system is isolated.) By design, the introduction of highly radioactive fluids to the system is precluded.

o Other systems containing radioactive materials which are excluded from this program (1) Radioactive Waste Drain System (RDS):

The RDS is isolated upon CIAS to ensure the liquid radwaste system will not become highly contaminated. due to post-transient, activities.

PVNGS LLIR III.D.3.3 IMPROVED INPLANT IODINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS Position (1) Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.

(2) Each applicant for a fuel-loading license to be issued prior to January 1, 1981 shall provide the equipment, training, and procedures necessary to accurately deter-mine the presence of airborne radioiodine in areas within the plant where plant personnel may be present during an accident.

PVNGS Evaluation Prior to fuel load, procedures will be developed for determin-ing airborne iodine concentration. Charcoal cartridges will be used in conjunction'ith a portable pump or a fixed vacuum system. The cartridges will be removed to the counting labora-tory for gamma spectrum analysis. Procedures will also define ALARA concepts for removal, transport, and analysis.

III.D.3.3-1

IO I

~

l)r l,

PVNGS LI-IR A.1.3 SHIFT. MANNING, Posi on Lim t Overtime

'1) t rative procedures shall be established to"l'imit g N Adminis maximum wo k hours of all personnel performing a s'afety-related func 'on to no more, than. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of continuous

. duty with at le t 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> between work .periods,. no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a 7-day period, and no more than 14 con-secutive days of work without:at .leasg 2.consecutive days off.

(2) Minimum Shift Crew The minimum shift crew for a it shall include three operators, plus an additional th ee,operators when the unit is operating. Shift staffing ay be adjusted at h

multi-unit stations to allow credit r operators holding

"'icenses on more than one unit.

In each control room, including common con rol rooms for multiple units, there shall be at all times licensed reactor operator for each reactor loaded with uel and a senior reactor operator licensed for each reacto that is 4

operating. There shall also be onsite at all times an additional relief operator licensed for each reactor, a A ~ vi 4 licensed senior reactor operator who is designated as

& II shift supervisor, and any other licensed senior reactor I.A.1.3-1 S

PVNGS .LLIR

\

operators required so that their total. number is at least one more than the number of control rooms from which a .,

reactor is being operated.

PVNGS Evaluation 4

Limit. Overtime PVNGS administrative procedures shall,'y fuel load, pro-vide provisions limiting maximum hours worked by licensed operators, non-licensed operators, and other key personnel who perform safety-related functions.

2. Minimum Shift Crew The minimum shift organization as described in FSAR Sec-tion 13.1.2.3 will be updated to require, commencing at fuel load, for each unit:

A licable Modes a License Cate or 1, 2, 3 and 4 d 5 and 6 SOP (SRO) 2(b) 1(b, c)

OP (RO)

Non-Licensed

('a) Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew member, provided immediate action I.A.1.3-2

PVNGS LLIR is taken to restore the shift crew composition to within the minimum requirements of this table.

Operational modes are as defined in CESSAR Table 1.1, Section 16.

(b) At least one of the unit senior reactor operators on site and on shift will be a shift supervisor. At least one senior reactor operator will be in the unit control room during mode 1 through 4 operations for each unit.

(c) Does not include the licensed senior reactor operator or senior reactor operator limited to fuel handling, supervising core alterations. A licensed senior operator is required to directly supervise any core alteration activity.

(d) A shift technical advisor shall be on site and on shift (available within 10 minutes) whenever one or more units are operating in modes 1 through 4.

PVNGS administrative procedures will by fuel load provide provisions governing required shift staffing and movement of key on-shift individuals around the plant.

Z.A.1.3-3

A ll II ii

PVNGS LLIR (5) Training with emphasis on reactor and plant transients.

3. Three Month Training On-Shift The training program described in FSAR Section 13.2.1.1.1 provides a combination of experience and training in the control room and on a simulator that provides license candidates appropriate control room experience.

4. Modify Training Heat transfer, fluid flow and thermodynamics will be included in Phase I training. Our response to item II.B.4 addresses training for mitigating core damage. Plant transients are currently covered in Phase IV training on the simulator.

5. Facility Certification The PVNGS operating organization is described in PVNGS FSAR Section 13.1.2. The Vice President of Electric Operations for Arizona Public Service. Company is the highest level of corporate management for plant operation and, as such, shall sign certifications pursuant to Sec-tions 55.10(a)(6) and 55.33a(4), and (5) of 10 CFR Part 55.

I.A.2.1-5

I S

PVNGS LLIR I.A.2.3 ADMINISTRATION OF TRAINING PROGRAMS Position Pending accreditation of training institutions, licensees and applicants for operating licenses will assure that training center and facility instructors who teach systems, integrated responses, transient, and-simulator courses demonstrate senior reactor operator (SRO) qualifications and be enrolled in appro-priate requalification programs.

PVNGS Evaluation Commencing at Unit 1 fuel load, permanent members of the training staff who instruct licensed operator candidates in systems, integrated transient responses, and control room simu-lator courses shall have 4 years of power plant experience, of which 2 years shall be nuclear power plant experience (includ-ing 6 months at PVNGS) with no more than 2 years being academic or related technical training. They shall successfully complete an examination equivalent to an SRO examination. They shall take part in the requalification training to maintain current on plant operating history, problems and changes to procedures and administrative limitations. If licensed as an SRO they shall participate in the requalification training program with the exception stated in PVNGS FSAR Section 13.2.2.1.3.

I.A.2.3-1

J PVNGS LLIR I.A.3.1 REVISE SCOPE AND CRITERIA, FOR LICENSING EXAMINATIONS Position (1) Increase Scope -(NRC letter, H. R. Denton 'to All Pow'er Reactor Applicants and Licensees, dated March "28, 1980)

(a) A new category shall'e added to .the o'perator wri'tten examination

\

entitled,,"Principles of, Heat Transfer and Fluid Mechanics."

(b) A new category shall be added to the senior operator written examination entitled, "Theory of Fluids and Thermodynamics."

(c) Time limits shall be imposed for completion of the p I I

written examinations:

1. Operator: 9 hours

2. Senior Operator: 7 hours (d) All applicants for senior operator licenses shall be

~

'equired to be administered an operating test as well as "'the written examination.

(e) Applicants will grant permission to NRC,to,inform their facility management regarding the results of the examinations for purposes of enrollment in requalification programs.

I I <> OQ O Q 21 22 23 MISSILE SHIELD PRE.ACCESS NORMAL AFU STEAM GENERATOR NO. 1 21 22 23 WASTE GAS DECAY TANK WASTE GAS SURGE TANK EVAIDRATOR SURFACE CONDENSER I ~ I I ~ Qi n n 24 EVAPORATOR VAPOR ROY I Qa 24 CEOM NORMAL ACU UNIT ANO FAN 25 WASTE GAS COMPRESSOR 25 FUEL CARNAGE WINCH II I I I I I ELEVATOR MACHI)E RM.

H2 AND 02 ANALYIER 26 8 OI ~ Q. OI I 27 TRANSFER SYSTEM CONTROL CONSOLE

~

Qi (3 s ttainn i Qn I 28 29 HYDRAUUC POWER PACK CLOSED HEAD UFT FVG ASSEMBLY AND CEDM CONTROL 9I 20 EXHAUST MONITOR DECOM AREA a 21 SUPERVSOR OFFICE 30 RBCTOR HEAD lAYDOWM AND SEAL RING STORAGE CONTROL BLDG. Oi 22 (UTCH(M I 31 STEAM GENERATOR NO, 2 ZONE 6 QO QO Ojj 32 CEA CHANGE PLTfM 23 24 PANTRY WOMEN'S TOILET D I 33 PRESSURIZER 25 MEN'S TOLEf Qi Qn 34 REFUEL MACH, Qpa QOD 26 CONFERENCE ROOM 27 COMPUTER ROOM I TURBINE 28 INSTRUMENT REPAIR BOOM Qn 29 Qii I Qi ~

20 21 HP TURBINE LP TURBINE 30 STOREROOM CONTROL ROOM 0 I OCS L AUXIUARY 22 23 24 REHEAIER DRAIN TANK NORMAL HR HANOUNG UNIT MAIM TURB. LO MIST EUMINATOR

'QO DESCRIPTION II

~

LEGEND'ONE 1 X ( 0.5 mR/hr 0.54X(25mR/hr

((

25 EXHAUST RAO. MONITOR 20 RESIN FILL (TYP. 6 PLCS) 26 GENBIATOR ZONE2 1

21 22 EXHAIST MONITOR OOMEST)C HOT WATER HEATER 27 COLLECTOR HOSE 1.

2.

HAfCH ll ZONE 3 2$ <X 15 mR/hr 23 SAMPLE RACKS TOLEIS DIESEL GENERATOR 3.

PIPE CHASE ELEC. CHASE ZONE 4 15 4 X 100 mR/hr 24 25 SHOWERS DRYING 3

4 4.

5.

HVAC. CHASE SVMP <X ZONE 5 100< (1000 mR/hr 20 EXHAUST SILENCBI 4 (5

(

26 2 6.

MEN'6 LOCKER ROOMS NO. 1 AND 2 21 COMBUSTNN AIR STAKE ELEVATOR ZONE 6 1 X R/hr 4LDG 27 28 29 WOMEN'S LOCKER ROOM CLOTHING SSVE, 22 23

'SSENTIAL EXHAUST FAM NORMAL EXHAUST FAM 7.

8.

MONORAIL VALVE GALLERY I ZONE 7 5 << X 10 R/hr 9.

ZONES 10~<X(100 MCC 30 JANITOR 24 AIR DRYER

'10. LOAD CENTER 25 FUEL OL OAY TANKS I 'I 1004 R/hI'ONES CLEAN STORAGE 31 STORAGE 3 26 STARTING AIR PACKAGE

~ SWITCHGEAR X R/hr 12 EOVIP OR ACCE OPENING 33 ODT ROOM 13. TRANSMITTER RACK Aohi ahcttontntootaotcotto 31 LUNCH ROOM I~ . VOLTAGE REG.

olh tolttct a otcncta ott To lliTAAAI 35 H. P. OFFICE aCAT OA tttott 15. RBAY CAB.

OKAATICIAtliilIO IIAA 36 HEALTH PHYSICS SPEN'f RIEL POOL UGHTING PANEL I

~ tstttn lls IOA IDCATaollo INST RESP MAINT ISSUE 21 SPENT FUEL HANOUNG MACHINE 16.

tta& AOADAACtl AATIDI 37 17. OSTRIBUTION PANEL tottttt noh toaa AllIKITOTCDADO h 38 39 OfACE SAMPUNG AREA; 22 CASK lOAONG PIT NEW FUEL STORAGE COVERS 18.

19.

JUNCTION BOX PIPE TRENCH

~ DENOTES VP U.M.O.

40 FIRST AIO 24 CASK UFTING IUG lAYDOWMAREA Cl SOCATES ONE-WAY DOOR OR CONTROLLI(0 41 PERSONNEL OECON AREA 2 25 DECON PIT f ROM USIOE OF BLDG H. P. SAMPLE COUNT. ROOM DECON PIT COVER 42 GRATE 27 GATE STORAGE 43 HOT LAB SAMPLE ROOM HOOD ANO SINK 28 NEW FUEL INSPECTNN PIT 44 INDICATES CCNTRDLLED 45 F(AMMABLESTORAGE NEW FUEL CONIAINERS LAYOOWN AREA 30 UPENDER HYD. DRIVE UNIT ACCESS AND/(IR ROLl. UP 46 C(UNDER STORAGE DOOR 47 PIPING AREA 31 POOL COOUNG PURIFICATIOM PANEL 48 COUNTYNG ROOM CANAL TOOL AND EOUIP. ROOM 33 NEW FUEL ELEVATOR 49 50 SPECIAL DECON.

51 OECON. ROOM 52 EMERGENCY SHOWN AND EYE WASH Pah Verde Nudear Generating Station 53 COLD (AB 54 CORRIDOR LLIR RADIATION ZONES LOCA WITH SUMP RECIRCULATION BETWEEN EL. 140'-0" fn 200'-0M Figure II.B.2-7 August 1981 Amendment 1

r m1 k '\

)

a

RP VNIT I I

id'CCI I 2 VNIT) ~/rJ'~

e e I I 47 e

Qi OO (o5'IN \  %%%%%

MAIN ~

BIT STEAM P Qi STRUC.

Qi

.o[ e Qcss-s 47 e

1 r

AUXIUARYSLOG.

ELEV.40 FT. 0 OL LBIELD ZONE 6 H iZO E6 D

Qi 1 Q E

AUXIUARY CONTAINMENT BLDG. .~i BLDG.

PO ZONE 8 1

ZONE 8 Qi

[~~]

yP@ aors i ocacrso coal oo IKIT~ Iioa>o Oars rouses al osarsca oos ro acAT oa lwsls 11> IAAa cacaaraa arsaro>saa F acraa Irs >oa occAT coaao raak aoaaasraaarca Ol I ~ o>sess Aa ao1 >KO!ceo il Qi Talc lia c ii Qi 1 I ZONE 9 LEGEND:

(0.5 ZONE 1 X 2ONE 2 0.5 4 ((

mR/hr X 2$ mR/hr

(

CONTAINNENT SLOG. f30 )

ZONE 3 2.5 ~< X 15 mR/hr 20 REACTOR VESSEL OESCRIPTl0$

HAlCH 2ONE 4 15 << X 100 mR/hr 2 PIPE CHASE I X(

(

21 CAVITY COOLING FANS 22 WET LAY.UP PUMPS 3 ELEC. CHASE 2ONE 5 100 m< 1000 mR/hr 4 HVAC. CHASE 23 RfACTOR DRAIN TANK 5 SUMP 2ONE 6 1 <~ X 5 R/hr 54X(10R/hr Ii D

20 MSSS (SI')

AUX. FEEOWATER TURBINf. PACKAGE I

20 AUX. 4LDG bBWEBI 40'ND SPRAY PUMPS 100'ONT.

6 7

8 0

ELEVATOR MONORAIL )

VALVE GALLERY MCC I ZONE7 ZONE 8 10 w< X ZONE 9 1004

( 100 h/hr X R/hr i 21 22 TUBINE ORNEN PUMP MOTOR DRIVBl PUMP 21 22 23 HPSI. PUMPS LPSI. PUMPS EOUIP. DRAIN TANK 10 11 12 LOAD CENT BI SWITCHGfAR EOUIP OR ACCESS OPENING 24 COOLING WATER HOLD.UP TANK 13 1RANSMITTLR RACK CONTROL SLOG. )74') 25 HEATER DRAIN PUMPS COOLING WATER HOLD UP TANK PUll PS LS 15 VOLTAGE REG.

RELAY CAB3 a- DENOTES UP U.N.O.

INDICATES ONE.WAY DOOR OR 27 CHEMICAL DRAIN TANK PUMPS 16 UGHTING PANEL 20 CONTROL BLDG. NORMAL AHU'S CONTROLLED FROM INSIDE OF 28 TRANSfORMERS 17 051RIBUTION PANEL 21 CONTROL BLDG. OUTSCE AIR WASHER UNIT BLDG 22 23 CONTROL BLDGE. ESF SWGR RM. ESSENTIAL AHU'S ESSENTIAL CHILLED WATER CIRCUIATNN PUMPS 20 30 AUX. STEAM CONDENSATE RECBVER TANK CHEMICAL DRAIN TANKS 16 10 JUNCTNN SOX 0 GATE 24 31 ESSENTIAL AIR CONOITONING UNITS CHEMICAL ADDITION TANKS C> SDICATES CONTROLLED ACCESS 32 ECW CHEMICAL AIXNTION TANKS 25 CONTROL ROOM ESSENTNL AHU'S ESSENTIAL CHILLEO WATER EXPANSIOtf TANK 33 ECW PUMPS AND/OR ROLL UP DOOR 34 ECW PUMP ROOM ESSENTIAL A.C.U.

AUXILIARYSLOG. 27 ESSBITIAL CHILLB5 35 ECW RANATNN MONITORS

Pi'LIR ELEV, 51 FT. 4 IN LEVEL C 36 REACTOR MAKE.UP WATER PUMPS 37 BORIC ACID MAKE.UP PUMPS 38 SHUTDOWN COOUNG HEAT EXCHANGB5 Palo Verde Nuclear Generating Station 33 GAS STRIPPER 40 GAS STRIPPER CONTROL CABINET 41 GAS STRIPPER INSlRUMENT RACK RADIATION ZONES - LOCA DEGRADED CORE - INTACT PREIMARY BETWEEN EL. 40'-0" 100'-0" 24 Figure II.B.2-14 August 1981 Amendment 1

I I I

J

PVNGS LLIR in excess of three rem whole-body or 18 3/4 rem to the extremities), to analyze samples within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for radioactive noble gases, iodines, cesium and non-volatile isotopes, and to analyze samples within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for pH, boron, chlorides, dis-solved oxygen and gaseous oxygen. These plant modifications, collectively referred to as the post accident sampling system (PASS) are illustrated in figures II.B.3-1 and II.B.3-2. Pro-cedures will be developed for obtaining and analyzing these samples.

Desi n Ca abilit Modification of currently existing plant facilities allows plant personnel to obtain samples, in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, of pressurized and depressurized reactor coolant, safety injection, containment air, all containment sumps, all auxiliary building sumps and volume control tank.

In-line analyses with automatic sequencing and computer control and readout will be backed up by a grab sample capability using manual controls. In-line analyses will include isotopic, gross gamma, pH, boron, dissolved hydrogen, chloride, dissolved oxygen and gaseous oxygen. Gaseous hydrogen will be sampled by the containment hydrogen control system.

The onsite facility design and procedures provide the following capabilities:

A. Containment air may be sampled under positive or negative pressures.

August 1981 II .B. 3-3 Amendment 1

PVNGS LLIR B. Provisions are made to purge sample lines, for reduc-ing plate-out,, to insure proper mixing, for minimizing leakage, preventing blockage, to back-flush, to blowdown, for appropriate sample disposal, minimizing crud traps, and for passive flow restrictions (when sampling high pres-sure systems).

C. The PASS sampling lines and components conform to Quality Group D requirements. Isolation valves are used with appropriate automatic closure signals where PASS piping interconnects with higher quality group classification piping.

D. Provisions are included to measure a wide range of isotopes for liquids from 10 pCi/ml to 10 pCi/ml and for gases from 10 pCi/ml to 10 pCi/ml. Sample dilution is avail-able for grab samples but, is not required for in-line analysis.

E. Provisions are included to measure dissolved hydrogen from 0 to 2000 cc/kg.

F. Provisions are included to restrict background levels of radiation at the PASS sample area such that the sample analysis provides results with an acceptably small error

( approximately a factor of 2 ) .

G. Provisions are included to facilitate plant procedures to identify the analyses required, measurement techniques and background level reduction methods.

Amendment 1 II.B.3-4 August 1981

PVNGS LLIR H. Chemical analysis capability is provided which can meet all requirements when exposed to the specified source term.

Provisions are included to be able to draw a grab sample while maintaining ALARA radiation exposures and not exceeding General Design Criteria 19.

1. Shielding calculations were performed using source term specified in NUREG 0737 and Regulatory Guide 1.97 Rev 2 (Refer to section II.B.2).

2. A result of the shielding review, piping used for backup grab sampling in the hot lab sample room area will be lead wrapped to keep operator doses ALARA.

3. High range portable and fixed survey instruments and personnel dosimeters are provided to permit rapid assessment of exposure rates and accumulated personnel exposure.

J. Design parameters are based on the full range of design pressure of the reactor coolant system and containment.

K. When plant modifications are completed and procedures are implemented, testing will be conducted to demonstrate the ability to obtain and analyze a sample by in-line and grab sample techniques.

August 1981 II.B.3-5 Amendment 1

PVNGS LLIR SANPLE INLET PIPING Samples are drawn from the following nine points in each unit:

A. RCS hot leg - north side of RCS B. Inlet to letdown system - south side of RCS C. Safety injection mini flow line - Train A D. Safety injection mini flow line Train B E. Containment, recirculation sump F. Containment radwaste sump G. Auxiliary building sumps H. Volume control tank Containment atmosphere via the containment hydrogen control system Isolation valves provide a quality and material classification break between the sample points and the PASS.

Sam le Return Pi in Samples are returned to the following three points in each unit:

A. Containment Atmosphere (via the containment hydrogen control system) for gas samples only-B. Reactor drain tank - for post accident liquid samples 1l Amendment 1 II.B.3-6 August 1981

~

'VNGS LLIR C. Equipment drain tank - for normal operation liquid samples I

Isolation valves provide a quality and material classification break between the sample return points and the PASS.

Sam le Station The PASS is a modular'system. The valve module is located in a piping penetration room. All pumps, heat exchangers, valves and other sources of possible leakage are located in this module.

The radiological and chemical module is located in the auxiliary building. This contains the in-line isotopic and chemical analysis equipment. The two modules are connected via short pipes through a shield wall that separates the auxiliary building from the piping penetration room.

This method of separation provides for:

A. Ease of routing sample piping to and from the PASS B. Isolation of any leakage during normal and post accident operation C. Ability to repair and maintain the spectroscopic and chemical instruments during normal operation D. Meeting space requirements in an existing building.

The valve module provides the ability to draw, purge, cool, degas, circulate, dilute,,and route samples in the PASS.

August 1981 II.B.3-7 Amendment 1

I PVNGS LLIR The radiological and chemical module provides isotopic, gross gamma, pH, chloride, boron, dissolved oxygen and gaseous oxygen analysis. (Hydrogen analysis is provided by the contain-ment hydrogen control system.)

Isotopic analysis is provided with a collimated Ge (Li) detector with at. least a 45 day supply of liquid nitrogen. All chemical detectors have the ability for remote reagent addition and calibration.

The entire sampling sequence is controlled by a valve sequenc-ing controller which in turn is run by a pulse height analyzer (PHA) with extended capabilities. The PHA is operated by a technician.

Grab samples are available from the hot lab in either a depres-surized, diluted or pressurized sample device. Instrumentation is provided for pressure, tempeiature and flow measurement.

The radiological and chemical module is provided with sufficient shielding to maintain ALARA criteria during normal and post accident operation.

Gas samples will be followed by a nitrogen purge to flush out remaining gases.

Liquid samples will be followed by a demineralized water purge followed by a nitrogen gas purge to flush out remaining fluids.

Nuclear cooling water will be provided for sample heat, exchangers.

Instrument air will be provided for pneumatically controlled valves.

1l Amendment 1 II.B.3-8 August 1981

PVNGS LLIR Control Panel The PASS Control Panel is part of the hot lab PHA computer system. This is located in a shielded room, adjacent to the hot lab. This includes:

A. The valve control assembly with manual switching capa-bility for all remote valves, pumps, and instruments B. Recording devices C. Mini status board D. Microcomputer with CRT display E. In-line instrumentation readout PASS 0 eration The operation of the system is dependent upon communication between the technician in the hot lab and operators in the control room. Prior to sampling a specific point, the hot lab technician verifies with the control room operator that the system isolation valve is open. This may involve overriding a CIAS to re-open a valve. The hot lab operator will then initialize the PHA computer and perform radiological and chemical analysis, directing the sample to a predetermined discharge point. Results will be available .on the PHA CRT and on the recorder printout.

August 1981 II.B.3-9 Amendment 1

PVNGS LLIR Meetin Re uirements The process sampling system discussed in FSAR section 9.3.2 reflects the addition of the post accident sampling system.

This fulfills the NRC requirements as outlined in NUREG 0578, NUREG 0737, and Reg Guide 1.97, Rev 2.

Amendment 1 II.B.3-10 August 1981

PVNGS-Z LIR E. Radiation Monitorin Response of Process and Area Monitors to sev'ere

', damages; behavior of detectors when saturated; method for. detecting'adiation readings by direct"measure-ment at detector output (overranged amplifier);

expected accuracy of detectors at different loc'a-tions; use of detectors to determine extent of core damage.

2; - Methods of determining dose rate inside conta'inment from measurements taken outside containment.

F. Gas Generation v I,

1. Methods of H2 generation during an accident; other sources of gas (Xe, Kr); techniques for venting or disposal of non-condensables.

2. H flammability and explosive limit; sources of 02 in containment or Reactor Coolant System.

(2) Complete Training The course shall be developed and these personnel shall participate prior to fuel loading of unit one.

PVNGS Evaluation (1) Develop Training Program A course will be developed to train shift technical advisors and operating personnel through the operations II.B.4-3

PVNGS KZ IR chain to the licensed operators in'he use of installed plant systems to control or mitigate an accident in which the core is severely damaged. The training program will include 'the topics suggested in the H.R., Denton letter of March 28, 1980.

Managers'and technicians in the Instrumentati'on and Control (I"5 C,), radiation protection and chemistry sections will receive training commensurate with their responsibilities that meets. the recpxirements of the H.R. Denton letter of March 28, 1980.

(2) Complete Training The above training will be completed prior to full power operation.

II.B.4-4

PVNGS.rrLL R I II.E' lt SYSTEM DESIGN, II:E.l.l, AUXILIARY'FEEDWATER SYSTEM EVAIUATION Position"

The Office of Nuclear Reactor Regulation is requiring reevalua-tion of'the auxili'ary feedwater'(AFW) systems for all PWR I

operating plant 1'i.censees and operating license applications.

2 This act'ion'.includes:

/.. <<

reliability analysis that

~

(1) Perform a simplified AFW system uses event-tree and e

fault-tree logic techniques to determine the potential for AFW system failure under, various loss-of-main-feedwater-transient conditions.

Particular emphasis is given to determining potential failures that could result from human errors, common I

causes, single-point vulnerabilities and test, and maintenance outages; (2) Perform a deterministic review of the AFW system using the acceptance criteria of Standard Review Plan Section 10.4.9 and associated Branch Technical Position ASB 10-1 as principal guidance; and (3) Reevaluate the AFW system flowrate design bases and criteria.

PVNGS Evaluation An auxiliary feedwater system review was conducted on August 21 and 22, 1980, and included NRC participation.

1I.E.1.1-1

~ 'I

PVNGS LLIR This review included a review of the PVNGS Auxiliary 'Feedwater System Reliability Analysis; A transcript of this review was provided to NRC on October 17, 1980. 'esponses to open items of the review were provided to NRC on March 6, 1981.

The PVNGS Auxiliary Feedwater System Reliability'Analysis was performed in response to the NRC letter of March 10, 1980, from D.F. Ross, Jr., to All Pending Operating License Applicants of Nuclear Steam Supply Systems Designed by Westinghouse and Combustion Engineering. The'final analysis was submitted to the NRC by APS letter dated February 10, 1981 from E.E. Van Brunt, Jr., to Director of Nuclear Reactor Regulation.

II.E.1.1-2

~ ~

PVNGS LLIR Essential systems are those systems critical to ensure the capability to mitigate consequences of accidents,,to ensure, the, integrity of the reactor coolant pressure boundary and to ensure. the capa-bility to shut down the reactor and maintain in it a safe shutdown condition., Table II.E.4.2-1 lists essential systems penetrating the PVNGS, containment.

No essential systems are functionally isolated by a containment isolation actuation signal (CIAS) with the exception of the hydrogen control system, as classified in the table. This system is not immediately required for accident mitigation. The I

isolation valves can be manually opened from the control room as part of the hydrogen recombiner startup procedures.

h Non-essential systems penetrating the PVNGS contain-ment are listed in table II.E.4.2-2. Non-essential systems are automatically isolated by a CIAS with the exception of the following systems:

A. Those containing locked closed valves or flanged closed connections.

B. Main steam and feedwater C. Steam generator blowdown and blowdown sample and safety injection drain D. RCP seal injection and CVCS charging II.E.4.2-3

'PVNGS LLIR The main steam and feedwater systems, while not essential, aid in heat removal during a small LOCA.

These systems should not, 'therefore, be isolated by a CIAS generated on'low pressurizer pressure. The steam and feedwater" systems are isolated for a main

.steam line break by a main steam isolation actuation signal (MSIS) on high containment pressure (5 psig) or low steam generator pressure (870 psig).

=

The steam generator blowdown and blowdown sample systems are isolated by either a MSIS, auxiliary feedwater actuation, signal (AFAS) or a safety injec-II

.tion actuation signal (SIAS). The safety injection k

drain is isolated on a SIAS. Plant par'ameters which generate SIAS also generate CIAS.

It is desirable to leave'CP seal injection and CVCS charging paths open to provide additional core pro-tection after an accident in which offsite power is available. In addition, the charging pumps can be transferred to emergency power at the discretion of the operator in the event that offsite power is lost. Conversely, it: is undesirable to lose charging or seal injection capability during, normal operation due to an inadvertent CIAS. The potential release of II.E.4.2-4

PVNGS LLIR

,fission. products, through the penetration is not a concern for the, following reasons:

(1) Flow is into the containment"and "RCS.

(2) 'heck valves i'nside the containment prevent

~ 'ackflow out of the containment if'the charging pumps stop."

(3) The connecting portions of the CVCS outside of containment are designed to Safety Class 2, Seismic Category I standards and have design pressure well in excess of containment design pressure.

(4) The operator has the capability of isolating these lines if continued charging or seal injec-tion proves to be unnecessary.

Override of a CIAS signal is available for each con-tainment isolation valve via the control switch for that valve. Resetting of a CIAS, does not result in the automatic opening of containment isolation valves.

Reopening requires operator action for each valve and does not compromise the containment isolation signal.

Item 1 above identifies 5 psig as the containment setpoint pressure that initiates containment isola-tion. Calculations are in progress confirming that the trip setpoint represents the minimum value com-patible with normal operating conditions.

II.E.4.2-5

PVNGS LLIR

6. Containment power access purge isolation valves satisfy the operability criteria 'set forth in Branch Technical Position CSB 6-4.

7. As stated in item 1 above, both the power access purge and the refueling purge isolate on high con-tainment purge radioactivity.

II.E.4.2-6

PVNGS LLIR II.F INSTRUMENTATION AND CONTROLS II.F.l ADDITIONAL ACCIDENT-MONITORING INSTRUMENTATION A human factor analysis will be performed to ensure that the displays and controls added for additional-accident accomplish this monitoring do not increase the potential for operator error (see section I.D.l).

II.F.l.l ATTACHMENT 1, NOBLE GAS EFFLUENT MONITOR Position Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions. Multiple monitors are considered necessary to cover the ranges of interest.

(1) Noble 'gas effluent monitors with an upper range capacity of 10 pCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.

(2) Noble gas effluent monitoring shall be provided for the total. range of concentration extending from normal condi-tion (as low as reasonably achievable (ALARA)) concentra-tions to a maximum of 10 pCi/cc (Ze-133). Multiple monitors are considered to be necessary to cover the ranges of interest,. The range capacity of individual monitors should overlap by a factor of ten.

PVNGS Evaluation FSAR section 11.5 provides detailed descriptions of the effluent monitors installed at Palo Verde Units 1, 2 and 3. This includes August 1981 II.F.1-1 Amendment 1

PVNGS LLIR the additional monitors that have been added specifically to address the NUREG-0737 and Reg Guide 1.97, Rev 2 requirements for radiation monitoring.

A. Wide-Range Effluent Monitor In order to cover the dynamic range required, a normal range monitor is used with a high range monitor. One decade is used for overlap when switching between monitors. Both monitors sample using controlled isokinetic flows. Their combined range

-6 xs 10 pCi/cc to 10 +5 pCi/cc. These monitors are located on the

~

plant vent, main condenser/gland. seal exhaust, and the fuel building vent. In addition to noble gas monitoring capability, these instruments have separate particulate and iodine sample chambers for both the low and high range, and use charcoal cartridges in conjunction with a portable pump. All cartridges will be removed to the counting laboratory for gamma spectrum analysis. High range monitor iodine samples are provided with a lead shield. Procedures will be developed to define ALARA concepts for removal, transport and analysis. These monitors have complete digital readout and control from the Health Physics Office and the main control room. The high range monitors automatically switch to a new particulate/iodine cartridge pair when the current cartridge reaches a preset radiation level.

Filter materials used minimize absorption of noble gases.

Samples are preconditioned as necessary to assure accurate results without damaging the sample assemblies. Each monitor Amendment 1 II.F.1-2 August 1981

PVNGS LLIR is controlled by a remote microprocessor. This microprocessor is linked by a "daisy chain" to a minicomputer which provides multiple informational displays on request by the operator.

A dedicated alarm status line is maintained on the CRT display.

This status line does not move with each change of CRT displays.

Thus alarms are provided regardless of the status of the displays in the Health Physics Office and Main Control Room.

Monitors are provided with an open structural construction that provides for easy maintenance and good heat dissipation.

Backup battery power is provided to assure continued micro-processor memory during a loss of external power sources.

Multiple detectors are used to achieve the dynamic range required. Hard copy readouts are available from dedicated printers in the Health Physics office and the control room.

B. Main Steam Line Monitor One area monitor with a collimating lead shield is mounted adjacent to each main steam line in the Main Steam Support Structure. These monitors measure direct dose rates from the main steam line to identify effluent from the atmospheric dump, main steam relief valves, and auxiliary feedwater pump discharge.

An extra 2 inches of shielding is placed on the containment side of the detector shield. There are a total of 4 detectors with one remote microprocessor for each 2 detectors. The ion chamber covers a range from lmr/h to '10 mr/h. The microprocessor is linked by a "daisy chain" to a minicomputer which provides August 1981 II.F.1-3 Amendment 1

PVNGS LLIR multiple informational displays at the request of the operator.

Alarms are provided regardless of the status of the displays in the Health Physics Office and the Control Room. Hardcopy readouts are provided from dedicated printers in the Health Physics Office and the Control Room. Backup battery power is provided to assure continued microprocessor memory during a loss of external power sources. The detector is designed to operate in a post-accident environmental condition with a background of 10 R/h.

C. Meeting the NRC Requirement These monitors operate in conjunction with the other monitors as discussed in FSAR section 11.5 and fulfillthe requirements as outlined in NUREG 0737 and Regulatory Guide 1.97, Rev 2.

II.F.1.2 ATTACHMENT 2, SAMPLING AND ANALYSIS OF PLANT EFFIUENTS Position Because iodine gaseous effluent monitors for the accident con-dition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the P

accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.

PVNGS Evaluation PVNGS response to this item is included in the evaluation of section II.F.l.l requirements.

Amendment 1 II.F.1>>4 August 1981

PVNGS LLIR II.F.2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING Position Licensees shall provide a description of any additional instru-mentation or controls (primary or backup) proposed for the plant to supplement existing instrumentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC).

.A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.

PVNGS Evaluation (1) PROCEDURES TO DETECT AND RECOVER FROM CONDITIONS LEADING TO INADE UATE CORE COOLING Emergency procedures are being developed which will pro-vide the reactor operators with directions to detect and recover from conditions that could lead to inadequate core cooling and will be in place prior to fuel load.

These procedures will be prepared by the NSSS vendor.

The procedures will be listed in FSAR section 14.2 later.

(2) INADE UATE CORE COOLING INSTRUMENTATION Inadequate core cooling monitoring requirements can be met by appropriately measuring and displaying margin to saturation, August 1981 II.F.2-1 Amendment 1

PVNGS LLIR reactor vessel water level above the core, and reactor core exit temperatures. In order to accomplish these functions, the PVNGS Unit 1, 2, and 3 design will utilize three instruments: (1) The Subcooled Margin Monitor to measure saturation/superheat margin, (2) The Heated Junction Thermo-couple System to monitor level/temperature in the region of the upper portion of the reactor vessel, and (3) Core Exit, Thermocouples to measure the temperature of the reactor coolant as it leaves the reactor core.

The Subcooled Margin Monitor is an instrument which indicates the degree of subcooling and superheat in the reactor coolant either in terms of temperature or pressure. The system is designed as a two-channel system, each channel measuring I four points within the reactor coolant system as follows:

Reactor Coolant Piping Hot Leg Temperature (1)

Reactor Coolant Piping Cold Leg Temperature (2)

Pressurizer Pressure The range of the Subcooled Margin Monitor is increased by incorporation of signals from the Heated Junction Thermo-couple System and the Core Exit Thermocouples such that the instrument can also calculate and display degrees superheat.

The Heated Junction Thermocouple System monitors liquid in the region above the core. Redundant strings of Amendment 1 II.F.2-2 August 1981

PVNGS LLIR II.K.3.17 REPORT ON OUTAGES OF EMERGENCY CORE COOLING SYSTEMS LICENSEE REPORT AND PROPOSED TECHNICAL SPECIFICATION CHANGES Position Several components of the emergency core cooling (ECC) systems are permitted by technical specifications to have substantial outage times (e.g., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one diesel-generator; 14 days for the HPCI system). In addition, there are no cumulative outage time limitations for ECC systems. Licensees should submit a report detailing outage dates and lengths of outages for all ECC systems for the last 5 years of operation. The report should also include .the causes of the outages (e.g.,

controller failures, spurious isolation).

PVNGS Evaluation Arizona Public Service Company will submit a plan for data collection of outage dates and lengths of outages for ECC systems prior to fuel load of Unit, 1.

II.K.3.17-1

I PVNGS LLIR II.K.3.25 EFFECT OF LOSS OF ALTERNATING-CURRENT POWER ON PUMP SEALS Position t

The licensees should determine, on a plant-specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers. The pump seals should be designed to withstand a complete loss of alternating-current (ac) power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Adequacy of the seal design should be demonstrated.

PVNGS Evaluation Refer to CESSAR Appendix B, Item II.K.3.2.5. In addition, the reactor coolant pump normal cooling water system. (nuclear cooling water system) is backed up by the essential cooling water system to supply cooling water to the seals during loss of ac power. On loss of ac power the nuclear cooling water pumps will stop, but the redundant essential cooling water system, powered from the Class IE redundant power supplies, is available to supply the reactor coolant pump seals with cooling water. The essential cooling water system is described in the PVNGS FSAR Section 9.2.2.

A C-E test program'as conducted to determine the effect on the reactor coolant pump seals due to a loss of component cooling water to the pump seal assembly. The test results demonstrate that the pumps will continue to operate for extended periods

/

without exceeding design seal leakage limits and seal August 1981 II .K.3.25-1 Amendment 1

PVNGS LLIR temperature limits. It has been shown that the System 80 reactor coolant pumps are capable of operating without component cooling water for 30 minutes without sustaining damage to the pump thrust bearings.

Amendment 1 II.K.3.25-2 August 1981

PVNGS LLIR (d) Remote interrogation of systems producing meteoro-logical data and effluent transport and diffusion elements.

PVNGS Evaluation (1) Comply with 10CFR50 Appendix E The PVNGS Emergency Plan submitted with the FSAR complies with Appendix E, "Emergency Facilities" to 10 CFR Part 50 and to Regulation Guide 1.101 as discussed in FSAR Sec-tion 1.8.

The Arizona Division of Emergency Services is currently preparing a comprehensive emergency response plan for fixed nuclear facilities which will incorporate actions to be taken by levels of government within the State of Arizona. This plan is expected to meet the essential elements of NUREG-75/111 and is expected to be forwarded to FENA for review prior to fuel loading of Unit 1.

(2) Comply with NUREG-0654 The PVNGS Emergency Plan is currently being revised and updated to address the emergency planning criteria contained in NUREG-0654, including the means for providing prompt notification to the public, staffing for emergencies, and an upgraded meteorological program. The comprehensive emergency response plan for fixed nuclear facilities cur-rently being developed by the Arizona Division of Emergency III.A.1.1-3

PVNGS LLIR Services is expected to address the emergency planning criteria contained in NUREG-0654.

(3) Conduct Exercise Prior to fuel loading of Unit 1, an emergency response exercise will be performed. The Arizona Division of Emergency Services will be invited to participate and major elements of both the PVNGS Emergency Plan and the state emergency response plan for fixed nuclear facilities will be tested.

(4) Meteorological Data The PVNGS meteorological system is designed to satisfy all requirements set forth in Regulatory Guides 1.23, Rev. 1; 1.97, Rev. 2; and NUREG's-0696, 0654, and 0737.

It will consist of two independent towers, sensors, signal conditioning, amplifiers and processing systems with backup power source. The system includes a highly reliable digital data processor system to develop real time dispersion and transport estimates. Additionally, auxiliary analog information is recorded for all parameters monitored in each independent system locally.

Real time validation of digital meteorological data will identify suspect data so that backup data will be used based on data reasonability by the comparison percentage method.

Amendment 1 III.A.1.1-4 August 1981

PVNGS LLIR Data acquisition of the two independent meteorological system signals may be accomplished for projected dose calculations, CRT displays and remote data transmission.

The same capability is provided for the nuclear data link.

As independant, redundant and validated signals are used, the system can assure a non-availability of 0.01.

Below is a summary table of the sensors used in each inde-pendent meteorological system to monitor the environmental parameters, and the common data processing available.

Sensor 195'ind 35'ind Rainfall Speed and Speed and 195' Monitor Wind Wind 35'5'spirated Aspirated Located irection Direction Temperature Dewpoint at. Ground Location Sensor Sensor Sensor Sensor Level Tower System A Provided Provided Provided Provided Provided Tower Not Not System B Provided Provided Provided Provided Provided Digital

& Analog 35'mbient Process- Provided Provided Onl Provided Provided ing 195'-35'T The environmental parameters monitored by each independent tower system permits highly accurate and reliable meteor-ological data necessary to cover all data for the Pasquill stability classes and transport protections needed for>the PVNGS site.

The auxiliary analog information for each tower system is provided on analog recorders located at the meteorological August 1981 III.A.1.1-5 Amendment 1 f1

PVNGS LLIR station in the respective tower system equipment trailers.

Analog information for each tower system is converted to digital data and transmitted by two serial links to the plant site where the data is reduced to fifteen-minute and hourly average meteorological parameters, and where all effluent and environmental parameters are recorded and available for time history displays in the control room, emergency response facilities, and at external locations.

Serial data is also provided to the radiation exposure management system to meet the reporting requirements of Reg. Guide 1.21. In addition, the signal is available to the multi-channel processing system for offsite and projected dose calculation, technical support center and the emer-gency operations facility CRT displays.

The models used for providing the estimates of offsite exposure are:

Plume exposure - Class A model will be used with 15 minute average meteorology data for initial transport and diffusion estimates within 15 minutes following the classification of the incident.

Ingestion Zone - Class B model will be used for relative concentration of the plume emergency planning zone (EPZ) and ingestion EPZ.

The output of the processor provides graphic displays of 2, 10, and 50 mile graphics with terrain, major population zones, and a display of environs for radiation monitors located in the plume EPZ. Capability is provided to view this display in the Technical Support Center (TSC),

alternate TSC's, and offsite Emergency Operations Facility (EOF). Additional capabilities exist to view this display at selected locations. Supplemental information displayed will be rain meteorological data, numerical outputs, etc.

Amendment 1 III.A.1.1-6 August 1981

PVNGS ZIIR The satellite TSC has the same display capabilities as the main TSC (see figures III.A.1.2-1 and III.A.1.2-2). An analysis will be performed to determine equipment to be incorporated into the TSC to monitor and display the status of the affected unit.

Display of data at the TSC will be in accordance with NUREG 0696.

The TSC will be habitable to permit occupancy following a loss of coolant accident. Monitoring equipment will be provided for direct and airborne radioactive contaminants to provide warning if radiation levels in the TSC are reaching dangerous levels.

The TSC will have permanently installed radiation monitoring and filtered ventilation systems. The nonredundant systems in conjunction with building shielding will ensure that the com-bined doses from airborne radioactivity and direct radiation meet GDC 19. As-built drawings and other appropriate records will be available in readily retrievable form at the site and will be accessible to the TSC.

The Palo Verde Nuclear Generating Station Emergency Plan will be revised to describe the TSC and its function.

C. Emer enc 0 erations Facilit EOF An Emergency Operations Facility (EOF) for PVNGS has been established and will be fully operational prior to PVNGS Unit 1 operating license. This facility is located outside of the protected area and occupies approximately 6000 sq. ft. of the basement floor in the Administration Annex Building (see figures III.A.1.2-1 and III.A.1.2-4). The EOF provides space for operations. personnel in support of the TSC. Plant personnel will be able to evaluate the magnitude and effect of radioactive August 1981 III.A.1.2-3 Amendment 1

PVNGS LLIR releases from the plant, and recommend appropriate protective measures. The EOF will have ventilation isolation capability and be shielded against direct radiation. The facility is habitable following a loss of coolant accident. A study will be conducted to determine analysis and monitoring equipment required. Display of data at the EOF will be in accordance II with NUREG 0696. Communications with the TSC will be provided.

The EOF provides space for recovery operations following an accident as well as training facilities for the site.

The PVNGS Emergency Plan will be revised to describe the EOF and it's functions.

Implementation of these items will be complete before PVNGS Unit 1 fuel load.

Amendment 1 III.A.1.2-4 August 1981

PVNGS LLIR cooling water system (through the shutdown heat exchanger) can be detected by installed radiation monitoring.

C. CVCS Char in and Letdown S stem The existing design incorporates all-welded piping.

The letdown system is isolated upon CIAS and SIAS.

Relief valves on the system relieve to the equipment drain tank.

The leakage from the CVCS charging pumps (positive displacement pumps) and other system equipment is ALARA as they are hard-piped to drains. The nuclear cooling water system is monitored for potential leakage from the CVCS through the letdown heat exchanger during normal operation.

D. Sam lin S stem The existing design of the normal sampling system incorporates "Swagelok" connections, however, the design will be upgraded to all-welded piping for sections which would come into contact with highly radioactive fluids. The system is isolated upon CIAS and SIAS. Relief valves relieve to the equipment drain tank. Leakage from the system is also mini-mized by the small size of the lines.

The post-accident sampling system will also be constructed of all-welded piping.

August 1981 III.D.1.1-3 Amendment 1

PVNGS LLIR E. Hi h-Pressure In'ection Recirculation HPSI The design incorporates all-welded piping. Relief valves on the system (external to the containment) relieve to the equipment drain tank. The vent and drain lines throughout the system are capped when not in use. Leakage from the HPSI pump seals and system valve stems is as-low-as reasonably achievable.

Miniflow connections to the refueling water tank (RWT) are isolated upon Recirculation Actuation Signal (RAS).

Manual cross over valves to the CVCS are normally locked shut.

F. Waste Gas S stem The waste gas system is isolated from the contain-ment upon CIAS. (The normal vent path from the reactor drain tank (RDT) and the reactor head vent system is isolated.) By design, the introduction of highly radioactive fluids to the syst: em is precluded.

G. Other systems containing radioactive materials which are excluded from this program (1) Radioactive Waste Drain System (RDS):

The RDS is isolated upon CIAS to ensure the lic{uid radwaste system will not become highly contaminated due to post-transient activities.

Amendment 1 III.D.1.1-4 August 1981

PVNGS LLIR (2) Fuel Pool Cooling System (FPC):

The FPC is normally isolated from potentially highly contaminated systems (e.g. SI in recirculation mode) by double, locked shut.

isolation valves.

As part of the system testing program, each of the above systems is hydrostatically tested to 150% normal operating pressure per the requirements of ANSI B31.1, Summer 1976 Addendum for ANSI B31.1 piping systems, and to 125% normal operating pressure per the requirements of ASME Boiler & Pressure Vessel Code, Sec-tion III, 1977 Edition, for ASME piping systems.

Leakage from potentially radioactive systems will be maintained ALARA by testing and maintenance. Required procedures, tests, and administrative controls will be developed prior to fuel load.

III.D.1.1-5

~ ~'

PVNGS LLIR III.D.3.3 IMPROVED INPLANT IODINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS Position (1) Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.

(2) Each applicant for a fuel-loading license to be issued prior to January 1, 1981 shall provide the equipment, training, and procedures necessary to accurately deter-mine the presence of airborne radioiodine in areas within the plant where plant personnel may be present during an accident.

PVNGS Evaluation Prior to fuel load, procedures will be developed for determin-ing airborne iodine concentration. Charcoal cartridges will be used in conjunction with a portable pump or a fixed vacuum system. The cartridges will be removed to the counting labora-tory for gamma spectrum analysis. Procedures will also define ALARA concepts for removal, transport, and analysis.

PVNGS response to this item is included in the evaluation of section II.F.1 requirements.

"August 1981 III.D.3.3-1 Amendment 1

I 0

,l

<4P'Oyg PVNGS LLIR, (g) layout.,of, control room, .air intakes, containment building, and chlorine, or other chemical. storage facility with dimensions: 4

, see,FSAR .Figures 1.2-7 and 6.4-1 (h) control-room sh'fielding including'radiation streaming from.penetrations, doors, ducts; stairways, etc:

shielding study in'rogr'ess (i) automatic isolation 'capability-damper closing time,

" '.idamper leakage'and area:

4 I closing time = 15 sec.

leakage = zero leakage (by bubble test) area = 12.6 sq ft (j) chlorine detectors or toxic gas (local or remote):

redundant in C.R. ventilation outside air intake plenum; smoke detectors alert operators for manual operation (k) self-contained breathing apparatus availability (number):

(l) bottled air supply (hours supply):

y III.D.3.4-3

~;g e4, PVNGS LLIR (n) control-room personnel capacity (normal and emergency):

PVNGS response to items k, 1, and n are undergoing review in light of the personnel manning requirements of section III.A.1.2.

(m) emergency food and potable water supply (how many days and how many people)':

Presently, the PVNGS control room design has emergency food and water supply for 6 people for 7 days (within the closed control room).

The commitment is being reviewed in response to'NUREG 0696 TSC requirements.

(o) potassium iodide drug supply:

This item is being reviewed in response to recent'studies which may significantly reduce the post-accident'iodine source terms.

(3) Onsite storage of chlorine and other hazardous chemicals:

NOTE: No onsite storage of liquid or gaseous chlorine. It is stored as sodium hypochlorite (solid).

(a) total amount and size of container:

hydrogen:

125,000 SCF 5 2200-2450 psi is stored in fourteen steel cylinders III.D.3.4-4

"'T GAPA*

aullll*4 cNP-001 EACTOH V 0 ( D-W )

Oak N 148K .,--a-9 e

018-SC44-4 CN CNP-001 CHARCINC ll)E V24 ~ 090) CP-008 OWE 3 I 0-0&7~ )

i9 Lp

~16 8 8 11166 Wc 8 J. CPC SJ CR 8 /sas j')

Ca OT5 ~95 e

p Fc 082-SC44-I FOA DET*118 T 112C4 G 7 0 CN* El 4 TYPICAL FOII DETAILS SEE llfv296 Sa ES 0 LOOP OI CAAH I

M M Qlo Ql03 PV ZSL L 8/SSE

/sn~

+IS ~lS OTf 3 toov Hr I.

039-8044-)".7 8 (8-8) ~83 e

I

%OF

I 8~OR ca Y---

11 P TO

'0'78-COO*-E PDT-124Z

-~THIS MS (0-<l

+*

80*

+4 884 I

TNS DWO

-s) I TCEONCO ca 101R(

Cooz R SET POINT loop v-loe LOOP 1-101 JsSE. /Sav ~His TDCPCS I 8~v( CR Tl 121CE g~g le T~

Olt-SC44 I 8 0 8 Pol cA 8 TIPIC4L 804 TIPIC41 FOR 050 cc84 ll L ~ujgagu c 4NHEL 8 TYPIc41 121 PLANT PROTECTION

~

OETAIL6. SEE DE T*IL8. SIE 112CS

~(8 4) sEcFDR OET4ILS 10 ever I'DT-IN 024 4

IS CR LO PZR PRESS.

RPS/ESFAO 8 PZR.PRESS.

AI IISS TRIP STETEM (PPE)

YAH IMI Lo PtR PRESS, TRIP RPS/ESF48 GET po EIPASS TNT p

8JPl 02 cR Tj@:: LOOP 0(ACA*

4 103 CCD4 IO 100)'IAG 8.002-8COL- 8' REACTOR l JsSs 8

~25 CN EL A TTPICAI E

24

-W -~TWS0 5 e

(D-*I loop oraca*

8 o I ova-seas-0/

I lolr-8 aa-a 8 0

--"o "" -~rws Iwo I v au Savaotn)PE 8 'E

, P-00'I DR4(H WIK FOA 06 TAILS

\

g)

SC

+~004 Zg, O.l ) 0 SEE 100 OIACRAN lo 8~49 8/SDS Qla 8

I. l 02 7 cs. rara 10181 e 804/~00*

(E-ls) 328(

Z3 l(ZN4 I 1 I 8

'PD II II e~lc R S

5 I 1 0

I 8 CP, RID c cAEt CPCE 0

8 '0 I

Fo I ,

8~MRS SAHPLE EPT 8 I CNkNNE DEG A TVP CA EE 0 0 6-ECA 010-EC*

- ll 008-DC*A- ll ll 0 JillI 8 Ecaa EcEa CNA NEL C TPICAL FOR DET4ILS EICHANCEA NQ 2 ll 8~04 CPC*

1(21I4 ek 4

CHANNEL 8 TYPIC4L 7

FOR SEE lOOP OET4IL8 DI4C448 I

- a.-- I CN N El 8 TYPIC4( FOR OET*IL6 SEE lOOP Ot4CRAN

'~04

.J 8A

+

I 2N 0 8~04 1

+24 'NNTX 3))

-8 OII-DC*4- ll v'-"Y SA '0 IN(

PADTECT ON D 6R 8 SEE IOOP 014CRAN TLP I 2 r @'

la a)

OR v'80 A

"U-NOTE 3+)

P 8 10.

6 1020 8 Fo JKMSV' Ol 02 105 PC

~SIP-002 Uv-864

'd

~

~~ ~~ ~ 56 c r~ 8 ~84 r I.o

'64 0 0 c Zs-scca-li 0 0 0 108 8 ~

~

F0 0 ~ ~I~

1~

<<I~ ~i I~ ANP- -CO

'6 00 ~ TO E Viol PS -Nl- 8 PT S

~PI cR V208 CHANNEL 4 TYPICAL 804 DETAILS SEE oar or 084 8~ ~l CEONCE I 6: 8 FO 8 Fo PSL 543 r -88 ~sip-002 (C-81 uv-Ora ~3)a-002 ~1060 C

t. IL l ( 9FO 8 01 e

1990 8

8 (S ca IS Ca 1278 701 8 0 (TTX ll(X I om-Seas-ZA'-.

Q 1

~SIP 002 -s a u ~9(a-002

~f-8) ' 1207 I T uv-OM ~9(a-00 0 UY-851 ~SI -OOZ IC-5 I

---

W-----N ,

T 8~OR

~lD FO LT Co IC 16) 8'c at ra ca I-NTQ 4 8 8 8 LT 1103 CO 101 ~00 ur-Ezv( ~ tv-OOK- TI (CGX)

Tol'

• Uv-002 ~s( 8-00 PPS 8 <<02) cs*-l 8 eI ~lo ,

eMUEcEASG l IC-10 )

A 072 CCOA-3 CHAHNEL 4 I ICAL 8~

I(OX-I s

'b -C3 UV-080 ~SIP-Do (0-111

( D. 4 t FOR DET41 8 SEE L04P 0140RI N

(.

LC MM

~PO IN(0 DNC TI-(11 ro secs E ecaj IPO 0

I -, -I I

8 'cN GR 8 10 2

~HS J DET I(2 LOOP SEE CA V294

. 64 5 D C LOOP IS 01 L,EOO 0C oas-Eckaar Iaaf 04 lla clEll aEF C MTf I ALL G M

SKAT OEE 4G HAVE FU fS SHOWN lu DZR T 2 S

(~-4 / l )0+)g"', 0 QLJ' (PO-8 PROPDRTI0 MIS (0 10)

D 2(4 I DIC*101 CONNECTIDH M*:- ....,.

SHOWN AITHISDRA 6 U LEIS 8,

NEk f CtHI HOL lit LOOP I 0(A 53D I 0 ERW IS IND MTED 8 scaat1cDA 40*CTOR SD 6 04 I 4NK 3. S LIED IPIHG DES 8

.07 -

le c PAD)'OAT(0 *I E*TER LC Ca D ll IS CR V2 0

-2 cua-ooa D 15)

~ 12 I I',:DR Ea 25DO PS 4

• D 6 5D) 4ND CDDf AEOVIREUENTS DF 01 T ~ IS (NO (8 I(0, /0 ) POV.IS ASUE SECT 0 IXI *SSL Ca 45-114-1 8~04 THDMS 2, LOOP 14  : /C3) / 8 g lb'(~TH(s Dw0 (7-5)

THIS P ~ ID SUPPLIED P110 E-14273-0 0-IS REDO*

P C6 FRI 1 E CE 04* INC cR 8 8 cR J=-----4-------L---I---L-------2--- -l.----- (CFG)

ST-(85 F

o89-cce*-li e 8 0 EVISIO 0.02 LL

~ca-002

@~CCN ST4HODI 044RCIHC HP F - 0 0 0 22 PV IP I 8 CH4HNEI TYPIC4L FOR LODP 014044N THD) 4 Sl

-~THIS DWS

( e - \I El*CHAN((ZCA

~ '(0,~38o (8- 8) 9 ,

I C SY

¹ ca VU CN4HHEL Tr ICAL I'IC /

C'ET4IL8 TYPIC4L AORS~CD FDA DET*ILS SEE

.--co -<O 10 fe.-.,

a aaacloa DA*(N T4HK cua-ooa ,,'8 M 55 lOOP SEE LTT \

yACFC CPC Ig 4

LDOP 01*CA*N l4-'-6 .) T3-M-RCP-OOT REV 2

(

8 PR 100 OR S~C P8HL

~IOO

'a C3 SSP-ODI S WLE HE4T EKCI4NCEE N4. 8 I

08.3) ~ T il ATE 11204

(~@..-oL 122CD 0

I 1 8 8 -

0 (2

f3)

SE I.D.

8 0

0- 0 Fo

--.='-.="-:M Palo Verde Nuclear Generating Station

'---CO 0(~4018 110 .Dip 028-8044-12 DR CHP-003

'E4CTOA REACTOR COOLANT SYSTEM Figure 2.1.11-2 l1 la~< wZL- a~

1(IIIL)ART

~ FRAY CHP 00)

Vi ~ I ACACTIN Dill)N IAWt

-SCIA- ~ CHP-Dol NAICINC LIN

~;-

0 CII PY IO E/P PV IOOE SN ISN 1

0

~ 0 I

TIP~

TIPICAL 0

I T~

I AI tart CC I

7 CE 17 0 ~0 0

'o-'-

CP V ~ 10 I DSTAII~

~ ~ LOOP 1

W,. ~,

CII D I ACIIAH I TTPICII

'gj~j CNRNNEL

~~a'II-SC*1-I I ~ Oll 0\TIIIC SEE LOCI CIICIIH O.,.-CO I ~ I ~

C ALVISS PY I

ZSL L WII QI ~

2 I s.v lOOF IIII~

I -CC ~ A-(0

-- CII I 0 II-SC*A-I T

AHP TC ICSDHC ~

~THIS T IR ~

lc-e)

L 0

Re(

~0 ~ ~00

• (I-et I 1

iV*4. 40T II ~ SIESFAS ALC pzi

~OINT SYPASC

~ I)ESC. '(RIP CVPASS PLANT PROTECTION STSTEN (PPS)

VAI 4L I~ T POINT ISO~PIE I'l)01 ~ H TIIIP Ci LOOP I-IDS TTPICAL FCI OCTA)IS ~ E ~

I041 0)iCIIAH LOOP I-)07 TYPIC*L.FOI DETAIL 4 4 LOOP D)111(HI

'I YI(0 ~IIII 010 CCIA S 0

L

~

ll HjS

(

CH

)

CH*I)NEL I

~Oll OCTP(l ~

IVIIC*l Lt((0 A V207 0 0 ci 0 0 lz TPOINTT'PS/EIFAI Lo PII.PIE04. Tll(P II~ IIHFA~ SVPASI 1 ID I T ~ ~TWG 0 VAR. I~ I-SCDA-)0 NEACTta CNAHNIL a vvNcal Ik DWG 0 '

DIIIIN TkNK FDi DETAIL4 TT IIIN4

~ D ~ EE LOOP OIACI*0 0A Oli (0

I S ~5)P.COI 3

0~II I I

0~01

~HIS I

DN I

I 0 CII 1 1( Pzi PR ~ 2. IRIIj ~ CAP~CIa 0 TC (EQ4) L 1-6 D (T

".9 (a I'a PSL 5R5

~IIP 002 I

0 0 po~ r

~ c*

~oa 0 j. caF

-~rwsoas ~

A ~to.o)

.I

~4 (0@sr I CHANNEL I TYPICki RW CCTIN)ISVE LCOP Iuiats CNANNEL C TYPICII FOR OCTAIIS

.0%~)

Sl)NFL ~ NEAT

~ llC)IIINCEII NC 2 44I'-401 0

THIS OWS tc. s (INTE I IKIB I ro QA TC 70 CPCC IE CHANNEI H'PC*

I TYPICAL Foll DETAIL 0 SE ~ lOOP OIACRA)I HJCC CNANNCL I TIPICAL Fol OET*ILC SEE lOOP 0(ACSAN f TT

~

0CII5CR Pa (

.-ra~:*"

NDTE 5)I ~

AA ~ E ~ LOOP It*CIRC l(1 l2 IIP~I 6 02 I C IS SCCA Sl t

0 4'1 o 0 0 o

0 ~ IP-002 0AD~I-(Il o 0 Uv-4 ~ 1 0

0

)WTK I) "

e-cc ACS sg

,;;L IA@

-CO 0 I~ Uk CII 0-T~ I I ~ ~ 0 (mrE EI Ggt f

'6 10 Fo YII~ F4 1 VIDE CN4NHKI I QFO -: ~tl P

IP-0 I J TTPICIIL FCI OET*li~ SE ~

0

~2 0 Tlo I 0(AQRFH0~

FD IE.II UV-44Q ~SI ~ -002 0 C)1 0 S o 0 roro-ccak. ~ '(3-IC-I) UV SCC (Es) ooz lo 1270 O.- S D

~XT 0

A(p PSL YQ5 I ~ A 0 ca CR V20C LVPL V201 \

~2(P.GDR -- VV-I 1 ~SIP-002 0 TDL

~a.I 1200 VIDI PP4 UV ~ 6)

(C- ~

~0( ~ -002 IC- ~ I I

0 CI 0 PT TE FC 0 PT LT UV"Ilk~SIP-00

~ID I (~f5] i l+70 Rtls 0 iC>32 i 0 0 E UV-I~ C ~0(P-002 --5 f

)0 Uv 0$ , ~2)P-00 (C~l I 0-5 I TDI 0 cR 0@Ill

~III'-DO 9

0 Pl 11-002 P

I C~ I-0', IC-lo I I ~. 7 . CSA-1(l" CN*NNEL A TYPICAL 10 It UV-4CS A ~SIP-00 0 0 Foll DETAILS 4E ~

IC III LCDP OIACA40 I-~TN!I Tl-IIII To 4IC ~ I CH*RCINC PU Nps J-~g DIIC HDTE.

1

~SIR DETAILS 4~ I Ii 0)LSSD

( IIL CL04E ALVES SHOWN IN THIS I

IH 1)%. ),

9 ~TV)BORS LOOP 0 I 4 C It 4)I VIIH LOOP 10 LOOP ICASPcoa cii-MIRE ~ UE\INc LEVEI Ss CRAW)HI SEAT RIVI FLOW UNDEI THE tF-N Vzl ~ IN)t)CATCI) 2. THE BIFfTY TRI(H DE ~ GHATOR I Q

~(Ag'"~-JL ~THIS DWS COHNECTIDH 4 )0 PRECEDE ALL LINE UNCERS EATCII 0 ,g.)S-QLB Clk SZO SNO I W T I ~ takW(NI UHLKSS Ill PDT 0 OTIIERW)SE )ND CITKD 0

0 TH)ID I sck~coa icacroo 1 Qo THIS CWC Ol(4)N Tt(NK '1 Cf SUPPLIED PIANO DES)CHID II 0 1-1) I )15 H.ort-t(CCI-I CNP 001 I FOI ~ Do PS)A U(D4 ~ Dl kl(D 100poisoHal Hcitoo 01, -~TS PER CODE RIOU RENEIIT4 OF "T,;-;~ ~ IE41 ~ A4

'W OH) 3 Y2)0 (D 15 D 2~ kSN1 SECTION Df, C kSS I CII 0 Ci CI HS (10 I CHP

~~(0 001 HS J=

N H N I 8AI ll51

~I)0 illA 'LDZI, -

(W(

WC I (VHG) lOOP )i  : jo>osc-ccc*-S'

~THIS CW4 (F-I) 0 1 1 I TN)4 PI)0 IS IEOIAIIN Pilot) THE C ~

SUPPL I ~ D P ~ lo PVI(CS 004V INC

~ -11272 110 I 1(L IEVI0 I ON NC D2 j IIJ 5 cvaHoev CHAACINC PUHP a f-~TH)S OW 5 1~00 f EP (I-It 0

~HID SW G

~CNP 001 LL OR INI NODE f I~ E 0.8) Jsas STANCSY CN))RCINC PUI(P CHIL TD Rl)S 1 0 Ei~P DSTAILS LOOP 045>> ~(0

~CHP-002 OI)ICII*ll Ci APERY J IEPCTOR DR)IIN TANK CHP-CDS (I-IS)

CT 55 CNAN)I ~ L4 TTP)c4L FDIIS 0 ~ T(IILS SEE LTTJ LOOP DIAGRA)I I 2 0

V' 0

cll I

T, J,,

TO CPC I I

I '@I sEK I oCV QTIT 0

,'l

~0 ~

Lkko, isis

~ II CHAN()EI I TYPICAL FOR DETAILS.S ~ E LOOP 0 I AC((41

~

'I 13-M-RCP-OOT REV 2 T'"-~'~.

1 I

0 OZS-SCCA-S('slo 4*itl~ NEAT I

0) QFd J 2

> '-;;ol ((NDTf I) 40 CIA 014 ))

~ I(CHANGER Ho.

'4 i

'.5 S I l2

-4CA*-\I" I ~ ACIDR cHP 000 00*IN TINK

( (E-IC)~~

REACTOR COOLANT SYSTEtt Figure 2.1.11-2