ML20097E667
| ML20097E667 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/14/1984 |
| From: | Douglas R, Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| IEB-81-03, IEB-81-3, NUDOCS 8409180269 | |
| Download: ML20097E667 (64) | |
Text
Public Serwce Electric and Gas Commm 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation September 14, 1984 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:
HOPE CREEK GENERATING STATION DOCKET NO. 50-354 DRAFT SAFETY EVALUATION REPORT OPEN ITEM STATUS Attachment 1 is a current list which provides a status of the open items identified in Section 1.7 of the Draft Safety Evaluation Report (SER). Items identified as " complete" are those for which PSE&G has provided responses and no confir-mation of Lcatus has been received from the staff. We will consider these items closed unless notified otherwise. In order to permit timely resolution of items identified as
" complete" which may not be resolved to the staff's satis-faction, please provide a specific description of the issue which remains to be resolved.
Attachment 2 is a current list which identifies Draft SER Sections not yet provided.
Enclosed for your review and approval (see Attachment 4) are the resolutions to the Draft SER open items listed in Attachment 3.
In addition, enclosed (see Attachment 5), is revised supple-mentary information to FSAR Section 13.4. This information supercedes the proposed HCGS Technical Specifications transmitted on September 13, 1984.
Also, enclosed (see Attachment 6), is Revision 1 to the response to IE Bulletin 81-03 (supercedes 8/12/84 submittal) as requested by-the Auxiliary System Branch.
The Energy Pennie 8409180269 840914 PDR ADOCK 05000354 -
E PM
Director:of Nuclear Reactor Regulation 2 9/14/84 A s'igned original of the required affidavit is.provided to document.the submittal of these items.
Should youi.h' ave any questions or require any additional information on these open items, please contact us.
Very truly yours, n .
ff-en 'b f Attachments / Enclosure C D. H. Wagner. _
USNRC Licensing Project Manager (w/ attach.)
W. H. Bateman ,
.USNRC Senior Resident Inspector (w/ attach.)
FB18 1/2 d
i 1
" N~ ,-r --,3 -v,~wwwe+c-e-- +-- y- ----wg-c,- -rw-,---+-w ,-e v e t w+w * ,y&*r---++.ye,----t-- ***7w-----err e= + + , g -v vw--~w--y----+-wt-
v- -
UNITED STATES OF AMERICA NUCLEAR-REGULATORY COMMISSION DOCKET NO. 50-354 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Public Service Electric and Gas Company hereby submits the enclosed responses to DSER open items and revised NRC requests for additional information for the Hope Creek Generating Station.
The matters set forth in this submittal are true to the best of my knowledge, information, and belief.
Respectfully submitted, Public Service Electric and Gas Company By: ~
x Thomas J. p rtin' Vice President -
Engineering and Construction Sworn to and subscribed before me, a Notary Public of New Jersey, this /'/ C! day of September 1984.
{ AlW -
/Af DAVID K. BURD NOTARYPUBUC 0F NEW JER$EY
- My Comm. Empires 10-23 85 MC 28 02 L
l IRTE: 9 / 1 4 / 84 AfDOBBFT 1 DSER R. L. M1TIL TO A. SOH G EER GWI SETIG6 LETIER DREED SUIDECT STATUS ITIBI Nt3HER l 1 2.3.1 DesigrHasis tesgeratures fw safety- Caplete 8/15/84 l related auxiliary syntes 2.3.3 Accuracies d meteorological Complete 8/15/84 l 2a (Rev.1) l measurements ;
2.3.3 Acoaracies of meteorological Caplete 8/15/94 2b (Rev. 1) maapurements 2.3.3 Accuracies d Estocrological Complete 8/15/84 2c (Rev. 2) '
measurements 2d 2.3.3 Accuracies of meteorological Caplete 8/15/84 measurements (Rev. 2) 2.3.3 Upgrading of onsite meteorological Cas@lete 8/15/84
, 3a (Rev. 2) measurements program (III.A.2) 2.3.3 Upgrading d onsite meteorological Ccaplete 8/15/94 l 3b (Rev. 2) measurements grogrant (III.A.2) 2.3.3 Upgrading d onsite meteorological MtC Action i 3c measurements progem (III.A.2i .
4 2.4.2.2 Ponding levels Caplete 8/03/84 2.4.5 Wave ispect and rurup on service Couplete gf13fg4 ,
Sa water intake structure (Rev. 3) 2.4.5 Wave ingect and runup on service Couplete 9/13fg4 Sb water intake structure (Rev. 3)
Wave impact ard rurup on service Ccaplete 7/27/84 Sc 2.4.5 water intake structure 2.4.5 Wave ingeet and runup on service Canplate 9/13/84.
5d
- water intake structure (Rev. 3) 2.4.10 Stability d erosion r A tion Complets 8/20/94 6a structures 2.4.10 stability d erosion grotection Couplete 8/20/84 l 6b
' structures i
2.4.10 Stability d erosion protection Ccaglete 8/03/84 i 6c structures M P84 80/121-g's j i i
e
- _ - . - ~ . . - , . . . - . , _ , . _ , - . , _ _ . _ . _ _ - _ . _ . . _ _ _ _ , _ _ . . . . _ , . _ - _ . . - ..--n----
Jur20Batt 1 (Cent'd) cent R. L. MmL m WWI SECTEM A. SM ERJElT 5501B MFMR IMEIS ITWE IDWR 2.4.11.2 Thesumi aspects of ultimate heat sink. Omplete W3/84 7a 7b 2.4.11.2 h aspects d ultimate best sink Quplete W3/84 8 2.5.2.2 Omico d ==d== earthganks flor New omplete #1VB4 England - Piedmont Tectonic Province 2.5.4 Soil damping values Ozplete V1/54 9
2.5.4 roundatim level response spectra complets 6/1/34 10 2.5.4 Soil shear moduli variation complets 6/1/34 11 2.5.4 Ocubinatim d soil layne properties complets 6/1/94 12 2.5.4 Lab test shear =
- 11 v=1 nam Ctuplets 6/1/84 13 2.5.4 Liquefaction analysis d river bottom Ozplete 6/1/94 14 sands 2.5.4 Tabulations of shear moduli Caplats 6/1/84 15 2.5.4 Drying and wetting offact on Complete 6/1/94 16 Vinconto m 6/1/84 Pouer block settlement monitoring Complets 17 2.5.4 2.5.4 Maxisua earth at rest pressure ocuplets 6/1/84 18 coefficient 2.5.4 Liquefaction analysis for service complete 6/1/84 19 -
unter piping 2.5.4 Explanation of cheerved pcuer block Complets 6/1/34 20
' settlement 2.5.4 Service u ter pipe settlement records Quplete 6/1/84 21 2.5.4 Cofferdan stability caplete 6/1/84 22 6seesy122-es
J!r ntO8 8 1 1 (Cost'd)
D88R L L. MmL M A. SOBBICER WW SENEN 175G15 IEmR tRIED 11 31 IEW W R SBJE2 34 3.6.2 tkuustrained whipping pipe inside C7 =to 1
7/18 /44 containment 3.6.2 ISI progrus for pipe welds in Ccup3ste E/29/84 35 huesk esclusion sons 3.6.2 Postulated pipe rupt: tires ccuplets 6/29/54 36 B d tar isolation check valve C a plete S/20/84 37 3.6.2 operability 3.6.2 Design cf pipe rupture restraints ccuplete 8/20/84 38 3.7.2.3 Sst analysis results using finite c e late 8/3/84 39 element method and elastic half-space appranch for omitainment structure 3.7.2.3 SSE analysis results using finite Ccaplete 4/3/84 40
' element method ani elastic half-spam aggranch for intake structure Steel containment buckling analysis Ccaplete 6/1/84
- 41 3.8.2 .
Steel contairument ultimate capacity Complete 8/20/84
! 42 3.8.2 (ase. 1) analysia
(
3.8.2 SitV/tDCA pool dynande .loada Cceplete 6/1/84 l 43 3.8.3 'ACI 349 deviations for internal Ccaplete 6/1/84 44 structures ACI 349 deviations for Category I Ccaplete 8/20/84 45 3.8.4 (Rev. 1) '
structures ACI 349 deviations for foundations Ccaplete 8/20/84 46 3.8.5 (ase.1) ,
same ant response spectra Complete 8/10/84 l 47 3.8.6 (asv. 1) nocking time histories Couplets 8/20/84 48 3.8.6 (aer. 1)
I l
M see 80/12 4 - go l -
JurD OSWff 1 (Cont'd) 55 R. L. NtrE. E WEI E!Im A. StBGEIR IRBWR stb M T ouuds Wr!ER 13t!ED FIDI 49 3.8.6 (kons concrete section Complete VE/84 (Rev.1) 50 3.8.6 vertical ficce flazibility response Completo 4/24/84 spectra (nor.1) 51 3.8.6 Ctaparison of Bechtel independent c e late 8/20/94 verification results with the desigt- (Rev. 2) l basis results 52 3.8.6 Ductility ratics das to pipe bred Ccaplate 8/.V84 53 3.8.6 Design d seismic Category I tanks Ccapiste 8/2Q/84 (Rev. 1) 54 3.8.6 Combination d vertical responses C e late 8/10/84 (Rev. 1) 55 3.8.6 Tersional stiffness calculation Couplete- 6/1/84 56 3.8.6 Drywell stick model dyrsicyment Ccmplete 8/20/84
' (Rev. 1)
~
57 3.8.6 Rotational tism history iguts Couplete 6/1/84
( 58 3.8.6 "0" reference point for auxiliary Camplete 6/1/84 building model 59 3.8.6 overturning soment d reactor Complets 8/20/84 building foundation est (Rev. 1) 60 3.8.6 MBAP element size limitations Complete 8/20/84 (Rev. 1) l 61 3.8.6 Seisade modeling d drywell shield C g lets 6/1/84 tell 62 3.8.6 Drywell shield well boundary Camplate 6/1/84 conditions 63 3.8.6 Reacter building dess boundary Ccaplate 4/1/84 conditlens !
N 986 80/12 5 - gs 1
- l
. 1 JurDQ8Sff1(Cont'd) c8st R. L. nrrE.10 SElT3tX A.StBSE3R i set '
STR!tB IE r M tutI E I IIBI IEEWR ' SERIE2 3.8.6 551 analysie 12 as oatoff frequency Ca plete S/21/N l 64 (Rev. 1) l l
3.8.6 Intahe structuse crans honey Iomd capiste 4/1/84 l 65 l
49 3.8.6 nupadance analysis for the irtake Caplete 8/10/84 ;
66 (Rev. 1) 1 structure 3.8.6 Critical loads calculation fer Complets 6/1/84 67 reactor building ene 3.8.6 Reactor building foundation mat Camp 1ste 6/1/N 68
- contact pressaares 69 3.8.6 Factors d safety against sliding and Ccuplete 6/1/N overturning of drywell shield wall 3.8.6 Seismic shear force distribution in Complete -
6/1/84 70 cylinder well overturning of cylinder wall Ccuplete 6/1/ 84
.71 3.8.6 ,
3.8.6 Deep basa design of fuel pool walls Ccuplete 6/1/84 72 3.8.6 ASHSD d: sus adel load iguts C e late 6/1/84 73 Tornado depressarization Complete 6/1/ 84 74 3.8.6 3.8.6 Auxiliary building abncenal pressure c e 1 ate 6/1/84 75 .
Tamert.ial shear stresses in &jwell Complete 6/1/84 76 3.8.6 shield wall and the cylinder wall Facter d safety.against overturning Cay 1ste 8/20/84 77 3.8.6 (Rev.1) d intake structure 3.8.6 Dead Zoed amiculations Complete 6/1/84 78 Post-1 modification solande loads for c e late 8/20/84 79 3.8.6 the torum (Rev.1) m 884 80/12 6 - es
l
. l
)
l gBGeerr1(Cont'd) ,
- n. L. u mr.10 nemt A. SM WW M3EE sIntes tatut amo rise use8R ar*M stmus fluid-structure interactions casdate 4/1/84 so 3.8.6 I i
seimic displaossent d torum Cceplate 4/20/M l 81 3.8.6 (Rev. 1) i Review d seismic Category I task Couplets 4/24/84 3.8.6 l 82 (mer.1) desik Factors d asfety for &yumil Caplate 4/1/N 83 . 3.8.6 hw*11rgi evaluatice i
3.8.6 Ultimets capacity d contaiment C g ista 8/2Q/54 i 84 '
(aer. 1)
' (astarials) ,
tend cabination consistency Ctsplate (/1/84 ;
! 85 3.8.6 caputer code validation Ctaplate S/20/N 86 3.9.1 Information on transients Caplete 8/20/84 87 3.9.1 I 3.9.1 Stress analysis and elastic-plastic Ca plete , 6/3/N 88 analysis Vibration lavals for IESS piping Complets 6/29/84 89 3.9.2.1 l systems-Vibration monitoring program during Complete 7/18/84 90 3.9.2.1 '
testing Piping supports and anchors Completa '6/29/34 91 3.9.2.2 Triple flus & hand contalment Complets 4/15/84 92' 3.9.2.2 '
penetrations 3.9.3.1 tend cabinations and alleueble Ctsplate 4/3/N ;
93 stress limits 6/3/84 Desiqpi d SINS and SRV discharge Cesplate 94 3.9.3.2 pipin0 _ ,
1 i
i M 704 80/12 7 - go r
-~~w-em- -
. m eg -
Im R. L. IETIL M M3EN A. Sm WW Starts ggrvum targo IM # = ""T _
' Fatigue evaluation a SRW piping Couplate 6/15/84 95 3.8.3.1 and ICCh damataners 1
3.9.3.3 IR IMemantian Isotico SM Ospists 4/3Q/84 96 ' (Ber. 1)
Buckling criteria used for otsponent Complete 6/29/84 97 3.9.3.3 apports i
Desip d bolts Ccaglate 6/15/84 C3 3.9.3.3 ,
Stress categories ant limits for. Campista 6/15/84 99s 3.9.5 core support structures Stress categcries and limits for Ccuplete 6/15/84 99b 3.9.5 I core agport structures 10CFIt50.55s peregraph (g) Ccuplate ( /29/1 4 100s 3.9.6 10Cf1t50.55s paragraph (g) Complets 9/12/8 100b 3.9.6 (Rev. 1)
PSI and ISI prograss for pumps and Casplate 9/12/84 101 3.9.6 (Rev. 1) valveg taak testire d presmre isolation Ccuplete b/12/84 102 3.9.6 (a.y. 13 valves Seismic ant dynamic qualification of Complets 8/20/84 103a1 3.10 machenical an$ electrical equipment l
Seismic and dynamic qualification d . Ccuplate 8/20/84 103s2 3.10 mechanical ans electrical equissent Seimic and dynamic qualification of Ccaplate S/20/84 103a3 3.10 nochenical ans electrical equipment Seismic and dynastic cpalification d Casplate 0/20/84 103s4 3.10 mechanical ans electrical equipment a ret se/13 e - es
---.e-ee me,-wr-- ----_ c--...,_. _ w ,, - -
l arncoeur 1 (ent'd)
DEER LL.NEHLf A.9tsmumn:
WW SECTEN ,
SEREM IEITER STIED _l FIBE MESER S R7ECT 3.10 seismic and dynamic gaalificat!ca of couplete 4/20/84 i 103a5 anchanical and electrical eqpipment seisude and dynande gaelificatim of complete 4/24/58 l 103a5 3.10 - l mechanical and electrial ==ir ?
seismic and dynamic gaalification of complete. 4/20/54
) 103a7 3.10
' sectanical and electrical eqpigment '
Seismic and dynamic gaalifi&Jm of cceplete 4/20/54 103h1 3.10 mechanical and electrical equipent Seismic and dynmaic gaalification d coq 1ste 4/20/84 103h2 3.10 mechanical and electrical agaipment ,
seismic and dynamic gaalificatics of - Complete 4/20/84 103h3 3.10 mechanical and electrical equignant .
Seismic and dynmaic qualification of Ccaplete' 8/20/84 103b4 3.10 mechanical and electrical equipment Seimaic and dynamic qualification of complete 8/20/84N 103b5 3.10 mechanical and electrical equipment Seismic and dynamic gaalificatics of Complets 8/20/84 103b6 3.10 mechariical and electrical equipment Seismic and dynamic qualification of ccuplete 8/20/84 103c1 3.10 anchanical and electrical equipment Seismic and dynande gaalification of ccuplete x8/20/54 103c2 3.10 mechanical and electrical equipment Seismic and dynamic gaalificat.icn of complete 8/20/54 103c3 3.10 marhanical and electrical equipent ;
r seismic and dynamic qualification of Ccuplets 8/20/84 103c4 3.10 mechanical and electrical equipment 3.11 awironmental gaalification of, leC Action 104 mechanical and eleci:rical equipment k
l a see se/12 9 - es ..
?
.p.__ . _ _ _ _ _ _ _ _ _ _ , _ _ _ . _ _ _ , _ . . . . _ _ . . . . , _ . - _,-__,_,..,.m.,_____,._._ , _ _-,. -- _ ._ _ -
l l
~
NFSOBBIF 1 (Omt'd) it. t IsmL m ,
m,," - ~ _ cast A. S M i gyg muernms .
M" 5555 I2r!ER IRIE l . ITut- _ IOMR _
s Plant-specific undlanical fracturing Couplete WE/S$
105 4.2 (Rev. 1) analysia s
Applicatdlity d meisade andd LOCA Omplate V2W54 106 4.2 (msr.1) l Iceding evaluatica i
Minizial post-irradiation fuel Caspiste V29/84 l ~197 4.2 surveillance progran l
Gadolina thermal cordactivity Ccaplate 6/29/84 l08 ~- 4.2 equation L
' DEI-2 Item II.F.2 Couplete 8/20/M 109a; 4.4.7 t
7
' DEI-2 Item II.F.2 Quplete V20/M 109b 4.4.7 W30/N Functional desip cf reactivity Complete ODa 4.6 (now. 1)
L control systems
)
Functional desip cf reactivity C<mplete 8/30/84 11 2 4.6 (Rev. 1) control systans ,
Preservice inspection gregram C W ate 6/29/84 r l111a 5.2.4.3
' (ccuponents within reactor pressure boundary)
Preservice inspection progran Ccaplete 6/29/84 111b 5.2.4.3 l
(ccuponents within reactor pressure boundary)
- l. '
l Preservice inspection program Complete 6/29/84 Illd 5.2.4.3 (components within reactor gressure I boundary)
! Reactor coolant pressure boundary Campists W30/84 l 112a 5.2.5 (Rev.1)
Isakage detection Reactor ecolant pressare boundary Couplete 8/30/84 112b 5.2.5 (Rev. 1) leakage detection s -
s '
O M W 10 p p
. Jm20 start 1 (cont'd)
DSER R. I IETIL WSI SECTIGI A. SOBBBCER ITEN IDIBER SIR 7ECT SUGUS IJIT rt R I m rto 112c 5.2.5 naar+ar coolant pressure boundary Ctzsplete W30/84 leakage detection (Rev. 1) 1128 5.2.5 Reacta coolant pressure M d ry-Ozplets 8/3Q/84 Imakage detection. (Rev. 1) 112e 5.2.5 apacter coolant pressure boundary Omplets 8/30/84 leakage detection (Rev.1) 113 5.3.4 GE pr M =1 applicability Complete 7/lVB4 114 5.3.4 Compliance witin NB 2360 of the Summer Omplete 7/18/84 1972 Addenda to the 1971 A! Bit Code 115 5.3.4 Drop weight and Charpy v-notch tests Omplete 9/5/84 for closure flange anterials (Rev. 1) 116 5.3.4 Omrpy v-notch test data fx base Omplete 7/18/84 materials as used in shell course No.1 117 5.3.4 ozpliance with NB 2332 of Winter 1972 Omplete 8/20/84 Addenda of the A! DIE 03de 118 5.3.4 Imad factors and neutron fluence fm Ccaplets 8/20/84 surveillance ?!les 119 6.2 1MI itan II.E.4.1 Omplete 6/29/84 120a 6.2 1MI Item II.E.4.2 Omplete 8/20/84 120b 6.2 1MI Item II.E.4.2 Omplete 8/20/84 l
121 6.2.1.3.3 Use of MJRIG-0588 Omplete 7/27/84 122 6.2.1.3.3 Tmuperature 3rofilm Couplete 7/27/84 i
i 123 6.2.1.4 Butterfly valve operation (post Omplete 6/29/84 accident) l M 384 80/12 11'- gs
AE'IROBBIF 1 (Cont'd)
R. L. MmL m DSER A. -
WWI SBCFICN IE3 W R SR7BCT 5555 IEr!ER IMEIED I2BE 6.2.1.5.1 Ere shield annulus analysia Ozq4sts 4/20/94 124a (Rev.1) 6.2.1.5.1 Ef7 shield annulus analysis Oceplets 4/24/04 12e (Ber.1) ocuplete 8/20/84 l 224c 6.2.1.5.1 154 b old annulus analysis (Bew.1) 6.2.1.5.2 Design dryus11 head differential Oumplets 6/15/54 125 pressure modundant position indicaters for ocuplets 8/20/84 126a 6.2.1.6 vacuim breakers (and control rocia alanus) 6.2.1.6 andundant position indicators for O g late 4/20/84
- 126b vacuum breakers (and control rocsi alarms)
Operability testing of vacuum breakers Ocgletin 8/20/84 127 6.2.1.6 (Rev. 1) complete 7/27/84 128 6.2.2 Air.i h tien Insulation ingestien Casplete 6/1/84 129 6.2.2 Potential bypass leakage paths Complete gf13fg4 130 6.2.3 131 6.2.3 Administration of secxmdary contain- Complete NN '
asnt genings 132 6.2.4 Containment isolation review W 1ste 6/15/54 omitainment surgs systen ocuplets 4/24/54 133e 4.2.4.1 Contairument surge systasi Ccuplete 8/20/84 f 133b 6.2.4.1 Containment purgs syntant Ocuplete 8/20/54 133c 6.2.4.1 n see se/12 12e os i - - - -- .- - _
ATD O M NF 1 (Cent'd)
R. L. MITIL '10 DSER A. SO M N2 R
@EN SECTEM SURTBCF SDGUS IETI1llR IWrED ITEM NLMIER 6.2.6 Containmet leakage testing Complete 6/15/84 134 6.3.3 IPCs and IPCI injection valve Complete 8/20/84 1 35 interlocks ,
6.3.5 Plant-specific IDCA (see Section Ccaplete 8/20/84 l 136 (Rev. 1) 15 .9'.13) 6.4 Ccmtrol roaa habitability Complete 8/20/84 137a 6.4 Centrol room habit' ability Ccaplete 8/20/84 137b 6.4 control team habitability Complete 8/20/84 137c 6.6 Preservice inspection program for Complete 6/29/84 138 Class 2 and 3 ccaponents 6.7 MSIV leakage control system Ccaplete 6/29/84 139 9.1.2 Spent fuel pool storage Ccaplete 9/7/84 140a (Rev. 2) 9.1.2 Spent fuel pool storage Canplete 9/7/84 140b. (Rev. 2) 9.1.2. Spent fuel pool storage Ccaplete 9/7/84 140c -
(Rev. 2) 9.1.2 Spent fuel pool storage Canplete 9/7/84 140d ,Rev.
( 2)
Spent fuel cooling and cimarup Couplets 8/30/84 141a 9.1.3 (Rev. 1) system 9.1.3 Spent fuel cooling and clearup Ccaplete 8/30/84 141b (Rev. 1) system Spent fuel pool cooling and clearup Ccaplete 8/30/84 141c 9.1.3 (Rev. 1) systen M P84 80/1213 - gs e
,,,v-,,v,w,mmme--o- ,-w,~,-~
ATUlOBENT 1 (Cent'd)
DSER R. L. M1TIL 10 SECTICM A. SOMW3R OPS 6 NLMIER SUETECT STAlfjS IEITER OfLG r111M 141d 9.1.3 Spent fuel pool cooling and champ Ccuplete 8/30/84 syste (Rev. 1) 141e 9.1.3 Spent fuel pool cooling ard cleamp Couplets 8/30/84 systemt (Rev. 1) i 141f 9.1.3 Spent fuel pool cooling and cleamp Couplete 8/30/84 systemt (Rev. 1) 141g 9.1.3 Spent fuel pool cooling and cleamp Complets 8/30/84 system - (Rev. 1) 142a 9.1.4 Light load handling systant (related Complete 8/15/84 to refueling) (Rev. 1) 142b 9.1.4 Light load handling system (related Ccsplete 8/15/84 to refueling) (Rev. 1) 143a 9.1.5 Overhead heavy load handling Complete 9/7/84 143b 9.1.5 overhead heavy Iced handling complete 9/13/84 144a 9.2.1 Station service water systen Complete 8/15/84 l (Rev. 1) l 144b 9.2.1 Station service water systant Ccsplete 8/15/84 (Rev. 1) 144c 9.2.1 Station service water systent Complete 8/15/84 (Rev. 1) 145 9.2.2 ISI progrant and functional testing closed 6/15/84 l
l of safety and turbine auxiliaries (5/30/84-cooling systems Aux.Sys.Mtg.)
l 146 9.2.6 Switches and wiring associated with Closed 6/15/84 HPCI/RCIC torus suction (5/30/84-Aux.Sys.Mtg.)
l l
l M P64 80/12 14 - go a
A1130 pelt L(cofd)
R. L. IEETE. W DSM '
W2Im A. N WE 5555 IEr1BR IME Im IONR 8tB352 9.3.1 cagressed air systems 0:mplete 8/3/88 ,
147a (ame 1) ;
9.3.1 capressed air systems Complete s/3/84 141b (Ast 1)
Complete 4/3/84 141b 9.3.1 C- -
- 7 e air systems (Asv 1) f 9.3.1 Compressed air systems ccuplate 8/3/84 147d (as,1) i 9.3.2 Post-accident sampling system Ccuplets 9/12/84 148 (Rev. 1)
(II.8.3) 9.3.3 sgdgment ard f1cer drainage system Caplete 7/27/84 lea 9.3.3 Equipment ard floce &ainage system Ccuplete 7/27/84 l 1eb Primary containment instrument gas Couplete 8/3/84 150 9.3.6 (Rev. 1)
' system i
9.4.1 Control structure ventilation system CompInts 8/30/84 151a (Asv. 1) f 9.4.1 Cetro1' structure ventilation systen C e late 8/30/84 151b (Rev. 1) 9.4.4 Radioactivity acnitoring elements C1cond 6/1/84 152 (5/30/84- ,
Aux.sys. Meg.) .
Ibigineered safety features ventila- C e t_ ate 8/30/84 153 9.4.5 (Rev 2) tica system complete 4/1/04 154 9.5.1.4.a Metal roof deck construction classificiation 9.5.1.4.b ongoing review cf safe shutdown leC Action 155 apability .
9.5.1.4.c ongoing review of alternate shutdoun s c Action 156 capability
- 4
_ _ q res Sc/12 15 - as-- - __ - _
ATU OBENT 1 (Cont'd)
DSER R. L. MITIL 10
@BE SECTIN A. SODeGR NUMIER SUB7BCF SUGUS TEFIER DMED ITEM 157 9.5.1.4.e Cable tray protection Complete 8/20/84 158 9.5.1.5.a Class a fire detection system Ca plete 6/15/84 159 9.5.1.5.a Primary ard secondary power supplies Caplete 6/1/84 for fire detection system 160 9.5.1.5.b Fire water pump capacity Caplete 8/13/84 161 9.5.1.5.b Fire water valve supervision Caplete 6/1/84
~162 9.5.1.5.c Deluge valves
- Complete 6/1/84 163 9.5.1.5.c Marnal toes station pipe sizing Complete 6/1/84 164 9.5.1.6.e Remote stutdown panel ventilation Completa 6/1/84 165 9.5.1.6.g Emergency diesel generator day tank Complete 6/1/84 protmetion 166 12.3.4.2 Airborne radioactivity monitor Cmplete '9/13/84 ,
positioning (Rev. 2) 167 12.3.4.2 Portable contirmous air nonitors Complete 7/18/84
! 12.5.2 Equipendt, training, and gcciddres Complete 6/29 /84 168 for inplant iodine instrumentation 169 12.5.3 Guidance of Division B Regulatory Caplete 7/18/84 Guides i
1 13.5.2 Procedures generation package Complete '6/29/84 J 170 submittal 171 13.5.2 TMI Item I.C.1 Complets 6/29/84 13.5.2 PGP Comunitment Ccaplete 6/29/84 j 172 13.5.2 Procedures covering abnormal releases Ca plete 6/29/84 173 of radioactivity M P84 80/1216 - gs l - ---_ - _ _
l JtrImagtINF 1 (Cont'd)
DSER R. L. MITIL 1
~ WEN SBCIIGI A. ersammrum ITBE DEMBER S M7ECT STRItB 2 TIER E5 TIED 174 13.5.2 Resolution explanation in FSAR of Omplete 6/15/54
' DEI Items I.C.7 and I.C.8 175 13.6 Physical security Open 176a 14.2 Initial plant test progran omplete 8/13/84 176b 14.2 Iditial plant test progras Omplete 9/5/84 (Rev.1) 176c 14.2 Initial plant test program Complete 7/27/94 4
176d 14.2 Initial plant test progran Omplete 8/24/84 (Rev. 2) 176e 14.2 Initial plant test progrant Caplete 7/27/84 176f 14.2 Initial plant test progrant complets 8/13/84 176g 14.2 Initial plant test program Complets 8/20/94 176h 14.2 Initial plant test programi 0:mplete , 8/13/84 1761 14.2 Initial plant test program ocuplete 7/27/84 177 15.1.1 Parti I feeduster heating Complete /20/84 (Rev.1) 178 . 15.6.5 IDCA resulting frca spectnam of tac Action posMilmari piping breaks within RCP 179 15.7.4 Radiological consequences of fuel IGC Action handling accidents 180 15.7.5 S. pent fuel cask &op accidents Inc Action 181 15.9.5 1MI-2 Itan II.K.3.3 Omplete 6/29/9 4 182 15.9.10 1MI-2 Item II.K.3.18 Caplete 6/1/84
. 183 . 18 Ikpe Creek DatDR omplete 4/15/84 l
M Pt4 80/12 17 - gs
- m. L. nrrs, so neum A. SWMEIR gg ervum sums grTm 3 30 ruum ag3ga www .
Ctuplete 4/1/94 184 7.2.2.1.e Failures in reactor vessel level (nov 1) sensig lines 7.2.2.2 Trip systne seasons and cablig in Ctsplats (/1/54 ISS turtine tasilding <
Testability d plant rm ion couplete 4/13/84 186 7.2.2.3 (Rev.1) systems at power Lifting d Isade to perfossi surveil- Caplete 4/3/94 187 7.2.2.4 lance testing Setpoint methodology Caplate 4/1/84 188 7.2.2.5 Complete 4/1/ 94 189 7.2.2.6 Isolation devices Regulatory Guide 1.75 Ctuplete (/1/94 190 7.2.2.7 Scrust disd.azgo volume Complete 6/29/84 191 7.2.2.8 Ca plete 8/15/84 192 7.2.2.9 Reactor mode aritch '
(Rev. 1) 7.3.2.1.10 Manual initiation d safety systems Caplete 8/1/54 -
193 Standard review plan deviations Complete 4/1/54 194 7.3.2.2 (Rev 1)
Freese-protectiern/umter filled Caplete 8/1/84 193e 7.3.2.3 ,
- instrument and sampling lines and cabinet temperature control Caplete 4/1/94 195b 7.3.2.3 ha r MicMuster filled instrument and sappling lines and cabinet ta perature control Sharing d causen instrument taps Couplete 4/1/94 196 7.3.2.4 Micrgrocessor, multiplaser and Camp 1ste 4/1/94 197 7.3.2.5 (Rev 1) ccuputse systams N 884 84/1218 - ge .
R. L. IETE. 2 DSM A. N WW SKTEN ,-
a m w "P biidus IJ5TER UtIED rrung agga 1MI Itsu II.K.3.10-ADB actuation Quplate 4/20/84 198 7.3.2.6 IE am1htin 7b27-Imse cf nesrclass Caplate 4/24/84 199 7.4.2.1 (Rev. 1)
IE te==*=tica and control power system tus enring geration Remote shutdoun system Cauplate 8/15/84
, 200 7.4.2.2 (nev 1) acIC/tWCI interactions cephte 8/3/84 201 7.4.2.3 Inval N errors as a result Caplate 8/3/84 202 7.5.2.1 cf enrircremental temperature dfects a level instnsuentation reference lag 7.5.2.2 Regulatory Guide 1.97 Caplate 8/3/84 203
'Dt! Itan II.F.1 - Accident sonitoring Couplete 8/1/84 204 7.5.2.3 Plart process camputer systen Complate 6/1/ 84 205 7.5.2.4 .
High pressure / low pressure interlocks Couplete 7/27/84 I
206 7.6.2.1 HE.Bs aru'i consequential control systen Couplete 8/24/84 2M 7.7.2.1 (Rev.1) failures Multiple control systaus failures Courista 8/24/84 208 7.i.2.2 (Rev. 1)
Credit for neresafety related systems Camplate 8/1/84 209 7.7.2.3 (Ari1) in Chapter 15 d the FSAR Transient analysis recording syst c. Caplate 7/27/84 210 7.7.2.4 Control red drive structural anterials complate 7/27/84 211a 4.5.1 Control red drive structural materials Campista 7/27/84 211b 4.5.1 ,
Control red drive structural antarials complate 7/27/84 l
211c 4.5.1 ,
4 n ses 80/12 1s - gs
AE1313 ert 1 (Q at'd) m . L . u m r. 2 asBR A. eru m mum WW m.
N SWW2 SIRIts IRITM tutIm IIM 211d 4.5.1 cetrol rod drive structural materials Osq4ste 7/27/94 4.5.1 control rod Eve structural materials Ozqplete 7/27/54 211e 4.5.2 anector internals antarials Quplets 7/27/54
. 212 5.2.3 Ranctor coolant grammure boundary Quplate 7/27/54 213 l matarial 6.1.1 Engineered safety features materials Q:mplete 7/27/54 214 10.3.6 Main stese and fes&satar systen Quplete 7/27/54 ,
215 metseials 5.3.1 anector vocool materials Quplete 7/27/14 216e 5.3.1 manctor vessel meterials Quglate 7/27/54 216b Fire protection organizaticn Quglete 8/15/54 217 9'.5.1.1 l
9.5.1.1 Fire hazards analysis Campista . 6/1/84
' 18 2
{ Fire protection ahdnistrative Ocuplets 8/15/54 219 9.5.1.2 controls l
22 9.5.1.3 Fire brigade and fire brigado Q$ets 8/15/54 training Physical separation.M Mfsite Oglets .
8/1/B4 221 8.2.2.1 trarmdssion linse Desip provisions for relish- Complete 9/14/84 222 8.2.2.2 (Rev. 1) ment M an offsite power source Independence d dfsite circuits Quplete 9/13/84 8.2.2.3 223 (rmy. Il between the switchyard and class M busse .
Common failure mode between ensite Quplate 4/1/94 224 8.2.2.4 and offaits pouer circuits r
a m avla no- -
e
MED8pir 1 (Cont'd)
R. r umr. so Dest A 8CEB83E N. N L"".
5555 tm M s
11BI 8.2J.1 Testaldlity d atomatic transfer cf Congdete WW 225 power from the noemal to gumferred pouer source l
Grid statdlity Onsdate V1V34 8.2.2.5 225 (Rev.1) l Cgacity and apability of offsite complets 8/1/94 227 8.2.2.8 drcuits 8.3.1.1(1) 931tage &q darirqi transient condi- Camplete S/1/04 '
23 tiorm 8.3.1.1(2) Basia 'for using bus voltaGa versus Caglets 8/1/94 f 229 actual connected Iced voltage in the voltage drg analysis .
Ccuplete 8/1/84 230 8.3.1.1(3) Clarification d Table 8.3-11 Cesplete 8/1/84 231 8.3.1.1(4) Undervoltage trip setpoints 8.3.1.1(5) Iced configuration used for the Ccq 1ste
- 8/1/84 232 voltage & g anelysis Complete 9/13/84 233 8.3.3.4.1 heiodic synten testing (Rev. 1)
Capacity and capability cf ensite Caplete V1/84 234 8.3.1.3 AC pcuer agp1',as and use cf ad.-
ministrative controls to gusvent -
overimMng cf the diesel generators Diesel generators load acceptance Cc g late 9/13/84
. 235 8.3.1.5 (Rev. 1) test complete 8/1/04 2 35 8.3.1.8 C4lancs with position C.8 cf sus 1.9 Decription d the load sequencer couplets 8/1/94 237 8.3.1.7 segaancirg cf leads on the Wisite ccuplate 8/1/84 ,
238 8.2.2.7 power systen i
l a See 80/12 a - en
i.,
ArEM N Wir_1 (Cent'd)
R. L. NMr.10 DSSR A. SWElm WW SEREM amm sums Igrra Ings reum geen Testing to verify SOS minimus Ozqdete V15/94 23 8.3.1.3 voltage Ompliason witik EIPPED-2 OcogGets Vl/84 240 8.3.1.9 Imed acomptance test after prolonged Ocepista 9/13/84 241 8.3.1.10 (Rev. 2) no Iced cperation d the diesel generator Casplete 9/13/84 242 8.3.2.1 ccagliance with position I cf Regula
' (Rev. ])
tory Guide 1.12 Ccagdete 243 8.3.3.1.3 PEctaction ce galification d Class 9/13/84 (Rev. 1) 1E ogalpment frtzi the effects of fire aggression systems Analysis anl test to demonstrate Ccaplete 9/13/84 244 G.3.3.3.1 (Rev. 2) adegaacy d less then Wfieri separation complete 8/15/ 84 r 2 45 8.3.3.3.2 The uso d 18 versus 36 inches d '
(Rev. 1) separation between racemeys 8.3.3.3.3 Specified separation d raceways by CapInte 8/1/84 246 i
- analysi's and test C e late 9/13/84 2 47 8.3.3.5.1 Capability d penetrations to with- (Rev. 1) stand long duration short circuits at less than enziman or womst mee -
short circuit complete 8/1/84 248 8.3.3.5.2 separation d penetration primary and baceti, protections ccuplete V1/84 2e 8.3.3.5.3 The uma d bygoneed theanal overiced
- p. +2ive devices for penetration protecticos 250 8.3.3.5.4 timatire d fuses in accordance with C4=te 8/1/84
- R.G. 1.43 e
I- n ses 80/12 22 - go l
L___._______.-_______-___ ___ ___
Jermoseer 1 (Omt'd) 08sa R. L. str n L m WWI SECITGI ,
A. SOBeOR W_ 3D2US IRTTER DEIED l
Fret S R7Ecr 251 8.3.3.5.5 Fault clarrent analysis for all Quplets 9/14/84
.--: _- %=*4ve penetration circuits (Rev. 1) 252 8.3.3.5.6 The uso d a single breaker to provide CW 9/14/84 l penetratjan p.2+2ian .
(Rev. 1) (
253 8.3.3.1.4 Ozmitment to protect all Class lE Omplate 9/13/84 equipment fram external hasards vermas (Rev. 1) only class lE equipment in ene division 254 8.3.3.1.5 Protection of class lE power supplies Omplets 9/14/84 from failure of unqualified class lE (Rev. 1) loads 255 8'.3.2.2 Battery capacity Ocaplete 8/1/94 256 8.3.2.3 Autcmatic trip of Iceds to maintain M tate 9/13/84
- sofficient battery capacity (Rev.1) l 257 8.3.2.5 Justification for a 0 to 13 second Ccaplets 9/13/84 Ioad cycle (Rev. 1) l 258 8.3.2.6 Desip and qualification of DC Omplete 8/1/84 system loads to operats between minissa and maximaan voltage levels 259 8.3.3.3.4 Use of an invertar as an isolation Completa 8/1/84 device ,
8.3.3.3.5 Use of a single breaker tri; ped l by Complet, 9/13/84 260 (E V- II a ICCA sipal used as an isolation device 261 8.3.3.3.5 Automat [ic transfer of Iceds and ompleta 9/13/84 interconnection between redundant (Eev 1) divisions 262 11.4.2.d Solid wasta control program omplets 8/20/84 O
a m an 23- .
ArtN3Mr 1 (Cent'd)
R. L. Iltf!L 3 DEER A 83 WGR WWI SECf!W . . .
SINEW WITER IMM 1151 1U8W ER7ET Fire pe+Miaa for solid redomste Quplate W13/td 263 11.4.2.e
- storage area sources of asygen . Quelate V24/54 264 6.2.5 EE Filter Testing complete V13/94 265 6.8.1.4 Field leek tests Casplete W13/94 ,
266 6.8.1.4 :
control roma traic etmanical 0:mplete V13/84 267- 6.4.1 detectors Air filtratica mit draine Quplate 9/13/84 268 29 5.2.2 code asas 35 242 and 16-242-1 complate M' Complate W20/84 270 5.2.2 code mee 1& 252 2.4.14 Closure of watertight doors to safety- Open TS-1 related structures -
Single recirculation loop operation Open TS-2 4.4.4 core fiow monitoring for crud effects Cagdets 6/1/84 TS-3 4.4.5 Imoso parts monitcring systen open TS-4 4.4.6 ion in nnemal Open TS-5 4.4.9 lintural circu3 .
operation secondary containment negative open TS-6 6.2.3 pressure Islaskage and draudam time in open 15-7 6.2.3 secondary contaissaant I
tankage integrity. testing open TS-e 6.2.4.1 ;
acts subsystne periodic ocuponent open Ts-9 6.3.4.2 testing i
- i l
N 904 SW12 24= 95 l
AfDGpert 1 (Qint'd)
R. L. NEEIL E DEER A. SM WWI SECHGI 5555 mM ITWI IDER BR75 T 13-10 6.7 IEIT leakage rate 15.2.2 Arallability, _^49Ma, and testing Open 1>11 af **=# dan bypass systant 1>12 15.6.4 Primary coolant activity I ruel rod internal pressure criteria Ocuplete 4/1/54 l E-1 4.2 Stability analysis sutaitted before Open g-2 4.4.4 x -:-4,7g1= cperation l ,
l l
G
. I N IS4 SIV12 2S= ge
l 1
ATTACHNENT 2 DATE: September 14, 1984 l DRAFT SER SECTIONS AND DATES PROVIDED DATE SECTION DATE SECTION 3.1 See Notes 165 3.2.1 11.4.1 3.2.2 11.4.2 See Notes 165 5.1 11.5.1 See Notes 165 4
i 5.2.1 11.5.2 See Notes 165 6.5.1 See Notes 165 13.1.1 See Note 4 8.1 See Note 2 13.1.2 See Note 4
- 8.2.1 See Note 2 13.2.1 . See Note 4 8.2.2 See Note' 2 13.2.2 See Note 4 8.2.3 See Note 2 13.3.1 See Note 4 ;
8.2.4 See Note 2 13.3.2 See Note 4 8.3.1 See Note 2 13.3.3 See Note 4 8.3.2 See Note 2 13.3.4 See Note 4 8.4.1 See Note 2 13.4 See Note 4 8.4.2 See Note 2 13.5.1 See Note 4 8.4.3 See Note 2 15.2.3 15.2.4 8.4.5 See Note 2 8.4.6 See Note 2 15.2.5 8.4.7 See Note 2 15.2.6 8.4.8 See Note 2 15.2.7 '
15.2.8 i 9.5.2 See Note 3 l 9.5.3 See Note 3 15.7.3 See Notes 165
! 9.5.7 See Note 3 17.1 8/3/84 9.5.8 See Note 3 17.2 8/3/84 .
1 10.1 See Note 3 17.3 8/3/84 j 10.2 See Note 3 17.4 8/3/84 -
10.2.3 See Note 3 '
10.3.2 See Note 3
- 10.4.1 See Note 3 10.4.2 See Notes 365 l 10.4.3 See Notes 365 l 10.4.4 See Note 3 ,
See Notes 145 Notes l
11.1.1 11.1.2 See Notes 165 11.2.1 See Notes 165 1. Open' items provided in 11.2.2 See Notes 165 letter dated July 24, 1984 11.3.1 See Notes 165 (Schwencer to Nitti) 11.3.2 See Notes 165
- 2. Open items provided in l
June 6, 1984 meeting i
- 3. Open items provided in ,
j i
April 17-18, 1984 meeting ;
Cradb I
- 4. Open items provided in Nay 2, 1984 mating l
- 5. Draft SER Section provided i
in letter dated August 7, 1984 (Schwencer to Mitti) l NF 84 95/03 01
m
- - t u
l ATTACHMENT 3 .
OPEN DSER
- ITEM SECTION SUBJECT 222 8.2.2.2 Design provisions for re-establishment of an offsite power source 254 8.3.3.1.5 Protection of Class lE power supplies from failure of unqualified Class lE loads 251 8.3.3.5.5 Fault current analysis for all representative penetra-tion circuits 252 8.3.3.5.6 The use of a single breaker to provide penetration protection 9
m 0
4 ATTACHMENT 4 e
9
1 DSER Open Ites No. 222 (DSER Section S.2.2.2)
DESIGN PROVISIONS FOR REESTABLISHMENT OF AN OFFSITE POWER SOURCE GDC 17 requires, in part, that each of the of faite circuits be designed to be evn11able in sufficient time following a lose i of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified accept- i oble fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. The description in the FSAR as to compliance with this part of GDC 17 is not i suf ficient to reach a conclusion of acceptability.
l' By Amendment 4 to the FSAR the applicant in response to a request for information, stated that in the event of relay operation, the relays can be reset and the equipment returned to service within one hour. This design provision description for reset of relays may be related to design provisions used for reesta-blishing an offsite circuit from the transmission system through the switchyard to the Class 1E system; however, the description by itself is not sufficient to reach a conclusion of acceptabi-lity nor is it responsive to the staff request for information.
mESPONSE .
- ,,p ,as,ao 8. a./.</ and Myare 8 2 t 4a ae bcen y ,, ,, , , ,t y a pro se de U e eegu es M o i>fo ma t',",
j I .
I i
l l
i
! I I
f i
i i
__ _ _ ___J
l l
/~
~
(
OUESTION 430.3 (SECTION 8.2)
GDC 17 requires, in part, that each of the offsite circuits be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the d reactor coolant pressure boundary are not exceeded. The description in the FSAR as to compliance with this part of GDC 17 is not sufficient to reach a conicusion of acceptability.
Describe design provisions for establishing an offsite circuit from the transmission system through the switchyard to the Class IE system describing some event in the switchyard or protective relaying that has tripped all 500 kV switchyard breakers. ,
RESPONSE g. g , g -g ,
Section8.2.1.4hasbeenrevisedtoprovidethisresponse, t
A i .
I 1
I l
1 gr p g 430.3-1 Amendment 4 ggg ex
- , . - - _ - . - . - - , _ , _ _-- --- _ - .. ~ . - - .
t
, f
'! ke.
provide an auxiliary switch contact for input to generating station computer systems via a data input / output (I/0) cabinet for status indication. For safety reasons, the control switches are provided with a lock-in handle. The generating station control room operator must release keys in his possession to i operate these switches.
i
. 8.2.1.4 Switehvard
) The 500-kV switchyard, located to the east of the Hope Creek i, plant, is designed with tapered tubular steel structures and
- rigid aluminum bus work. This yard consists of two breaker-and-1 a-half bays containing five SF-6 circuit breakers connected to two 500-kV main buses, 10K and 20X, as shown on Figure 8.2-2.
! Bus 10X is protected by primary and backup differential relays.
Breaker failure relaying detects a failure-to-trip or failure-to-i i s interrupt condition at the line terminal and trips associated i
breakers necessary to isolate the line.
1 , i Generating station auxiliary services are supplied via two l
,- 13.8-breaker bays by four 500/14.4 kV, 42/56/70-MVA, oil-immersed,
- ' - self-cooled / forced-air-forced-oil-cooled (OA/FOA/FOA).three-phase transformers connected to the 500-kV busses 10X and 20X, as shown l
l on Figure 8.2-2. Station power transformers T1 and T4 each supply two 13.8/4.16-kV and one 13.8/7.2-kV station service transformers. The remaining two transformers, T2 and T3, each supply one 13.8/4.16-kV station service transformer and one 14.4-kV/208V station light and power transformer. Each 13.8-kV breaker bay consists of three breakers in series. To prevent paralleling of the transformers, one of the breakers is normally open. This breaker is closed in case one of the transformers is out of service.
As shown on Figure 8.2-2, there are six 13.8-kV, 1500-MVA oil circuit breakers. Breaker failure protection detects the failure to trip or failure to interrupt conditions at the lineand Primary terminals backup and electrically isolates faulty equipment.
relay protection on the 500/14.4-kV station power transformers is provided by the use of harmonic restraint differential relays.
The 13.8-kV system is ungrounded and connected to the delta side of all station power and station service transformers. To detect a phase-to-ground fault in the system, a 13.8-kV/208-V grounded-
- wye grounding transformer is installed or the secondary side of each station transformer. The neutral of the grounding DSER OPEN ITEM gAA 8.2-3
XEROX TELECOP!ER 495113- 9-848 7125PM 2 30165241C19 3# 7 I
C INSERT 'A' The 500 KV circuit breakers are pneumatically operated and ;
each breaker has suf ficient stored air for a minimum of three operations without compressor actuation. The compressor motor is j fed from the breaker A.C. distribution panel, which is provided with two independent A.C. circuits from the switchyard control l
house. ,
l The control room and the switchyard control house have
$ independent and simultaneous control of the 500 KV circuit breakers.
The electric system operation center, located in Newark, N.J., has limited control of the line breakers 51x, 60x, and 61x and the tie
! breaker 50X, and no control of the generator breaker, 52x.
Restoration of the 500 KV lines would generally consist o the following procedural steps:
The system load dispatcher would be contacted to verify availability of 500 KV circuits.
Verify 4 KV & 7.2.KV non-1E bus infeed breakers are opened.
Verification of 500/13KV transformer and 13 KV ring bus breaker positions aligned to restore of fsite power.
The load dispatcher is contacted for final clearance to reclose 500 NV breakers.
Once 500 KV power is reestablished, 4 KV & 7.2 KV power is provided to the respectives non-TE Puses, loading of these non-IM l busos can than comenan. '
- Final tranfer of class IE loads from the stand-by to the i
perferred power source can be made when plant conditions are stable.
1
.._.m._, . , _ , , , . - - - . , _ , _ _ _ _ _ _ ._ , _ _ . - , . , _ _ _ _ . . . , . . _ _ _ _ . . _ _ . . - _ _ . . - - ~ . _ _ _ , _ . .
I g ,
500 KV TRANSMISSION LINES r T U N WEENEY MEW FREEDOM Soo - 13 KV SECT cx h TRAN5 FORMERS (4)
)
% A) VM T- 3 T-s N.C., ' N.O. N.C. fly S/p
/
13.8 KV g N.C. SWITCNYARD N.O. -- --
- SLP2 l C _3hSLPI _
XFMR;rC , J (XFMR N C. 'N.C. 500 KV A SWITCHYARD
/
0 0 --
Sh
',/ ,
N.C. N O. N.C. C U E T-4
. ("Y D T-2 w:..
/"YD pp
- v. -
\ \\\ ob
//// SECT 20X
- uw LAA. W q evy m 24 - 500 KV MAIN TRANSE
{ { { lf l l lf ] j 13.8 kV FEEDS TO STAT!ON SERvlCE TRANSFORMERS GEN HOPE CREEK GENERATING STATION FINAL SAFETY ANALYSit REPORT i
DSER OPEN ITEM M ol ONE.LINE DIAGRAM F10 0 R E 8.2 2
OSER Open Item No. 254 (DSER Section 8.3.3.1.5)
PROTECTION OF CLASS 1E POWER SUPPLIES FROM FAILURE OF UNQUALIFIED CLASS 1E LOADS I'n Section 8.1.4.6 of the FSAR, it is stated that Class 1E equipment is qualified to perform its f unction during applicable design basis j accidents. The terminology " applicable design basis accidents" is of concern Sections 4.2 and 4.7 of IEEE standard 308-1974 requires l that Class 1E equipment be designed and qualified to perform their
- f unction during jan design basis event. If a class 15 component is subject to the ef fects of a design basis event environment, that cos~
ponent must be designed and qualified to f unction in that environ-ment irrespective' of the fact that the component may not be directly required to mitigate the design basis event.
By Amendment 4 to the FSAR, the applicant indicated that safety-related equipment that is not qualified (because it does not have !
' to perform a safety function to mitigate the design basis event ,
condition to which it is being subjected) are identified in Table 3.11-6 of the FSAR.
In justification of this design, the applicant further indicated that this identified equipment is connected to its power supply
, by a class 1E circuit breaker. The circuit breaker will operate to clear any f ault caused by the failure of unqualified equipment.
Thus, under the single failure criterion only one Class 1E circuit breaker is postulated to sail. The failure of this one circuit The breaker can degrade only its associated power supply bus.
redunda nt power supply and load will be available to perform the j safe ty load.
Further justification or assurance that Class 1E power supplies will not fail as a result of failure of unqualified equipment !
and results of analysis that provide a positive statement to the ef fect that the unqualified equipment failure position will not af fect station shutdown capability will be pursued with the 4 <
applicant.
RESPONSE
Additional justification has been provided in the revised response t to Question 430.)(.
.31 ;
l
}
i i
n~.--,---,., ---,-----,,v,w_- _ - - - - -_
HCGS TSAR -
1/84 r , . .
QUESTION 430.37 (SECTION5.3.1and$.3.2)
In Sections 8.1.4.6 of the FSAR you state that Class IE equipment is qualified to perform its function during applicable design basis accidents. The terminology " applicable design basis accidents" is not clear. Section 4.2 and 4.7 of IEEE standard ,
308-1974 requires that Class 1E equipment be designed and qualified to perform their function during any design' basis event. If a Class IE component is subject to the effects of a
) design basis event environment, that component must be designed and qualified to function in that envitonment irrespective.of the fact that the component may not be directly required to 'nitigate the design basis event. For each design begis event defined in Table 1 of IEEE standard 308-1974:
~
- a. Identify each Class. -1E componeit that does not meet the 4 design and qualification guidelines of Sections 4.2 and 4.7 of IEEE Standard 308-1974, and '
f ,
- b. Provide an analysis that demonstrates compliance with the single failure criterion assuming simultaneous fdilure of all CompJnents identified above with their associated power supplies. >
/
/
RESPONSE _
"applica e gn basis ac dent" is u d to more I precia y describe the stulated D which the afety- '
i relat component _s ne ded to mitiga e that DBE ill be l requ ed to operate , and thus d. scribe the enditions the equ ment must be q alified to. " is is tan on:plaince with NU G-0588, Part 2 (3)(a) which states in et, "should b
! q lified by test o demonstrat its opera lity for the -
me required in he environse tal condtt naresulting' dom l ,
accident." t is PSE&G's position t comply with t es requirement fo each piece safetty-re ted (Class IE '
equipment.
PSE&G agree that safety- elated con nents (Class )
should be esigned and alifled~to unction for e h DBA.
case of c ponents not r utred to However, unction in t mitigat a DBA is the equirement ot to fail in manner detrim tal to plant afety as sg cifled in NUR -0588, ,'
Part .1(3)(b). Th is interpr6ted to mean t t if the
- com nent is not r utred tc o rate during t DBE, the ;
qua ification regi rement for uch component is to d onstrate that ; hey will n t fail l'n a ma ner which woul 1 event safe sh tdown under the postulate DBA environmen al
- ondition. -
N% .
ossa orsw ITsM j f / 430.37-1 Amendment 4
-~~ ___
The a cific design asis events co sidered for CGS are disc sed in Chapt 15. These ev nts are comp rable to the gen ic postulat events of Tabl I of IEEE-3 -1974.
S ety-related uipment that es not have be qualified determined functionality reviews and/ DBA condition ave been ide tified in Tabl 3.11-6 of th FSAR.
- b. The equi t identified a ve is connec ed to its power supply bu by a Class IE reuit break . The power su ly bus and e circuit brea e are quali ed for the DBE enviro nt in which th are locat . The Class IE ircuit break will operate t clear any f it caused by t fail e of the equipe nt identifie above. Under e single fai re criteria a ication, on1 one Class IE c cuit ,
be kor is postula to fail., e failure of t b eaker can degra only its as ciated power s ply bus. scircuit,I this event, a ombination o the redundant wer suppl us and load is vallable to eform the safe y function j/
IC d. CA.No.C.h eh e e.V i $ e. of Me 3 f 0"3 C .
e e
ossa open ITut A JII 430.37-2 Amendment 4
- +
k RESPONSE To 43o.37
- There are no unqualified Class lE components in use at Hope -
Creek Generating Station. Each Class lE component is qualified in accordance with the requirements of 10CFR50.49 i
" Environmental Qualification of Electric Equipment Important ., ,
to Safety for Nuclear Power Plants", NUREG 0588,,and g '
All Class lE devices'are powered from Class lE power supplies and are separated from these Class lE supplies by qualified. -
Class lE circuit breakers or interrupting devices. .
/
o p E m %
\
w 0
e 4
9 h
4 k
DSER OPEN ITEM e s
, - . . -. -- - -,- - ~, . - - . - . . . . . . . . - - - , - . _ . . , , , -- -, ,~q,..,---- - . . , v.
DSER Open Iton C% 251 (DsER sesetica 8.3.3.5.5)
FAULT CURRENT ANALYSIS FOR ALL REPRESENTATIVE PENETRATION CIRCUITS By Amendment 4 to the FSAR, the applicant indicated that coordi-nated f ault-current versus time curves for representative pene- l tration conductors and their protective devices are included in Fig ures 420.46-1 of the FSAR. Based on a review of these figures, the staf f concludes that representative curves for motor dif-forential relay, current transformer, and instrumentation circuits were not included in Figure 430.46-1. Inclusion of these circuits as well as other circuits such that the coordinated fault-current versus time curves is representative of all penetration circuits will be pursued with the applicant.
RESPONSE
The response to Question 430.46 has been revised to provide additional f ault current versus time curves.
-4L I
f, l
l.
s I
lT j.
.A
.s.
r k
n I .
L i
~*w d--*-- " > m.wamy-.e m ,,iw-.-wwe,+wem.e-.m- w-w-ewwesw-nm--ym w----a,m.y-me.e9---m-ewp q er ' m r W - w g-r'e y a y y .- ww- wm y evtqmgy-'M 'Mm W v'W W Spr- pee e NT 9-T F"7***
.- ?
BCGS FSAR :
, 1/34
~
2arger than the circuit conductor. ,.
requested information is as follows: The remainder of the '
- a. '
HCGS complies with position 1 of Regulatory Guide 1.63 as stated in Section 8.1.4.12. In addition, the penetration assemblies are designed to withstand, without loss of mechanical integrity, the maximum short-circuit current vs.
time conditions that could occur, given single random -
failures of circuit overload protection devices. Time current characteristic curves, based.dn tests, of the penetration conductors have been estabt'shed by the -
penetration supplier; these curves show the maximum duration of symmetrical short circuit current. Based on these curves the primary and backup protective devices are selected to ensure that the mechanical integrity of the penetrations is maintained. This is further demonstrated in Part b, below.
The testability of the primary and backup protective devices is addressed in the response to Question 430.88.
b.
Coordinated fault-current versus time curves for representative penetration conductors and the protective devices are included in attached Figures 430.46-1. ..
- c. The test report that substantiates the capability of the l electrical penetration to withstand fault current without i
seal failure been for worst submitted under case a separate environmental cover. conditions has s
E
- O s rvi ' " E t, w M v o t.cr GE PE N E.N t N5 ASS CI ATED vitTH EACTOR 12.E. C 1 C.ULATtOM PLIM P MOTot ,ARE, pg.oTE Eb SY iWO C. A S. S lE me _ . m . ,, es . ~ - >
s e c,. T I O W
~
6 14. t HA5 .
EEW E.E H ED To IN C.L Of_ 7g g g g ,g g g g g,,
A-ad inse d -
Et DSER OPEN ITEM 88/
4 30 A
.I; e s erk The 1000 kc mil medium voltage penetrations associated with reactor recirculation pump motors, are protected by two ,-
Class lE circuit breakers in series as shown on revised FSAR Figure 8.3-4. Section 8.1.4.12 has been revised to include this response. Figure 430.46-1 Sheet 21 is a typical coordination curve for a #16 AWG penetration for RTD and thermocouple circuits. The curves show that the instrument penetration is protected for the maximum short-circuit current.
I i
, s . ... . . . ...... . .
.m
. __ ..... m a
_a a e s assel i
,L I IIliggl 1
. ._. - p.. ,
g 1 .1 I
I I I .
E, E E,
- ,-- . - - .,2a. . .
1 I 1, ,
- r l ,
.,a si gr. ; .. s ; .
. . r. .c-- =*
= m ;, . .mi
= . 4 1, ,+ ..y.
' , -_ -. .n , .- - c .. ., ; w a
- , ..d MM J.i E Z ~- f 4
,' l Mj t i f [ l td'"f-dt---- k + -- - =.
m u_Qw- ;.m .... p .1 a <, i
== -
- s. . m: -,
o m- +. , 4 , - n.-
2 >
.m -= =-
n i 2.,
- 7 '.-
e a : = , _ =4. =.g _3- L .
'7, c_E.. 1..
r
= -
-c ; ta_.= _ _ _ -
- =
q=-_.
g- _- ;: : .=z_
= :::-.4 .s .H== : . .
- . l .i.
,= - .._= *
,. I J _=. .w-
- m:: .
-._=
.:= ' i=
_:::m s t
- 1.1. 4. .w
.mmai
..-h..
mos .m : .
=- . , . , ,. ,
.,. . ., m , .. 6-
--H- - . ,
- ,. m
,o .
m - :- . mi
=
.- -. ni v + - n a , - .-a ,
n T- =*
" a 8 >
... n s s .
-_ .A : ' 44 + - + ;6- # 1 m _ _-;.e .. . 1
-4E..,' s iy-r
= - ,
__-.--;. _-,m _
m 44: --. ; -
=
.g - 9g ,; .2 .
gas 7
_%u .
- ,_ a
- w ,
n
- v. -: m
._ #. m a" ww=
- n4 r .
r
+ <
t t
w, - - -
sr-
%g =E
=
_z . ,i . ,
ar _- : g -
sq wM= ;;9 ;4 . g .3,,.pij
. 4,i ,.
, , . ,_.z.' 4 g,
.s ss w A :-'_=. m ;p i..'g 5 =.
i= .r 4 dQ e- b .. - =; i; x . -
= == = -2..-
= , .- ;p y:=.
.=-.-:._= u
. : t =_
= .,w. 2 := 1= .:.._ _T. s . - ..v - -
. , _ 1m_ ;:;= -
r . .=:., r : - i
. m__
--4 :j .4@.p.
ti.j-t. m. ., .
t
-- x 1
- t :l -A u . . . : tnt . . . .
4 r,-
.a_
-u_
_-4;r 4
m >
1_._.
y r-.-_.M.
x.,
hw :--1, i.
^
.< 1. ni > ,.
.g u
. : 6ei- .,
==# .. -$ 6 e . ,.- t4.,-
l
[.. . . . .
i :. W.
ia o . t_ 2
- -;, _. i '
, il i ?, --... . ,
{ ,~ . . - . ' - ---
- - - - -i-*3 ; ._
_ : ,., f'-- . ,
a _ -4 6 + i , - + .1 I'?-. 4 ; L.,. ..
p pJ.1,. s-L 4
f !T ;4 e 1i .i s .3 s
.. )
3=:j .
_-4 4 g:: 22m .
,& i
,' ", q ., ._ _ @. . :@ ' :
- -G}_- . U. 4, . - 1. . _. ; . :@-2
. h. , ' . .h'. - 4 .
, g gi
[2; e
- _ 4'a
- -
.1 3 --1n -' -i:L141 -. 71 ~_gd - T. .t:.q:t:4 h.1. p3 :r..;. .a ; ,. '
,.!!....! . ./y.p:- , . .
. )_ jf 1; --
_%;a - ., _~, _,,,
1 mu m s
. ,_ , 4- .
,w.
,. %. ,. e, - .- . _ . _ & - i i1 . ..o -
. . . g
,e
. , . .m .c _+.ev. .
- e.-
., . . . . _-__m.m_-
w~ ,
o m-.w.- -e
. 2,#. . -,: ,.
t ._.,;.,,-.
.,+, -
.3. . .. >
-m.c m . . . ---
m , ---- . - - . +q 4- 4 m .
&WSi.s-
- ..,3 6 m_ --._-,2_-
i ri ,u.3 ., f3 t. .a
- ~g., -
f ! inammm - -
O L k- E. & L_ - -
die 7 W 8sMuh48M^ :$A 1 !.
1 .;
l gg -a _ ain .
. 1-' #
- s .
[ ?4. .U.C=5
. _. _- :-:4 2 . . . ..
.p.,_ - - . . .r' . ---
--*-s- , ; .
~+~. ., . - - r; .
J.m t1 :____~
e E '- n T
. , _ _ -- ma
. *_.. ' - ?
..>-o , i 4 w i i __
4 f1 I = . -
4a '
b -.< f I .a4-A -*T m J.E
- i .
I"'" !#8T I 4 J . 1] J '~~-.-- ,
%- 41 .4 J 3 -h -
, , ' . , , , , , , _% ./1_ _
~' = + W i 1 y
w-_'1 .; . m . . .. -
! sn.
c__,.___o,..-..- y_,,=. _=___
~
- ._ m , ,.
1 ,,.- -
m . . . .
w - . -_.
,,_ m m '
< ~ .
c:w
- '- -, u .,i-w , '.m . e_
I . . i ..
- .n. em .m R s, W m.% .
E rt-irJ :~5_'
- ~ ~-
~ E E j - - - -. .
= =-
T ] ,..
vs
- a*
,f* , _=#=_
s n
w :m' l
u ._
__.,_-i_
- _ g..
_, . . ~m . _ . _
_,,,3-
-g
_ _ _ __., 3
._4_._,__.4.. 2 u i _ _ _ ; .,,,
g i i i glil[
.e. s asessi s... . . . . . . . ...... . . . ... a E
-l ,
Il I
1 i- 1 1
[ [ i l
I DSER OPEN ITEM J2 $'/ 1 I
O 1 I
. s.
. d
. . . . . . ._. a a a e. .a e snel.,-- I _I_ I_I_II_Ill -
_ y ,,,, .
, ..y . .-.. m _ . . . - .
-.-- - s .
=
_. 2;..;: n itt :
e., vder.;ga,w,d, 1 .- m o.
1
=6 ,
i-- " ==,
f- :-:rlv. . W
- s. , .
,.j, ,4, ~ ' -
T
=.b n O MWW-C;;$mn W1 . ., . .
== .
u ._. ,# n :,- 1. - -- .. . u. .
- m. :., .t. ,.
- # .t. .
- a. .2. . m ._,.
a ..h -4 . _d W ti W & H l fi W- M E - -
1JNM O Mii i .E 9 .
a i.;) ...
ye. ;..t......
,,,, .- e, i i ;t
,4 ~i e e rr v .y--v-
-i 71 J
j j .. . -
j:-
IL,t 75f p!d '
.I D'. u.; 7 p e4 - . .
-i 3 ,
.k + m - i f ., 7. . ,,J 4'
.: 6
- es i L J' an .. : ;.
-u .n: i,,h rnmi' H m .:-
e-i .
m., ,., .-
4: u. .p;
- i. -
Is : ;-
Wie. .ir - uen a. *:u -4:
li l ; ;, I a ip;>i: .,-,
.g rau r
N.':pla g -
T u.n r .,-. ;, tj e;:: u
- . . . . - I -
- n. i.. u [; .
11:
e,. :y:.
i - t y:
in. j,' 1
[1;t
..]_. .
e- in,ti L.. 1 l a m __
?
l"ll3 7 aq _...,_..
, .l 4, t 3M yf g
. , f,,i,
..o'--..
mg n, , .
a.r._
a i
=
- m. _ .
u
. y m-
=H
. s!cw --
. .- . , 1
> m . ., t- '--
=
m A... -
,. :,o ,
y7 ; .
=
% w.,+..T .n ,t ___-
..p + h _m. _. , i. T_
- , ...$._ ;- _J-n nnmmat
. ... . i _..
- + M3C.W.i t15. w ,. t & 't r" __
-- i " ,
. T e 4 b o p.bsr4 - -M W *i M.WWne.s i? WP t + - it
- ;- 1, !
.L t L,..:.y, ], . [_:-_
J '
. im ~ .
p Apt a:a . '.0%.#% .zimw ,,- yu .i .
G.
an s.i m 3, 1c m . T' e ..
i a l
=
y !
9 &
n se Desn -t- p Jvgg_~ -
i et b v :-.; T M %,ii ;dT .~.' r " -nli * . /..;.;.;.'i
% v .' .h .
m M'gT
-n ~: : . r- , ,
g -
(a.. .l} hip p. A, r W ..u.
- -' h w . . -,d, u a .a ;
.M..;
4, .J.h.14.a, n i .
p kw i . .) .:::.
.. yM,::
~ ; - .
ip.4 , m _,_ ._, -
-t- &e p- .m
,, . r _ < .
.m si, .y e'. 7 1. :
. 7--
,s . . . ., m c- --
,m... ,,, 9
..,. e * :e,r
... .n
. _ _ .a. . - 1 .. o, ,-
. ,___.- ... ._ , . 4 p. p . .a. ..y , -
u .- .. p .
- n. .__-_-
laF ; ,j] I T4.. PEntflAtw %,.g#d 4-" -F. 4-e h4-
'-- a m - - -- -h --1- '-- '
, - - i
._H'-.
- r- -
5%.
L9 : _;.
- - an p . ---o-..-.- ybW F y-j i G F w. 3-Q (M. K.11+ . iili u w + n W wwdm. Q u w w .d : :v ;.: e + r 1
- w. "i h .2 ,-
~ -. t,i n - -- ? ~- .i (p,. . 8- : . e. ..:p,y e ; p -l ' hh iC > s:i e . .- ..
- ,a,lm .1pA. li Ri g#. : g.. , p 4.
sqa r .h4 md,c I l - =..,... . . . . . , . . . , . . . . , , , , ,,,,,- g g g gggggg b -. . -m M !!csk. ,.a .* m == ser a = I E .yr mem n mim=c n
- peast SOPoues, eF 'YesS piMwar. etsN W is es g g) ,attes , , , , , ,
ear.off .e6 ***st.se. ,, are e' vet sawsuweer8P'*W 8sesAN1'a o .at 88 a M B ES B l l . HCGS FSAR = 1/34 ,
- f. 120-V ac lighting circuits
, g. Motor differential relay current transformer circuits
- h. Low voltage instrume,ntation circuits
- 1. Communication circuits.
The following system features are provided to ensure compliance ~ with the Regulatory Guide position on single random fai2ures of ~ circuit overload protection devices:
- a. Medium voltage penetration assemblies: The only siedium voltage circuits routed through the penetration are the 3.92-kV circuits for the two reactor recirculation pump motors. Each motor is supplied from a variable The maximum fault frequency motor-generator set. i current available for a fault inside theandcontainment the circuit is limited by the generator contribution resistance. puM Aty A gg gA cgyp prt.oyscTtow 1:ot THE 1000 Ecm' d , PE. weit.Aitou 15 PtovtDED %Y TWO CLAS S 1E CIE. Cult S2.E AKER5 1W SEE.155 AS SHovJM 1M F S AR FtG. S ,3-4. . E ACM ClECUlT SE.EAKEE, 15 Pit ovietb WMMAM OVSEcut.RE M t LE.L AY *TV E SE E5' AVS ArE. SE.T To TRIP THElt EESPECT)vG
'C \E NT - BE.E Ax1RS F Ic,. u o . y g, sagg7 3i TIM.E - c.u tttt.E N T c. Ap Agi t.i T y SMCWS CV THE 1000 THAT 'THE ,d PEMETEATion is c,gEATEE E cra MAN ANT M AC Wlula SH O E.T CIP, cu t T qwy I 9 5. TIME C.o4 ditto N TgA7 c.oV LD o c e. v t. . I' 480-V ac motor feeder circuits: The 480-V ac loads 4 5. % inside the containment consist of C2 ass IE and non-l class 1E motor-operated valves All andloads these non-class are iE supplied E continuous-duty motors.
- from 480-V motor control centers (MCCs).
E a The magnetic-only circuit breaker used in the g n combinatian starter for the motor provides primary " protection for penetration conductors.
- A thermal-l 3.1-13 Amendment 4 l t
.. ...,..w. . m . . . ,an- .r? - mt e.n.. ** - 3 i k
- M.x.c/;a . .,- , -N$ge . ;- f g.p .%fi. . . J,,..
.3. ;. *,'rg. . . 3.4.;.. s .m , a a .; . . : 'g .::. e. 4 g i?1;:5.y ::w..d 5~ .. , w .-T sf.,m% , :....,.... &. . ;, .2. m;,%.E ...o....- n .w. , , . . = llg ,. a . . .. a I . g liell!EIN::SM@.d.158 ~ ~v qw.y. m.E . } .. . . . . k . x.b. .2n~. , 1p. \ : (.~pm v c ::. - 1}
- -: w.. .w
. ~ *
- - I j..
\. M ?;G. kC..c':'r - g - _- f >t.bWf Ti+.%% - s. .. f , .- l h ( . l . . w,,.mg u ,.w;.... lt. & - .a . - ,im,, ,, ,o.. . n n Dt. re; .. e _.rw ' -- ~ ' - -g : <. mam , -- .:... . . .. .m y.. t, .:.: g _t p.:. .; ; g._t a .. ;.s.a.. , ... ; o., . . i :. - o, 4 - c .. ...y.~::u..:. .~. F ! ! ., (. +b.l S .. - iPri. .. .T..i. LM5 B ' ek . - i vv - ~ .,1 - ni . i.
- t. +qt _.__. gj .pi y~ - - - m, "gJ g j
- m.,21) ,- !
I ! ,1 - F..t8.Sif _ f i 6 llIJ g.Q.ek,s,,,'
- a llj gh.b,e,r,,h '~T (
5 .6 - 2. i .Y @$ Wl Ydi a - M. C L*'~"f YT' K~ N I r- a ' li Bt 5 .[.--- h i !. ('h, ljI : -E _ 5 * , I ( t_ EM.3- -. r e- .- ia.IF -Pl v i C.A-! u . a *- * -" f. d' --l Rr-g , . ._ . . .. I 1 :' -- .........3. t -' * # I* v- I 8 Il1 ( +$ . t '~
- g. I*cl .b % .
.tly
- In( !L. ~ . -.j A.
- '.'*h x *., ..... e .
...!3... 7f ...il.....p. ! - l e iL@_ : jai
- 4. - . ... . t:f"' O I!j ,[, -* :
~ - 2,. 5 I pI ,g f g i I .\n ;_ll i +I , , , ' gg,.I.,. :' . . ]'o 4g--(n...f r T.. m" n ..{ . , Esp,, 4 l :.*y. .. . . $. ;... . p.I! . "i r1 . el h [I %. .T-t YY ] *. . . L.-- .k, , y 1.... ...u_ y-l _ p. m . ., ; >. . . _ ~ . . . _ _ .. .d : r ~ 43 ' $l l '- ,- t 'lL'- f. . - s=r=w@ L.lEl - . . .J { tt m . . -. a i a ? .. f, .-T <
- m[i L .pf i 5h...?
8
- \'-h.yV..d.'f. :r. . .1.4 y.6'-4?2i.ij,&k,
- *?
. ?-$il$ YU - - - .f j .-i'f .;
- q cQ;: ; *; 7 -
.?'-Y: k )Y.)f e - i = .s a 4 1 J [ .mf
- m. ,.n.....- g . . . . . . .. .. .
u -- 4.z.- i 4;: l <. . .: p. . . .+$$, S& ?* -: ~ . .. ..j. .. .
- l.. .,..,- M & % ik V -g n - - - & 2 !. %. -
_._.,. ~. ._'. .... . .. ... _ muu , mieue.wis w-sty DSER Open' Item No. 252 (DSER Section 8.3.3.5.6) , 'THE USE OF A SINGLE BREAKER TO PROVIDE PENETRATION PROTECTION By Amendment 4 to the FSAR, the applicant has indicated that penetration protection for the two reactor recirculation pump motor circuits is provided by a single breaker that is tripped by primary and backup relaying. This design does not meet the requirements of position 1 of Regulatory Guide 1.63. Justification for noncompliane will be pursued with the applicant.
RESPONSE
Figure 430.46-1, Sheet 11, of Amendment 7, has been revised to show two breakers.
The only penetrations with instrument class circuits that are protected by a single circuit breaker or fuse are as follows:
- 1. Vibration Monitoring
- a. Circuit Breaker is 7 amps.
- b. Maximum short circuit current is 0.8 amps.
- c. Penetration is 916 wire with a continuous rating of 15 amps.
- d. These penetrations have a continuous rating in excess of 18 times the maximum short circuit current they may be expected to experience.
- 2. Neutron Monitoring System
- a. Circuit protected by a 1/4 amp fuse.
- b. Maximum short circuit current is 0.2 amps.
- c. Penetration is #16 wire with a continuous ratinq
! of 15 amps.
- d. These penetrations have a continuous rating in excess of 75 times the maximum short circuit current they may be expected to experience.
- 3. Acoustical Monitoring System
- a. Circuit protected by a 2.5 amp. fuse
- b. Maximum short circuit current (0.1 amp.
(The 330kJL resistor would limit the short circuit to 0.1 amp even if the rest of the circuit
- impedance was zero.)
- c. Penetration is #16 wire with a continuous rating l of 15 amps.
- d. These penetrations have a continuous rating in s excess of 150 times the maximum short circuit current they may be expected to experience.
l-kJk/ l
I Pago two
- 4. Thermocouple Circuits
-a. Thermocouples cannot generate any conceivable short circuit challange to a penetration.
- 5. P.A. Voice Circuits
- a. These circuits carry millivolt signals only when they are actually transmitting a voice communication.
The system cannot generate any conceivable short circuit challange to a penetration.
The above cases illustrate that the intent of Reg. Guide 1.63 is
- met. No single failure of a circuit over current protective device could cause a penetration failure. Refer to the repre-santative curves of Figure 430.46-1.
/
l EsJl
l,,:
v...
gI e i IL, nb ,
II
- . o, r,.
IIIIIIl'I I insis a e n i : .... . . . . , . . . .
wm y
ng? nW F m7 A.) 1m ma NJ :
m oa t p
': ' EnG3Hl E i
.M Ml
% g ,lgpl jrq 4 v l ! jf M j g g j(
i H@ Womi DR Mj ( . ,
M..t . p,m g Wg_ET, !,!,$..[. M,i$d,...q. l @as i$.
i..l $
l L . .
1
?
.f l w 2-k .I.-- '.
k' ,, a m h$ftk
- h. 1
..u -
rl ' i' 8*
'I I 1 ' d'
.;"-M' "
r - ' I' l( 7{oTT9 I
.a l"""':;8 b I"'
L
""ri" II t' t
.,. ;i u; -
, P 4t*'**"
2 u
- u ..'""? g, h CT'T 'b'* oir- . ..,
l' v **
'.i.
d *F sqinwns y m it.g ,
ITF
' ~ ~ ~ '
.,.... i
~t'f m-C'*h". M" .P.n
- 7%g,p w
..i ai' d"6'.
- a-.r. o. is mi !
, it Ot< 4 *,-
r ; .? 'e r
?{:'d"-T" -r.... vt 9't.! [r~t** .c .n .L... y- .,I t..a#
.. #-eq i
o, i..
n Ce m ,.u,,= .
w w
$.4 .tly
~ - ..,
, , sr ,
.n ; ,.,-. y :
uM ,i .6 _
F 7 Ty C - ~. . . . . ,
W'iil": P 9% h.
.- ryg. -
e, _ c h, z "
l't. m ;W,hMJ k srd ib '-
hql,dl 4
g't 2 "hf"*4 %. l Il -
}"l ; ; N dptd
, M'"; hf {j . gi
.y q "Aq .o f]il i -mmm..-
i I ,
W ur b .m y .
&q n
m A qp n .
j y%,m@. ',
H.E
.; '; K, $ g p,y s o, w W gd $.77 m p gi e,m .
. m n
T j
- g a .a o u . -
+E P9%{%q '
M* m:. e WW
, m r
,, i -.
I @ Gh .
p .:,e p, @$T f ;j",ypEqQM1i 30 ; ,
A ..p w%.w g
,;;,;g,gg]4,
, ', !p c ,
3 sj L. 1 um,.re . 3 ,
m e.q,.;.,,t,n
, gg
.n q,,m.
,,,, 3
., p .e
_ m o ,
, - , , w .m. 94-qg --H M, ,
- " Qp-- ie n s. t t ,c+c
- i-y
, ~
0 "
QA l.] :" c., ,
N hhpdNE '
. I Il "h
- s. fhh. v b
a l3 -
Yo _h_l g@ ,-@j!{
i l%gi MH{
>N 1 .
o@$ L W 'g I
N P"G@Sl@ l ( "
J 3, Mt ":d i l L mm ;swn y g
' gy mk : -q..
p,4 .s
- l, m qw mA s me m 6g .m _w. om mI. ,
. at% - * *"" .
o
,o r o , n nn io s m
m n
g J; c .ui
- ; : = ' - ,
a w v m %. . . ,n, .., . w ._,ni r hg
,i m q 7, f ..;,,,,, g 4._.73 g y, j l,, n .y,,p q
- 1 9 ; w,ai w
- w.euam, :
ri,
,o ._ y h ,e
- i ,, , ny
. v
,,i'. ' l% ; l *';
..,,,,,.~.., ,
l
,' , l l:- i ; d% 4- p h o E @M@e* 'Tn MrOW
":: L
[ -
k M.,o1 3MNh .h h%M N,$ N_b!h b .
_. f "hh hi$$_ ll
.. .. J- $._N.N..
a Lh.N..
- t@.i ).i%! $ 1 _..a. hun M rr .
l . : T. hr: WM$
g Wild 59 mi! !! NhWM}' 4. 4.. ~'
w"A l
e mes
=
llg ,,
s.w, .n. %
o c,www n i g: a .,s e.,a ,,,, 3, %.., # g .j qmm "
% s .,,W'r*ti-
- ,un
" yr* pay mm i.g,q,, ww p I
% b o
r +v
....h 1q s- nnpm Y * $- M i I i
[
. [ 7_
.g '_" ,
M., M..-
- m. .
4_ ..
mn ne.w-w hO um r
-=
wre w, w-
- M E2 !- e hu,bish y Jhfle f y
gg,.,
i P'
g d M0 P3MT \u4 U.TUII p li !
d QQ [
L i
, hgj "l f"'I" E (""Cil0 O 1 '"
l
'5 ,ab' f,0 m.ypen UE. killI '
i..
i,&
i pp {AM i M.tg. ,N e@@
g- ,;T """ s i, 3 c w nw + ...,
w +.% w - p,.p - :: m : : x,.s yeiinni j i ii:i.+- W m
e i p,, .i1 in
, ,g ,
tui es.c+-..n R1 mbmnq,1 .:p n ,g wii -
n i .
4 .,
-.m ,
ygi , . . _
y y e _ 1,,..
e .g ag ,, 7, , , , ,
" ' m" " 'g
_ . s na .
- e a m.r 1 ' " 't:uP-g t >?*+- rP'
,I ' '
'E !""- 1 ..
3""= "==== """
lilliI i i i 18838 8 8 8 , on 8
E en G S 9
. . _ , . . - _ , _ . _ . , _ _ _ . , _ _ _ . _ , . . - , . _ _ . _ _ , , , . . . _ . - , , , _ , _ . _ _ - _ . . . . . , . _ _ _ e ,, .
/
l ATTACHMENT 5 e
e
AcnuA ' s cLcu.n s ces ws, A - v-o- , A4 4Ar re ~s -s ;a g
, PROPOSED HCGS TECH SPEC 6.5 REVIEW AND AUDIT -
6.5.1 STATION OPERATIONS REVIEW COMMITTEE (SORC)
FUNCTION 6.5.1.1 The Station Operations Review Committee shall ,
function to advise the General Manager . Hope Creek operations I on operational matters related to nuclear safety.
COMPOSITION 6.S.I.2 The Station Operations Review Committee (SORC) shall be composed of:
i Chairman:
i Assistant General Manager - l Hope Creek Operations l Member and Vice Chairman Operations Manager l 4
Member and Vice Chairman: Technical Manager Member and Vice Chairman: Maintenance Manager Member Operating Engineer Member: I & C Engineer Member Senior Nuclear Shift Supervisor Member Technical Engineer Member: Maintenance Engineer
, Member: Radiation Protection Engineer l Member Chemistry Engineer Members Manager - on site Safety Review Group or his designee.
ALTERNATES .
l 6.5.1.3 All alternate members shall be appointed in writing j by the SORC Chairman. -
L a. Vice Chairmen shall be members of Station l, management.
I
- b. No more than two alternates to members shall l participate as voting members in SORC activities at any one meeting.
- c. Alternate appointees will only represent their l respective department.
I
- d. Alternates for members will not make up '
part of the voting quorum when the member the I
alternate represents is also present.
l l
l ,
1
)
MEETING PREQUENCY 6.5.'l.4 The SORC shall meet at least cnce per calendar month and as convened by the SORC Chairman or hi: designated alternate.
QUORUM 6.5.1.5 The minimum quorum of the SORC necessary for the performance of the SORC responsibility and authority provisions of these technical specifications shall consist of the Chairman er his designated alternate and five members including alternates. No more than two alternates to members shall participate as voting members in SORC activities at any one meeting.
RESPONSIBILITIES 6.5.1.6 The Station Operations Review Committee shall be responsible for
- a. Review of: (1) Station Administrative Procedures and changes thereto and (2) Newly created procedures or changes to existing procedures that involve a
- significant safety issue as described in Section 6.5.3.2.d.
- b. Review of all proposed tests and experiments that affect nuclear safety.
- c. Review of all proposed changes to Appendix "A" Technical Specifications.
- d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety. .
- e. Review of the safety evaluations that have been completed under the provisions of 10CFR50.59.
- f. Investigation of all violations of' the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Vice President - Nuclear and to the General Manager -
Nuclear Safety Review.
- g. Review of all REPORTABLE EVENTS. .
- h. Review of facility operations to detect potential nuclear safety hazards.-
NR82/02 2
- i. Performance of special reviews, investigations or analyses and reports thereon as requested by the General Manager - Hope Creek Operations or General :
Manager --Nuclear Safety Review. l
- j. Review of the Plant Security Plan and implementing l procedures and shall submit r(.ommended changes to the General Manager - Nuclear Safety Review.
- k. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the General Manager - Nuclear Safety Review.
- 1. Review of the Fire Protection Program and implementing procedures and shall submit recommended changes to the General Manager - Nuclear Safety Review.
- m. Review of all unplanned on-site releases of radioactivity to the environs including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence an'd the forwarding of these reports to the Vice President - Nuclear and to the General Manager - Nuclear Safety Review.
- n. Review of changes to the PROCESS CONTROL MANUAL and the OFF-SITE DOSE CALCULATION MANUAL.
SURC REVIEW PROCESS 6.5.1.7 A technical review and control system utilizing qualified reviewers from within the station organization shall be established to perform the periodic or routine review or procecures and changes thereto. Only those items that have a safety significance will be reviewed by SORC. Details of this technical review process are provided in Section 6.5.3.
50RC reviews will concentrate on safe and reliable operation of the station. Independent reviews for determination or verification of USO shall be performed by the Nuclear Safety Review Department (NSR) and the results of NSR reviews will be provided to SORC.
AUTHORITY 6.5.1.8 The Station Operations Review Committee shall:
y a. Recommend to the General Manager - Hope Creek
, Operations written approval or disapproval of items considered under 6.5.1.6 (a) through (e) above.
NRB2/02 3
1
- b. Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President - Nuclear and the General Manager -
Nuclear Safety Review of disagreement between the SORC and the General Manager - Hope Creek Operations; however, the General Manager - Hope Creek Operations shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.9 TheStationOperationsReviewCommitteeshalk maintain written minutes of each meeting and copies shall be i provided to the Vice President - Nuclear, the General Manager
- Nuclear Safety Keview and the Manager - Off-Site Review.
6._5.2 NUCLEAR SAFETY REVIEW FUNCTION 6.5.2.1 The Nuclear ' Safety Review Department (NSR) shall function to provide the independent safety review program and audit of designated activities.
COMPOSITION 6.5.2.2 NSR shall consist of a General Manager, a Manager of the On-Site Safety Review Group (SRG) supported by at least four dedicated, full-time engineers located on-site, and a Manager of the Off-Site Review Group (OSR) supported by at least four dedicated, full time engineers located off-site.*
The OSR staff shall possess experience and competence in the general areas listed in section 6.5.2.4.- The General Manager and Managers will determine when technical experts shall be used to assist in reviews of complex problems.
NSR shall establish a system of qualified reviewers from other technical organizations to augment its expertise in the disciplines of section 6.5.2.4. Such qualified reviewers shall meet the same qualification requirements as the NSR staff, and will not have been involved with performance of the original work. .
- Since the Nuclear Department is located on Artificial Island site, the terms on-site and off-site are intended to convey the distinction between inside and outside of the station fence.
l NRB2/02 4
Establishment of the Manager - oft-Site Review and Staff is guided by the provisions for independent review of Section 4.3 of ANSI N18.7 (ANS-3.2), and the qualification requirements for the review staff will meet or exceed those described in Section 4.7 or ANS-3.1. The Manager - On Site Review and staff will meet or exceed the qualifications describec in Section 4.4 of ANS 3.1.
CONSULTANTS 6.5.2.3 Consultants shall be utilized as determined by the NSR General Manager to provide expert advice to the NSR.
4 4
4 4
f NRB2/02 5 j
UFF-SITE REVIEW GROUP 6.5.2.4 The Off-Site Leview Group (USR) shall function to provide independent review and audit of aesignated activities in the areas of:
- a. Nuclear Power Plant Operations
- b. Nuclear Engineering
- c. Chemistry and Radiochemistry
- d. Metall'urgy
- e. Instrumentation and Control
- f. Radiological Safety -
- g. Mechanical Engineering
- h. Electrical Engineering
- i. Quality Assurance -
- j. Nondestructive Testing
- k. Emergency Preparedness It shall also function to examine plant operating cnaracteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources which may indicate areas for improving plant safety.
REVIEW 6.5.2.4.1 The OSR shall reviews
- a. The Safety evaluations for
- 1) Changes to procedures, equipment, or systems and
- 2) Tests or experiments completed'under the provision of Section 50.59, 10CFR, to verify that such actions did not constitute an unreviewed safety question.
- b. Proposed changes to procedures, equipment, or systems that involve an unreviewed safety question as defined in Section 50.59, 10CFR.
NRB2/02 6 -
- c. Proposed tests or experiments that involve an unreviewed safety question as defined in Section 50.59, 10CFR.
- d. Proposed changes to Technical Specifications or to the Operating License.
- e. Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety sign (ficance.
- f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
- g. All HEPORTABLE EVENTS
- h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures, systems or components.
- i. Reports and meeting minutes of the Station Operations Review Committee.
AUDITS 6.5.2.4.2 Audits of fac.ility activities that are required to be performed under the cognizance of OSR are listed below:
- a. The conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.
- b. The performance, training, and qualifications of the entire facility staff at least once per 12 months.
- c. The results of actions taken to correct deficiencies occurring in facility equipment, qtructures, systems, or method of operation that affect nuclear safety at least once.per 6 months,
- d. The performance of activities required by the Operational Quality Assurance Program to meet the Criteria of Appendix "B", 10CFR50, at least once per 24 months.
NRB2/02 7 T.
- e. The Facility Emergency Plan and implementing procedures at least once per 12 months.
f.
The Facility Security Plan and implementing procedures at least once per 12 months.
- g. Any other area of facility operation considered appropriate by the General Manager - Nuclear Safety Review or the Vice President - Nuclear.
- h. The Facility Fire Protection Program and "
implementing procedures at least once per 24 months.
i.
An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified off-site licensee personnel or an outside fire protection firm.
3 An inspection and audit of the fire protection and i
loss prevention program shall be performed by a qualified 36 months.
outside fire consultant at least once per
- k.
The radiological environmental monitoring program and the results thereof at least once per 12 months.
The above audits shall be conducted by the Quality Assurance Department or an independent consultant. Audit results and recommendations shall be reviewed by NSR.
ON-SITE SAFETY REVIEW GROUP 6.5.2.5 The On-Site Safety Review Group (SRG) shall function to provider the review of plant design and operating experience for potential opportunities to. improve p? ant l
safety; the evaluation of plant operations and maintenance I activities; and advice to management on the overall quality and safety of plant operations. ,
I. f l' The SRG will make , recommendations for revised procedures, equipment modifications, or other means of improving plant safety to appropriate station / corporate management.
RESPONSIBILITIES
) 6.5.2.5.1 The SRG shall be responsible fort 1
l
! NRB2/02 8
- a. Review of selected plant operating characteristics, NRC issuances, industry advisories, and other appropriate sources of plant cesign and operating experience information that may indicate areas for improving plant safety.
- b. Review of selected facility features, equipment, and systems.
- c. Review of selected procedures and plant activities including maintenance, modification, operational problems, and operational analysis,
- d. Surveillance of selected plant operations and maintenance activities to provide independent verification
- that they are performed correctly and ,
that human errors are reduced to as low as reasonably achievable.
NSR AUTHORITY 6.5.2.6 NSR shall report to and ad, vise the Vice President - Nuclear on those areas of responsibility specified in Sections 6.5.2.4 ana 6.5.2.5.
REC'ORDS 6.5.2.7 Records of NSR. activities shall be prepared and maintained. Reports of reviews and audits shall be distributed as follows:
- a. Reports of reviews encompassed by Section 6.5.2.4.1 above, shall be prepared, approved and forwarded to the Vice Presicent - Nuclear, within 14 days following completion of the review.
- b. Audit reports encompassed by Section 6.5.2.4.2
. above, shall be forwaraed to the Vice President -
- Nuclear and to the management positions responsible for the areas audited within 30 days after completion of the audit. e 6.5.3 TECHNICAL REVIEW AND CONTROL ACTIVITIES 6.5.3.1 Programs required by Technical Specification 6.8 and other procedures which affect plant nuclear safety as
- Not responsible for sign-off function NRH7/07 0
_-, ___._..._._..-.______________.,,_-,__._,,.m . . - - - - , _ a.
determined by the General Manager - Hope Creek operations, and changes thereto, other than editorial or typographical changes, shall receive an independent operability and technical review and be subjected to an independent USQ determination.
PROCEDURE RELATED DOCUMENIC 6.5.3.2 Procedures, Programs and changes thereto shall be reviewed as follows:
- a. Each newly created procedure, program or change thereto shall be independently reviewed by an individual knowledgeable in the area affected other than the individual who prepared the procedure, program or procedure change, but who may be from the .
same organization as the individual / group which prepared the procedure or procedure change.
Procedures other than Station Administrative procedures will be approved by the appropriate station Department Manager or by the Assistant General Manager - Hope Creek Operations. The General Manager - Hope Creek Operations shall approve Station Administrative Procedures, Security
. Plan implementing procedures, Emergency Plan implementing procedures, and Fire Protection Program implementing procedures.
I b. On-the-spot changes to procedures which clearly do not change the intent of the approved procedures shall be approved by two members of the plant staff, at least one of whom holds a Senior Reactor operator's License. For revisions to procedures which may involve a change in intent of the approved procedures, the person authorized above to approve the procedure, shall approve tne revision.
- c. Individuals responsible for reviews performed in accoraance with item 6.5.3.2a above shall be memL'ers of the station staff previously approved by the SURC Chairman and designated as a Qualitied Reviewer. A system of Oualified Reviewers shall be maintained by the SORC Chairman. Each review shall include a determination of whether or not additional cross-disciplinary review is necessary. If deemed
- necessary, such review shall be performed by the appropriate designated review personnel.
l l
NRB2/02 10
, . . - . . . , . _ _ _ , . . _ _ - _ , _ . , - . . _ , _ . _ . - . , , . , . _ _ _ . _ _ , . - _ . . . . . . . . . _ _ _ _ _ - _ _ . . . _ _ . _ __......-.-.....,..._..._...._.w.
- _ _ . .=. - _
t s
- d. If the Department t1anager determines that the documents involved contain signiticant safety issues, the documents.shall be torwarded for SOHC review and also to.NSR for an independent review to determine whether or not an unreviewed satety
, question is involved. Pursuant to 10CFR50.59, hMC approval of items involving unreviewed safety questions or Technical Specification changes shall be obtained prior to implementation.
NON-PROCEDURE-RELATED DOCUMENTS 6.5.3.3 Tests'or experiments,'hanges c to Technical Specifications, and changes to equipment or systems shall be reviewed in a manner similar to that described in items 6.5.3.2a, c, and d above with the exception that the recommendations for approval are made by SORC to the General -
Manager - Hope Creek Operations. Independent safety reviews for determination or verification of unreviewed safety questions will be performed by NSR and the results of NSR reviews will be provided to SORC. NSR reviews will be performed not only by using its own staff, but, when needed, also through the use of a system of qualified reviewers established throughout the corporate organization to support NSR. Pursuant to 10CFR50.59, NRC approval of items involving unreviewed safety questions or Technical Specification changes shall be obtained-prior to implementation.
RECORDS -
6.5.3.4 Written records of reviews performed in accordance with item 6.5.3.2a above, including recommendations for approval or disapproval, shall be maintained. Copies shall be provided to the General Manager - Hope Creek Operations, SURC, NSR, and/or NRC as necessary when their reviews are required.
6.6 REPORTABLE EVENT ACTION .
6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
- a. The Commission shall be notified 'and/or a report ~
submitted pursuant to the requirements of Section 50.73 to 10CFR Part'50, and >
I b. Each REPORTABLE EVENT shall be reviewed by the SORC and the resultant Licensee Event Report submitted to the NSR and the Vice President - Nuclear.
NRB2/02 11 -
6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions snall be taken in the event a safety Limit is violated:
1
- a. The unit shall be placed in at least HOT STANDBY within one hour,
- b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice President - Nuclear and General manager - NSR shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- c. .A Safety Limit violation Report shall be prepared.
The report shall be reviewed by the SORC. This report shall describe (1) applicable circumstances .
preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence,
- d. The Safety Limit Violation Report shall be submitted to the Commission, the General Manager - Nuclear Safety Review and the Vice President - Nuclear
, within 14 days of the violation.
E r
)
NRB2/02 12 .,
Y i
a-9 ATTACHMENT 6 C
e a
f 0
a f _ , - - _ _ ,, , - _ , , . .
f CV ASB OPEN ITEM IE BULLETIN 81-03 Hope Creek has been requested to address the applicability of IE Bulletin 81-03: Flow Blockage of Cooling Water to Safety Components by Corbicula sp. (Asiatic Clam) and Mytilus sp. ( Musse l) .
RESPONSE
Experience at the site has been shown that the referenced organisms are not indigenous to the local esturay. However, biofouling by similar species could potentially occur.
At Hope Creek, the only safety related heat exchangers which receive esturine water are the safety auxiliaries cooling system (SACS) heat exchangers. The balance of safety related heat exchangers are cooled with condensate quality water which is cooled on the shell side of the SACS heat exchangers.
Biofouling will be controlled by the continuous injection of sodium hypochlorite in front of the service water pumps.
Should this control be temporarily disrupted, sodium hypochlorite can be injected at a higher rate to assure the ,
cleanliness of the system.
c m.a.lenloSo n o ( Whe h'*I Biofouling would' be detected by monthly ezeurcment Of g_ g,gg [tg differential pressure across the SACS heat exchangers. The heat exchangers will also be visually inspected during F0 9: O refueling outages. The SACS heat exchangers are tubed with 3/4 inch diameter titanium tubes. Titanium is not subject to erosion from contact or turbulent flow.
Since the service water system incorporates redundant equip-ment with piping cross ties, it would be possible to physically clean a SACS heat exchanger while operating.
Chlorine discharge for the service water system is not a concern since the service water system discharges to the closed loop circulating water systems. Blowdown from the l
circulating water system will be dechlorinated.
JES vw l MB 18 01-A 1 ,
l