ML20098C921
| ML20098C921 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/24/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8409270282 | |
| Download: ML20098C921 (127) | |
Text
{{#Wiki_filter:I O PS. Pubhc Serwce G comoany Electric and Gas 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 I September 24, 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 L Division of Licensing Gentlemen: HOPE CREEK GENERATING STATION ~ DOCKET NO. 50-354 I DRAFT SAFETY EVALUATION REPORT OPEN ITEM STATUS 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 status 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 resol' red to the staff's satis-faction, please' provide a specific description of the issue which remains to be resolved. 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. Also, enclosed (see Attachment 5) isacopyogmodifications to FSAR Section 17.2 and SRAI (1) as requested by the f Quality Assurance Branch. These modifications will be included in Amendment 8 to the FSAR. del 8409270292 840924 1 I I DR ADOCK 05000354 The Energy People E PDR l 95 4912 (4M) 7 83 s .m
-Director of Nuclear Reactor Regulation-2 9/24/84 Also, confirming discussions with the Core Performanco Branch, HCGS reactor vessel water level indication is I analog.. Should you have any questions or require any additional-information on these items, please contact us. Very truly yours, A/. M/#/Aa Attachments / Enclosure C D. H. Wagner USNRC Licensing Project Manager (w/ attach.) W. H. Bateman USNRC Senior Resident Inspector (w/ attach.) J. Spraul USNRC Quality Assurance Branch FB18 1/2 e (..
t ? IRIEs 9/24/84 ~ ATDL9ENT 1 R. L. MITIL TC gggg A. SGMHCER CPER SECICH SD@S IETTER E9CED _ ITDI NUtER StN ECT 1 '2.3.1 Design-basis tange'ratures fer safety-Cmplets 8/15/84 related mriliary systans ~ 2a 2.3.3 .WMes d metectoIcgical Caplate 8/15/84 (Rev. 1) measurements 2b 2.3.3 Accuracies cf meteorolcgical Caplete 8/15A4 (Ref. 1) maapurements 2c 2.3.3 Accuracies cf metecrological Ccmplete 8/15 /84 (Rev. 2) -mments Accuracies cd met $creiogical Ccnglata 8/15/84 2d 2.3.3 (Rev. 2) amasurements 3a 2.3.3 LWig cf cnsite metecrological Cc=@lete 8/15/84 (Rev. 2) (III.A.2) measurements m a. l 3b 2.3.3 LWim cf ensits metscrological Ccuplete 9/15A4 (Ref. 2) -]_ measurements m am (III.A.2) NRC Action l 3c 2.3.3 LWing cf cnsite meteorological measurements prcgram (III.A.2) Ccmpleta 8/03/84 c' 4 2.4.2.2 Pmdirq 1evels Sa 2.4.5 Wave impact and rurup cn service C mplete gf13/g4 (aev. 3) water intake structure Wave f: ;ect and n:nup en service Cenplate 9/13/84 Sb 2.4.5 (3,v, 3) water bitake structure Sc 2.4.5 Wave ingact and nuup cn service ccupIsta 7/27A4 wter intake struc*m 5d 2.4.5 Wave ignet and runup en service C=nglace 9/13/s4, (Rev. 3) water intake structure l' .6a 2.4.10 stability cf eresicn protecticn Ccmplete 8/20/84 structures 6b 2.4.10 Stability cf ercsien ;rotecticn Ccuplets 8/20/84 structurns 6c 2.4.10 Senh414 ty cf eresicn prctecticn Ccnglets 8A3/84 st=ucturae M P64 80/12 1-qs i i f l
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@St SEC22CN SDRS LETTER DRTED fras utseet SER7ECF 7a 2.4.11.2 2homent angsets cd ultimats heat sink. CbspIsts 4/3/84 7b 2.4.11.2 1hammal aspects cd ultimata beat sink c:spIsta 8/3/84 8 2.5.2.2 Choice d==w4== est+%'*= for New ompista 8/15/54 EhgLand - Piedscrrt Tectcnic Province 9 2.5.4 soil dauging values Complets 6/1/84 10 2.5.4 Ramdatica level rompense We C:splets 6/1/84 11 2.5.4 Soil sbaar WM variatica Ctaplets 6/1/84 Nih cf soil layer prcperties Ctspists 6/1/54 12' 2.5.4 13 2J.4 Lab test abeer @ 'tt values Mic 6M 14 2.5.4 LM=f=etien analysis of river bottcm Ccaplets 6/L44 serufs 15 2J.4 Tabulaticas of shear =*11 Cbeplete 6/1/34 16 2.5.4 Drying and wetti:q effect m Ccapista 6/1/54 Mt3* l 17 2.5.4 Peuer bicck settlement acnitcring Ccaplets 4/1/84 i ) 15 2.5.4 Maximm earth at rest pr==ws ccuplets 6/1/84 l coefficient 19 2.5.4. U4aefacticn analysis fer servics C: splats 6/1/84 weer piping 3 '2.5.4 Esplanatica cf ctserved scher bicck Ccoplets 6/1/84 settlement 21 2.5.4 Service water pipe setetan=nt roccrds emplete 6/1/84 22 2.5.4 Csifordams stability Ccaplets 6/1/84 itP8480/122..gs
JtF50 tert 1 (Cent'd) lt. L. METE ?J WiER A. SQ MIEC23L CrSt TI1DI NGSER SWUECF STMUS IETER DPGTD 2.5.4 Clarificatica d FSAR Tables 2.5.13 Ccaplete 6/1/84 23 and 2.5.14 24 2.5.4 Scil depth nodels for intake Casplete 6/1/84 structure 25 2.5.4 Intahm structure soil :sedeling Caplate 8/10/84 26 2.5.4.4 Intake structure sliding stability Caplate 8/20/84 CcapInts 6/1/84 27 2.5.5 Sicga stability 28a 3.4.1 Flced r L-2im Ccapists 8/30/84 e (Rev. 1) 28b 3.4.1 F1ced grotecticn Cmplate 8/30/84 (Rev. 1) 28c 3.4.1 Fload -p r uicn Ccipista 8/30/84 (Rev. 1) Canplate 8/30/84 28d 3.4.1 Ploed grotectica (Rev. 1) l 28e 3.4.1 Ficcri forsicn Caiglets 8/30/84 (Rev. 1) C=rplate 7/77/84 28f 3.4.1 F1ced gretecticn Ccupista 7/27/84 28g 3.4.1 Flced gretecticn l-29 3.5.1.1 Internally generated missiles (cutside Coplate 8/3/ 84 (Rev. 1) centainment) Cicned 6/1/34 ~30 3.5.1.2 Internally generated missilas (inside (540/84-ccatainment) Aux.5ys.Mtg. ) 31' 3.5.1.3 W rtine missiles Caplace 7/18/84 32 3.5.1.4 Missi1~me generstad ty naturs1 p%w 4 Cmplets 7/27/84 33 3.5.2 5tructures, systems, and wents to Conglaes 7/27/84 to grctacted frca externally generated missilme J. e-
d ....e o. 1' MEDO9eff 1 (Cert'd) R. L. NETE. 2 J DBER A. SOBENCER WW N SD:rus trrrza otrED n1nrT TIse vuesR I 34 3.6.2 Omrestrained 4 W rg pipe inside ccupInte 7/12/54 containment 3.6.2 Ist program for pipe welds in e 7 =to 6/29/54 1 35 treen - -in= fen sene ccuplate 6/29/54 36 3.6.2 Postulated pige rupturns Mtse isolaticn dack valve Casplets 8/20/54 37 3.6.2 gerability 38 3.6.2 Design cd pipe rupturn restraints ccuplets 8/20/84 39 3.7.2.3 SSE analysis results usire finite Cang1ste 8/3/84 element sothat ard elastic half-space ggscach for centainment structurs 40 3.7.2.3 Sst analysis results using finits cc@a 8/3/84 elamart methcd ard elastic half-space 4 fee intake structure w Steel centainment bv+1bg analysis ccuplets 6/1/ 84 41 3.8.2 Steel ccmtainment ultimats Mty Cc:rglets 8/20/84 (Rev. 1) 42 3.8.2 analysia I 43 3.8.2 SRV/IJ A pcol dynamic.1 cads Carplats 6/1/84 44 3.3.3 'M3 349 deviaticra fcr ir: tarnal Corplata 6/1/84 structures 45 3.8.4 M3 349 caviaticna for Catagery I C=nglate 8/20/34 (Rev.1) structures 44 3.8.5 ACI 349 deviaticns fu fomdaticos Cenplate 8/20/54 (Rev.1) Ccuplate 8/10/54 47-3.8.6 Basemat ws We (Rev. 1) Congists 8/20/a4 44 3.8.6 Rocking ti:ne N.as (Rev.1) + G ,w, ._.,.__,,,_,--m,,_--,,,,-._,ny_,m..r,.,_.,_ , m mm moy..
JITDogett 1 (Cmt'dt R. T NETT.1D M A. - Gel MTrru SDES mrD!R CRED n15E M3eER StBM!CF 49 3J.6 Gm:ss ccncrete section c e face 8/20/84 (ase. 1) l- ~ 50 3.8.6 vertical floce flexibility roepense Ccaghte 8/20/1 4 Oter.1) spectra 51 3.8.6 ccagariscm cf Bechtal 14A-2 Caiglete 8/20/84 (Dev. 2) verification remilts with tta desigrr-basis rumults 52 3.8.6 9_*4Mty ratica da to pige break CcngInte 8/3/84 53 3.8.6 Desigt d! seismic Ca'vi I tarts Canplete S/20/84 (Rev. 1) 54 3.8.6 Cabinatica cf vertical.=v:-c:: Ccaglate 8/10 / 84 (Rev. 1) 4 55 3.8.6 Tarsicnal stiffness afm12tica Ccaglats-6/1/84 56 3.8.6 crywell stick nedel devolc;xtent Ccaplets S/20/84 (Rev. 1) 57 3.8.6 Retaticnal time histcef nputs 'Ccagiste 6/1/84 i 58 3.8.6, "o* refe5ence point for auxiliary Ccmplete 6/1/84 witiate medel 59 3.8.6 over*W mcment cf reacter Ccaglets 8/20/84 (Rev. 1) building foundatica mat 60 3.8.6 SEAP element size 11:sitaticru Carglets 8/20/84 (Rev. 1) 61 3.8.6 saisic sedeling <f drfwell shield cmf=ta 6/1/34 wall Ccupiste 6/1/84 62 3.8.6 Orywell shield wall txzndarf conditicos 63 3.8.6 asecter building ocza bcundarf Ccs,Imen 6/1/s4 ccnditicns J e e,,-- -.,.--.r. ,,,,-ew,,.,.,-, n, ,._.--,,,,,_,,.,m,---w..,-.w_.-n,e-w-,-.-,,_,,,_.,,._e._w,_,,-.,_,_w.-,,+,n
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TEN M3MR 64 3J.6 SEE analysis 12 as osteff frequency Ccuplate 8/20/84 i: (Rev. 1) e 3.8.6 Intake structura crans heavy Iced CcapJets 4/1/s4
- qp 64 3.8.6 Y=ce analysis fee the irtake ccuplets 8/10/54 (Rev. 1) structure 67 3.8.6 Critical Icads ~1mf2ticn fer Ccaplets 6/1/84 reactor h'41Mg ens 68 3.8.6 Reactor t'41Mg fcundatica mat Caglets 6/1/84 contact pr.assarus 69_
3.8.6. Factors d sadety against slidig and ccuplete (/1/84 overturning cd drywell shield wall 6/1/84 70 3.8.6 Seismic shear force distrihiticn in CagIste cylinder a ll 71 3.8.6 onetturning cf ef nder all Ccuplace 6/1/84 li 72 3.8.6 camp bean desigt of fusi gool alls Ccuplets 6/1/84 ASESD b model Iced iguts Caspiste - 6/1/84 73 3.8.6 74 3.8.6 2rnado W=<e4-iticn Ccuplets 6/1/S4 75' 3.8.6 Auxiliary N11Mng.brM gressure C=nplets 6/1/84 76 3.8.6 Tangential shmar stresses in efw11 ccuplets 6/1/ 84 shield wall and the eflirx$ne wall 77 3.8.6 Facter d safner, against overturnig Ccuplets 8/20/54 1 (Rev.1) d irtake structurs 78 3.8.6 Dead Icmd atm12ticns Ccaplete 6/1/54 79 3.8.4 Pest-medificatien seismic Iceds ter Caglets 8/20/ 84 (Rev. 1) the tcrus J, A
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3,- ~' . :.2.. - a u - - 1 (c m e'd)_ It. L. Mmr. E EB8R A. - Om m SDatB IZr2R Oc2D MN leegt SEB2H2 80 3.8.6 Ttzus fluiN_r tm interactions cczehte 6/1/84 81 3.8.6 Seismic < Mapl % cf terus Complets 8/2D/84 Oter. 1) 82 3.8.6 Review d seismic Categ:ry I tar
- Cczylate 8/20/84 (Rev. 1) desiq):
83 3.8.6 Factezz d safety for defsell Ccapiete 6/1/84 h
- 145 ev= Wiry 84 3.8.6 Ultimata cagacity cf centafrummt Cczplete 8/20/84 Utst.1)
(==*= rials) 85 3.8.6 toad ccabinatica c:rmistancy complete 6/1/84 -86 3.9.1 c i tae ecds validation Ccaplate 8/2D/84 87 3.9.1 Infcenatica crt trarmients Complete 8/20/84 88 3.9.1 Stress analysis and elastic 7 astic Cciplete, 6/29/84 1 analysis 89 3.9.2.1 Vibraticn levels for HSSS piping Casplate 6/29/84 syntass-90 3.9.2.1 fibration acnitorim ry d.: rig Ccuplete 7/18/84 testing 91 3.9.2.2 Pipig wi and anchces Complets '6/29/84 I 92 3.9.2.2 Tripla flusc R eed cental. m Ccaglets 6/15/84 p.ticns 93 3.9.3.1 Imad a:abinations ard allcmeble Cczylate 6/29/84 stress limits 94 3.9.3.2 Desigt cd SRYs and SRV discharge Couplets 6/29/84 AA PGM 5 M PG4 80/12 7 - gs
.~ 4 maapert 1 (Cent'ot R. L. MIT2L 1D IMER A. m @tn m SD5tB IETTER DCED _ TMN MBEER SEMLT 95 3J.3.2 - Fatigue erahatica en SRV piping Caglete 6/15/84 and ICC1 damccmars 96 3.3.3.3 II IW-th Notics 83-80 Caplats 8/2Q/84 Otsy. 13 97 3.9.3.3 PMifng aitaria used for +m Ccuplets 6/29/84 sqports Caplats 6/15/84 98 3.9.3.3 Design d bolts 99a 3.9.5 Strema = M ;-les ami limits fer. C a pists 6/15/84 core sq; port structu:ss 99b 3.9.5 Stress categories ant li:mits for ccuplets 6/15/84 r core sgport structures 100m 3.9.6 10 Crit 50.55a per,@ (g) Capists (/29/54 i loch 3.9.6 10C11t50.55a paracaph (g) Caplate 9/12/84 (Rev. 1) 101 3.9.6 PSI and ISI rw fcr pmps and Ccq1sts 9/12/64 (Rav. 1) valves 102 3.9.6 taak tasting cf press.tre isolaticn Cciglete /12/84 9 (a y.1) valm Seismic ani dynamic qualificatien cf Caglets 8/20/194 103a1 3.10 ( mechanical ans electrical e:pignant ( Cc9 sts 8/20/84 1 Seismic ani dynamic qualificatien cf 103a2 3.10
- W 1 and electrical equipment Seismic ami dynamic q=1Hicaticn cf Cczplats 8/20/84 103a3 3.10 mechanical and electrical equipment Seismic ani dynamic qualificatien cf Cenplets 8/20/84 l
103a4 3.10 mechanical ant electrical equi;annt l i t r N
e g AmcDeert 1 (0:nt'd)_ R. L. NETTL DSER A. *Me88'F' N srtius TJerTest Dts ~ CNEN sua7Ber mm teent salade and dynamic qualificatica cf 0:aplete 8/20.4 4 103a5 3.10
- 4cn1 and elsetrical ansi 4 m seismic and dynande ?=11Ma*4= of M ets 3/20/54 f
103as 3.10 m_ -f=1 ard electrical ar**fM sefsmic and dynamic qualifimef m cf Miate 8/32/54 103a7 3.10 =mehenical and electrical nif M Seismic and dynamic qualification of Ccaplets 8/20/54 103b1 3.10 W i~ t and electrical agif M seistic ard dynamic q=1tficatica of C:nglets 8/20/84 103b2 3.10 =arhanical and electrical n>4 M Seismic and dynamic galificaticn of Conglets 8/20/54 103h3 3.10 =mehanical and electrical gif M Seismic and dyn==4e qualificaticn cf couplets' 8/20/84 103b4 3.10
- 1c=1 ani electrical equi;:nent Seismic and dynande q=14Heatien cf Ccuplete S/20/84 103b5 3.10
==rhanical and electrical agaipment Selmie and dynamic qpalificatica cf C::splets 8/20/84 103b6 3.10 mechariical and electrical equi;ssent Seismic and dynamic qualificaticn cf Cc:9 r,ta 8/20/84 1 103c1 3.10
- 4c=1 ard electrical eq 1;nsent Seimic ard dynamic qualificatien cf C::nglets 8/20/84 103c2 3.10
- 4~1 ani electrical MiMaant Seismic and dynmaic galificaticn cf C:ngInte 8/20/84 103c3 3.10
- 4mt ard elsetrical a<*if-'=nc Seismic and dynamic qualificaticn cf C:mplets 8/20/84 103c4 3.10 anchanical ard electrical equignent NRC Acticn terrimamental c=tificatica cf 104 3.11 mechanical and electrical eq11poent
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e 4 1 iuG 1 (Cent'd)_ h' : R. f NI21L 10 j ENER A, - ~ WWI SECED3B SD12tB tJ2TER Ot22D __ ' SUBJECT 4 I!BI IOEER 105 4.2 Plase-specifis sedunnical WWrqi Couplets 8/ 21/84 (ase. 1) analysis 4 I 105 4.2 Mud"ty d seimda andd IcCA Ccaplete 8/20/54 (ase.1) 2cading evaluatica Minisfal p:st-irradiatien fuel Ccuplete 4/29/84 107 4.2 marveillancs progran Ga&line thammi conduc'tivii:y Ccuplets 6/29/84 108 4.2 aguatica Ccuplets 8/20/84 109a 4.4.7 1NI-2 Itsu II,.F.2 Ccaplate 8/20/84 109b 4.4.7 1NI-2 Itsa II.F.2 110m 4.6 Functional dusigt of reactivity Ccapiste 8/30/84 (ase.1) a.scL systems 11 2 4.6 functicnal desigrt d reactivity Ccaglets 8/30/84 (Rev. 1) w.sLi systans lila S.2.4.3 Preservice insgecticn ;sn:x;ran Ccuplets 6/29/84 l (ccagements within reacter grassure L M
- 7) 111b 5.2.4.3 Preservice inspecticn program Ccaplete 6/29/14 j
I (ccupcments within reacter grassuru i l boundary) lile 5.2.4.3 Preservice inspecticn psw Ccaplete '6/29/84 (cesgenents within reactoe ;rassure h y) 112a 5.2.5 Reacter ecolant pressure bcundary Compleen 8/30/84 (ame.1) 1eakage detectica l 112b 5.2 J anactor ecolant grossare b:undarf C=uplets 8/30/84 (asv. 1) l leakaqp detectica i l t -a t WD to a n i
l JtrSODENT 1 (Ctat'd) R. L. MITI: DEER A. SOBENQ CPSI SECT!DI ITBt IEMBER SGUECT SDtfUS I2TITit mn ~ 112c 5.2.5 emar+ar v12r* Exessure b:x:ndary Ca plete 8/30/54 leakage detection (ase. 11 H2d 5.2J Desctcr a:clant pressure bcundary C:nglets 8/30/9 4 Jaakage detectias, (Rev. 1) 112e 5.2.5 ppectcr m 12re pressure teundary C:aplete F30/34 leakage detectica (ase. 1) 113 5.3.4 GE g+:-M agglicability Ctaplets 7/18/84 U4 5.3.4 ' M14*rx:e with NB 2360 cf the Summer C:nglets 7/18/114 1972 M$enda to the 1971 Ase: Cbda 115 5.3.4 tat:p wight and Charpy v-notch tests C:mplate 9/5/84 fu closure flange materials (ast. 1) 116 5.3.4 Charpy v,et.ch test data fa base C:nglets 7/18/5 4 materials as used in she n course No. 1 117 5.3.4 ompliance with NB 2332 cf Winter 1972 C:nglete 8/20/34 Addenda cd the JdiME C:de 5.3.4 Imed facters and neutren fluence fx C:nglets 8/20/54 118 ausmillanna capsules US 6.2 Dt! itan U.S.4.1 Caplets 6/29/8 4 120e 6.2 1MI Itma. n.E.4.2 Caplete 8/20/84 120b 6.2 ' 1MI Itse U.E.4.2 C:spiste 8/20/34 121 6.2.1.3.3 Use cf NtatID-0588 Caplete 7/27/54 7/27/ 14 1 122 6.2.1.3.3 Tuperature ; refile C:splete 123 4.2.1.4 matterfly valve eseratien (pest Ctzglets 6/ 29/114 accident) l ? i
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r238 Ntsamt SR7ECT 1 SDtmB t.EFTEst E3tr2D WW SE22t3 caplete 8/20/54 4.2.1.5.1 mv = Maid anruhas analysis (Env.1) 124a 4.2.1J.1 Erv ahield anrulus analysis Me 4/20/56 (Eme.1) 12e 124c 6.2.1.5.1 Rev shield annulus analysis empInte 8/20/84 (Dev.1) 125 6.2.1.5.2 casign drp 11 head differential omqdets 6/15/54 pressure 126a 6.2.1.6 pedundant p:mitiat inflest.ws fer Ctaplete 8/20/54 vacuum breakars (and centzel reca alarms) 126b 6.2.1.6 andundant pcsitdat indicators fee cuplets 8/20/84 vastas breakers (and centzel reos alacas) 8/20/84 cperability testing ci vacam treakers Complat's ( Asv.1) 127 6.2.1.6 128 6.2.2 Air S tien Omplets 7/27/54 129 6.2.2 Insulation ingestien Ccaplets 6/1/54 130 6.2.3 Pctantial bypass leakage paths Ctaplet* 9/u/m f1N .%istraticn cd seccndary contaire. Caglets 131 6.2.3 ment cpenings 132 6.2.4 E-r*L a isolatica review Ctsplete 6/15/54 caplete 8/2V34 133a 6.2.4.1 hr =L
- purge systna Czplete 8/20/34 133b 6.2.4.1 contairment surge system Complets 8/23/84 133c 4.2.4.1 C
- stainment guzgo systw
o ~ I t: ATDOpert 1 (Cmt'd) R. L. Mrr!L 'IO DEER A. SO N NCER GEN SRCFIGI 5"DdUS IIIT!R DM1!D ITD0 NUMEER SUETDCT 134 6.2.6 Ce=truert leakage testing Ccaglete 6/15/84 1 25 6.3.3 IPCs ard IPCI injecticri valve Ccaglete 8/20/84 intericcks 136 6.3.5 Pla.a.-WJic ICOL (see Sectiers Ccagleto 8/20/84 (Rev. 1) 15.9'.13) 137a 6.4 Ccritrol reas habitability Camplete 8/20/84 137b 6.4 Centrol reca habitkility Ccaglete 8/20/84 137c 6.4, Ccritrol room habitability Ccaplete 8/20/84 138 6.6 Preservice irspectica gregram for Ccuglate 6/29/84 Class 2 and 3 ccagcziones 139 6.7 WIV leakage centrol systen Ccuglets 6/29/84 Ccuplate 9/7/84 140a 9.1.2 Spent fuel pcol stxrage (Rw. 2) ^ Caglets 9/7/84 140b 9.1.2 Spent fusi pol storage (Ray. 2) ccupleta 9/7/84 9.1.2 Spent fuel peci storage (Rev. 2) 140c Couplete 9/7/ 84 140d 9.1.2 Spent fuel geol stcrage ,Rev. 2) ( 1414-9.1.3 Spent fuel ecolirq ard clearup Ccaplete 8/30/84 (Rev. 1) systen 141b 9.1.3 Spent fuel cec 11rq ard clearup Ccuplete 8/30/84 (Rev.1) system 141c 9.1.3 Spent fuel peal ceeling aid clearup Cczplata 8/30/84 (Rev. 1) systen t e M P64 80/1213 - gs ~ m --,-_-._._,....,_...--__-..........,,,,__m., .,m-.-_--w-,-
9 EncueNr 1- (Cent'd) R. L. METIL M DSSL A. SQ5EN2R CFEN SEI:TIQI r!DE NtM55t SUMECT STAMS IETI15t CG!D 141d 9.1.3 Spent fuel Ipol cooling and cleamp Canplete 8/30/84 (Rev. 1) system 141e 9.1.3 Spent fuel pool cec 11m ard clearup Ccapista 8/30/84 (Rev. 1) system Spent fuel pol cooling and cleamp Ccuplete 8/30/84 l 141f 9.1.3 5 (Rev. 1) system ~ 141g 9.1.3 Spent fuel pool cooling ard clearup Ccaglets 8/30/84 (Rev. 1) system 142a 9.1.4 Light Iced hardling systen (related Couplete 8/15/84 (Rev.1) to refueling) 142b 9.1.4 Light lead handling system (related Ccmpleta 8/15/84 (Rev. 1) to refueling) 143a 9.1.5 overhead heavy lead handlig Complete 9/7/84 143b 9.1.5 overhead heavy 1 cad hardlig complete 9/13/84 144a 9.2.1 Staticn service wtar systen Ccaplete 8/15/84 (Rev. 1) 144b 9.2.1 Staticn service water system Ccagiate 8/15/84 (Rev.1) 144c 9.2.1 Staticn service water systern Canplets 8/15/84 (Rev. 1) l 145 9.2.2 ISI program ard functicnal testing Closed 6/15/84 cd! safety ard turbine alxiliaries (5/30/84-Aux.sys.Ntg.) e cooling systame 146 9.2.6 Switches and wiring.taanHated with Cicaed 6/15/84 HPCI/ItCIC t=rus sactica (5/30/84-Auz.Sys.Mtg.) l L f M P64 80/12 14 - gs
.. a d' ani '61 (Cent'd) R. L. MITIL 1D ~ DMR A. SQBelt3R WW m"MN ITBI M3eER S BJECT SDt!ts IZrTBtCNTED _ I 147a 9.3.1 Ccagensand air systes Ctzplats 9/21/s4 (Rev. 2) Ccmqt2sts 9/21/s4, 141b 9.3.1 M air systems (Rev. 2) 147c 9.3.1 Ccapressed air systems ccup2sta 9/21/s4 (Rev. 2) Ctagressed air systmas ccaplats 9/21/84 147d 9.3.1 (Rev. 2) les 9.3.2 Post-accident 714 rig systen Ccuplets 9/12/s4 (w. 1) (II.a.3) lea 9.3.3 squipment and ficce drainage systma Complets 7/27/84 leb 9.3.3 Egaigment ard ficcr drainage system Caspista 7/27/54 150 9.3.6 Primary centainment instr.unent gas Ccaplets 8/3/84 (Itsv.1) system 151a 9.4.1 Centrol structurn ventilatien system Ccaplets 8/30/84 (Rev. 1) e i 151b 9.4.1 Centro 1' structure ventilaticn system Cceplats 8/30/84 L (Bev. 1) C1cond 6/1/34 132 9.4.4 Pdelvity :ncnitcring elssents (5/30/84-Auz.sys. Meg. ) 153 9.4.5 Engineered safety featuras ventilm-Ccaplets 8/30/84 (Dev 2) tica syntas 154 9.5.1.4.a Metal rect &ck ccnstr.actica Cczplets 4/1/ 84 classificiat.fsn NIIC Actica 155 9.5.1.4.b ongoing review cf safs shutdbe P "tY Mic Action 9.5.1.4.c onping zuview cf alternata stutinn 154 p_ Hty
o ' ~. -. J.: : .s ' t. ~ ~ ATDopert 1 (Cmt'd) R. L. MITIL TO MiER A. SONNCER GEN SECTTQI STATUS U. ITR DMTD T ITDI NOMIElt SUETECT 157 9.5.1.4.e Cable tray pmien Caplete 8/20/84 158 9.5.1 J.a Class a fire detectim systse cm plete 6/15 / 84 4 i. 159 9.5.1.5.a Pri: nary and d4 pcwe s2pplies Caplete 6/1/84 fm fire detectim system 5 160' 9.5.1.5.b Fire water pLap capacity Casplete 8/13/84 161 9.5.1.5.b Fire wter valve sapervisicrt Caplete 6/1/84 i Caplete 6/1/84 162 9.5.1.5.c Deluge valves 9.5.1.5.c Marual hose station pipe sizing Cmplete 6/1/84 163 9.5.1.6.e Renete stutdows panel ventilaticut Camplete 6/1/84 164 165 9.5.1.6.g E2mergency diesel generator day tar
- Casplete 6/1/84 protection 166 12.3.4.2 Airborne rwmivity :ncrtitar Ccuplete
'9/13 /84 (Bav. 2) positiating 167 12.3.4.2 Portabla centitucus air a:rtitors Complete 7/18 /84 168 12.5.2 Equignertt:, trainirg, ard pe-- tm Caplete 6/29/ 84 fcr iglant iodine instrantaticn 169' 12.5.3 Guidance of Divisicri B Reg.tlaterf Ccuplete 7/18/84 mifA 170 13.5.2 hh generatim package Couplete '6/29 / 84 i sdamittal 171 13.5.2 TME Item I.C.1 Caplete 6/29/84 Ccaglets 6/29/84 172 13.5.2 PG Ccamit=ent 173 13.5.2 PWires coverirg atmormal teleases Caglete 6/29/84 d! radioactivity I M P84 80/1216 - gs ) l
.. i t a AfDCBUNF 1 (Cbnt'd)_ R. I NIITI. i D88R IIBI PEMB8R StBJECT SD2V5 IZrTER IRE: 174 13.5.2 Desoluticn explanaticn in FSAR of Cbsplete 6/15/84 130' Items I.C.7 and I.C.8 175 13.6 Physical security Cten 176a 14.2 Initial plant tsst program cesplats 8/13/14 3 : 175b 14.2 Iriitial plant test program c: splats 9/5/14 (Dev. 1) 176c 14.2 Initial plant test pregram Ccaplete 7/27/84 176d 14.2 Initial plant test program Ctuplete 8/24/84 (Rev. 2) 176e 14.2 Initial plant test program Ctaplete 7/27/84 1762 14.2 Initial plant test progra Ctaplets 8/13/54 176g -14.2 Initial plant test pecgram Ctsplets 8/20A14 176h 14.2 Initial plant test pro;rma Ccaplete., 8/13/14 1761 14.2 Initial plant test pregram Ccuplets 7/27/1 4 L 177 15.1.1 Parti $1 feedheter hasting Ccuplete 8/20/1 4 (Rev. 1) 178 15.6.5 IICA resulting f::ca spectnza of NBC Acticn ,=1*M piping breaks within IU 179 15.7.4 Dadiological wqm cf fuel NRC Acticn handling accidents NRC Acticn 180 15.7.5 spent fuel cask drep accidents 181 15.9.5 3tI-2 Ita II.K.3.3 Ctaplete 6/29/84 182 15.9.10
- DtI-2 Itas II.K.3.18 Casplete 6/1/54 183 18 scge Creek DC3cR Ctaplets 8/15/14 M N 90/1217 - es i
1 --n--T 1 (ccrit'dt R. L. NETW. 2 EMS A. - esi S2EN 515trt5 12TTIst attED SER E T temen legga 184 7.2J.1.e Failures in reacter vassal Insel Ctzplete 8/1/84 (Rev 1) sensig lines 135 7.2.2.2 Trip syntes sensors and cablig in Cceplats 4/1/84 turbins h=4MN 186 7.2.2.3 Testanility d plant grttectica complete 4/13/84 (Rev. 1) systems at peuer 187 7.2.2.4 I fting d Isads to perfoca surveil-Ccagiste 8/3/54 lance testing Camplets 4/1/84 188 7.2.2.5 Setpoint W 1 m i t= 8/1/84 f ISS 7.2.2.6 Imlatim devices Comp 1ste 4/1/84 190 7.2.2.7 Dagulatory Guide 1.75 191 7.2.2.8 Scram dischazgo volume Ccuplete 6/29/84 192 7.2.2.9 Reactor :mde stitch Ccaplate 8/15/84 (Anv. 1) 193 7.3.2.1.10 Manual initiatias d safety systems Ccaglets 8/1/34 194 7.3.2.2 Standard review plan deviatiens ccupista 8/1/84 (Est 1) 195a 7.J.2.3 Freese-;tecectienbetar filled ccuplata 8/1/84 instrument and samplig lines and caninst tasserature cente. 195b 7.3.2.3 L - g e m % tse f111sd C=mplate 8/1/84 instement ard suppling lines arsi caninst ensgerature ccmezel Casplete 8/1/84 194 7.3.2.4 Sharing d ccumn iratn:mont taps Miw =~, multiplazer ard Ccuplate 8/1/94 e (Ret 1) 197 7.3.2.5 cragutse systems M ~5H 'm/12 13 - p
t' r 1 h 7,. I 1 (Cbnt'd1 R.' L. METE.1C ^ M A. enemu GEN SET!Of 515t!GB IETT25t D32D WMT ITIN 43EER ISS 7.3.2.6 1NI Itsu II.E.3.lS-ADS actuatica Ccsqdete V20/84 199 7.4.2.1 II hallatin &27-Issa d, nezeclass Cagdets 8/24/84 (Rev. 1) II inammarentica ard e. J pcuse system tus dring cgen*D Couplate 8/8/84 200 7.4.2.2 Besets stutdown system (Der 1) 201 7.4.2.3 BCIC/EPCI LWfons Caglets 8/3A4 202-7.5.2.1 Inval w errers as a result C:splete 8/3A4 cd enrizermental tamperatura affects cn level instzusentatica reference Ing 203 7.5.2.2 P=ca'h W Guide 1.97 ccupiste 8/3/84 204 7.5.2.3 'Dt! Itsa II.F.1 - Accident senitcrirq Ccuplete 8/1/84 205 7.5.2.4 Plart grocess ccuputer systzus Ccaplate 6/1/84 l 206 7.6.2.1 algtt gressure/lcw gressure intericcks Ccuplate 7/27/84 8/24/84 HEas ars't ccnsequential centrcl systen C=giste 207 7.7.2.1 (Rev. 1) failurus 208 7.7.2.2 Italtiple centrol systs is11uras CagInts 8/24/84 (Rav. 1) 209 7.7.2.3 Czedit for neresafety related systans ccuplace 8/1/84 (Bev 1) in Chapent 15 d the F5AR 210 7.7.2.4 Tranniere analysis m:ording systus complets 7/27/11 4 i c.ast red drim structural satarials Cczplate 7/27/84 211a 4.5.1 7/27/84 Centzel red drive structural satsrials Ccmplate 211b 4.5.1 7/27/84 C <. awl red driw structural seterials Cespists 211c 4.5.1 j l I 9.
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R. L. M m L 4 em st22t3B ' A. stamIGB issess smact snm3 Intuit tutts ~ ~ ~ 21M. 4J.1 Qatrol med ein structural materials Quptsee 7/27/54 211e 4J.1 catrol roi ein structural materials Cagdets 7/27/54 212 4J.2 ansetz intmensis materials Q@ 7/27/54 8 213 5.2.3 Ensctor conlant gressure %==*==y MW 7/27/54 e l M ~ 214 6.1.1 mgineered smisty features matarials Q:qdses 7/27/94 i 215 10J.4 lain staan and w=r systan Ompista 7/27/94 j ,,,,,s t. 1 218e 5.3.1 Beacter immoni smearials Quplets 7/27/54 216b 5.3.1 anacter vassal untarials Qugdete 7/27/54 217 9'.5.1.1 Fire p 2 2'= M *=*'% Ozqdats 4/15/54 213 9J.1.1 Fire hasards analysis Cagdate. (/1/34 [ 219 9.5.1.2 Fire a +7i - akinistrative cugdate 8/15/94 r 1 autrels 1 a I 23 9.S.I.3 Fire brigade and fire brigade cmplets 8/15/54 t_ymi nim, 8/1/94 221 8.2.2.1 Physical =parecim.af offsita Quplets transmissica lines 1 222 8.2.2.2 Design guarisions far a *11ste caglets 9/14/84 (Rev. 1) meet at as aftsite peur marce i 223 8.2.2.3 Independamm of attsita circuits cuplate 9/21/84 between the swi4M asui class II (Rev. 21 f i busse c 224 8.2.2.4 Ozman failure made between casite Czqdate 9/21/84 i and offsita pmer circuits (Rev. 1) l
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M A. 9na"* WW. W2EN 23 8.2.3.1 TestabilitF d atomatic transfer d ccuplets 9/21/84 <a.v. 13 pouer fzco the noensi to preferred pouer source Ctuptste 4/13/84 23 8.2.2.5 Grid **411ty (now. 1) 6 227 8.3.2.4 CMtr and capshility d effsite ccuplets 8/1/84 circuits 8.3.1.1(1) voltage &g daring' transient cordi-Cesqdete 8/1/34 23 tiens CTate 8/1/94 8.3.1.1(2) Basis 'far using hun voltage wesus 229 anual connectad Iced witage in the Mtage deg analysis Ccuplate 8/1/84 8.3.1.1(3) Clarification d Tabis 8.3-11 230 ccuplets 8/1/84 8.3.1.1(4) crdervoltage trip setpoints 231 Ccaplets 8/1/84 8.3.1.1(3) 1.ced configuration used for the 232 voltage & g analyelm Ccaplete 9/21/s4 233 8.3.3.4.1 periodic systne tasting (Rev. 2) 234 8.J.1.3 cWty and capability d craite ccuplets 8/1/84 Ac pcasar suppline and use ed ae-ministratin centrols to grownt cworicadig cd the diesel generaters Ccuplete 9/21/84 13$ 8.3.1.3 Dissal generators Iced acceptance (Pav. 2) test 238 8.3.1.4 C M ance with position C.4 ed ccuplate 8/2/ 94 m I.s Ccuplete 9/21/84 237 8.3.1.7 Decriptim d the 1 cad sequencer (Rev. 1) 238 8.2.2.7 sagaencing d loads en the dfsits ccuplete 9j,1/04 <p.v, t3 Pouer p tas t
i k Jiremert 1 (Cent'd1 R. L. NETE X M A. - trr!!!m tz nto @E M auuum -7 T!se lessa 23 9.3.1.3 Testirg to wrify 804 minimm Caplete 8/15/84 240 8.3.1.9 Ct @ with Et>F5 >2 camphtm 8/1/94 241 8.3.1.10 Iced acceptance test aftur gelenged Caplate 9j21j84 ( m, 33 no Imad cperatica d the diamel gammastor 9/13/84 242 8.3.2.1 Cay 11anas with penitfen 1 W Dagpla-Cenplate (Rav. D tory Gaide 1.128 Pr*% ce gatification d Class, Campists 9/13fg4 243 8.3.3.1.3 la egdpment from the effects d (gey, 13 fise sappressica systamos 244 8.3.3.3.1 Analysis and test to demonstrate Ccaplate 9/13/g4 (Rev. 2) adapaacy d less thart specified. separatica 245 8.3.3.3.2 'Ihn uso d la veaus 34 inches d ccupleto 9/24/84 (Rev. 2) sepasstlen he m raceways 244 8.3.3.3.3 Specified separatica d raceways by Ccaglate 8/1/34 analyst's and test 247 8.3.3.3.1 capability d penetraticns to with-Cengtsta 9/13/84 (Rev. 1) stant long daratim shcrt circuits at lass than===4== ce wesst came short circuit 248 8.3.3.5.2 Segerstlen d penetratica primary Caplate 8/1/84 and tash, pectacticns 2e 8.3.3 J.3 'the uso d bypassed thermal cariced Caplate 8/1/84 ,. x m in devices tw penetration protecticas 250 8.3.3J.4 Tustirg d tuses L1 ed.a:n with C=mpleem 3/1/84 R.G. 1.82 14 4=.
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A. someK2R retrus' LETTER Du!ID g 251 8.3.3JJ Fault carrant analysis h au Caphts 9/24/84 (Rov. 3)
- 4se penetration ciztuits 9/21/e4 SJ.3J.4 The uso d a single breaker to proride Ctspiste f
(Rev. 2) 33 paastrat.ica p e M = i 253 s.3.3.1.4 Ozmaitment to pM all Class 12 Qupiste 9/24/84 (Rov. 2A) elpaipment frta external hasards versus only class 1E ognip. ant in one division 9/14/84 Protection d class 1E poser sgplies capista (Rev. 1) 254 8.3.3.1.5. tras failure d W=ti# fed class 1E Iceds b @ ta 255 8'.3.2.2 asttary capacity .t ti.e.,d we=
- t*
w~ gism 2,. 8.3.2.3 adficient battery Wty Justification fx a 0 to 13 M W 82 hkk 13 8.3.2 J 257 1 sed cycle i i 8.3.2.6 Desigt and qualification of DC caplats S/1/34 254 systen Icaen to cperate tweneen minima and anximum voltage levels i Use of an imortar as at isolatica c:splets 9/24/84 (Rov. 1) 239 8.3.3.3.4 devia 9/13/84 280 8.3.3.3.3 cas of a single breaker tri; ped bf cmqplets (Rev. 1) a tcch siqpal uend as an isolatica l devie l 2s1 8.3.3.3.4 htamatic % . ed Icede and caplets 9/13/84 l (Eev. 1) interamnaction totween refundant divisions W 8" N 252 11.4.2.d solid wesen omtrel W-I 1 M iv 30/12 23= SS
.= ~ l mencoeur1 (Omt'ot R. L. Nt!!L 1 Im A. sasauas l cras.
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, srzras tama nunt rues amesa 7 253 11.4J.e Fire p-J 7 = der solid radmasts. Quplete V1.V54 8 samenge area Quplete V20/14 i 254 4JJ sesens ed amygum, 285 6.8.1.4 EW. Filter Testing Quplete V13/14 Ouglets 8/13/44 I 266 4.4.1.4 Fiald leak tests 257 6.4.1 Omtrol razas toda chemical Caplets 5/13/44 detacenes i 4 ALr 51tration W E & mins cupleta g/13/g4 'b' 2dB ads cases >242 and >242-1 ' Quplate ~ 269 5.2.2 caplete V20/14 270 5.2.2 Qde case > 252 clamare of wtsrtight d:x$rs to safety-cpan l . TS-1 2.4.14 related structures Single ruebteim Iccp cperatica Cten 15-2 4.4.4 6/1/54 Que dow acni$cring fer crud offsets Caplets TS-2 4.4.5 15-4 4.4.4 tm as parts monitoring systen Cpen 13 S '4.4.9 Natural cizculatica in neenal Cpen opeastian sessadary amtainment negati'm Cpen Tb4 4.2.3 9"**** 29-7 4.2.2 mi d =r= and draudeus time in open W omatainment 1>4 6.2.4.1 tankage intsgrity testing ogsn 15-4 4.3.4.2 333 maksyntas periodic w eit open tu 0 4 44 f ,ne-,-.-- - -,., ,_--_-n-...nn.--,an_~, .,_+-,n, .m.
,r s 4 t . ar=were 1 (omt'd1 R. L AtIEL M A. SWWEEE ces m sena 12r!ER tums:
- ga7tcr reme g3ega 19 10 6.7 IsIT 3aakage rats Araumbuity, -V5**, and testing Cput 13-u 15.2.2 of tuntdne Iqpass systen l
19 12 15.4.4 Prisery coolant activity fuel red internal pressure critsefa capim M Ig-1 4.2 Ic-2 4.4.4 stakulty analysis simittad badme cpan assand cyc.te cgaration i I t i r i l ? i I a i I i ( I
s l' AT'1AGNENT 2 DATE: 9/24/84 DR5FF SEA SECTIONS AND DATES PROVIDED SECTION .M M ,gg1 l 3.1 11.4.1 See Notes 145 3.2.1 11.4.2 see Notes 165 3.2.2 11.5.1 See Notes 145 5.1 11.5.2 See Notes 145 5.2.1 4. 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 l 4.3.1 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 l 8.4.2 See Note 2 13.5.1 See Note 4 8.4.3 See Note 2 15.2.3 8.4.5 See Ncte 2 15.2.4 8.4.6 See Note,2 15.2.5 l l 8.4.7 See Note 2 15.2.8 8.4.8 See Note 2 15.2.7 9.5.2 See Note 3 15.2.8 9.5.3 See Note 3 15.7.3 See Notes 1&5 9.5.7 See Note 3 17.1 8/3/04 i 9.5.4 See Note 3 17.2 8/3/84 10.1 See Note 3 17.3 8/3/84 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 3&S 10.4.3, see Notes 345 10.4.4 See Note 3 11.1.1 See Notes 145 Notest 11.1.2 see Notes 165 11.2.1 See Notes 145
- 1. Open
- items provided in 11.2.2 see Notes 145 letter dated July 24, 1984 11.3.1 See Notes 145 (Schwencer to Nitti) 11.3.2 See Notes 1&S
- 2. Open items provided in June 4,1984 aceting
- 3. Open items provided in April 17-18, 1984 seating i
CTidh
- 4. Open items provided in Nay 2, 1984 seeing l
S. Otait SER 5ection provided l in letter dated August 7, 1984 (Schwencer to Mittil
D2tqJ 9/24/84 DSER~ 'DSER SECTION SUBJECT -ITIDi 245' RB.3.3.3.2 ~The use of 18 versus 36 inches of separation between raceways 251 8.3.3.5.5 Fault current analysis for all representative-penetration circuits. 2'531 ^ 8.3.3.1.4 Commitment to protect all Class lE equipment from external hazards versus only. Class 1E equipment.in one division . 259L . 8.3.3.3.4 Use of an inverter as an isolation- ' device s 9 1 W f W T y- -,e.we-wr --
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4 4 ATTAcausar 4 l l
kOW D 2fo s 3 3 2) DSER Open Item No. 245 (DSER Section 8.3... ETWEEN RACEWAYS THE USE OF 18 VERSUS 36 INCHES OF SEPARATIO l of the FSAR it is In Sections 1.8.1.75 and 8.1.4.14.3.1ble trays in the
- i.
stated that separation between redundant ca l y room, cable spreading area, control equipment room,, re a i rtically and main control room are separated by 18. inches ve i f separation as opposed to the recommended 36 inches o 384-1974. required by IEEE Standard indicated that this l by Amendment 4 to the FSAR, during the
- 18. inches of separation was approved by the staf f The applicant, The staff's k plant.
preliminary design review of the Hope Creefor this item states f, 2 preliminary safe ty evaluation report that: 11 claims these separation distances are cabling will be 'The applicant ty pe adequate because a high gradespecified and results h that no cable degradation or flamelower tray, separated by 12 i i tes." tray, is exposed to a gas flame for 15 m nu j et i degradation The results of these tests, that demonstrate no h above the tray ex-to cables located in the trays 12 inc esposed.co t h applicant.
RESPONSE
to Question 430.51 8.1.4.14.3.1' a nd c[t.he re spo ns e I justification for have been revised to provide additional Se ction the separation distance. r l \\. t-I l j'~ l DB/em J . F70(8) j e . j{& e's H.
7 [htrw 26 OSER 24 C b.L HCGS FSAR QUESTION 430.51;(SECTION 8.3.1 and 8.3.2) In Sections 1.8.1.75'and 8.1.4.14.3.1 o'f the FSAR you state that . separation between : redundant cable trays in the cable spreading area,' control equipment room, relay room, and main control room are separated tur 18 inches. of separation required by IEEE Standard 384-1974. Provide analysis substantiated by test that demonstrates the adequacy of 18 inches of separation.. RESPONSE-The HCGS~ PSAR was approved with 18 inch ver tical separation between redundant cable trays. ' A copy of the test repor t that substantiated.the use of this vertical separation has been submitted under separate cover (le tter f rom R. L., Mitt,.PSE&G, to A. Schwencer, NRC, dated August 15, 1984). ' Revised section 8.1.4.14.3.1 provides the analysis based on this -test to demonstrate the adequacy of 18 inches separation. InL addition to the above test, an additional cable tray test will be performed that; tests shorting of electrical cabling utilizing the 1811nch ' vertical separation. This test plan is being submitted under separate cover. i: DSER'OPEN ITEM'245 Rev. 2 430.51-1
gey, L -~ i-NCGS FSAR 4 And 3.1.4.14.3.1 Cable Spreading Area, C ntrol Equipment' Room 4 R:fla. h, and Main Control Room andM W maf-Ht m e. E The cable spreading area, control equipment room
- N b c~e, and sanin control room do not contain high energy equipment such as snel' chgear, transformer, rotating equipment, or potential sources t
of missiles or pipe whip, and are not used for storing flammable materials. Power supply circuits are limited to those serving These 208/120-V power these areas and their instrument systems. Conduits containing redundant cables are installed in conduit's. cables are separated by a minimum of 1 Lpch. Conduit couplings, clamps, locknuts, bushings, etc, shall mot be considered in For conduits determining the required separation distances. carrying redundant neutron monitoring cables, boxes also shall not be considered in determining the required separation. Redundant cable trays ~are separated by at least 18 inches vertically and 12-inches horizontally. The configurations, for l ' : -traj: can not be separated by which the redundant { distances specified above, will either be analyzed or tested to demonstrate the compliance with the. intent of Regulatory Guide Separation distance requirements between Class 1E and non-1.75. Class 1E raceways are the ame as for the separation among redundant channels. % KSEAT A Strict administrative control of operations and maintenance activities is developed to control and limit the introduction of potential hazards into these areas. 3.1.4.14.3.2 Limited Hazard Areas Limited hazard areas are the general plant areas from which l potential nonelectrical hazards s'uch as missiles, pipe whip, and L The hazards in this area are exposure fires are excluded. limited to f ailures or f aults internal to the electrical ,These areas include elevations 77, 102, equipment or cables. 124, ~130, and 137 feet in the auxiliary building wing areas and Minimum separation in elevation 87 feet in the radwaste area. these M..g.i-yjreas is as follows: Conduits containing redundant cables are separated by a a. minimum of 1 inch, unless consideration of haz'ards indicates greater separation is required. Conduit couplings, clamps, locknuts, hushings, etc, shall not be considered in determining the required separation For conduits carrying redundant neutron dista7ees. M[ g1 DSER OPEN ITEM 3.1-21 Amendment 4 l
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05R 245 .(I M 20 ABSTRACT-PHYSICAL SEPARATION TEST PLAN 1.0 SCOPE 'Thisjdocument is a test plan for the purpose of testing physical separation between redundant Class '1E cables and Class lE and non-Class lE cables with respect to electrical f aults in configurations representative of: HCGS. 1.1 OBJECTIVE. The purpose of this procedure is to present the requirements, procedures, and sequence for testing the design adequacy of the Hope Creek cable tray-to-cable tray separation. Figure 1 identifies the-tray-to-tray separation test configuration. 1.2. APPLICABLE DOCUMENTS IEEE Std 384-1981 IPCEA S19-81 .HCGS FSAR Section S.1 1.3 EQUIPMENT DESCRIPTION This test procedure encompasses testing of control cable and instrumentation cable as described below: Item No. Description 1.0 Okonite 600VAC, two= conductor, size # 14 AWG (HC No. CO2) 2.0 Okonite 600VAC, two conductor, size'# 12 AWG (HC No. P12) 3.0 Ea to n - 600VAC, two conductor, size # 16 AWG (HC No. 102) - 2. 0, TEST REQUIREMENTS i 2.1 Acceptance Criteria -2.1.T Insulation Resistance Tes t ~ Measured insulation resistance on all " target" cables and any' other cable, in the targe t raceway, that might sustain significant damage to its insulation system shall1he greater than 1.6 x 106 ohms with an applied potential of 500 VDC for sixty (60) seconds.
.I'- D56X 14f (I b u) Page two 2.1.2 -High ' Potential Test There shall be no evidence of insulation breakdown or flashover with: an ' applied potential of 2200 VAC for sixty (60)' seconds on all " target" cables and any other cable,- in the target raceway, that might sustain significant damage to'its-insulation system. 2.1.3 - Cable-Continuity Test Energized non-f ault specimens in the " target" raceway shall. conduct 100% rated current at 120 VAC throughout the overcurrent test. N C 9 4 r
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N1 an Rev. 3 DSER Open Item No. 251_ (DSER Sesc' tion 8.3.3.5.5) FOR ALL REPRESENTATIVE PEHETRATION CIRCUITS FAULT CURRENT ANALYSIS the applicant indicated that coordi-By Amendment 4 to the FSAR, nated f ault-current versus time curves for representative pene-tration conductors and their protective devices are included in Based on a review of these figures, Fig ure s 420.46-1 of the FSAR.
- urves for motor dit-representative the staff concludes that trans former, and instrumentation circuits forential relay, current were not included in Figure 430.46-1.
Inclusion of thase circuits the coordinated fault-current as well as other circuits such thatcurves is representative of all penetration circui versus time will be pursued with the applicant.
RESPONSE
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MT HCGS FSAR 1/84 I~ OUESTION 430.46 (SECTION 8.3.1 and 8.3.2) Section 8.1.4.12 of the FSAR simplies, through the use of the term " penetration conductor", that primary and backup circuit protection is provided~to protect the circuit conductor versus the penetration. This design for. containment electrical penetration protection does not meet the guidelines of position 1 of Regulatory Guide 1.63. Position 1 requires primary and backup protection where maximum available fault-current exceeds the . current-carrying capability of the penetration versus capability of the conductors. a. Provide justification for noncompliance with the guidelines of position 1 of Regulatory Guide 1.63. b. Provide coordinated fault-current versus time curves for each representative type cable that penetrates primary containment. For each cable, the curves must show the relationship of the fault carrying capability 'between the electric penetrations, the primary overcurrent-protective device, and the backup overcurrent protective device. c. Provide the test report with results that substantiates the capability of the electrical penetration to withstand the total range of time versus fault current without seal failure for worst case environmemental conditions.
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NT 1 =HCGS FSAR l 8.1.4.10 Reculatory Guide 1.53, Application of the Sinole Failure Criterion to Nuclear Power Plant . Protection Systems, June 1973 The~ electric power system is designed to comply with Regulatory Guide 1.53 as discussed in-Section 1.8. All four Class 1E power system channels are designed and located in accordance with the separation ~ criteria for the plant. Routing of cables and location of equipment is designed so that a failure of any kind in any channel cannot propagate to any other redundant channel. t Consistent with the single failure criterion, only one failure is assumed to occur in the system following a DBA. 8.1.4.11 Reculatory Guide 1.62, Manual Initiation of Protective Actions, October 1973 HCGS complies with Regulatory Guide 1.62 as discussed in Section 1.8. 8.1.4.12 Reculatory Guide 1.63,' Electric Penetration Assemblies in Containment Structures for Licht-Water-Cooled Nuclear Power Plants, July 1978 Design of HCGS penetration ~ assembly systems is in compliance with Regulatory Guide 1.63, wi% fhe exceptions inJ2c,Jed m g3 h,and y 8-h elcW. /\\ The. types of circuits that'go through penetration assemblies are as follows: Power feeders for medium voltage 3.92-kV motors .a. b. Power feeders for 480-V ac motors 480-V ac and 208-V ac miscellaneous power feeders c. ~d. 120-V ac control circuits .e. 125-V de' control circuits 8.1-12
t .e 1/34 HCGS F3AR s i 120-V ac lighting circuits f. Motor differential relay current transformer circuits s g. Low voltage instrume,ntation circuits h. 1 Communication circuits. 1. The following system features are provided to ensure compliance f with the Regulatory Guide position on single random f ailures o circuit overload protection dev' ices: The only. medium Medium voltage penetration assemblies: voltage circuits a. 3.92-kV circuits for the two reactor recirculation pump Each motor is supplied from a variable The maximum fault motors. frequency motor-generator set. current available for a fault inside the containment is limited by the generator contribution and the circuit resistance. pg.\\M Av.y A yp gg cy u p p g.o-n;;.cytm pog. L THE 1000 V.cmi.1, PE.MEit.Aitou ts Ptov10ED BY TWO h CLAS S lE CIECUlT BR.E AKER5 1M SE E.lE S AS SHOWM 15 I N F S AR F g cqq g, 3. [+ E A CM C12.C.u lT S E.EAK EE U*DED Wi"iM AM oVE.2.c um.tt.E wT c.s.t Ay. vy c se ~ hIb M S A7 E. SG.T TO Tg t p T H E12. R E S P E CTt yc-ue e *T - se.sxwsss. av g.3.n,,xe,y n 5%CW S T H AT -THE, TIM.E - c.ute M E M T c Ap4pj d y OV TWE. \\oco Y cM FEh!ETE.ATIOM 15 c. E.E AT ER. DAM AW M A%t M UM SM c E.T cig.ct )l 7 c.v g g i 9 5. TIMP.:. C.04DiTt D N yyAT c.o v t.D o c c y g,,, The 480-V ac loads 480-V ac motor feeder circuits: inside the containment consist of Class 1E and non-b. Class 1E motor-operated valves and non-Class IIAll these loads a i g continuous-duty motors. from 480-V motor control centers (MCCs). e E 3 The magnetic-only circuit breaker used in the combination starter for the motor provides primary m protection for penetration conductors. A thersal-5" 3,. dang g
1 j UT \\ i HCGS FSAR 1/84 magnetic breaker in series with the starter breaker- _provides_ backup protection for these penetration. conductors. These primary and-backup breakers used for the protection of penetration conductors are both located in the same cubicle.of the MCC. The primary breaker is set to provide _only short circuit- -protection. It does not provide locked-rotor protection, which is provided by overload relays in the MCCs for non-Class lE motor-operated valves and continuous-duty motors. For Class 1E motor-operated valves (MOVs), the overload relay is bypassed for emergency plant operation to increase the availability of these valves in accordance with Regulatory Guide 1.106. For these Class 1E MOVs, the backup breakers are selected to allow for - sustained locked rotor. current and penetration conductors are selected to ensure that the thermal limits of the penetration are not exceeded during this condition. The thermal-magnetic backup breaker has a nonadjustable trip setting, which is rated on the following basis: 1. The time-current characteristic curve remains -under the thermal damage curve of the penetration conductor over the range of postulated temperatures so that the_ breaker trips on overcurrent before the thermal limit of the penetration, conductor is reached. 2. The breaker allows locked rotor current of non-Class 1E motors for at least 10 seconds and 1000 seconds for Class 1E motors. These breaker settings prevent nuisance-tripping of non-Class 1E motors during starting and allows ample time for the motors to start. c. 480-V and 208-V miscella?ieous, feeders: Non-Class 1E 480-V MCCs provide power for hoists, reactor recirculation pump motor space heaters, and welding outlets in the.drywell. The primary and backup protections for these feeders are provided by two thermal magnetic breakers in series. Both the breakers have the-same ratings and are located in the same cubicle of the MCC. The ratings of both the breakers 8.1-14 Amendment 4
t MT + HCGS FSAR~ 1/84 are selected.so that on overcurrent, the breakers' trip before_the thermal limit of associated penetration conductor.is reached. 208-V ac miscellaneous feeders from a.208/120-V ac power panel provide power for source range. monitoring (SRM).and intermediate range monitoring (IRM) systems. The primary protection for the.208-V ac circuit is provided by fuses in each circuit conductor. These fuses are located in GE control panels ~. The main 20-ampere thermal-magnetic breaker, located in the power panel, provides the backup protection.for these circuits. The time-current characteristics.of both the fuses and circuit breakers are selected so that both the devices trip before the thermal limit of the associated penetratiqn conductor is reached. e d. 120-V ac control circuits: 120-V ac circuits are powered from 480/120-V ac control transformers located in the MCC cubicles._ Two fuses, with the same rating L in series for each circuit, located-in the associated cubicles of MCCs, provide both the primary and backup protection. For a fault, the fuses blow before the thermal limit of the associtted penetration conductor is reached. 120-V ac contr'o1 circuits fed from uninterruptible power supply (UPS). distribution panels are provided with two fuses in series for.each circuit. Primary protection is provided by the fuses located in GE control panels. Backup protection is provided by the main fuse with a rating higher than the primary fuse located in the UPS panel. For a' fault, the fuses blow before the thermal limit of the penetration conductor .is reached. e. 125-V de control circuits: Each circuit powered from the 125-V de control bus in the switchgear is provided with two fuses of the same rating located in the associated switchgear cubicle. These two fuses wired in series provide both primary and backup protection for the associated penetration conductor. Each circuit ~ powered from the control bus in the GE control panels _is provided with a fuse in that panel to ensure primary protection for the penetration 8.1-15 Amendment 4
~ NT 'HCGS FSAR 1/84 conductor. Backup protection is provided by the feeder breaker supplying the control bus. In both cases above, either the primary or backup protection is capable of clearing the fault before the thermal limit of the associated penetration conductor is reached. f. '120-V ac lighting circuits: All lighting circuits going through the penetrations are 120-V ac. Each circuit is provided with two thermal-magnetic breakers in series. . The primary protection for the penetration conductor is provided by breakers located in breaker panels. Breakers located in the lighting panels wired in series circuit with breaker panels provide the backup protection for the penetration conductor. Both the primary and backup protection are capable of clearing the fault before the thermal limit of the penetration conductor is reached. g. Motor differential relay current transformer circuits: The only circuits in this category are the current transformer circuits for differential protection of the reactor recirculation. pump motors. No protection is necessary for the penetration conductors associated with these current transformer leads because the maximum possible relay current for a sustained fault in the medium voltage cable is only 37 amperes. The ampacity of the penetration conductor is 41 amperes. Furthermore, the relay current decays to 1.7 amperes after 80 seconds because of the fault current decrement. These current transformer circuit cables are designated control cables and are routed in separate raceways from power cables. This eliminates ((p/fgj y'/4b the possibility of a short circuit between power and g control cables. w h. Instrumentation circuits: Instrument circuits are all low-energy circuits carrying only a few mil 11 amperes. Also, these circuits are routed in separate raceways from power cables to eliminate the possibility of a short between ,a p,Jysppyment circuits. "The r ent in The instrument circuits C4M not exceed the ampacity of penetration conductors under any faulted condition. In addition, l 8.1-16 Amendment 4 c _.
o NT Insert A: i The-differential relay fails safe:for shorts or opens in the current F transformer circuits. If the differential leads were to short while carrying their normal' load of 3.17 amperes, the differential' relay would_ operate and trip the generator drive motor in 144 i . milliseconds and. the 3.17 amperes ioad would drop down_ to 1.7 amperes in 80 seconds. The penetration is rated for 41 amperes continuously. 1
a D: tG HCGS FSAR 1/84 the. instrumentation circuits are protected from . overloads by primary overcurrent protective devices which are integral with their power supply and by g /l. ' backup overcurrent protective devices located upstream ofthepower.suppliey Y i. Communication circuits - Communication circuits consist of 120-V ac power and signal circuits. Each power circuit has two fuses in series. One. located in the ~ distribution panel provides the primary protection, and another located in a terminal box near the penetration provides backup protection for the associated penetration conductors. Both of these are capable of clearing the fault before the penetration conductor reaches its thermal limi d 8.1.4.13 Reculatory Guide 1.73, Oualification Tests of Electric Valve Operators Installed-Inside the Containment of Nuclear Power Plants, January 1974 HCGS complies with Regulatory Guide.l.73 as discussed.in 'Section 1.8. 8.1.4.14 Reculatory Guide 1.75, Physical Independence of Electric Systems,~ September 1978 HCGS-complies with Regulatory Guide 1.75. Clarifications and 1 exceptions are noted-in Section 1.8. 8.1.4.14.1 General Separation Criteria Electrical equipment.and wiring for the engineered safety feature " systems (ESF), reactor protection system (RPS), and neutron monitoring system'(NMS) are segregated into separated channels / divisions as shown.in Table 8.1-1, so that under DBAs no single credible event is. capable of disabling sufficient equipment to prevent-reactor shutdown, decay heat removal from the core, or mitigation of~ accidents. The ESF systems, RPS, and 'NMS are separated electrically and physically from one another, and each is further' separated into four channels. 'The degree of separation provided is commensurate with the potential hazards in a given area. 8.1-17 Amendment 4
L .w. 1 g Insert B. iJ TheLonly.penetrat'io'ns with; instrument class circuits that are -protected-by a single circuit breaker or: fuse are as follows: Li 1. fVib' ration Monitoring (a) Circuit breaker is 7 amperes. (b) Maximum short circuit current is 0.8 amperes. ic) Penetration is-#16LAWG wire with a continuous rating of,15 amperes. (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 ampere fuse. .(b) Maximum short circuit current is 0.2. amperes. (c) Penetration is $16 AWG wire with a continuous rating g of 15 amperes. z . (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 1 Monitoring System (a) Circuit protected :by a 2.5 ampere fuse. ~ -(b) -Maximum short circuit current <0.1 ampere.' (The 330KA. resistor would limit the short circuitLto 0.1 ampere even if the rest of the circuit impedance:was zero.) (c) Penetration is ' il6 ' AWG wire with a continuous rating of 15 amperes. (d)'.These penetrations have a continuous rating in excess of 150 times the maximum short' circuit current they may be. expected to experience. -4. Thermocouple Circuits (a) Thermocouples cannot-generate =any-conceivable short circuit challenge to a penetration'.- 9 2.
Insert C The P.A.. voice circuits carry millivolt signals only .when'they are actually transmitting.a voice communication. The system-cannot generate any conceivable short circuit challenge to a penetration. 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 on tests, of the penetration conductors have been established 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. Coordinated fault-current versus time curves'for representative penetration conductors and the protective devices are shown in Figures 8.3-17, Sheets 1 to 22. The test report that substantiates the capability of the electrical penetration to withstand fault current without seal failure for worst case environmental conditions has been submitted under a separate cover. The testing of all penetration over-current protective devices will be incorporated in the HCGS Technical Specifications.
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TR e, < 2A DSER Open Item No. 253 (DSER Section 8.3.3.1.4) COMMITMENT TO PROTECT ALL CLASS lE EQUIPMENT FROM EXTERNAL BAZARDS VERSUS ONLY CLASS lE EQUIPMENT IN ONE DIVISION In Se ctio n 6.1.14.3.3 of the FSAR, it is stated that where neither compartmentalization nor the construction of barriers a is possible (to protect Class lE circuits or equipment fr om hazards such as pipe break, flooding, missiles, and fires) an analysis is pe r formed to demonstrate that none of the hazards disables redundant equipment, conduits, or trays. Based on this statement, the staff concludes that at least one of the redundant Class lE systems and components at Hope Creek need not be protected from external hazards. The de sign, thus, does not meet the protection requirement of Criteria 2 and 4, nor the single failure requir ement of Criterion 17 of Appendix A to 10 CFR 50. Justification for non-compliance with criteria 2, 4, and 17 will be pursued with the applicant. 4
RESPONSE
f The response to Question 430.38, and section 8.1.4.14.3.3, have been revised to provide a discussion of protection of Class IE t l systems and components against external hazards *. I
s ) 44 3 ~ - %a +: Rev. LA 1 V .HCGS-FSAR p .OUESTION 430.38-(SECTION 8.3.') and 8.3.2) n ~
- In --Shet ions /9.1 ~. 4.14. 3. 3 of the FSAR-you state'that where neither compartmentalizationinor.the construction of barriers is possible
.(to protect? Class-IE circuits or equipmentLfrom hazards such as pipe break,, flooding, missiles,-and fires) an analysis-is fparformed:to demonstrate that none of.the hazards disables
- D ircdundant equipment,. conduits, or trays.
Based on this statement citcappears that ateleast one of the' redundant' Class 1E systems nnd. components at Hope Creek may not be protected from external- ~hotards. The design, thus,-does not meet the protection ,rcquirement of Criteria 2 and 4 nor the singla failure jrcquirement of. Criterion 17 of Appendix A to 10CFR50. -Justify 1 T < nan-compliance-with Criteria 2, 4,-and 17.-
RESPONSE
i Section 3.5. indicates tha t-class.1E equipment is protected f rom - pos tula ted' missiles by use of-plant, arrangemen t or suitable physica l barriers such that ~a single missile.cannot simultaneously damage a critical. system component and its backup system. This is accomplished Lby. !1oca t ing' redundan t systems. in different areas of the plant or
- separation by missile-proof ' walls.
There are no class 1E electrical equipment and components that can be damaged by missiles generated externally to the plant.- ~ as part of the design basis for Section 3.6.1.1 indicate s tha t, l protection against dynamic.ef f ects associated.with the. postulated
- rupture of piping, a
- single active component failure is assumed to occur 11n systems used to. mitigate the consequences of the postulated' 1 piping'ruptureLand'to shut down the reactor.
A thorough review of nthe' plant using-the design bases provided in Section 3.6.1.1 was conducted and no cases were found where the piping f ailure would _ prevent safe shutdown.(Reference Question / Response 410.~ 2 3 ). pRt.yot rys pattows sunM: p Section 8.1.4.14.3.3-has been revised to (The HCGS separation review -(hazard analysis) confirms that no' I J ~ external hazard originating in a non-safety related system or component cansprevent safe shutdown of the plant, even when the loss of of fsite pcwer and the worst single active f ailure of any -safety related system'or component ls assumed. a tnistre r "a" / .muf
- -~.~~.
1
c. $q Rev. 2P I (DSER 253/ QUESTION 4 30.38) Insert "A" . 'I PITO J GW1MJ UO '3 I The ' design: of. the ' Class 51E electrical power and control systems o. is such that two or more Class lE current interrupting devices are provided for primary ' and back-up protection for all Class lE cable _S ould=an_ external missile or hazard cause a shorting h or. equipment. failure of an _ electrical cable (s) or device (s) (as a result of an
- impact on-a cable raceway or-electrical component) the Class lE protective devices will operate to isolate that specific cable or component.- This design. feature enables HCGS to accommodate a failure of a Class lE electrical component due to an external Lhazard with a single failure in a redundant Class lE train.
The diesel generators, 4.16 kV Class lE switchgear, 480V Class lE unit sutstations' and the '125V de switchgear and distribution cabinets are located in areas protected f rom external hazards. Only the following types of Class lE lsolation. devices are required to operate-to isolate failed components: 1. 70 ampere fuses in 125V de distribution cabinets. 2. AKR - 30 breakers in 125V de switchgear. 3. Molded case breakers in 480V motor control centers. 4. AKR - 50 breakers in 480V ac unit substations. 5.. 20' ampere fuses in instrument distribution cabinents. 6. 300 ampere fuses in 20kVA uninterruptible power source (UPS)_ systems. 1 7. 200 ampere fuses in 10kVA UPS systems. To insure that the above Class 1E overcurrent protective devices required to ensure this capability'will operate correctly, HCGS will -include surveillance testing'of these devices in the Technical Specifications. +.
a
- t39
~ Rev.1A HCGS FSAR 1/84 monitoring' cables, boxes also chall not be considered in determing the required separation, b. In case of open ventilated trays, redundant trays are separated by 3 feet horizontally and 5 feet vertically, respectively. If the redundant trays cannot be separated by the distances specified above, solid covers for-trays are provided as designated in Section 6.1.4 of IEEE 384-1981. Separation requirements between Class IE and non-Class IE circuits are the same as those required between redundant circuits. 8.1.4.14.3.3 Hazardous Areas ~ ~ - These are areas where one or more of hazards such as pipe break, flooding, missile, and fire can be postulated. R;uting of redundant Class 1E circuits or the locating of rcdundant Class 1E equipment in hazardous areas is avoided. The preferred separation between redundant Class IE circuits or cquipment in these areas is by a wall, floor, or barrier that is otructurally adequate to shield redundant raceways from potential- 'hazardsLin the area. Where neither compartmentalization nor the construction of barriers is possible, an analysis is performed to demonstrate that no missile, fire, jet stream impingement, or pipe whip hazard disables redundant equipment, conduits, or trays. In no case, regardless of the distance of physical separation, are redundant equipment cable trays located in the direct line of Oight of the same potential missile source. The plant design for fire protection separation of electrical cables and equipment is reviewed against 10 CFR 50, Appendix R, which is discussed in Section 9.5.1. The NCGS separation review (hazard analysis) confirms that no external hazard originating in a non-safety related system or
- omponent can prevent sa fe shutdown of the plant, even when the loss of offsite power and the worst single active f ailure of any saf ety related system or component is assumed.
iMSERT"N'
F-Vb' d 34 f Rev. lA p (DSER 253/ QUESTION 430.38) y p. 1 Tb S ctw o 4 f.l.4.19.3/3 4 Insert "A" The) design of' the. Class lE electrical power and control systems is such that1two or more Class lE curren't interrupting devices are provided for. primary and back-up protection for all Class lE cable cor equipment. Should an external missile or hazard cause a shorting _ failure of an electrical cable (s) or device (s) (as a result of an impact on a ca.ble raceway or electrical component)' the Class lE protective devices will operate to isolate that specific cable or component. .This design f ea ture enables HCGS to accommodate a failure of _a Class lE electrical component due to an external hazard with a' single f ailure in a redundant Class lE train. The diesel generators, 4.16 kV Class lE switchgear, 480V Class 1E unit sutstations and the 125V de switchgear and distribution cabinets are located in areas protected from external hazards. Only the following types of Class lE isolation devices are required to operate to isolate failed components: 1. 70 ampere fuses in 125v de distribution cabinets. 2. ' AKR - 30 breakers in 125V de switchgear. 3. -Molded case breakers in 480V motor control centers. 4. AKR - 50 breakers in 480V ac unit substations. 5. 20 ampere fuses in instrument distribution cabinents. 6. 300 ampere fuses in 20kVA uninterruptible power source (UPS) systems. 7. 200 ampere fuses in 10kVA UPS systems. To' insure that the above Class lE overcurrent protective devices required!to ensure this capability will operate correctly, HCGS will include surveillance testing of these devices in the Technical . Spec i f i ca t ion s.
Rev. 1
- 41-
- DSER Open Item No. 259 (DSER Section 8.3.3.3.4)
USE OF AN INVERTER AS AN ISOLATION DEVICE By. Amendment'4 to the'FSAR, the applicant indicated that the non- "o - Class lE public address system distribution panel shown on _ sheet 2 - of Figure 8 3-11 of the FSAR is supplied power f rom the Class 1E dc system.through' an inver ter. The applicant further stated that this-inverter is an acceptable isolation device per IEEE-384-1981, Section-7.1.2.3. The staff does not-agree. Test and analysis to demons tra te the adequacy of an inverter as an isolation device will be pursued with the applicant.
RESPONSE
s
- The' response to Question-430.33 has been revised to state that the inverter -will be tested as an isolation device.
In the event that the tests are not successful, the non Class lE loads will be removed or - the cables will be re-routed. f
HCGS FSAR 00EST1oM 430.33 (SECTION 3.3.1 and 8.3.2) Section 8.3.1.1.2 of the FSAR indicates that the Class 1E system Non-Class IE* loads are provides power to non-Class 1E loads. connected to the Cl The single breaker tripped automatically by a LOCA signal.
- tripped by a LOCA signal provides acceptable isolation between Class 1E and Non-Class 1E circuits for the design basis accident However, for other design basis accidents or operating occurrances that do not generate a LOCA signal (such as loss of
- LOCA. offsite power, design basis exposure fire, seismic events, etc), it is the staff concern that a single breaker may not provide Provide an analysis, in accordance with acceptable isolation. 308-1974, that the guidelines of Section 4.9 of IEEE Standard demonstrates that f ailure of anyone or simultanous combined failure of all non Class II loads will not prevent any of the four channels of Class 1E power from performing its safetyT capacity and capability of onsite and of f site power supplies function. and their associated distribution system to supply power to (1) Class 1E loads within their design ratings for all modes of p operation, (2) an analysis of diesel generator loadings for loss 384-1981, (3) of of f site power similiar to that presented in Tables 8.3-2the failure of (4) through 8.3-6 of the FSAR, system that supplies control power to the subject non Class Ioads, and (5) non-Class IE loads are connected.
RESPONSE
The following discussion demonstrates the adequacy of employing a single circuit breaker tripped by a LOCA signal as an isolation device between a Class 1E power bus and a non-Class 1E load for design bas ) event that do not generate LOCA signals. 4 shows the two configurations that employ a circuit breaker tripped by a LOCA signal as an isolation device. Tigure 430.33-1 The two configurations are: A Class 1E unit substation supplies a non-Class IE a motor load through a. motor control center (NCC) et O Class IE circuit breaker B. N3 lA Class 1E motor control center supplies t 01 b. 5 C panel. The Class II circuit breakers B and D are qualified to operate z for HCGS seismic and environmental parameters for all d t 1 basis events. g ) Amendment 4 430.33-1 - ~ ~ - - - - - T*W"-=+-tuw,
NCGS FSAR respective Class IE power supply buses f rom the non-Class IE This applies loads in the event the non-Class 1E loads fail. 1 whether the plant is supplied f rom an of fsite source or an ensite the f ailure of the non-Class 1E loads supplied 1E power supply buses will not prevent any of the four
- Thus, source.-
. channels of Class 1E power supplies f rom perf orming its saf ety from.Clas y,gs,sNt A PAed f44E 4% % - 7.4 I""C" COMPLIANCI WITH GUIDELINES OF SECTION 7.1.2.1 0F IEEE 384-198 Protective device coordination studies for devices shown in have shown that the time-overcurrent trip 430.33-1 C, and D are such thats Figure characteristics of circuit breakers A, B, Circuit breaker 3 will trip to clear a fault current prior to initiation of a trip of circuit breaker A. a. Circuit breaker D will trip to clear a fault current prior to initiation of a trip of circuit breaker C. b. Both the onsite and offsite powers supply sources are separately capable of supplying the necessary fault current for sufficien 2/ loss of function of Class 1E loads. STANDBY DIESEL GENERATOR LOADINGS FOR LOSS OT OTTSI tabulates the loads, their XW ratings, and loading Table 8.3-1 and loss of offsite sequences for design basis accident (DBA)It can be verified by inspecting scenarios. power (LOP) that DBA loading of the SDGs is the lir.iting case Table 8.3-1 with respect to the loading capability of the SDGs. 'TAILURI OF THE NON-CLASS 1E DC SYSTEH THAT SUPPLIES C TO THE SUBJECT NON-CLASS 1E LOADS (described above) the circuit breaker B For configuration (a) supplying a Non-Class 1E MCC or a motor load is centro 11ed byFor a non-Class Class 1E 125 V de control power supply.a non-Class IE circuit breaker / f
- load, This non-Class IE circuit breaker (GE-AXR circuit breakp r B.
is conttolled by a non-Class 1E'125 V de control power. GE-AKR type circuit breakers are directihg acting trip devic type) a Therefore, the failure of the de electrical fault conditions. control power supply does not prevent the circuit br in response to the f ailure of non-Class 1E motor load. /Kor IHSat? C Pum PM,esq36.n-t6 4 osan ores I;c1 #60 Amendment 4 430.33-2
WS6$T A ...Ac CI4ss, t e. MSlic d c Scortes a ad 44 e a#sif e reever seweces med #slr Ahtelbuilen sysism a re e{ suHielen+ c<pci4y ang t +e b e+ k cl.s u t t a,g non. cfa,, og r< f< Mt:4 9 to sart y power lords during att piel condifadn s, 2 n 44 e e,ie n t o f' < t o c 4 y <4:n 3 o l A le~ b e<< K.d r dime - overcurroni +rlf, sill Le forf*rned t o a! ms,,s4mt g./Ag 44e e hA ra cierh-he s t i u circuif be:A.rer trig -{ansilen rt*m<in s ovi4hin quired ll mils, "Te.l>l e A M,5 3-l idenfile'es +ke nu - Class 12. loads -tW a re saff ed 4h reyh eirtul+ beautes 5A.ndDofft}uetAst.55.g, ll osa orn z:m,2 /,0 . ~ ~. % v.,1
HCGS FSAR lbl5ER.T ' C " OUESTION 430.33 ANALYSIS FOR SUPPLYING NON-CLASS lE FROM CLASS 1E DC SYSTEMS Figure 8.3-11 shows non-Class lE public address system distribution panel 10J496 supplied from a Class 1E de power bus 10D410 through a Class lE inverter in UPS unit 10D496. The inverter is an acceptable Section 7.1.2.3. The re f ore, a isolation device per IEEE-384-1981, failure in the non-Class lE distribution panel 10J496 will not degrade Class lE de system bus 100410. the adequacy of The HCGS UPS system will be tested to demonstrate The test will an inverter beino applied as an isolation device. demons tra te that voltage, current, and frequency on the Class 1E side of the UPS are not degraded below acceptable levels when a maximum credibic voltage or current transient is applied on the non-The tests to be performed will Class lE side of the UPS system. simulate all operating modes for which the HCGS UPS system is designed. The tests will include the following, types of f aults at the UPS output locations a. Phase to ground b. Neutral to ground c. Phase to neutral without ground d. Hot short (460 Vac) The A test plan is submitted separately for the staf f's review. test results will be test report and any associated analysis of the submitted in December 1984. An analysis has been perf ormed to support the values used for the This analysis shows that the acceptance criteria for voltages. voltages specified will not cause misoperation or loss of any electrical equipment connected to the supply buses. the acceptance values for the test currents are well In addition, infeed breakers to the UPS below the level that would cause the supply '
- s to trip.
These values are as follows: Infeed breaker Acceptance Setting Circuit Current __ 0-55 amperes continuous 600 amperes Normal 400 VAC with a maximum peak of Pick-up supply 132 amperes for not longer than 10 mS 0-78 amperes continuous 600 amperes Back-up 480 VAC with a maximum peak of 500 Pick-up Supply amperes for not longer than 10 ms 2000 ampero Alternate 125 VDC 0-364 amperes Fuse Supply PA&c 43o.33 28M
+ r. e l N SET &'T "C " .Page ' two If the' testing can not demonstrate adequacy of the UPS as an isolation device, then.an isolation transformer will be added between the inverter and-the distribution panel. The test plan for .the isolation transformer is also submitted separately for the R-staff's review. In the event of failure of both tests the non-Class lE loads ' associated with the UPS system will be removed from the Class lE ' buses or the. cables to these loads will be re-routed so as to be t-separated f rom class IE cables associated with other Class 1E channels. e e s b W \\ ~ 4 r . j i s >s _.p. .,/., ' ,/ s*, / ,~ / i ] s { e { } A l' i om me.n-ts(2) i
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- Aadasaatt MCC, 104343
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T48LL 430.93-l L O AT t n a s.P ~ ,,g4ga g g,,4,,, t A m 4er Acer Mt.c osz.st. S t H1821 St 14v3al its4r o Acacf er 50/14.*3 vhusf F43 51-4 @ E is Ju;, "**5c:, W <r=toaqrs I; ' 5 1-4flar.3 106451 Pubite. A4/rtss Sysitm Lverier 2f ic 048t 4 b IE Y
- Ibh*
3d s t-ems nos4,i s,mr.4 y s.y.+,-, w a,c om, y 7-D 5 2i ieE4CI u..,-M er h v e rf t- \\ i1: + 98 e DSER OPEN I'1'IM Mhd t f-1 ,_.-. ~.. c.._,.... .-....__..---.-...._..,_-,.,_m,.-
i i 'FOR MOTOR LOADS,1N ADDITION TO CIRCUIT DREAKER B, THERE IS NON CLASS lE CIRCulT BREAKER DOWNSTREAM OF BREAKER B. s 1 4.16 KV CLASS lE BUS )A LLD frm 480 V CLASS lE UNIT SUBST A. 4? ll O )C ) B TRIPPED BY LOCA au NON CLASS lE CABLE V 'NON CLASS lE MOTOR OR MOTOR CONTROL CENTER 480 V CLASS lE MCC ll ) D TRIPPED BY LOCA ' NON CLASS lE CABLE d V-BACK UP POWER SUPPLY FOR UPS SUPPLYING.
- NON CLASS lE DISTRIBUTION PANEL.
HOPE CREEK GENER ATING STATION FINAL SAFETY ANALYSIS REPORT ISOLATION BETWEEN CLASS lE POWER SUPPLIES AND NON CLASS lE LOADS-TRIPPING CIRCulT BREAKER FIGURE 430.33 1 AMENDMENT 4,1/84 ~~
-i
- j -
TEST PROCEDURE, ISOLATION VERIFICATION S/N 9743 1E 20KVA UPS (INSTRUMENTATION AC POWER SUPPLY) FOR PUBLIC SERVICE ELECTRIC & GAS CO. HOPE CREEK GENERATING STATION P0. 10855-E-154 (Q)-AC '0B JE C T IV E : TESTING TO ESTABLISH THE UPS SYSTEM AS A CIRCUIT ISOLATION SYSTEM. PASS CRITERIA: DEFINITION OF ISOLATION DEVICE OR SYSTEM: A DEVICE OR SYSTEM IS CONSIDERED TO BE A CIRCUIT ISOLATION DEVICE IF IT IS APPLIED SUCH THAT THE MAXIMUM CREDIBLE VOLTAGE OR CURRENT ' TRANSIENT APPLIED TO THE NON CLASS 1E SIDE OF THE DEVICE WILL NOT DEGRADE THE CLASS 1E CIRCUIT ON THE OTHER SIDE OF 'THAT DEVICE. CIRCUIT NORMAL VARIATION ALT. DC. SUPPLY 150-140 VDC 0-364'ADC NORMAL AC SUPPLY 48 0 +10%. V( L-L ) 3 PHASE 0-55A, 0-132AP FOR 10 MSEC BACK UP AC. SUPPLY 480+10% V 1 PHASE 0-78A, 0-500AP FOR 10 MSEC ANY VARIATIONS OUTSIDE OF NORMAL VARIATIONS SPECIFIED,'WILL BE ANALYZED ON A CASE BY CASE BASIS. l'
l FAULT LOCATION AND TYPE - FAULTS WILL'BE APPLIED TO UPS SYSTEM OUTPUT TERMINALS BY CLOSING A SWITCH AS REQUIRED. FAULT l TYPES: 1. PHASE (HOT).TO GROUND. 2. NEUTRAL ~TO GROUND. 3. PHASE-T0~ NEUTRAL W/0 GROUND 4.. 480VAC APPLIED ACROSS UPS OUTPUT W/0 GROUND (HOT SHORT) THE CONDITION OF THE THREE CLASS 1E SOURCES WILL BE MONITORED THROUGH SUITABLE SIGNAL CONDITIONERS, BY G0ULD INC., 2000W SERIES HIGH FREQUENCY RECORDING SYSTEM. 4 b L 8 1 2
v. ~~(EST PROCEDURES 1.0. GENERAL NOTES 1.1 BEFORE STARTING TEST DETERMINE AND RECORD ALL SIGNAL CONDITIONER TRANSFER RATIO (MULTIPLIER) VALUES. 1.2 NORMAL SYSTEM OPERATION DURING EACH TEST A. CONNECTION PER FIG. 1. B. OUTPUT LOAD 10KVA 0.08PF (66.7 AMP RESISTIVE AND SO AMP INDUCTIVE) 0'120VAC NOMINAL. C. UPS POWERED BY " ALTERNATE" DC SOURCE (BATTERY) AND ONE OR BOTH AC SOURCES, " NORMAL" & "BACK-UP". D. STATIC SWITCH IN " PREFERRED" POSITION. E. ALL BREAKERS & SWITCHES CLOSED, BOTH BYPASS SWITCHES IN " NORMAL" POSITION " TEST" SWITCH - CENTERED " RETURN MODE" SWITCH - IN "AUT0" POSITION " ISOLATION" TOGGLE SWITCHES - ON " SYNC" TOGGLE SWITCH - ON 1.3 TEST INSTRUMENTATION A. GOULD INC., MODEL 2800W HIGH FREQUENCY RECORDING SYSTEM. EIGHT CHANNEL, INDEPENDENT SCALE SELECT A.0SO TO 500 VOLTS FULL SCALE. B. POTENTIAL TRANSFORMER 480V, 60HZ PRIMARY 120V SECONDARY (4:1 RArIO), C. CURRENT TRANSFORMER 1000:1 RATIO WITH 10 OHM BURDEN RESISTOR. -(.01V/A). D. WIDEBAND DC ISOLATION AMPLIFIER, GOULD INC. MODEL 13-4615-10 OR EQUIVALENT. 3
1.4 - TEST FACILITY AND EQUIPMENT JA. DC SUPPLY - C&D 4LCW-15 BATTERY (60 CELLS, 80KW FOR 30 MIN.) AND-BATTERY CHARGER. B. AC-SUPPLY - 480V, 3 PHASE, 4W, 60 HZ, 1200A GROUNDED NEUTRAL. . 'C. AC LOAD BANK 30KW OR 0-30KVA 0 0.8PF. D. FAULT APPLICATION DEVICE - G.E. CIRCUIT BREAKER TJC.36400G 400A, 3P. MAGNETIC ONLY. E. , HOT FAULT SOURCE - TRANSFORMER, 1 PH 480:120V 30XVA OR LARGER. .2.0 TEST PROCEDURE 2.1 BASE LINE DATA START UP.THE UPS WITH ALL SOURCES AVAILABLE. SET UP " NORMAL OPERATION" PER 1.2 AND ALLOW SYSTEM TO WARM UP FOR AT LEAST 30 MINUTES. 1A1. METERING AND' CONNECTIONS PER FIG. 2 AND " BACKUP SOURCE" BREAKER OPEN. RECORD IN " STORE" MODE AT 20KHZ TIME BASE. COPY MEMORY TO PAPER. A2.- REPEAT Al EXCEPT-USE 500HZ TIME BASE. Bl. WITH. METERING AND CONNECTIONS PER FIG. 2 AND " NORMAL SOURCE" BREAKER OPEN. RECORD IN " STORE" MODE AT 20KHZ TIME BASE. ~ COPY MEMORY TO PAPER. B2. REPEAT 01 EXCEPT STATIC SWITCH TRANSFERRED TO BACKUP. B3. REPEAT B1 EXCEPT USE 500HZ TIME BASE. -B4 REPEAT B2 EXCEPT USE_500HZ TIME BASE. 4 u
2.2 FAULT TESTING CO. HETERING AND CONNECTIONS PER FIG 2, RECORDER IN MANUAL TRIGGER MODE. APPLY FAULT BY CLOSING " FAULT" CB AND AT THE SAME TIME (OR 0 TO 10 MILLISECONDS BEFORE) TRIGGER THE RECORDER.IN " STORE" MODE. REMOVE THE FAULT AND RECORD THE MEMORY ~TO PAPER. AFTER EACH FAULT APPLICATION CHECK THE UPS FOR DAMAGE. REPAIR THE UPS IF REQUIRED BEFORE PROCEEDING. C1. INSTALLi JUMPER " A" T0 "F AULT" CB WITH "B ACKUP SOURCE" CB OPEN WITH RECORDER AT 20KHZ TIME BASE APPLY FAULT PER C0. C2. REPEAT C1 EXCEPT WITH 500HZ TIME BASE. C3. OPEN " NORMAL SOURCE" CB AND CLOSE " BACKUP" WITH RECORDER 20KHZ TIME BASE APPLY FAULT PER C0. C4. REPEAT C3 EXCEPT WITH 500HZ lIME BASE. C5. REPEAT C1, C2, C3 & C4 WITH JUMPER "B" INSTEAD OF "A" CONNECTED T0 " FAULT" CB. C6. . REPEAT C1, C2, C3, & C4 WITH JUMPER "C" INSTEAD OF "A" CONNECTED TO " FAULT" CB. C7. REPE AT C1, C2, C3,. & C4 WITH CONNECTIONS TO HOT FAULT SOURCE (UPS RUNNING AT NO LOAD). 2.3 COMPLETE TEST
SUMMARY
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~ 4 i TEST PROCEDURE, ISOLATION VERIFICATION S/N 9743 1E 20KVA UPS (INSTRUMENTATION AC POWER SUPPLY) IN SERIES WITH A POWER CONVERSION PRODUCTS ISOLATING TRANSFORMER MODEL # - RTF-120/120-30 FOR .PUBLIC SERVICE ELECTRIC & GAS CO. HOPE CREEK GENERATING STATION P0. 10855-E-154 (Q)-AC OBJECTIVE: TESTING'TO ESTABLISH THE ISOLATING TRANSFORMER IN SERIES WITH A UPS SYSTEM AS A CIRCUIT ISOLATION SYSTEM. PASS CRITERIA: DEFINITION OF ISOLATION DEVICE OR SYSTEM: A DEVICC OR SYSTEM IS CONSIDERED TO BE A CIRCUIT ISOLATION DEVICE IF IT IS APPLIED SUCH THAT THE MAXIMUM CREDIBLE VOLTAGE OR CURRENT TRANSIENT APPLIED.T0 THE NON CLASS 1E SIDE OF THE DEVICE WILL NOT-DEGRADE THE CLASS 1E CIRCUIT ON THE OTHER SIDE OF THAT DEVICE. CIRCUIT NORMAL VARIATION ALT. UC. SUPPLY 150-140 VDC 0-364 ADC NORMAL AC SUPPLY 48 0+10% V(L-L) 3 PHASE 0-55A, 0-132AP FOR 10 MSEC BACK UP AC SUPPLY 48 0+10% V 1 PHASE 0-78A, 0-500AP FOR 10 MSEC ANY VARIATIONS OUTSIDE OF NORMAL VARIATIONS SPECIFIED, WILL BE ANALYZED ON A CASE BY CASE BASIS. 1
D FAULT-LOCATION AND TYPE FAULTS WILL.BE APPLIED T0 ISOLATING TRANSFORMER OUTPUT TERMINALS BY CLOSING.A SWITCH AS REQUIRED. FAULT TYPES: 1. PHASE (HOT) TO GROUND 2. NEUTRAL TO GROUND 3. PHASE TO NEUTRAL W/0 GROUND ~4. 480VAC APPLIED ACROSS UPS OUTPUT W/0 GROUND (HOT SHORT) THE CONDITION OF THE THREE-CLASS IE SOURCES WILL BE MONITORED THROUGH SUITABLE SIGNAL CONDITIONERS,.BY GOULO INC., 2000W SERIES HIGH FREQUENCY RECORDING SYSTEM. 2
f5 T TESTcPROCEDURES 1.0 GENERAL' NOTES 1.1 BEFORE.STARTlHG TEST DETERMINE AND RECORO ALL SIGNAL CONDITIONER TRANSFER' RATIO (MULTIPLIER) VALUES. ^ 1.2 NORMAL SYSTEM OPERATION DURING EACH TEST A.
- CONNECTION PER FIG.
l.- B. OUTPUT LOAD 10KVA @.08PF (66.7 AMP RESISTIVE AND 50 AMP INDUCTIVE) 0 120VAC NOMINAL. C. UPS POWERED BY " ALTERNATE" DC SOURCE (BATTERY) AND ONE OR
- BOTH.AC SOURCES, " NORMAL" & "DACK-UP".
D. STATIC SWITCH IN " PREFERRED" POSITION. E. ALL DREAKERS & SWITCHES CLOSED, BOTH BYPASS SWITCHES IN " NORMAL" POSITION " TEST" SWITCH - CENTERED "RETUP.N. MODE" SWITCH - IN "AUT0" POSITION "lSOLATION" TOGGLE SWITCHES - ON " SYNC" TOGGLE SWITCH - ON 1.3 TEST INSTRUMENTATION A. GOULD INC., MODEL 2800W HIGH FREQUENCY RECORDING SYSTEM. EIGHT CilANNEL, INDEPENDENT SCALE SELECT 1.050 TO 500 VOLTS FULL' SCALE. B. POTENTIAL TRANSFORMER 480V, 60HZ PRIMARY 120V SECONDARY (4:1 RATIO). C. CURRENT TRANSFORMER 1000:1 RATIO WITH 10 OHM BURDEN RESISTOR. (.0lV/A).
- D.
WIDEBAND DC ISOLATION AMPLIFIER, GOULD INC. MODEL 13-4615-10 OR EQUIVALENT. J 3
i ii
- .i 1.4 TEST FACILITY AND EQUIPMENT.
A. DC SUPPLY - C&D 4LCW-15 BATTERY (60 CELLS, 80KW FOR 30 MIN.) AND' BATTERY CHARGER. ,B. AC SUPPLY - 480V, 3 PHASE, 4W, 60 HZ, 1200A GROUNDED NEUTRAL. C.. AC LOAD BANK 30KW OR 0-30KV A 0 0.8PF. D. FAULT APPLICATION DEVICE - G.E. CIRCUIT BREAKER TJC 36400G 400A, 3P. MAGNETIC ONLY. E. HOT FAULT SOURCE - TRANSFORMER, 1 PH 480:120V 30KVA OR LARGER. 5.0 TEST PROCEDURE 2.1 BASE LINE DATA START UP THE UPS WITH ALL SOURCES AVAILABL2. SET UP " NORMAL OPERATION" PER 1.2 AND' ALLOW SYSTEM TO WARM UP FOR AT LEAST 30 MINUTES. A1. METERING AND CONNECTIONS PER FIG. 2 AND " BACKUP SOURCE" BREAKER OPEN. RECORD IN " STORE" MODE AT 20KHZ TIME BASE. COPY MEMORY TO PAPER. A2. REPEAT.Al EXCEPT USE 500HZ TIME BASE. Bl. WITH METERING AND CONNECTIONS PER FIG. 2 AND " NORMAL SOURCE" BREAKER OPEN. RECORO IN " STORE" MODE AT 20KHZ TIME BASE. COPY MEMORY TO PAPER. B2. REPEAT 01 EXCEPT STATIC SWITCH TRANSFERRED TO BACKUP, B3. REPEAT B1 EXCEPT USE 500HZ TIME BASE. 84. REPEAT B2 EXCEPT USE 500HZ TIME BASE. = 4
~~ i
- 2. 2 - FAULT TESTING CO.
METERING AND CONNECTIONS PER FIG 2, RECORDER IN MANUAL TRIGGER MODE. APPLY FAULT BY CLOSING'" FAULT" CB AND AT THE SAME TIME (OR 0 TO 10 MILLISECONDS BEFORE) TRIGGER THE RECORDER IN " STORE" MODE. REMOVE THE FAULT AND RECORD THE MEMORY TO PAPER. AFTER EACH FAULT APPLICATION. CHECK THE UPS FOR DAMAGE. ' REPAIR THE UPS IF REQUIRED BEFORE PROCEEDING. C1. INSTALL JUMPER "A" T0 " FAULT" CB WITH " BACKUP SOURCE" CB OPEN WITH RECORDER AT 20KHZ TIME BASE APPLY FAULT PER CO. C2. REPEAT C1 EXCEPT WITH 500HZ TIME BASE. C3. OPEN " NORMAL SOURCE" CB AND CLOSE " BACKUP" WITH RECORDER 20KHZ TIME BASE APPLY FAULT PER CO. C 4 '. REPEAT C3 EXCEPT WITH 500HZ TIME BASE. C5. REPEAT C1, C2, C3 & C4 WITH JUMPER "B" INSTEAD OF "A" CONNECTED TO " FAULT" CB. C6. REPEAT C1, C2, C3, & C4 WITH JUMPER "C" INSTEAD OF "A" CONNECTED TO " FAULT" CB. ^ C7. REPEAT C1, C2, C3, & C4 WITH CONNECTIONS TO HOT FAULT SOURCE (UPS RUNNING AT NO LOAD). 2.3 COMPLETE TEST
SUMMARY
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.m ATTACHMENT 5 MODIFICATIONS TO FSAR SECTIONS 17.2 AND SRAI (1) O l.
HCGS FSAR 6/84 c.22 Activities covered by the QA program are delineated in Table 17.2-1 and include inplant I, radiation monitoring under o " Control of Radioactivity." c.23 The HCGS position on TMI Item II.d.3.4 is given in Section 1.10. This item is not a " structure, system or component" requiring entry in Table 3.2-1. Control of this activity is provided by appropriate procedures. Chapter 17 describes the Quality Assurance Program coverage of procedural controls. The following information is provided for additional clarification: a) The nonsafety-related, non-ESF internal epmponents include the steam dryer, the shroud head and steam separator assembly, the in-core guide tubes and stabilizers, the differential-pressure and liquid-control lines inside the RPV, the fuel orifices, and the feedwater spargers. In all BWR 4, 5, and 6 designs, these components are not 0-listed because they are neither required for safe shutdown of the plant, nor would their failure jeopardize the safety functions of other safety-related internal components. During the operating phase of the HCGS, the same high-quality design, procurement, and installation control practices, as were applied during the design and construction phase, will,be applied to any changes to these components. As Section 3.2.1 and notes (13) and (50) for Table 3.2-1 indicate, the quality assurance controls for non-ESF RPV internal components and for seismic Class II/I equipment are described in Section 1.8.1.29. In addition, the reactor pressure vessel internal structures which are accessible are included in the ISI program, which is covered by the operational QA program. b) Reactor building penetrations are not required to be Q-listed unless the piping system is 0-listed. A non-Q piping system penetrating the reactor building is not required to have a Q-listed penetration. t elb Q-U.ste<{ or I m pnuever reactor faile{iny w, e+< die n.s a re-chss. fred as Sessn.:e.lrfr a s s/*w n in relvned 'n.4/s 3e 2. -l
- c)
The spent fuel pool liner does not perform a safety function and therefore is not Q-listed. However, the spent fuel pool does meet the quality assurance requirements of 10 CFR 50, Appendix B, and has been noted as such in Table 3.2-1, Item XIX.e. d) Shore protection of the intake structure does not have a safety function and therefore is not 0-listed (Item XVIII.j SRAI (1)-12 Amendment 6
- .. s H
=... HCGS FSAR ~1/84 t ~ 6. ~ Vendor services b. Coordinates PSE&G system / component turnover documentation review . Coordinates resolution of support-related problems c. ~. d. ._Provides test and startup technical support and ~ assistance, as required. In implementing this responsibility,-he directs assigned group-leaders and personnel. i' 14.2.2.2.4 Integrated Testing Coordinator The integrated testing coordinator is responsible for providing . technical expert se to-the startup program. In this capacity,. i s his responsibilities include: Development and direction of the test program on a. integrated system tests b. Technical procedure review Acquisition of baseline data. c.- In implementing this responsibility, he directs assigned group' leaders and personnel. 14.2.2.2.5 Quality Assurance Startup Engineer -(OASE) kanW A.S.rox n45 The quality assurance startup engineer performs a sta[f function Public Service Ocrperate quality - E J / f,s E E E t h D v, @: n. m.. M,,2 /cg43to which he is technically and 8,rence cpar m coministratively responsible. He will identify to the Startup Manager, or his designee, all qu'ality related problems associated with O/F-listed activities. His responsibilities include: t g 14.2-6 AmenCnent /
HCG'S.FSAR 4/84 I f.- 17.2 OUALITY ASSURANCE DURING THE OPERATIONS PHASE _r. Public Service Electric and Gas Company (PSE&G) is responsible for assuring that the oparation,. maintenance, refueling and modification of Hope Creek Generating Station (HCGS) is accomplished in a manner that protects public. health and safety and that is in compliance with applicable regulatory requirements. To carry out this responsibility, PSE4G has developed and implemented a comprehensive quality assurance . program that is applicable to the design, construction, and testing phases. The description of the quality assurance program provided herein parallels the operational quality assurance program currently being implemented at the Salem Generating Station. This operational quality assurance program is documented in the nuclear department manual. This description is maintained by . nuclear
- rsti n; quality assurance (NOA).
The program provides - measures to assure the control of activities.affecting the safety-related function of structures, systems, and components. 'The quality assurance program encompasses fire protection of safety-related areas and other activities enumerated in Regulatory Guide 1.33. A planned monitoring and audit program assures that specified requirements of the operational quality assurance program are met. The program provides coordinated and centralized quality assurance direction, control, and documentation, as required by the NRC criteria set forth in 110 CFR 50, Appendix B. Applicable NRC Regulatory Guides, codes, and standards, as well as the policy statements contained in the .Mv.hs, QMMP E C q ;11ty ;;; ;;n; manual, are used by PSE&G organizations . performing activities affecting safety to prepare appropriate ' implementing procedures. To assess the effectiveness of the PSE&G quality assurance program, independent auditors from outside the company periodically audit the program for compliance with 10 CFR 50, Appendix B, and other regulatory comm!tments. ' Independent audits shall be conducted at least every two years. Reports of such audits are made directly to upper management. 1 OA. policy statements are issued by key management representatives including the Company Board Chairman / President, by the Senior Vice President - In;zg, C.pply and Engineering and by the Vice l President - Nuclear and as such, are mandatory throughout the i ' Company. WO N l.'. 8 17.2-1 Amendment //
~ HCGS FSAR 4/84 Key policy elements, as thep apply to nuclear safety, include the following: .a. Nuclear safety-is of the highest priority and shall take precedence over matters concerning power production. b. The public's health and safety is the prime consideration in the conduct and support'of Public Service Electric and Gas Company's nuclear operations and shall not be compromised. All decisions which could affect the health and safety of the public shall be,made conservatively. c. A-Quality Assurance' Program is an essential part of the PSE&G commitment to safe and reliable nuclear power operation.. Applicable. program. requirements shall be strictly adhered to in the performance of activities covered by the Quality Assurance Program. l 1 PSE4G requires its suppliers and contractors to assume responsibility for establishing and implementing QA/DC programs, as applicable, to meet 10 CFR 50, Appendix B. However, the responsibility for the overall OA program is retained and 2 excercised by PSE&G. NOA rev.iews those programs and conducts appropriate monitoring and auditing as required to assure that the suppliers are properly implementing their DA/QC programs. The nuclear operations quality assurance program verifies that requirements necessary to assure quality are properly included or referenced. In addition, these suppliers' procurement documents include applicable PSE&G quality assurance requirements for items and services provided by their suppliers. 17.2.1 ORGANIZATION c The nuclear
- rrtir r QA program, referred to hereafter as the OA program, assures *that adequate administrative and management controls are established for the safe operation of the Hope Creek Generatin'g Station.
Implementation is assured ongoing review, monitoring and audit under the direction of the ' Manager-Nuclear GyeseMons Quality j Assurance who reports to the Vice President-Nuclear. 1 8 Amendment [ 17.2-2
~ HCd5 FSAR 4/84 p.- Company organization is shown on Figures 13.1-1, 13.1-2 and 17.2-1. Responsibilities for activities affecting safety are described in the following sections. 17.2.1.1 Nuclear Deoartment Thekieefresident-helearisresponsibleformanagingand directing the nuclear activities of the company. ":pe,ctin general sneger - ng;g t: i th: "i-a re-8 dent n.cle:r :: th-
- aac e-ruir e, ;ener
- 1 22n:g::
nuclear sop A L, geumal e.; n;;:: - p-ler e r retiene, ;fnt::1 ;;n;.;;c
- p; C n. d ; ; :r st-ic-r, end
=== ;er
- lity rrecrance n ;1eec c,; ccativua.
Overall duties and responsibilities of the nuclear department are, included i prow 1 4 n. rel i nwi n,,..re.<m-- i ; g 2 ---. : --4--- --- --- - g U%ene3 h8d beMpb[tiMN- '"'N in Section 13.1. implementatiohofqualityassurancerehuirementsbytheirstaff.' These OA requirements are contained in the men shinistratif: twele; ,,m,....... .na oth,r a.m ..on.1. ek,p;WV M mzm d ' A ~[n mcliMchs/ 0 dd w w 9 1 The Vice President - Nuclear regularly assesses the scope, status, adequacy, and compliance of the QA program to 10 CFR 50, g-Appendix B through: Frequent contacts in staff meetings, NOA audit reports, a. cooperative management audit reports, NRC inspection reports, department status reports. b. An annual assessment of the OA program is preplanned and documented. This assessment addresses the scope, status, and adequacy of the QA program. Corrective action is identified, and tracked. 1 1.1.1 Nuclear Department - Nuclear Service The general a er - nuclear services i esponsible for providing techni support to stati organizations in the areas of radiation protect 1 site pro tion, including fire, security, and emergency p ness; site maintenance; nuclear procurements and materia on
- and personnel training.
17.2.1 Nuclear. Department - Nuclear Su i 5.' 8 17.2-3 Amendment [ l
HCd5 FSAR 4/84
- p. -
Company organization is shown on Figures 13.1-1, 13.1-2 and 17.2-1. Responsibilities for activities affecting safety are described in the following sections. 17.2.1.1 Nuclear Department icefresident-)helearisresponsibleformanagingand The directing the nuclear activities of the company. ";pecting to tr.: -! ~ pr--ident nec1 :: ::: the geners! r e;er - nu;1eer r--=ic-r,
- earr
- 1
.:ne;;: nucle;c supm, veuwel 2:n;;;r - ?-ler eper=ti:n, gfncr:1 ;;;;;;;
- sp; Cccck ;;rraticar, --a
==- ;er - ;; 1it; rreurance nucle;r escreLuus. Overall duties and responsibilities of the nuclear department are 8- '"'-'_ _ _ --. ' 4 prov; dd +w. rmiinuinn ...-*4--- 2---.: 6: -2 in Section 13.1. -helenr3I~NhN implementatiohofqualityassurancereh[uirementsbytheirstaff.e MpMN e These OA requirements are contained in the e,tatien ad-inistrati : nw/c. =na other A-- --an-le QWO m m & 'Y--* -,'o'*fm c6 Osck +/ N& m s u n, m,. - - .t The Vice President - Nuclear regularly assesses the scope, status, adequacy, and compliance of the QA program to 10 CFR 50, Appendix B through: a. Frequent contacts in staff meetings, NOA audit reports, cooperative management audit reports, NRC inspection reports, department status reports. b. An annual assessment of the OA program is preplanned and documented. This assessment addresses the scope, status, and adequacy of the OA program. Corrective ( action is identified, and tracked. l t l-1 1.1.1 Nuclear Department - Nuclear Service i L The general m er - nuclear services i esponsible for l providing techni support to stati organizations in the areas of radiation protect 1 site pro tion, including fire, security, and emergency p ness; site maintenance; nuclear procurements and materia on
- and personnel training.
17.2.1 Nuclear Department - Nuclear Su 8 Amendment [ 17.2-3
r HCGS FSAR 4/84 j ( The gene ager - nuclear support is r le for providing suppor e station areas of reactor j engineering, engin' erin sigo, fuel management, licensing e and risk assessment and regulato 1 y, nuclear anal 17. .1.3 Nuclear Department - Hope Creek Operations The general m ger - Hope Creek operations is responsible r the safe and effi nt operation of the plant, and fo e general direction o e station operating, mai
- nee, radiation' protection, an echnical suppor partments.
The general manager - Hope Creek rati a so directs the activities of the station oper review committee (SORC) and is responsible for assur at plan itions are staffed by fully qualified an ined personnel, as 1 as being responsible e implementation of quality surance require s by station staff. These quality as ance r ements are contained in the station administr ive rocedures and other station department manuals. I ( 17.2.1.1.j[ Nuclear Gew& tiene Quality Assurance l Th$ mana er - nuclear escr;ti;.7s quality assurance is responsible 5 for defining, formulating, implementing, and coordinating the quality assurance program. He has been delegated the authority and has the independence to interpret quality requirements, identify quality problems and trends, and provide recommendations or solutions to quality problems. He is responsible for approval of the nuclear QA department manual to be used during the operations phase of Hope Creek. He also is responsible for assuring compliance with established requirements for the quality assurance program through document review, inspection, monitoring, and audit. NOA provides a centralized coordinating l function for quality assurance and quality control activities applied to the operation phase. P The, manager - nuclear.,___..... quality assurance has the authority and responsibility to: a. Stop work when significant conditions adverse to quality require such action. J 17.2-4 Amendment /f
e j \\
- ~
HCGS FSAR 4/84 The PSE&G po es and. organization structure assure that the [W i . quality assurance nuclee; peratiens has manager - sufficient orginizational freedom and independence to carry out his responsibilities.- 17.2.1.1.4.1 Nuclear c ece:.Las. Quality Assurance Personnel r hsf fo&M sd w,h o% % <f v>G b www u. m. Q m pThe, manager - NOA and engineers reportipg directly to him nust es% ~ have a combination of 6 years of experience in the field of At least 1 of these 6 years of quality ascurance and: operations. mentation of a nuclear . experience must be in'the,overall imple power plant quality asstirance program.1 A minimum of 1 year and a maximum of 4 of the 6 years of experience may be fulfilled by related technical /or academic training. Personnel performina inspections, examinations, and test,activitiesvare certified as Leve.1 III as appropriate'to their f. Level I,:LevellII, 'r.esponsibilities, also in. accor' dance with-Re'gulatory' Guide-l'i58',) F ~ as noted. '.(L.a..,s ~,s a,: 4 - Q a The, manager - nuclear ---- "" quality, assurance fulfills the above qualifications with the addition of the following i Knowledge and experience in quality assurance, l a. High level of leadership with the ability to :ommand b. the respect and cooperation of company personnel, vendors, and construction forces Initiative and judgment to establish related policies c. to attain high achiever.ents and economy of operations. O M */8h/uo 17.2.1.1.5 Jadv.ndum. Revie-G eup; m-Threeadvisory.groupsfareresponsibleforreviewingand evaluating items related to nuclear safety. The overall responsibilities of these groups are included ir th= f al!c"ing - f>y*M "ere detail-A da-criptirre sre centein:f in i aa- - 7* tion 13 4 Sec r h th. 2d;iew y group. Ceepesed of kcy 2.; F*' i-e- -i".r-4484esponeiM111.i.5 incl.J. spi;
- f ;1r P etetit
- perreriel i
5 [ %. 3.h e,,,.m~~m W -1,1 csouyp n=g=cnR,,,, A t n ~. ~,...... . o _, S,.c.,,.,._ _ gee-se,--w- -w e-w&,+- =,*---w--,y+.-._w y.w_,,.y.,,-,y-y,-w. m v- ,e,,-- __,,,..,r,ew.,.w-,, --.e.,w,+,-p--%rw-mwww-wy--,m--y-.,---i.-=w--e-,,y.,-m-,
NdGS FSAR ~4/84 ~ h h [W sanager - ; g and. organization structure assure t at t e The PSE&G po quality assurance wec.eer grratirn: has sufficient organizational freedom and independence to carry out s his responsibilities. p 17.2.1.1.4.1 Nuclear Cr.. m.... Quality Assurance Personnel o$ ken sd w,JWn & 9v>4 h sose w u. m % t Ad/w Q WThe manager - NOA and engineers reportipg directly to him must eng have a combination of 6 years of experience in the field of At least 1 of these 6 years of i quality. assurance and operations. mentation of a nuclear experience must be in the.overall icple power plant quality assurance program. 1 A minimum of 1 year and a maximum.of 4 of the 6 years o,f experience may be fulfilled by Personnel performino related technical or academic training. inspections, examinations, and test,activitie F,are certified as ' Level I,: Level'II, Level III as, appropriate ~to their ' as noted. ',ities, also in. accordance with Jtegulatory' Guide ~ responsibil .(c.a..,4. yc,n h m-h. :,a c.a.g - '^ The, manager - nuclear ---- "- quality. assurance fulfills the above qualifications with the addition of the followings Knowledge and experience in quality assurance, a. High level of leadership with the ability to command b. the respect and cooperation of company personnel, vendors, and construction forces Initiative and judgment to establish related policies c. to attain high achievements and economy of operations. OM >/ 88v/em 17.2.1.1.5 Jal.c..J m Review GieuP; Threeadvisory.groupsfareresponsibleforreviewingand evaluating items related to nuclear safety. The overall fr d ' e Falic-in; responsibilities of these groups are-included in th= "rre detriled d--r-iptiene tre ceatsin:d in _ pr* u== i I Section 13.4. i f f Th; ENt* i-== la ?tsti;n OdJiews.y sseap. Osepweed Of key l ststic: perre-el, its reepen;iMililee i..cl Os spi;
- f plert~
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1 - 4.. HCGS FSAR 4/84 invae*4 0-ti:n of Tcui.nical
- ti:n:, r=partable occurrences-pr=rieA=
soecificatinn viniatinne ruito rece-enA=H nne en _ran"----re), s-d pear =A"r
- vieu: f er art f ui *i =e =ffecting nuclear -=fety_
Fecc-r d:tions ;f this ed"iracy gra"p -ere fero = rand to the nonoral emn=g-r - Mepe Cre=k nparnfinne, with ccpic: t: th: th:ir :n of the n;;l::: revice heard ' Frui. The t f SOAE is invited to all SORC meetings and receives the minu es n He attends the meetings periodically as part of the meetings. the planned surveillance program. 'SRC) This .T;..euv..d :d"innrv ornnp ie the ::f t, Jcview vovue ,gr;;p, -hee = primary -=epeneihility ig Lv imptuve pl.nt : fety7 ec.;cus reorres of clant dasigr nad egutoLing ::periene= - Lafecasti n, ?"A performe errveill;ncc ef planc opet. Lien and SC*ivitiae
- pe"It" Of '
maint nzn 0 serreiance 2;e repprtad te the 20;r: ri'*r g;..cuol mou-ger. .g. F i M f sd ift I" **f** W We*J -Gun Or, 1~},, os Q f
- Phe-tnire aavisory 9tvue is th; !!TI, 7.iica aavises de
. cc pessideau nucle: in ;;tterr =#f-ct =7 ' UI* r EEf*IY ??~ r r=1 =H g te pl wi vect euion oc mvdif gtien te try eloni uwaign. ceview or "T.c !!T is moyun51bic for p;rformir. -....ucecuuencFrafis responsible for selected pl?-t acti"it! :. Ir 2dditi:n, eplanned,' independent audits *of plant operations in accordance These audits are ..th Technical Specification requirements. 4?.s mm..cg e r - generally conducted by the NOA under 4RB-cognizance. - MC?. i;. amab;; cf thie ke ? e+. 05UR i 17.2.1.2 Research and Testino Laboratory The Research and Testing Laboratory is a part of the PSE&G Research Corporation, which is an independent entity. j The Research and Testing Laboratory performs calibrations, ~ analyses, and evaluations on systems, equipment, and materials, as requested by PSE&G departments, and maintains compliance with-i is quality assurance program. 17.2.1.3 Fuel Supply Department 4 epd' f / The general manager - feel supply reports to!the vice president % fuel supply who, in turn, reporir to the see m vice president - crrr;y rrpply ?nd en;i rering. The fuel supply department is sponsible for arranging for procurement of uranium ore, l assembly fabrication l wanverston and enrichment services, and fue l 1 17.2-6 Amendment ~
$kE kw&pwJ<- Goolad Alwim feww A fond S&s+t-~~s, SM& " d i . T i +s. q, HCGS FSAR 4/84 7:.* Regulatory Guide 1..'123, Quality Assurance Requirements q. (- for Control of Procurement of Items and Servir:es for l Nuclear Power Plants \\ r. Regulatory Guide 1.137, Fuel-Oil Systems for Standby Diesel Generators 7 i p.f Regulatory Guide 1.144, Auditing Quality Assurance Programs for Nuclear Power Plants [.u-RegulatoryGuide1.146,QualificationofQuality Assurance Program Audit Personnel for Nuclear Power Plants f.v BTP 9.5-1, Appendix A, Guidelines for Fire Protection ~ for Nuclear Plants Docketed Prior to July 1, 1976. t Commitments to Regulatory Guides, with respect to revision level, exceptions, etc, are contained in Section 1.8. < ~,. Substantive changes to the' quality assurance program described h'erein will be submitted to the NRC within 30 days of implementation. Nonsubstantive changes will be identified in the annual FSAR updates. The overall quality' assurance program is described in the nuclear department manual. This description is prepared and maintained by nuclear apr;ti = quality assurance. PSE&G organizations performing activities affecting nuclear safety prepare and maintain ~ implementing procedures and instructions. These procedures and instructions, and subsequent revisions thereto, are subject to NQA review and approval to the extent necessary to verify compliance with the quality assurance program and.the applicable quality-related Regulatory Guides and standards identified above. NOA will monitor the preparation and issuance of required procedures. to assure that they are in place at least six months prior to fuel load. p The general manager - Bope Creek operations has~ instituted and will maintain an administrative procedures (AP) manual for Hope (, Creek Generating Station (HCGS). O Amendment [ 17.2-10 ,,w, ---,,,,,=.,,,--.-4,r.,, ,w,,,.,,,, -,.e.,4,v-.. ,_,,m,,..-..,w.w-,,,-, ,-e%.,---,_-,%-+,,.-.w ,me--rmm.-#-eme,.e-,.-.,,,.%,w,,-,,-w,.,e-,,
~ HCG 5 FSAR 4/84 ( The' station administrati.ve procedures and all subsequent I revisions thereto are prepared by the technical staf f, are I' reviewed by ths Technical Engineer, Technical Manager, NOA and SORC, and are approved by the General Manager-Hope Creek Operations. Regulatory Guide 1.33 requires that plant activities affecting quality-related items and services be conducted in accordance with written administrative controls prepared by management. The procedures and instructions by which plant activities are performedarepreparedbyggg,e nsible station organization as required by station APs,.... i.mu y the organization responsible q for the activity reviewed by NOA for quality requirements, j reviewed by the SORC (for procedures affecting safety), and approved.by the department manager. In the absence of a department manager, procedures will be approved by the assistant general manager or his designee. Procedures cannot be implemented unless the review / approval process is accomplished. Station administrative procedures provide a means to accammodate ~ on-the-spot changes to subtier implementing procedures. The routine practice for revising a procedure is to repeat the original review and approval sequence. Implementation of the quality assurance program is verified by means of independent inspections, monitoring, and audits conducted by NOA. NOA reviews and analyzes problems affecting safety that occur i during the operational phase. Items subject to review include: a. Documented nonconformances occurring at the vendor's facility and those during receiving, storage, installation, test, and operation, e.g., Deficiency Reports, Nonconformance Reports, Licensee Event Reports, etc b. Documented corrective actions taken on significant noncompliances and on audit findings c. NRC inspection findings, notifications, bulletins, etc. P TheAmanager - nuclear er rr*ian: quality assurance, or his -designee, has the authority to stop work through the issuance of a stop work order where continuance of an activity would ( 17.2-11 Amendment /
l l 4/84 HCGS FSAR ( criously compromise safety or constitute a persistent and deliberate failure to correct a serious deficiency. Designees include the station quality assurance engineer for activities conducted at the station and the engineering and procurement Gngineer for supplier activities. l NOA reports significant problems affecting the quality assurance ' program to respective management along with: Measures taken to improve quality assurance program a. controls Appropriate recommendations to achieve compliance with b. applicable requirements. Management policy and implementing procedures provide all personnel awareness and direction for reporting of defects and noncompliance pursuant to 10 CFR 21. te quality assurance program requires that activities affecting nuclear safety, including activities affecting the fire protection of safety-related areas, be accomplished under cuitably controlled conditions. The program takes into consideration the need for procedures, special controls, tools, and skills cleanliness, special processes, test equipment, to obtain the required quality and the verification of quality by examination, monitoring, and independent review inspection, test,These activities include, but are not limited to, and audit. designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining,frepairing, refueling, and modifying. vsroad% personnel who have the responsibility to implement the nuclear quality assurance program also have the responsibility _rr ::ti:n: and authority to escalate unresolved quality problems to the level of management necessary to effect a resolution of the Escalation is applied by NQA personnel to increasingly l problem.- higher levels of management, up to the vice president - nuclear, r L _as required. Personnel. performing'0- and F-designated activities are trained l mnd/or indoctrinated as necessary to assure that suitable Personnel outside the OA l roficiency is achieved and maintained. organization performing inspection and test are trained and 8 17.2-12 Amendment / i l
W T#a.,,,M ca. M& b MO' vosvm4 ierM m+ Pm O 'Y k h***'"**~ h ' W D % Personnel requiring certification are evaluated to establish i their qualifications for their respective level and discipline. Recertification is based upon demonstrated continued proficiency ( or requalification, if necessary. Personnel requiring certification in accordance with Regulatory Guide 1.58 are limited to NOA personnel who perform inspection and test activities,end members of the Operational Test Group (OTG) who perform post-design modification testing ' ^ ~ - ^ rh 4 4 eeHbeeMon personnel receive a periodic training needs assessment to identify additional supportive training needs as well cs to evaluate individual post-training performance. The e assessment period is three years o'c less. Inspection and test activities not requiring personnel certification per Regulatory Guide 1.58 include Technical Specification surveillances and periodic inspection and test of fire protection equipment. These personnel are qualified and retrained in accordance with applicable requirements of Regulatory Guide 1.8. Training programs of supporting organizations are described in their manuals, which are required to comply with the' quality assurance program. The Nuclear Training Center is responsible for the licensed operator training and retraining, in addition to other technical and supervisory training programs, including General Employee Indoctrination, which is required for all personnel having access to the station. 17.2.3 DESIGN CONTROL + The design control program includes activities such as field design engineering, associated computer programs, compatibility 4 of materials, and accessibility for inservice inspection, maintenance, and repair. During the operations phase, issuance of new drawings and l revisions to existing drawings require the implementation of a l-design change. L M eJw l The nuclear cup;;rf i m ms# procedures, approved by the l-anager - nuclear. crat!~"- QA, provide implementation guidance for the intent of Regulatory Guide 1.64 " Quality Assurance Requirements for the Design of Nuclear Power Plants." WRten 1:. t divici:., The nuclear engineering section has the following i responsibilities: DTh sc ps. & k& e&h,lpws% i,,s4u & de-s & & o & 3*! N s,oh % (f v4S w d WM f e j corrssk +wu W -f yg es+We. m u& tlo&, t r1tAucl< N k l l 1 % w.p ne ms~to los % k s S I ma d eSu +%~, po sa w "JYwA d4w ". b 17.2-14 Amendment f 7
~ 4/64 HCGS FSAR Prepare and update detailed engineering and design documents, including drawings and specifications, for a. ( all systems, components, and structures Specify applicable codes, standards, regulatory and b. quality requirements acceptance standards, and other design input in design documents Identify systems, components, and structures that are c. covered by the quality assurance program Per(orm design verification for systems, components, d. and structures Perform safety evaluations of proposed design changes e. Prepare documents for procurement of equipment, f. materials, and components Recommend eng'ineering consultants and laboratories for procurement services and coordinate their activities g. Review design documents submitted by suppliers h. (including the nuclear steam supply system (NSSS) supplier) and contractors i. Specify, or approve as required, inspections and/or tests A Designate whether they will see the service of other j. 3 qualified engineering organizations - The cognizant engineer is responsible for the identification and i. The purpose of design analyses is completion of design analyses.to assure that the technical design is acco controlled, and correct-manner. Types of design analyses include, but are not limited to, reactor physics,~ stress, seismic, thermal, hydraulic, radiation, and accident. 8 Amendment / 17.2-15 r- ,-n--.,-n-->,,~mm,,-,
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HCGS FSAR 4/84 I, In the event that the verification method for design .t. modifications is only by test, procedures and instructions will be written which include measures to ensure that: . s. a. Criteria are provided to specify when verification should be by test. ~ b. Where applicable, prototype, component or ' feature testing will be performed prior to installation of plant equipment. In those cases where this cannot be met, the testing will be deferred but not beyond the point when the installation would be irreversible. d. Tests wil1' be performed under conditions that simulate the most adverse design conditions, as determined by. l analysis. 4 Drawings are prepared by, or under the supervision of a designer from information received from the responsible engineer, manufacturer's drawings, etc. The drawings are reviewed and initialed-as being checked by another designer.or design supervisor. The drawings are approved by the Chief Designer or his assistants. Specifications and changes thereto for items covered by the quality assurance program are prepared by nuclear engineering and are reviewed and approved by NOA for quality assurance content. The NOA review assures that the documents are prepared, reviewed, and approved in-accordance with company procedures and that the ' documents contain the necessary QA requirements such as inspection and test requirements, acceptance requirements, and the extent of documenting inspection and test results. i' The station operations review committee (SORC) reviews proposed changes affecting nuclear safety and makes recommendations concerning implementation of the change to the_ general manager - Hope Creek operations. The design change process provides for sign-off of the design change by station department heads for the purpose of identifying required procedure change. If the proposed modification involves-a Technical Specification change, or is considered by-the SORC to involve an unreviewed safety question (10 CFR 50.59), the matter is submitted to the erle:r e J.(- s' h & Y e u ! w y Yt* i 8 17.2-17 Amendment)I
.~ HCGS FSAR 4/84 (05b - - 1. " before a license d;Eange request is submitted for NRC approval.for a determinatio rrei;; i:;s "KRL ( NOA reviews design changes for quality content both before and after implementation. Design changes are reviewed prior to implementation for-adequacy of inspection, test, and supplier . requirements, as applicable. The. design change package is reviewed following implementation for acceptability of documentation. During the development of a design, nuclear engineering section, i.e., electrical, mechanical, controls, structural / civil, design, is identified as the sponsor. The sponsor is responsible for the interface with affected disciplines within nuclear engineering, manufacturers, consultants, and PSE&G organizations outside of nuclear engineering, identified in documents such as contracts, specifications, purchase orders, design data sheets, and drawings. The primary interface between nuclear engineering and ~ external organizations is through the cognizant engineering section head or his assigned representative. Updating of records, including drawings, blueprints, instructions ( and technical manuals, and specifications resulting from design changes, is the responsibility of the general manager - nuclear uupy ;;. JDesign change procedures provide for the timely update of affected drawings'following design change implementation to reflect as-built configuration. 17.2.4 PROCUREMENT DOCUMENT CONTROL Procurement documents and changes thereto for the purchase of Q-and F-designated material, equipment, or services are reviewed and approved by NOA prior to issuance by the purchasing department to the prospective supplier. NOA review assures that spare and replacement parts are procured using controls which are commensurate with current operational QA program requirements. l l The review also assures that procurement documents adequately and correctly: a. Identify applicable quality assurance program requirements b. Reference applicable regulatory requirements, codes, and standards i j 8 17.2-18 Amendment / i -. ~
'8/84 HCGS FSAR the station operations review committee (SORC) for thchnical i ' content, by NOA for quality assurance requirements, and are -spproved by the responsible station department manager or his designee. c9neerYrg-. i The. general manager - nuclear 9n. p ct is responsible for issuing specifications, drawings, blueprints, and instruction and technical manuals associated with D-and F-designated structures, Approved and implemented modifications systems, and components.and design changes are incorporated to these referen Master lists of current editions or for the life of the station. revisions of these documents are periodically issued by the general manager - nuclear m ;g -t to the general manager - Hope i Creek operations to periodically ssure that only current and approved referenced documents a e used at the station. eyamang NOA reviews and approves station inspection plans and procedures including testing, that implement the quality assurance program, and repair. Changes to calibration, maintenance, modification these documents are also reviewed and approved. In addition, NOA is responsible for review and approva i PSE&G specifications, test procedures, and results of testing. M 17.2.6 DOCUMENT CONTROL Instructions, procedures, drawings, and changes thereto are reviewed for inclusion of appropriate quality assurance requirements and are approved by apppropriate levels of management of the PSE&G organizations producing such documents, l l and distributed on a timely basis to using locations. Measures are provided for the timely removal of obsoleted or superseded documents from the using location. Supplier documents are controlled according to contractual agreements with suppliers. i ,,, ream The-following is a generic listing of documents for the operational phase, showing organization responsibility for review and approval, including changes thereto: egeody [ Design specification - Nuclear Support, NOA l or a. 8.Jih%eien L b. Design,^ manufacturing, construction, and installation drawings - Nuclear -Support wl i wdur ser vtce$ Ucpc MC. t opW5, /Y%% 17.2-20 Amendment 7 l i-4 --,r ,,..m .,.,--.r --.r ,,~.~.---rm. ..-,.~ ~,,,., ~ m -,,,w_,e,-.v., _,,.-w,--,-,-ww-,.,,,--e,,,.y,,-,-,---
l ~ HCGS FSAR 4/84 - j. . (i d. Satisfactory'past history of-providing similar items y e.. Survey of supplier's facility. - The evaluations ~of the prospective suppliers are conducted using standard checklist form designed to include the 18 quality criteria of 10 CFR 50, Appendix B, as appropriate. Surveys of suppliers' capabilities
- include evaluation of management systems manufacturing processes and adherence to QA/QC procedures.
The results of supplier evaluations are documented by the appropriate checklist form and filed. Supplier control is maintained through a planned inspection, ~ l monitoring, and-audit program by NOA NOA and the responsible engineer conduct a review of the [ - manufacturing process for complex manufactured items, such as pumps, valves, heat exchangers, vessels, electrical panels,'etc. This review establishes
- critical inspection points and establishes a notification point program for the identified inspection or-surveillance activities.
The established inspection or surveillance activities are implemented by qualified NOA personnel or NOA agents. Commercial grade items l are not normally included in the notification point program. Receipt inspection and subsequent installation and test controls l provide the basis for acceptability of commercial grade items. 1 vawed Monitoring of suppliers / con ractors during fabrication, installation, modification, repair, inspection, testing, and shipment of Q-and F-designated materials, equipment, and services, is conducted by qualified NQA personnel or NQA agents at'the supplier's/ contractor's facility or at the generating station. Surveillances are conducted in accordance with written procedures and are designed to assure conformance with procurement requirements, in accordance with the safety significance of the item or service. Periodic evaluations of the supplier / contractor quality program are also conducted, consistent with the importance or complexity of'the item or service. Dependent upon the evaluation, additional audits or corrections by the supplier / contractor may be required. Supplier's certificates of conformance are periodically evaluated by audit, inspection, or test to assure 6 17.2-23 Amendment 8'
~ \\ 4e'8 4 ~ - HCGS FSAR s T. r84u'db Material identification and traceability is maintained for }; I . repairs, crpircr rrt, and modifications throughout operation. Organizations which implement requirements for.the identification and control of mate al ar.ts and components include Nuclear ope Creek Operations and NOA for Services,- Nuclear procurement document controls, and Nuclear Services, Hope Creek inspection Operations and NOA for receipt, storage, installation, - and test activities. 17.2.9 CONTROL OF SPECIAL PROCESSES Special process controls provide for the use of qualified s procedures, equipment, personnel, and documentation of satisfactory completion of an activity.- Special processes are generally those processes where direct inspection is impossible or disadvantageous. Procedures have been established for special processes such as welding, brazing, soldening, concreting, protective coating, cleaning, heat treating, and nondestructive examination (NDE) to The assure compliance with codes and design specifications. general manager - nuclear support is responsible for preparing I' These specifications are special process specifications. reviewed and approved by NOA for neces NOA monitoring and audits assure that qualification of special processes, equipment and personnel have been satisfactorily 4 performed. J Procedures for implementing the requirements of the specifications are prepared either by the nuclear department orare app by supplier persong gi p or his designee, (with the exception of q 4 manager - nuclear.up;_.m special process procedures prepared by code suppliers holding an Procedures prepared by suppliers are also reviewed "N" stamp). and approved by NOA. and personnel Qualification records of procedures, equipment, associated with special processes are retained as stated in -Section 17.2.17. 17.2.10 INSPECTION A planned inspection program is conducted and documented by personnel appropriately qualified ir accordance withThe inspection 2 Section 17.2.2. o 17.2-25 Amendment / ,y -nn.g,ve-.g-,.-, ,,_.--9,_,,--n.-nw., ,,,,-,._mnn_,,.-n_ _g__., _, w m n..w
~ 1, 4/84 4-HCGS FSAR b established procedure, code, or standard, consistent with the .ccm's or activity's importance to safety. 7 -Inrpection instructions include, as required, characteristics to b2 inspected, method of inspection, acceptance rejection t. criteria, required measuring and test equipment, and required rcierence documents. Documentation includes inspection idantification and results of inspection operation. inspection hold points are.ine.luded in the When required, procedure.or instruction and performed by NQA personnel. l Station department heads are also responsible for inserting ~ inspection hold points for critical activities in procedures they l The These inspection hold points are witnessed by NOA. -cpprove. ctation operations review committee (SORC) may recommend .cdditional or different hold points to the general manager - Hope ' Creek operations as a result of their review. Safety-related procedures are reviewed byADA prior to issuance and additional The ' inspection hold prints may be added to a procedure. inspection hold points cannot be passed without authorization representative. Typical critical 'l from the applicable QA/QC (NOA) e ttivities include: Visual inspection and NDE of AS!E pressure boundary a. welds Verification of cleanliness prior to closing safety-b. related systems Verification of reactor trip and engineered safety c. feature (ESF) initiation setting after adjustment Packaging and loading of radioactive material for d. shipment tems
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Hydrostatic testing of Q-and F-designat Acceptance testing of major modifications and repairs m f. to Q-and F-designated structures, systems, and components. 17.2-26 Amendment [ L
m HCGS FSAR 4/84 I ) Periodic inspection may be performed by qualified individuals other than those who performed or directly supervised the ( activity being inspected. These typically include periodic inspections of: a. Storage areas b. Housekeeping (general) c. Fire protection equipment d. Special handling tools and equipment NDE visual inspection required by the inservice e. insp)ction program Inspection of operating activities, i.e., work functions associated with the normal operation of the plant,. routine mainte::ance, and certain technical services, may be conducted by second line supervisory personnel or other quclified personnel not assigned first-line. supervisory responsibilities for conduct When inspections are so conducted and the activity of the work. involves breaching a pressure retaining boundary, the quality of i the work is demonstrated through appropriate testing unless restrictions, such as ALARA considerations, prevent such testing. l MA> nd retest requirements necessary to The applicable inspection assure that modifications,or repairs have been accomplished correctly are included in the design' change package, work order, The inspection and retest requirements for or procedur4 modificationfand repair are based on the original inspection and ~ test program, as well as the nature and scope of the modification or repair activity. Evaluation and review of inspection results are conducted by Level II individuals. A planned and documented monitoring program is conducted for 0-and F-designated activities. Monitoring of the implementation of the quality assurance program by station and site contractor NOA conducts enitcrin of personnel is conducted by NOA. supplier and (renritd contractor a eff-rifs 9 17.2-27 Amendment /
a. ~ /84 4 HCGS FSAR 3 Discrepancies found during the conduct of monitoring are brought to the attention of the management responsible for the activity. g h cyv>hh M m " % ! W TheMOAE)orhisdesigneeroutinelyattendsandparticipatesin plant work schedule and status meetings to assure that they are kept abreast of day-to-day work assignments throughout the plant and that there is adequate OA/QC coverage relative to procedural and inspection controls, acceptance criteria, and QA/QC staffing and qualification of personnel to carry out QA assignments. 17.2.11 TEST CONTROL Q-and F-designated equipment and components that must be tested periodically'to assure satisfactory performance, or have been replaced, modified, or repaired, are tested by qualified personnel in accordance with written procedures that provide design and procurement documents. Provisions are implemented that assure that nonconformances are corrected or resolved prior to the initiation of the preoperational test program on the item. Retest requirements are provided by engineering specifications and/or the responsible engineer, as were the original test The operational test group is responsible for requirements. preparation of test procedures incorporating the engineering Nuclear engineering reviews applicable test parameters. procedures; NOA reviews test procedures and test results. Test procedures prescribe, as applicable: Prerequisites, including completeness of test item (s) a. Instructions for performing the test b. Instrumentation and equipment for conduct of the test c. adequate to the test objective i Suitable environmental conditions and adequate test d. l methods .r. pp Amendment / 17.2-28
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