ML20098H043
| ML20098H043 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/03/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20098H044 | List: |
| References | |
| NUDOCS 8410090313 | |
| Download: ML20098H043 (199) | |
Text
,
O PS G Cornpany Pubhc Service E!ectnc 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 October 3, 1984 Director of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention:
Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen HOPE CREEK GENERATING STATION DOCKET NO. 50-354 DRAFT SAFETY EVALUATION REPORT OPEN ITEM STATUS 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 those items closed unless notified otherwise.
In order to permit timely resolution of items identified as
" complete" which may not be resolved to the staff's satis-faction, please provide a specific description of the issue which remains to be resolved. 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 previously submitted on October 1, 1984.
Pursuant to discussions with the Licensee Qualifications Branch, enclosed for your review (see Attachment 5) is a copy of revised FSAR Section 13.2 concerning training programs previously submitted on October 1 and 2, 1984.
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8410090313 841003 (h00 PDR ADOCK 050003S4 E
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The Energy People V 4 )Q (yh 4 R4
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Director of Nuclear Reactor Regulation 2
10/3/84 t
' Also enclosed.(see Attachment 6) is one copy of "An. overview of PSEAG Technical Quajafications and Management Capability in! Support of the Operation of Hope Creek Generating Station" previously transmitted on July'18, 1984, in a letter-from E. Liden, PSE&G, to F. Allenspach, NRC.
A signed' original of the required affidavit is provided to-document the submittal of these items.
LShould you ' have any questions or require any additional information.on these items, please. contact us.
Very truly yours,
'e Yf I httachments/ Enclosure C
D. H. Wagner USNRC Licensing Project Manager (w/ attach.)
W. H.
Bateman USNRC Senior Resident Inspector (w/ attach.)
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-354 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Public Service Electric and Gas Company hereby submits.the enclosed responses to DSER open items and revised FSAR Section 13.2 for the Hope Creek Generating Station.
The matters set 'forth in this submittal are true to the best of my knowledge, information, and belief.
Respectfully submitted, Public Service Electric and Gas Company
/
By:
)
' Thomas J.
drtin Vice Pre ident -
Enginee ing and Construction Sworn to and subscribed before-me, a Notary Public of ' New Jersey, this 3 d day of October 1984.
k Obl R4f(
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n DAVID K. BURD NOTARYPUBUC OF NEW JER$l'
- ,,;l My Comm. (spires 10-03-C5
<=,grp MP84 154/04 4-vw
DATE: 10/3/84 ATDosert 1 DSER R. L. MITIL TO OPEN SECIICN A. SOfWENCER ITEM NUMBER SUEDECr SDGUS LETIER DATED 1
2.3.1 Design-basis tanperatures for safety-Ccmplete 8/15/84 related auxiliary systens 2a 2.3.3 Act:uracies cf noteorological Canplate 8/15/84 measurm ents (Rev. 1) 2b 2'. 3. 3 Accuracies of meteorological Ccmplete 8/15/84 naasurements (Rev. 1) 2c 2.3.3 Accuracies cf nateorological Canplete 8/15 /84 measurements (Rev. 2) 2d 2.3.3 Accuracies of meteorological Ccmplete 8/15/84 measurments (Rev. 2) 3a 2.3.3 Upgrading cf onsite noteorological Ccmplete 8/15/ 84 measurements progra (III.A.2)
(Rev. 2) 3b 2.3.3 Upgrading of onsita meteorological Ccuplete 8/15/84 measurements program (III.A.2)
(Rev. 2) 3c 2.3.3 Upgrading cf onsite noteorological NRC Action measurements progran (III.A.2) 4 2.4.2.2 Ponding, levels Ccmplete 8/03/84 Sa 2.4.5 Wave impact ard rurup on service Ccmplete 9/13/84 Water Intake Structure (Rev. 3)
Sb 2.4.5 Wave impact and runup on service Canplete' 9/13/84 water intake structure (Rev. 3)
Sc 2.4.5 Wave inpact ard turup on service Ccmplete 7/27/84 weer intake structure 5d 2.4.5 Have inpact and runup on service Canplete 9/13/84 water intake structure (Rev. 3) 6a 2.4.10 stability cf erosion protection Ccmplete 8/20/84 structures 6b 2.4.10 Stability cf ertmion protection Canplete 8/20/ 84 structures 6c 2.4.10 Stability cf erosion protection Ccmplete 8/03/84 structures
~
M P84 80/12 1-gs 9
ATDGMENT 1 (Cont'd)
DSER R. L. MITIT., TO OPEN SECTICN A. SCHWENCER ITEM NUMBER SUETECT STATUS LETIER DMED 7a 2.4.11.2 Thermal aspects of ultimate heat sink C'aplete 8/3/84 c
7b 2.4.11.2
'Ihennal aspects of ultimate heat sink Ccuplete 8/3/84
^
8 2.5.2.2 Choice of maximum earthquake for New Cmplete 8/15/84 England - Piedmont Tectonic' Province 9
2.5.4 Soil danping values Couplete 6/1/84 10 2.5.4 Foundation level response spectra Conglete 6/1/84 1
11 2.5.4 Soil shear moduli variation Conglete 6/1/84 12 2.5.4 Ccabinaticn of soil layer properties Ccuplete 6/1/84 13 2.5.4 Lab test shear moduli values Ccmplete 6/1/84 t
14 2.5.4 Liquefaction analysis of river botton Ccuplete 6/1/84 sands 15 2.5.4 Tabulations of shear noduli Ccuplete 6/1/84 16 2.5.4 Drying and wetting effect on Ccmplete 6/1/84 Vincentown 17 2.5.4 Power block settlement monitoring Ccuplete 6/1/84 18 2.5.4 Maxima earth at rest pressure Caplete 6/1/84 coefficient 19 2.5.4 Liquefaction analysis for snevice Caplete 6/1/84 water piping 20 2.5.4 Explanaticn of observed power block Ca plete 6/1/84 settlement 21 2.5.4 Service water pipe settlemerit records Canplete 6/1/84 22 2.5.4 Cofferdam stability Conglete 6/1/84 i
M P84 80/12 2 - gs
.l I
ATEACIMNr 1 (Cont'd)
R.
I., MITIL 10 DSER A. SOMN 3R CPEN SECTICN ITEM NGEER SUEk7ECT STATtJS IETER CATED 23 2.5.4 Clarification of ESAR Tables 2.5.13 Canplete 6/1/ 84 and 2.5.14 24 2.5.4 Soil depth nodels for intake Canplete 6/1/84 structure 25 2.5.4 Intake structure soil nodeling Catplete 8/10/84 26 2.5.4.4 Intake structure sliding stability Canplete 8/20/84 27 2.5.5 Slope stability Cotylete 6/1/84 28a 3.4.1 Flood protection Carplete 8/30/84 (Rev. 1) 28b 3.4.1 Flood protection Canplete 8/30/84 (Rev. 1) 28c 3.4.1 Flood protection Canplete 8/30/84 (Rev. 1) 28d 3.4.1 Flood protection Canplete 8/30/84 (Rev. 1)
'28e 3.4.1 Flood protection Corplete 8/30/84 (Rev. 1) 28f 3.4.1 Flood protection Canplete 7/27/84 28g 3.4.1 Flocd protecticn Canplete 7/27/84 I
29 3.5.1.1 Internally generated missiles -(cutside Canplete 8/3/84 (Rev. 1) containment) 30 3.5.1.2 Internally generated missiles (inside Closed 6/1/84 contaianent)
(5/30/84-Aux.Sys.Mtg.)
31 3.5.1.3 Turbine missiles Caiplete 7/18/84 32 3.5.1.4 Missiles generated by natural phenonena Carplete 7/27/84 33 3.5.2 Structures, systems, and components -to Carplete 7/27/84 be protected fran externally generated missiles M P84 80/12 3 - os
- - - + - -
---+m__-,_
i ATEACHPENT 1 (Cont'd)
DSER R. L. MITII. TO CPEN SECTICM A. SODENGR ITEM NUMBER SUIL7ECT STATUS IEITER DATED 34 3.6.2 Unrestrained whipping pipe inside Cmplete 7/18/84 containment 35 3.6.2 ISI program for pipe welds in Ccuplete 6/29/84 break exclusion zone 36 3.6.2 Postulated pipe ruptures Ccnplete 6/29/84 37 3.6.2 Feedwater isolaticri check valve Ccmplete 8/20/84 cperability 38 3.6.2 Design cf pipe rupture restraints C m plete 8/20/84 39 3.7.2.3 SSI analysis results using finite Caplete 8/3/84 element method ard elastic half-space l
approach for cmtainnent structure 40 3.7.2.3 SSI analysis results using finite Ccmplete 8/3/84 l
element method and elastic half-space approach for intake structure f
41 3.8.2 Steel contairment bucklirg analysis Ccmplete 6/1/84 42 3.8.2 Steel containnent ultimate capacity C mplete 8/20/84 i
analysis (Rev. 1) 43 3.8.2 SRV/tCCA pool dynamic loads Ccnplete 6/1/84 44 3.8.3 ACI 349 deviations for internal Ccmplete 6/1/84 structures 45 3.8.4 ACI 349 deviations for Category I Cmplete 8/20/84 structures (Rev. 1) 46 3.8.5 ACI 349 deviations for fcundations Ccmplete 8/20/84 (Rev. 1) 47 3.8.6 Base mat response spectra Cmplete 8/10/d4 (Rev. 1) 48 3.8.6 Rocking time histories Ccmplete 8/20/84 (Rev. 1)
M P84 00/12 4 - gs a
ATTACHMENT 1 (Cont'd)
DSER R. L. MITE TO CPEN SECn Oi A. SOINENCER ITTM NUMBER SUBJECT STATUS IETTER QMED 49 3.8.6 Gems ccmcrete section Cmplete 8/20/84 (Rev. 1) 50 3.8.6 Vertical floor flexibility response Ccmplete 8/20/84 spectra (Rev. 1) 51 3.8.6 Cmparison cf Bechtel independent Canplete 8/20/84 verificatica results with the design-(Rev. 2) basis results 52 3.8.6 Ductility ratica due to pipe break Ccmplete 8/3/84 53 3.8.6 Design ci seismic Category I tanks Carplete 8/20/84 (Rev. 1) 54 3.8.6 Carbination cf wrtical responses Ccnplete 8/10/84 (Rev. 1) 55 3.8.6 Torsional stiffness cala21ation Ccnplete-6/1/84 56 3.8.6 Drywell stick nodel develcpnent Ccmplete 8/20/84 (Rev. 1) 57 3.8.6 actational time history irputs Canplete 6/1/84 58 3.8.6 "O" reference point for auxiliary Ccmplete 6/1/84 building model 59 3.8.6 Overturning mcment cf reactor Ccnplete 8/20/84 building foundation mat (Rev. 1) 60 3.8.6 BSAP element size limitations Ccuplete 8/20/84 (Rev. 1) 61 3.8.6 Seismic modeling cf drywell shield Ccmplete 6/1/84 wil 62 3.8.6 Drywell shield wall boundary Canplete 6/1/84 conditions 63 3.8.6 Reactor building dcme boundary Ccuplete 6/1/84 conditions M P64 80/12 5 - gs
ATDOiMENT 1 (Ccet'd)
DSER R. L. MITIL 'IO CPEN SECTICN A. SOfWENCER ITEM NLNBER SUILTECT S1XIUS IETIER DMED 64 3.8.6 SSI analysis 1211 a.itoff frequency Canplete 8/20/84 (Rev. 1) 65 3.8.6 Intake structure crane heavy load Caglete 6/1/84
&m 66 3.8.6 Impedance analysis for the intake Canplete 8/10/84 structure (Rev. 1) 67 3.8.6 Critical loads calculation for Corplete 6/1/84 reactor building cbne 68 3.8.6 Reactor txtildirg foundation mat Canplete 6/1/84 contact pressu*:M 69 3.8.6 Factors d safety against slidirq and Canplete 6,1/84 overturning of drywell shield wall 70 3.8.6 Seismic shear force distribution in Carplete 6/1/84 cylinder wil 71 3.8.6 Overturning d cylinder wall Conplete 6/1/84 72 3.8.6 Deep beam design d fuel pool walls Canplete 6/1/84 73 3.8.6 ASHSD done nodel load irputs Canplete 6/1/84 74 3.8.6 Tornado depressurization Canplete 6/1/84 75 3.8.6 Auxiliary building abnonnal pressure Canplete 6/1/84 76
,3.8.6 Targential shear stresses in drywell Canplete 6/1/84 shield wall and the cylinder wall 77 3.8.6 Factx d safety against overturning Carplete 8/20/84 d intake structure (Rev. 1) 78 3.8.6 Dead load calculations Canplete 6/1/84 79 3.8.6 Post-modification seis:lic loads for Canplete 8/20/84 the torus (Rev. 1)
M P64 80/12 6 - gs i
ATTJOlMENT 1 (Cont'd)
DSER R. L. MIT!L TU
' CPEN SECTICM A. SOMNCER ITEM POWER SUIUECT STATUS IErfER IRIED 80 3.8.6 Tbrus fluid-structure interactions Ccaplete 6/1/84 81 3.8.6 Seimnic displacement d torus Ccmplete 8/20/84 (Rev. 1) 82
- 3.8.6
-Review d seismic Category I tank Ccuplete 8,'20/84 design (Rev. 1) 83 3.8.6 Factors d safety for drywell Ccmplete 6/1/84 buckling evaluation 84 3.8.6 Ultimate capacity d contairnent Ccuplete 8/20/84 (materials)
(Rev. 1) 85 3.8.6 Icad conbination consistency Couplete 6/1/84 86-3.9.1 Ccaputer code validation Ccmplete 8/20/84 87 3.9.1 Information on trarmients Canplete,
8/20/84 88 3.9.1 Stress analysis and elastic plastic Ccmplete 6/29/84 analysis 89 3.9.2.1 Vibration levels for NSSS piping Ccaplete 6/29/84 systems-i 90 3.9.2.1 Vibration nonitoring program during Ccmplete 7/18/84 testing 91 3.9.2.2 Piping supports and anchors Ccaplete 6/29/84 92 3.9.2.2 Triple flued-head containnent Ccuplete 6/15/84 penetrations 93 3.9.3.1 Iced conbinations and allowable Ccaplete 6/29/84 stress limits 94 3.9.3.2 Desi@ of SRVs and SRV discharge Ccmplete 6/29/84 i
P ping e
M P84 80/12 7 - gs p
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ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL TO CPEN SBCrIm A. SOHN3R TIDI NLNBER SUIL7ECT STATUS IEITER DNTED 95 3.9.3.2 Fatigue evaluation cm SRV piping Canplete 6/15/84 and IOCA downoamers 96 3.9.3.3 IE Information Notice 83-80 Ccmplete 8/20/84 (Rev. 1) 97 3.9.3.3 Buckling criteria used for canponent Ccmplete 6/29/84 supports 1
98 3.9.3.3 Design cf bolts Ccnplete 6/15/84 99a 3.9.5 Stress categories ard limits for Carplete 6/15/84 core support structures
'99b 3.9.5 Stress categories ard limits for Canplete 6/15/84 l
core support structures 100a 3.9.6 10CFR50.55a paragraph (g)
Ccnplete 6/29/84 100b 3.9.6 10CFRSO.55a paragraph (g)
Complete 9/12/84 l
(Pav. 1) 101 3.9.6 PSI and ISI programs for pungs and Ccmplete 9/12/84 valves (Pav. 1) 102 3.9.6 Imak testing of press 2re isolation Canplete 9/12/84 l
valves (Pav. 1) 103a1 3.10 Seismic ard dynamic qualification of Ccmplete 8/20/84 nochanical and electrical equipment i
103a2 3.10 Seismic and dynamic qualification of Canplete 8/20/84 mechanical ard electrical equipment i
103a3 3.10 Seismic are dynamic qualification of ccup,lete 8/20/84 mechanical ard electrical equipment 103a4 3.10 Seismic and dynamic qualification cf Canplete 8/20/84 mechanical and electrical equipment M P84 80/12 8 - gs
ATDGMENT 1 (Cont'd)
DSER R. L. MITIL TC OPEN SECTICN A. SOfWENCER IT1!M NUMBER SUETECT STATUS IffIER IWrED 103a5 3.10 Seismic and dynamic qualification of Ccaplete 8/20/84 mechanical and electrical equipnent 103a6 3.10 Seismic and dynamic qualificaticn of Ccaplete 8/20/84 mechanical and electrical equipnent 103a7 3.10 Seismic and dynamic qualification of Ccmplete 8/20/84 mechanical and electrical equipnent 103bl 3.10 Seismic and dynanic qualificaticn of Cenplete 8/20/84 mechanical and electrical equipnent 103b2 3.10 Seisnic and dynamic qualification of Ccuplete 8/20/84 mechanical and electrical equipnent 103b3 3.10 Seismic and dynanic qualificaticn of Ccaplete 8/20/84 mechanical and electrical uguipnent 103b4 3.10 Seisnic and dynamic qualification of Caplete 8/20/84 mechanical and electrical equipnent 103b5 3.10 Seismic and dynanic qualification of Ccmplete 8/20/84 mechanical and electrical equipnent 103b6 3.10 Seismic and dynanic qualification of Caplete 8/20/84 mechariical and electrical equipnent 103c1 3.10 Seismic and dynanic qualification of Ccuplete 8/20/84 mechanical and electrical equipnent 103c2 3.10 Seismic and dynanic qualificaticn of Ccmplets 8/20/84 mechanical and electrical equipnent 103c3 3.10 Seismic and dynamic qualification of Ccuplete 8/20/84 mechanical and electrical equipnent 103c4 3.10 Seismic and dynanic qualificaticn of Ca plete 8/20/84 mechanical and electrical equipnent 104 3.11 Environmental qualification of NRC Action mechanical and electrical equipnent em M P64 80/12 9 - gs
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ATmaterr 1 (Cont'd) i R. L. MITII. TO DSER A. SOMENCER CPEN SECTICN I*I1!N NUMBER SUBJECT
,S*IATUS IfrIER DAIED 105 4.2' Plant-specific nadianical fracturing Caiglete 8/20/84 analysis (Rev. 1) 106 4.2 Applicability cf seismic andd ICCA Ccmplete 8/20/84 (Rev. 1) loading evaluation 107 4.2 Minimal post-irradiation fuel Canplete 6/29/84 surveillance program 108 4.2 Gadolina thermal conductivity Ccuplete 6/29/84 equation 109a 4.4.7
'IMI-2 Item II.F.2 Canplete 8/20/84 l
109b 4.4.7
'IMI-2 Itan II.F.2 Canplete 8/20 / 84 i
110a 4.6 Nnctional design cf reactivity Canplete 8/30/84 control systems (Rev. 1) 110b 4.6 mnctional design cf reactivity Conglete 8/30/84 control systems (Rev. 1) lila 5.2.4.3 Preservice inspection program Canplete 6/29/84 (ccmponents within reactcr pressure boundary) 111b 5.2.4.3 Preservice inspection program Ccuplete 6/29/84 (conponents within reactor pressure boundary) llic 5.2.4.3 Preservice inspection progran Ccmplete 6/29/84 (wv iwnts within reactor gressure o-boundary) 112a 5.2.5 Reactor coolant pressure boundary
. Ccmplete 8/30/84
, leakage detection (Rev. 1) 112b 5.2.5 Reactor coolant pressure boundar/
Ccuplete 8/30/84 leakage detection (Rev. 1)
M P64 80/1210
,gs e
ATDGMENT 1 (Cont'd)
D6ER R. L. MITIL 'IO OPEN SECTICH A. SOMNCER FITM IN3EER SUEL7ECT SIA'IUS LETIER DATED 112c 5.2.5 Reactor coolant gressure boundary Carplete 8/30/84 leakage detection (Rev. 1) 112d 5.2.5 Reactx coolant pressure MJrvimg Carplete 8/30/84 leakage detection (Rev. 1) 112e 5.2.5 Reactor coolant pressurs boundary Canplete 8/30/84 leakage detection (Rev. 1) 113 5.3.4 GE procedure applicability Carplete 7/18/84 114 5.3.4 Carpliance with NB 2360 cf the Sumer Corplete 7/18/84 1972 Addenda to the 1971 ASM Code 1 15 5.3.4 Drty wight ard Charpy v-notch tests Canplete 9/5/84 fx closure flange noterials (Rev. 1) 116 5.3.4 Charpy v-notch test data for base Carplete 7/18/84 materials as used in shell course No. I 1 17 5.3.4 Carpliance with NB 2332 of Winter 1972 Canplete 8/20/84 Addenda cf the ASMC Code 118 5.3.4 Imad factors and neutron fluence for Carplete 8/20/84 surveillance capsules 119 6.2
'DtI iten II.E.4.1 Canplete 6/29/84 120a 6.2
'IMI Itan II.E.4.2 Canplete 8/20/84 120b 6.2
'DtI Itaa II.E.4.2 Canplete 8/20 / 84 121 6.2.1.3.3 Uso d' NURf!3-0588 Canplete 7/27/ 84 j
122 6.2.1.3.3 Tuiperature profile Canpleto 7/27/84 123 6.2.1.4 Butterfly valve cperation (post Canplete 6/29 /84 accident)
M.P64 80/12 11 - qs 4
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ATDGMENT 1 (Cont'd)
R. L. MITIL 10 I
DSER A. SCHWENCER l
CPEN SECTICM ITEM NUMBER SUBIECT STAIUS TETIER DAIED 124a 6.2.1.5.1 RPV shield annulus analysis Cmplete 8/20/84 (any. 1) 124b 6.2.1.5.1 RPV shield annulus analysist Ca plete 8/20/84
( Rev. 1) 124c 6.2.1.5.1 RPV shield annulus analysis Cmplete 8/20/84 (Bev. 1) l 125 6.2.1.5.2 Design drywell head differential Ca plete 6/15/84 1
pressure 126a 6.2.1.6 Redundant position indiators fcr Caplete 8/20/84 vacuum breakers (and control roca alarms) j i
i 126b 6.2.1.6 Redundant position indicators for Cm plete 8/20/84 vacuun breakers (and control roan alarms) 127 6.2.1.6 Operability testing of vacuun breakers Caplete 8/20/84 (Rev. 1) 128 6.2.2 Air ingestion Coup,Lete 7/27/84 1-129 6.2.2 Insulatica ingestice Caplete 6/1/84 8
130 6.2.3 Potential bypass leakage paths Caplete 131 6.2.3 Administraticn of secondary contain-Caplete 7/18/84 ment openings 132 6.2.4 containment isolation review Cmplete 6/15/84
.j 133a 6.2.4.1 Containment purge system Caplete 8/20/84 i
l 133b 6.2.4.1 Containment purge systera Cmpleta 8/20/84 133c 6.2.4.1 Contairment purge systein Caplete 8/20/84
-i I
M P84 80/12 12-gs 7
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ATDC19ENT 1 (Cont'd)
DSER R. L. MITE. 10 a'
CPEN SECIICN A. SCHIENCER MTM NU1BER SUIDE.T STATtJS IETITR DATED 134 6.2.6 Catairment leakage testing Ccmplete 6/15/84 135 6.3.3 IPC5 and LPCI injection valve Ccmplete 8/20/ 84 interlocks 136 6.3.5 Plant-specific IOCA (see Section Ccmplete 8/20/84 15.9.13)
(Rev. 1) 137a 6.4 Control roan habitability Cmplete 8/20/84 137b 6.4 Cmtrol rocm habitability Ccmplets 8/20/84
!,y 137c 6.4, Centrol roon habitability Ccmplete 8/20 / 84 138 6.6 Preservice inspection gregram for Ccmplete 6/29/84 Class 2 and 3 s@wnts 4
139 6.7 MSIV leakage control system Ccmplete 6/29/84 s
140a 9.1.2 Spent fuel pool storage Ccmplete 9/7/84 (Pav. 2) 140b 9.1.2 Spent fuel geol storage Ccmplete 9/7/84 a
(Rev. 2) d40c 9.1.2 Spent fuel pool storage Ccmplete 9/7/84 V;
.)\\
(Pav. 2) 140d 9.1.2 Spent fuel pool storage Ccmplete 9/7/84 (Pav. 2) a' 141a 9.1.3 Spent fuel cooling ard clearmp Ccmplete 8/30/84 system (Rev. 1) 141b 9.1.3 Spent fuel cooling and clearup Ccmplete 8/30/84 system (Rev. 1) 141c 9.1.3 Spent fuel pool cooling and clearup Ccuplete 8/30/84 systen (Rev. 1) i',
~ M P84 80/12 13 - gs p
si c.-
ATDCHMENT 1 (Cont'd)
DSER R. L. MITIL 'IO OPEN SErrICH A. SODENCER
_ ITEM NLMBER SUBJECT STA'ItJS IfrITR DATED 141d 9.1.3 Spent fuel pool cooling ard clearup Ccmplete 8/30/84 system (Ruv. 1) 9 earup c 7 sta 8/30/84 141e 9.1.3 Spent fuel pool coolirg ard 1
1 system (Rev. 1) 141f 9.1.3 Spent fuel geol coolirg arx* clearup Ccmplete 8/30/ 84 systen (Rev. 1) 141g 9.1.3 Spent fuel pool coolirg ard clearop Ccznplete 8/30/84 systen
~
(Rev. 1) 142a 9.1.4 Light load hardlirg system (related Catplete 8/15/84 to refueling)
(Rev. 1) 142b 9.1.4 Light load handlirg system (related Ccanplete 8/15/84 to refuelirg)
(Rev. 1) 143a 9.1.5 overhead heavy load hardlirg Cceplete 9/7/84 143b 9.1.5 overhead heavy load hardlirg Ccrnplete 9/13/84 l
144a 9.2.1 Station service eter systen Canplete 8/15 / 84 (Rev. 1) 144b 9.2.1 Station servim water systen Ccrrplete 8/15/84 (Rev. 1) l 144c 9.2.1 Station service wter systen Ccmplete 8/15/84 (Rev. 1) 145 9.2.2 ISI program and functional testing Closed 6/15/84 cf safety ard turbine auxiliaries (5/30/84-coolirg systems Aux.Sys.Mtg. )
146 9.2.6 Switches and wirirg associated with Closed 6/15/84 HPCI/RCIC torus suction (5/30/84-Aux.Sys.Mtg.)
M P84 80/12 14 ga
ATIACINENT 1 (Cont'd)
DSER R. L. MITIL TO GEN SECTICH A. SOMENCER ITIN NUMBER SUE 7ECT STATUS LEITER DATED 147e 9.3.1 Carpressed air systens Canplete 9/21/84 (Rev. 2)
Carplete 9/21/84 147b 9.3.1 Carpressed air systems (Rev. 2) 147c 9.3.1 Corpressed air systems Canplete 9/21/84 (Rev. 2) 147d 9.3.1 Canpressed air systens Canplete 9/21/84 (Rev. 2) 148 9.3.2 Post-accident sangling systen Caiglete 9/12/84 (II.B.3)
(Pav. 1) 149a 9.3.3 Equipnent and floor drainage systen Canplete 7/27/84 149b 9.3.3 Equipnent ard floor drainage systen Canplete 7/27/84 150 9.3.6 Primary ccntairment instrunent gas Canplete 8/3/84 system (Rev. 1) 151a 9.4.1 Control structure ventilation systen Carplete 8/30/84 (Rev. 1) 151b 9.4.1 Centro 1' structure ventilation systsu Canplete 8/30/84 (Rev. 1) 152 9.4.4 Radioactivity nonitorirq elements Closed 6/1/84
?
(5/30/84-Aux.Sys.Mtg.)
153 9.4.5 Engineered safety features ventila-Carplete 8/30/84 tion syst.en (Rev 2) 154 9.5.1.4.a Metal roof deck ccmtruction Canplete 6/1/84 classificiation 155 9.5.1.4.b Ongoing review d safe shutdown NBC Action capability 156 9.5.1.4.c Orgoing review d alternate stutdown NRC Action capability M P64 80/1215 - gs
ATDGMENT 1 (Cont'd)
D6ER R. L. MITIL TO '
OPEN SECTIQi A. SQiWENCER ITDI NGEBER SUELTECT STATUS LEITER DAIT:D 157 9.5.1.4.e Cable tray protection Ca plete 8/20/84 158 9.5.1.5.a class B fire detection system Cmplete 6/15/84 159 9.5.1.5.a Primary and secondary power' supplies Complete 6/1/84 for fire detecticn systen i
160 9.5.1.5.b Fire water pung capacity Cmplete 8/13/84 l
161 9.5.1.5.b Fire water valve supervision Conglete 6/1/84 162 9.5.1.5.c Deluge valves Cmplete 6/1/84 163 9.5.1.5.c Manual hose station pipe sizing Coupleta 6/1/84 164 9.5.1.6.e Remote shutdown panel ventilation Cmplete 6/1/84 165 9.5.1.6.g energency diesel generator day tank Cmplete 6/1/84 protectim
- 166 12.3.4.2 Airborne radioactivity monitor Caplete 9/13/84 positioning (Rev. 2) 167 12.3.4.2 Portable continuous air nonitors Caplete 7/18/84 168 12.5.2 Equipment, training, and procedures Ca plete 6/29/84 for inplant iodine instrunentation 169 12.5.3 Guidance of Division B Regulatory Caplete 7/18/84 Guides 170 13.5.2 Procedures generation package Caplete 6/29/84
= hmittal 171 13.5.2-TMI Item I.C.1 Caplete 6/29/84 172 13.5.2 PGP remnitment Cmpleto 6/29/84 173 13.5.2 Procedures covering abnormal releases Caplete 6/29/84 of radioactivity 0
F M P84 80/12 16-gs
---.---.,-,,,,m.,..._m..
wm,.
.w.-
-.-y-,y..
,..--,,--ey,,.,-.,y-
ATD C MENT 1 (Cont'd)
DSER R. L. MITIL 70 j OPEN SECIICN A. SODENCER ITD4 NLMBER SUEL7ECT STATUS LEITER DMED 174 13.5.2 Resolution explanaticn in ESAR of Couplete 6/15/84 TMI Items I.C.7 and I.C.8 175 13.6 Physical security Open 176a 14.2 Initial plant test progra Carplete 8/13/84 176b
~14.2 Initial plant test progran Corplete 8/13/84 176c 14.2 Initial plant test program Corplete 7/27/84 176d 14.2 Initial plant test program Carplete.
8/24/84 (Rev. 2) 176e 14.2 Initial plant test program Canplete 7/27/ 84 176f 14.2 Initial plant test program Couplete 8/13/84 176g 14.2 Initial plant test pecgram Canplete 8/20 /84 176h 14.2 Initial plant test pecgram Canplete 8/13/84 1761 14.2 Initial plant test program Canplete 7/27 / 84 177 15.1.1 Partial' feedwater heating Canplei.e 8/20 / 84 (Rev. 1) 178 15.6.5 IDCA resulting fran spectrun of NRC Action postulated piping breaks within RCP 179 15.7.4 Radiological ccmsequences cf fuel NRC Action j
handling accidents 180 15.7.5 Spent fuel cask drcp accidents NRC Action 181 15.9.5 TMI-2 Item II.K.3.3 Corplete 6/29/84 182 15.9.10 TMI-2 Item II.K.3.18 Carpleta 6/1/84 183 18 Hope Creek DCROR Catpleta 8/15/84 M P84.80/12 17 - gs
,.-.,s,.,
. -. +,,.,,,.,.,,, _ _,
a,
---.-..,,,,,,,-m
0
'+ $:
^-
4 ATDolMDir 1 (Cont'd)
DSER R. L. MITIL 'IO
- CMN SECTICN A. SODENCER ITai NGEER SUETECT STA'ItJS IEITER DATED 184 7.2.2.1.e Failures in reactor vessel level Canplete 8/1/84 sensing lines (Rev 1) 185 7.2.2.2 Trip systen sensors and cablirg in Carplete 6/1/84 turbine building 186 7.2.2.3 Testability cf plant protection Canplete 8/13/84 systems at power (Rev. 1)
'187 7.2.2.4 Lifting cf leads to perform surveil-Carplete 8/3/84 lane testig 188 7.2.2.5 Setpoint e thodology Canplete 8/1/ 84 189 7.2.2.6 Isolation devices Canplete 8/1/84 190 7.2.2.7 Regulatory Guide 1.75 Canplete 6/1/84 191 7.2.2.8 Scram discharge volune Canplete 6/29/84 192 7.2.2.9 Reactor mode switch Canplete 8/15/84 (Rev. 1) 193 7.3.2.1.10 Manual initiation cf safety systems Carplete 8/1/84 194 7.3.2.2 Standard review plan deviations Carplete 8/1/84 (Rev 1) 195a 7.3.2.3 Freeze protection / water filled Canplete 8/1/84 instrument and sangling lines and j-cabinet tengerature control i
f 195b.
7.3.2.3 Freeze protection / water filled Canplete 8/1/84 instrument and sappling lines and cabinet tenparature control 196 7.3.2.4 Sharing cf common instrunent taps Canplete 8/1/84 197 7.3.2.5 Micrcprocessor, multiplexer ard Canplete 8/1/ 84 ccup ter systems (Rev 1) t M PS4 80/1218 - gs e
3
,-.,..,,,,#,-..---,,,,-,.,-,--,v-,,.-.w,-w,
,,w-.,
e,,--6-,---,,,-,,-,,--.wwy-,-n v-,--,-----*%-,,,.a.
ATDOBENT 1 (Cont'd)
DSER R. L. MITIL 'IO GEN SECTICN A. SODENCER ITDI NUMBER SUIk7ECT STAItJS IETIER DMED 198
' 7.3.2.6
.1MI Iten II.K.3.18-ADS actuation Canplete 8/20/84 E
199 7.4.2.1 IE Bulletin 79-27-Icss cf ncn-class Canplete 8/24/84 IE instrumentaticm and control power (Eev. 1) systen bus during cperation.
200 7.4.2.2 Remote stutdown systen Canplete 8/15/84 (Rev 1) 1 201 7.4.2.3 RCIC/HPCI interactions Canplete 8/1/84 202 7.5.2.1 Invel measurement errors as a result. Ccmplete 8/3/84 of environnental taiperature effects cn level instrunnntation reference leg 203 7.5.2.2 Regulatory Guide 1.97 Canplete 8/3/84 204 7.5.2.3
'IMI Iten II.F.1 - Accident nonitoring Canplete 8/1/84 205 7.5.2.4 Plant process conputer systen Canplete 6/1/84 206 7.6.2.1 High gessure/ low gessure interlocks Canplete 7/27/84 207-7.7.2.1 HELBs and consequential control systen Canplete 8/24/84 failures (Rev. 1) 208 7.7.2.2 Multiple control systen failures Ccuplete 8/24/84 i-(Rev. 1)
.209 7.7.2.3 Credit for ncn-safety related systens Canplete 8/1/84 l~
in Chapter 15 cf the FSAR (Rev 1) 210 7.7.2.4 Transient analysis recording systen Ccmplete 7/27/84 l
L 211a 4.5.1 Control red drive structural materials Ccuplete 7/27/84 l
211b
- 4.5.1 Centrol rcx1 driw structural materials Canplete 7/27/84 211c 4.5.1 Control rod drive structural noterials Ccuplete 7/27/84 l
l l-M P84 80/1219 - gs I -
--e
,.,,. n -
-n-_nn,
.,_,--_,,,en.,,,-..
' i.
-l ATDOiMENT 1 (Cont'd)
DSER R. L. MrITL'T CPM SECTIQi A. SCHWDiCER i
ITDI NUMBER SUE 7ECT SIAIUS TEITER DARD 211d 4.5.1 Control red drive structural m terials Cm plete 7/27/84 211e 4.5.1 Control rod drive structural materials Complete 7/27/84 s
212
_4.5.2 Reactor internals materials Cmplete 7/27/84 213 5.2.3 mactor coolant pressure boundary Complete 7/27/84 material l
214 6.1.1 Engineered safety features materials Couplete 7/27/84 215 10.3.6 Main steam and feedwater systern Ca plet-0 7/27/84 l
materials l
216a 5.3.1 anactor vessel caterials Cm plete 7/27/84 216b 5.3.1 anacter vessel materials Ca plete 7/27/84 217 9.5.1.1 Fire protection organization Caplete 8/15/84 218 9.5.1.1 Fire hazards analysis Ccmplete 6/1/84 219 9.5.1.2 Fire protection administrative Couplete 8/15/84 controls 220 9.5.1.3 Fire brigade and fire brigade Caplete 8/15/84
.i.
training 221 8.2.2.1 Physical separation of offsite Cmplete 8/1/84 i
transmissicn lines i
222 8.2.2.2 Design provisions for ;ar establish-Ccaplete 9/14/84
.l.
ment of an offsite power source (Rev. 1)
,j.
l 223 8.2.2.3 Independence of offsite ::ircuits Caplete 9/26/84 l
between the switchyard and class IE (Rev. 3) t L
buses 224 8.2.2.4 Ocmunon failure mode between ensite Ca plete 9/26/84 j
and offsite power circuits (Rev. 2) l-l M P64 80/12 20- gs l
I
p ATDd9ENT 1 (Cont'd)
DSER R. L. MITIL TO GEN SECTIO!
A. SOMN R
[
ITEM' NLMBER
~
Si.1, !k7ECT STATtJS IETIER DAIED
- 225 8.2.3.1 Testability cf automatic transfer d Cmplete 9/21/84 power fran the normal to preferred (Pav 1) power scurm 2 26 8.2.2.5 Grid stability Canplete 8/13/84 (Rev. 1) 227 8.2.2.6 Capacity and capability cf offsite Caplete 8/1/84 circuits 2 28 8.3.1.l(1) Voltage decp chring transient condi-Canplete 8/1/84 tions 229 8.3.1.1(2) Basis 'for usirg bus voltage versus Cmplete 8/1/84 actual connected load witage in the voltage drcp analysis 230 8.3.1.l(3) Clarification of Table 8.3-11 Caplete 8/1/84 231 8.3.1.1(4) Undervoltage trip setpoints Caplete 8/1/84 232 8.3.1.1(5) Load configuration used for the Cmplete 8/1/84 witage decp analysis 233 8.3.3.4.1 Periodic system testing Canplete 9/21/84 (Pav. 2) 234 8.3.1.3 Cgacity and capability cf cnsite Canplete 8/1/84 AC power supplies and use cf ach ministrative controls to prevent overloading cf the diasel generators 235 8.3.1.5 Diesel generators load acceptance Ccmplete 9/21/84 test (Pav. 2) 236 8.3.1.6 C mplians with position C.6 of Canplete 8/1/84 IG 1.9 237 8.3.1.7 Decripticn cf the load sequencer Cmplete g8{
238 8.2.2.7 Sequencing cf loads cn the offsite Ccuplete 9/21/84 power systen (Pav. 1)
M P84 80/12 21 - gs
'l.
!i i
.,s ATDOBerr 1 (Cent'd) i R. L. MITIL TO DSER A. SCBMN2R GWI SECITGI ITDI NUMMt SURRCr SDatB IETTER DATED l.
239 8.3.1.8 Testing to verify 808 mini: sus Caplete 8/15/84 l
voltage 240 8.3.1.9 t'<=pH.ance with BIP-PSB-2 Caplets 8/1/84
~
'l
.241 8.3'.1.10 Ioad acceptance test atw ganged Ca plete 9/21/84 no Iced cperation cf the diesel (Rev. 3)
I generatax I
{
242 8.3.2.1 cmpliana with position 1 cf Regula-Couplete 9/13/84 (Rev. 1) tory Guide 1.129 i
l 243 8.3.3.1.3 Protection or qualification cf Class Cm plete 9/13/84 (Rev. 1) 1 18 equ!psamt from the effects of j
fire suppression systems 244 8.3.3.3.1 Analysis and test to demonstrate Complete 9/29/84 I
I
=4arriary cf less than specified (Rev. 2A) separation 245 8.3.3.3.2 The uso d 18 versus 36 inches cf Complete 9/28/84 separaticn between raceways (Rev. 2B) g.. t I
246 8.3.3.3.3 Specifi,ed separation cf raceways by Caplets 8/1/84 analysis and test
-l
' 2 47 8.3.3.5.1 Capability d pe strations to with-Complete 9/13/84 stand long duraticm short circuits (Rev. 1) at less than maxima or worst cs.se I
short circuit l
248 8.3.3.5.2 Separation cd pnetration primary Casplete 8/1/84 i
and tackup protections 249 8.3.3.5.3 The use cf bypassed thermal overIced Ccaplete 8/1/84 p
[
protective devices for penetration protections 250 8.3.3.5.4 Testing d fuses in accordana with Caplete 8/1/84 f-R.G. 1.63 l
M P64 80/12 22 - gs
t-ATr>GMENT 1 (Cont'd) f
.I DSER R. L. MITTL It j
OPEN SECTICN A. SQWENCER t
ITEM NUMBER SUBJECT STATUS LETITR DATED l
251 8.3.3.5.5 Fault current analysis for all-Ca plete 9/24/84 representative penetraticn circuits (Rev. 3) 252 8.3.3.5.6 The use of a single breaker to provide Caplete 9/21/84 penetration protection (Rev. 2) 4 j-253 8.3.3.1.4 Ca mitment to protect all Class lE Cmplete 9/28/84
.I equipent fran external hazards versus (Rev. 3A)
I only class 1E equipnent in one division 1
l 254 8.3.3.1.5 Protection of class lE power supplies Caplete 9/14/84 from failure of unqualified class lE (PaV. 1) loads j
255 8.3.2.2 Battery capacity Ca plete 8/1/84 r
L 256 8.3.2.3 Autimatic trip of loads to maintain Caplete 9/13/84 sufficient battery capacity (Fav. 1)
-257 8.3.2.5 Justification for a 0 to 13 second Ccaplete 9/13/83 load cycle (Pav. 1) 258 8.3.2.6 Design and qualification of DC Ca plete 8/1/84 system loads to operate between
,]
minimum and maximum voltage levels t
259 8.3.3.3.4 Use of an inverter s an isolation Caplete 10/3/84 de..tce (Rev. 3) 260 8.3.3.3.5 Use of a sirgle breaker tripped by Ccmplete 10/3/84 a IDCA signal u ed as an isolation (Rev. 2) devi.ca 261 8.3.3.3.6 Autmatic transfer of loads and Ca plete 9/13/84 interconnection between redundant (Fav. 1) l divisions i
262 11.4.2.d Solid waste control progran Caplete 8/20/84 l
=i M P84 80/12 23-gs 1.
ATDOIMENT 1 (Cbnt'd)
DSER R. L. MITIL TO OPEN SECTICN A. SCHkHiCER ITEM NUMBER SUBJECT SIATUS LETIER DA2i!I) 263 11.4.2.e Fire protection for solid radwaste Ccuplete 8/13/84 storage area 264 6.2.5 Sources of o:ygen Cm plete 8/20/84 265 6.8.1.4 ESF Filter Testing Cmplete 8/13/84 266 6.8.1.4 Field leak *ests Cm plete 8/13/84 267~
6.4.1 Control rusa toxic chemical Cmplete 8/13/84 detectors 268 Air fil zation unit drains Caplete 9/13/84 (Rev. 1) 269 5.2.2 Code cases N-242 and N-242-1 Cmplete 8/20/84 270 5.2.2 Code case N-252 Ca plete 8/20/84 TS-1 2.4.14 Closure of watertight doors to safety-Open related structuras TS-2 4.4.4 Single recirculaticn locp operaticn Open TS-3 4.4.5 Core flow monitoring for crud effects Cm plete 6/1/84 TS-4 4.4.6 Inose parts monitoring systen
@n l
TS-5 4.4.9 Natural circulaticn in normal Open operation i
i TS-6 6.2.3 Secondary containment negative open pressure TS-7 6.2.3 Inleakage and drawdown time in Open secondary containment TS-8 6.2.4.1 Imakage integrity testing Open TS-9 6.3.4.2 BCCS subsystem periodic wwwnt open l
testing M P64 80/12 24-gs
ATIACHMENT 1 Kbit'd)
DSER R. L. MITIL TC OPEN SECTICN A. SQlWENCER ITEM NUMBER SUBJECT STATUS LETIER DAITD TS-10 6.7 MSIV leakage rate l
TS-ll 15.2.2 Availability, setpoints, and testing Open of turbine bypass system TS-12 15.6.4 Primary coolant activity IC-1 4.2 Fuel rod internal pressure criteria Caplete 6/1/84 IC-2 4.4.4 Stability analysis submitted before Open second-cycle operation h
1
?
4 I
t M P84.80/12 25-gs
ATTACHMENT 2 DATE - 10/3/84 DRAFT SER SECT. IONS AND~ DATES PROVIDED SECTION DATE SECTION DATE 3.1 3.2.1 11.4.1 See Notes l&5 3.2.2 11.4.2 See Notes 165 5.1 11.5.1 See Notes 1&S 5.2.1 11.5.2 See Notes 165 6.5.1 See Notes 1&5 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 l
8.2.4 See Note 2 13.3.2 See Note 4 i
8.3.1 See Note 2 13.3.3 See Note 4 8.3.2 See Note 2 13.3.4 See Note 4 8.4.1 See Note 2 13.4 See Note 4 8.4.2 See Note 2 13.5.1 See Note 4 8.4.3 See Note 2 15.2.3 8.4.5 See Note 2 15.2.4 8.4.6 See Note 2 15.2.5 8.4.7 See Note 2 15.2.6 8.4.8 See Note 2 15.2.7 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 8/3/84 9.5.8 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&5 l
10.4.3
.See Notes 3&5
[
10.4.4 See Note 3
~
11.1.1 See Notes 165 Notes:
11.1.2 See Notes 1&5 11.2.1 See Notes 165
- 1. Open items provided in
~
11.2.2-See Notes 165 letter dated Jul'y 24, 1984 i
j 11.3.1 See Notes 1&5 (Schwencer to Mitti) 11.3.2 See Notes 165
- 2. Open items provided in June 6, 1984 meeting
- 3. Open items provided in April 17-18,198 4 meeting CT:db
- 4. Open items provided in May 2. 1984 meting
- 5. Draft SER Section provided in letter dated August 7, 1984 (Schwencer to Mittl) 1HP 84 95/03 01
DSER OPEN ITEMS 259 8.3.3.3.4
'Use of an inverter as an isolation device 260 8.3.3.3.5 The use of a single breaker tripped by a l
LOCA signal as an isolation device 1
l l
I l
l
t i
l ATTACHMENT 4 e
i 9
Rev. 3 041 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-l 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 lE de system th', rough an inverter.
The applicant further stated that this inverter is an acceptable isolation device per IEEE-384-1981, Section 7.1.2.3.
The staf f. does not agree.
Test and analysis to demonstrate the adequacy of an inverter as an isolation device will be pursued with the applicant.
RESPONSE
- 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.
5 1
v - +
-ws~,
rm-w--,,ew-,--.v-g-,-
,~v
,w
,w--...-e%.s.-----,
wem,m,-.,.-,v.-=.
,,-,.e----
,s.+..r.w-s-.,---~--
, g Gue.s4.'cn 430.33 DSER Open Item No. 260 (DSER Section 8.3.3.3.5)
TER USE OF A SINGLE BREAKER TRIFFED BY A LOCA SIGNAL AS AN ISOLATION DEVICE
]
section s.3.1.1.2 of.the PSAR indicates that the Class 1E systee provides power to non-Class 1E loads..Non-Class 1E loads are I
connected to the Class 1E system through a single breaker that is tripped automatically by a LOCA signal.
The single breakse tripped by a LOCA signal provides acceptable isolation between Class 1E and non-Class lE circuits, for the design basis accident--
LOCA.
Bowevers for other design basis accidents or operating occurrences that do not generate a LOCA signal (such as loss of of f site power, de sign. basis exposure fire, seismic events, etc.),
is is the staff concern that a single breaker may not provide acceptable isolation.
By Amendment 4 to the FSAR, the applicant indicated that protect-tive device coordination studies show that the single britake r time ove rcurrent trip characteristics will trip to clear a fault prior to initiation of a trip of a upstream breaker.
Identifi-cation of all non-Class 1E circuits being isolated using a single breaker trip by LOCA signal, periodic testing of breaker coordi-nation, and capability of breaker to trip prior to any versus only upstream breaker and for all versus only circuit faults, wil:
be pursued with the applicant.
d
RESPONSE
Re s ponse to Question 430.33 has been revised to provide the requested info rmation.
b
^
i
/n. 3 NCGS FSAR oursT1oM 430.33 (SECTION 3.3.1 and 3.3.3)
Section 3.3.1.1.2 of the FSAR indicates that the Class 1E system Fon-Class 1E* loads are provides power to non-Class 1E loads. connected to the Class 1 The single breaker tripped automatically by a LOCA signal.
! tripped by a LOCA signal provides acceptable isolatlon 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
- 14CA.
?
offsite power, design basis exposure fire, seismic events, etc),
F 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 failure of anyone or simultanous combined failure of all non Claes 1E loads will not prevent any of the four channels of Class 1E power from performing its safetyThe an capacity and capability of onsite and offsite power supplies function.
and their associated distribution system to supply power to (1)
Class 1E loads within their design ratings for all modes of plantth (2)
(3) an analysis of diesel generator loadings for loss operation, of of fsite power similiar to that presented in Tables 8.3-2the failure of the N 384-1981, (4) through 8.3-6 of the FSAR, system that supplies control power,to the subject non non-Class 1E 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 event that do not generate LOCA signals.
design ba g
shows the two configurations that employ a Figure 430.33-1 circuit breaker tripped by a LOCA signal as an isolation device.
The two configurations ares A Class IE unit substation supplies a 'non-Class 1E motor control center (MCC) or a motor load through a.
R Class 1E circuit breaker B.
g lA Class 1E motor control center supplies t'hro b.
l panel.
The Class 1E circuit breakers B and D are quali[ied to operate 5
for HCGS seismic and environmental parameters for all des g
basis events.
a Ameadsent 4 430.33-1
~
k0V secs rsAn respective Class IE power supply buses from the non-Class 1E loads in the event the non-Class 1E loads fail.
This applies whether the plant is supplied from an offsite source or an onsite Thus, the failure of the non-Class IE locds supplied IE power supply buses will not prevent any of the four source.
from Clas channels of Class 1E power supplies f rom performing its safety function.
g g g,n.p 384-1981 COMPLIANCE WITH GUIDELINES OF SECTION 7.1.2.1 OF IEEE Protectiive device coordination studies for devices shown in Figure 430.33-1 have shown that the time-overcurrent trip characteristics of circuit breakers A, B, C, and D are such that Circuit breaker B will trip to clear a fault current a.
prior to initiation of a trip of circuit breaker A.
Circuit breaker D will trip to clear a fault current b.
prior to initiation of a trip of circuit breaker C.
Both the onsite and offsite powers supply sources are separately capable of supplying the necessary fault current for sufficient time to ensure the proper prp ective device coordination without loss of function of Class 1E loads.
DBY DIESEL GENERATOR LOADINGS FOR LOSS OT OTTSITE POW ST tabulates the loads, their KW ratings, and loading Tab 3e 8.3-1 (DSA) and loss of offsite sequences for design basis accidentIt can be verified by ins.pecting scenarios.
power (LOP) that DBA loading of the SDGs is the limiting case Table 8.3-1 with respect to the leading capability of the SDGs.
l TAILURE OF THE NON-CLASS 1E DC SYSTEM THAT SUPPLIES C TO THE SUBJECT NON-CLASS 1E LOADS (described above) the circuit breaker B For configuration (a) supplying a Non-Class II MCC or a motor load is controlled byFor a non-Class IE m Class 1E 125 V de control power supply.a non-Class IE circuit break
/
I
- load, This non-Class IE circuit breaker (GI-AKR circuit breakger B.is controlled by a non-Class 1E 125 V de control power.
GI-AKR type circuit breakers are direct %g acting trip dev type) i Therefore, the failure of the <'
electrical fault conditions. control power supply does not prevent the circuit br rip @.JS
[p in response to the f ailure of non-Class 1E motor load.
- IIUegT.
- C Fw pm,esqs6.syt6 esta orts Itzx M Amendment 4 430.33-2 l
i p,.y wCGAT A
........ Th e Cl4 S b (L #4SI4C a f. SeOrdt S a nd '84 f effc!4 g gas,fr WPCC $
mal 4kslt dlstrlbuilsn system aee *< suHtelen+ cepCel4y a.g fa both Clau Il a nd n on. g(4,, gg
'A fr Mtl4 3 4e safyty fewercondiitsn s.
2 n 1 4 e e,re n f o" f < L o c 4 i* Ads daring a// plani loads are nafem<t'u.n/fy < riffed &om -t 4 ske., son. c/nss IL
^ lass IL buses to, ateobee w;.4h Post 4 tan c.I of Aeyat.<.4,,,
- k'b IiS.
A op> rt 0+J, C.p B l.6 s F/tont Tile CLACs.M (fsisa6 70 1'YE N*A'~C LA'f Ta V oA.ps p ac pB t.o ta p
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CA G LS S
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h kW OPEsarna OKS/aw C J/d"*x Co~ raat Pao w d.
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EF'ffc.7' Mr, Tid f
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coo.,A worH 7H15 RE Q UInGon BN T, gpppy/00 ALLY; Typ fB47twnwT wisL B2
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pcs/ Vox Pr e u nt enir.s I N/S N S 0 "'" # '" ' ^'
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,79 p
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- TAblt-4 %.3>2rl ide.nfifies 4ke non -4less i2 lands *ffd 4 re b u f f t k N h r 1 ) In eirY. ult breAStrs E a.nd D o.( fu urg AM 55.g, II DSEE OPEN ITEM M 45,c,W 2A
. QUESTION 430.33 Insert "C" ANALYSIS FOR SUPPLYING NON-CLASS lE FROM CLASS lE DC SYSTEMS Figure'8.3.ll shows non-Class lE public address system distribution panel 10J496 supplied from a Class lE de power bus 10D410 through a Class lE inverter in UPS unit 10D496.
The inverter is an acceptable isolation device per IEEE-384-1981, Section 7.1.2.3.
Tharefore, a failure in'the non-Class lE distribution panel 10J496 will not
' degrade Class 1E dc system bus 10D410.
The MCGS UPS system will be tested to demonstrate the adequacy of an invester being applied as an isolation device.
The test will demonstrate that voltage, current, and frequency on the Class lE side of the UPS are not degraded below acceptable levels when maximum credible voltage or current transient is applied on the non-Class lE side of the UPS system.
The tests to be performed will simulate a11' operating modes for which the HCGS UPS system is l
designed.
The tests will include the following types of faults at the UPS output location:
{
a.
Phase to ground j
b.
Neutral to ground c.
Phase to neutral without ground j
d.
Hot short (460 Vac).
l A test plan is submitted separately for the staf f's review.
The test i
report and any associated analysis of the test results will be i-submit $ed in December 1984.
i An analysis has been performed to support the values used for the acceptance criteria for voltages.
This analysis shows that the voltages specified will not cause misoperation or loss of any l
electrical. equipment connected to the supply buses.
The results of this analysis for the ac systems is stated in FSAR Section 8.3.1.2.1 and the calculated results a; a shown in Table 3
i 8.3-11.
The results of the de analysis are cor.**.ined in FSAR section 8.3.2.
These results indicate that the 125 volt de system has an acceptable operating capability with battery voltage variations
[
of 35 volts (140 volts dc to 105 volts dc).
The test acceptance j
criterion limits the bus voltaga variation to 105-135 volts.
In addition, the acceptance values for the test currents are well below the level that would cause the infeed breakers to the UPS supply buses to trip.
These values are as follows:
Acceptance Infeed breaker Circuit Current Setting Normal 480 VAC 0-55 amperes continuous
'600 amperes supply with a maximum peak not Pick-up to exceed 132 amperes and no value above 55 maperes
{
shall persist for longer than 10 mS Page 430.33-2B(1)
Insert "C" Af v. 3 Pcgo two Acceptance Infeed breaker Circuit Current Setting Back-up 480 VAC 0-78 amperes continuous 600 amperes Supply with a maximum peak not Pick-up to exceed 500 amperes and no value above 78 amperes shall persist for longer than 10 ms Alternate 125 VDC The bus voltage variation 2000 ampere Supply of 105-135 volts will hold fuse for the following cases:
(1) With the UPS energized but without load the input current should not exceed 56 amperes (2) With the UPS input current at 56 amperes the input current should not exceed the range of 0-56 amperes (3) With the UPS input current at 158 amperes the input current should not exceed the range of of 0-158 amperes The fo,llowing is justification that the above acceptance current values'do not adversely effect the Class lE buses.
The 480 volt ac back-up feed is supplied from a 480 volt Class lE motor control center which in turn is supplied from a 480 volt Class lE unit substation.
The infeed breaker to the MCC is an AKR - 50 which has a 600 ampere pick-up setting for its time delay trip setpoint.
This allows the lcrgest motor loads on the MCC, in combination with the maximum acceptable current spike of the UPS acceptance values (500 amperes for not longer than 10 mS), to persist for 25 seconds.
Since the 500 ampere spike is completed in 10 mS, the largest motor loads then have 55 seconds to accelerate.
This is 48 seconds longer than the time delay for the primary protective device for the largest motor and, therefore, it is not possible for any of the Class lE loads to be disabled.
The inrush current of the normal ac feed is 132 amperes for 10 mS which is less than the 480V ac backup supply.
The normal 480V supply breaker is the same type and size as the 480V back-up supply breaker.
Therefore the Class lE loads on the MCC's from the normal and backup 480Vac supply are not affected by any short circuits on the output of the inverter.
The alternate 125V de supply full load amperes are already included in the 125 volt battery load profiles.
The maximum current duty on any of the 125 volt Class lE batteries is 451.1 amperes (battery 1AD411).
The impedance of the conductors from the battery to the 125 volt de bus is such that the voltage drop for the specified load profiles does not cause the 125 volt bus to drop below 105 volts.
Page 430.33-2B(2)
N 14SIGs 1
Pcgo 3
)
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 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 separated from Class lE cables associated with other Class lE channels or an isolation means acceptable to the staf f will be employed.
Page 430.33-2B(3)
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'FOR MOTOR LOADS,IN ADDITION TO CIRCulT BREAKER S.
THERE IS NON CLASS lE CIRCulT BREAKER DOWNSTREAM OFBhEAKER5.
4.14 KV CLASS 19 SUS 4 ;
f
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4 O
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'NON CLASS lE MOTOR OR, MOTOR CONTROL CENTER 480 V CLASS lE MCC
) D TRIPPED BY LOCA i
a NON CLASS lE CASLE 1 I BACK UP POWER SUPPLY FOR UPS SUPPLYING.
NON CLASS lE DISTRIBUTION PANEL.
HOPE CREEK GENERATING STATION FINAL SAFETY ANALYSIS REPORT ISOLATION SETWEEN CLASS lE POWER SUPPLIES AND NON CLASS lE LOADS-TRIPPING CIRCUlf BREAKER FIGURE 430.331 AMEND *, LENT 4,144
i TEST PROCEDURE. ISOLATION VERIFICATION i
S/N 9743 IE 20KVA,UPS (INSTRUMENTATION AC POWER SUPPLY)
FOR PUBL'IC SERVICE ELECTRIC & GAS'C0.
HOPE CREEK GENERATING STATION P0. 10855-E-154 (Q)-AC 08 JECTIVE i.'
TESTING T0 ESTABLISH THE UPS SYSTEM AS A CIRCUIT ISOLATION SYSTEN.
PASS CRITERIA:
DEFINI. TION 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 Y0LTAGE OR CURRENT TRANSIENT APPLIED TO THE MON CLASS 1E SIDE OF THE DEVICE WILL~NOT DEGRADE THE CLASS 1E CIRCUIT ON'THE OTHER 51DE OF THAT DEVICE.
NORMAL VARIATION CIRCUIT ALT. DC. SUPPLY 105-115 VDC 1-0-FULL LOAD Apc NORMAL AC S'!PPLY 48 0+101 V(L-L) 3 PHASE 0-55A. 0-132AP FOR 10 MSEC
-8ACK UP AC SUPPLY 48 0+101 V 1 PHASE 0-78A.'0-500AP FOR 10 MSEC ANY VARIATIONS OUTSIDE OF P.ORMAL VARIATIONS SPECIFIED. WILL BE ANALYZED ON A CASE BY CASE BASIS.
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FAULT LOCATION AND TYPE FAULTS WILL BE APPLIED TO UPS SYSTEM 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 (H0T SHURT) l THE CONDITION OF THE THREE CLASS 1E SOURCES WILL BE MONITORED THROUGH' SUITABLE SIGNAL CONDITIONERS, BY GOULD INC., 2000W SERIES HIGH FREQUENCY RECORDING SYSTEM.
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Rn.5 TEST PROCEDURES 1.0 GENERAL NOTES l
1.1 BEFORE STARTING TEST DETERMINE AND RECORD ALL SIGNAL CONDITIONER TRANSFER RATION (MULTIPLIER) VALUES.
1.2 NORMAL SYSTEM OPERATION DURING EACH TEST A.
CONNECTION PER FIG. 1.
B.-
THE LOAD ON THE UPS SHALL BE ADJUSTED FOR EACH OF THREE SEPARATE TESTS FOR EACH UPS INPUT SOURCE:
1)
NO LOAD
-(2)
(
OUTPUT LOAD AT.08 PF TO ACHIEVE 56 AMPERES INPUT CURRENT WHEN FED FROM 125 VOLT DC.
LOAD SHOULD REMAIN THE SAME FOR AC INPUTS (3)
OUTPUT LOAD AT.08 PF TO ACHIEVE 158 AMPERES INPUT CURRENT WHEN FED FROM 125 VOLT DC.
LOAD SHOULD REMAIN THE SAME FOR AC INPUTS 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 B.YPASS SWITCHES IN
" NORMAL" POSITION
" TEST" SWITCH - CENTERED
" RETURN MODE" SWITCH - IN "AUT0" POSITION
" ISOLATION" TOGGLE SWITCHES - ON
" SYNC" TOGGLE SWITCH - ON 1.3 Tf3T IhSTRUMENTATION A.
GOULD INC., MODEL 2800W HIGH FREQUENCY RECORDING SYSTEM.
EIGHT CHANNEL. INDEPENDENT SCALE SELECT 1.050 T01500 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.
(.01Y/A).
D.
WIDEB AND DC ISOLATION AMPLIi J.R. GOULD INC. MODEL 13-4615-10 OR EQUIVALENT.
~
80/.}
1.4 TEST FACILITY AND EQUIPMENT A.
DC SUPPLY - C&D 4LCW-15 BATTERY (80 CELLS. 80KW FOR 30 MIN.)
AND BATTERY CHARGER.
,8.
AC SUPPLY - 480V 3 PHASE. 4W 60 HZ,1200A GROUNDED NEUTRAL.
C.
AC LOAD BANK 30KW OR 0-30KVA 9 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.
2.0 TEST PR0CEDURE 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.
A1.
METERING AND CONNECTIONS PER FIG. 2 AND " BACKUP SOURCE" 8REAKER 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.
82.
REPEAT 01 EXCEPT STATIC SWITCH TRANSFERRED TO BACKUP.
83.
RE-PEAT 81 EXCEPT USE 500HZ TIME BASE.
i 84 REPEAT B2 EXCEPT USE 500HZ TIME BASE.
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$W.3 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 (0R 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.
REPA!R THE UPS IF REQUIRED BEFORE PROCEEDING.
C1.
INSTALL JUMPER "A" T0 " FAULT" CB WITH " BACKUP SOURCE" CB 0 PEN WITH RECORDER AT 20KHZ TIME BASE APPLY FAULT PER CO.
C2.
REPEAT C1 EXCEPT WITH 500HZ TIME BASE.
' OPEN " NORMAL SOURCE" CB AND CLOSE " BACKUP" WITH RECORDER C3. '
-20KHZ TIME BASE APPLY FAULT PER CO, C4, 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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FAULT LOCATION AND TYPE FAULTS WILL SE AFPLIED TO ISOLATING TRANSFORMER OUTPUT TERMINAL 5 SY CLO5tNG A-5 WITCH AS REqu!REO.
FAULT TYPES:
1.
PHASE (H0T) TO GROUND 2.
NEUTRAL TO GROUND 3.
PHASE TO NEUTRAL W/0 GROUNU 4.
4bOVAC APPLIED ACROSS UPS OUTPUT W/0 GROUND (HOT SHORT)
THE CON'd! TION OF THE THREE CLASS IE SOURCES WILL BE MONITORED THROUGH SUITABLE SIGNAL CONDITIONERS. 8Y GOULD INC.. 2000W SERIES HIGH FREQUENCY RECORDING SYSTEM.
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TEST PROCEDURES 1.0 GENERAL NOTES' 1.1 BEFORE STARTING TEST DETERMINE AND RECORD ALL SIGNAL CONDITIONER TRANSFER RATION (MULTIPLIER) VALUES.
1.2 NORMAL SYSTEM OPERATION DURING EACH TEST A.
C'ONNECTION PER FIG. 1.
8.
THE LOAD ON THE UPS SHALL BE ADJUSTED FOR EACH OF THREE SEPARATE TESTS FOR EACH UPS INPUT SOURCE:
-(1)
NO LOAD
(2)
OUTPUT LOAD AT.08 PF TO ACHIEVE 56 AMPERES INPUT CURRENT WHEN FED FROM 125 VOLT DC.
LOAD SHOULD REMAIN THE SAME FOR AC INPUTS (3)
OUTPUT LOAD AT.08 PF TO ACHIEVE 158 AMPERES INPUT CUPRFNT WHEN FED FROM 125 VOLT DC.
LOAD SHOULD REMAIN THE SAME FOR AC INPUTS UPS POWERED BY " ALTERNATE" DC SOURCE (BATTERY) AND ONE OR C.
BOTH AC SOURCES
" NORMAL" & "8ACK-UP" D.
STATIC SWITCH IN " PREFERRED" POSITION.
E.
ALL BREAKERS & SWITCHES CLOSED BOTH B.YPASS SWITCHES IN
" NORMAL" POSITION
" TEST" SWITCH - CENTERED
" RETURN MODE" SWITCH - IN "AUT0" POSITION
" ISOLATION" TOGGLE SWITCHES - ON
" SYNC" TOGGLE SWITCH - ON 1.3 TfST INSTRUMENTATION A.
GOULD INC., MODEL 2800W HIGH FREQUENCY RECORDING SYSTEM.
EIGHT CHANNEL. INDEPENDENT SCALE SELECT i.050 TO 1500 VOLTS FULL SCALE.
8.
POTENTIAL TRANSFORMER 480V, 60HZ PRIMARY 120V SECONDARY (4:1 RATIO).
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.
W 3'
_ _ 1_
~-
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1.4 TEST FACILITY AND EQUIPMENT A.
DC SUPPLY - CSD 4LCW-15 BATTERY (80 CELLS. 80KW FOR 30 MIN.)
AND SATTERT CHARGER.
8.
AC SUPPLY - 480V, 3 PNASE. 4W. 60 HZ.1200A GROUNDED NEUTRAL.
'C.
AC LOAD BANK 30KW OR 0-30KVA 9 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.
2.0 TESTyROCEDURE 2.1 SASESINEDATA 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.
A1.
METERING AND CONNECTIONS PER FIG. 2 AND "SACKUP SOURCE" SREAKER OPEN. RECORD IN " STORE" MODE AT 20KHZ TIME BASE.
COPY MEMORY TO PAPER.
A2.
REPEAT Al EXCEPT USE 500HZ TIME 8ASE.
81.
WITH METERING AND CONNECTIONS PER FIG. 2'AND " NORMAL SOURCE" BREAKER OPEN.
RECORD IN " STORE" MODE AT 20KHZ TIME BASE.
COPY MEMORT TO PAPER.
82.
REPEAT 81 EXCEPT STATIC SWITCH TRANSFERRED TO 8ACKUP.
83.
REPEAT 81 EXCEPT USE 500HZ TIME BASE.
84.
. REPEAT 82 EXCEPT USE 500HZ TIME BASE.
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'f*Ty-i, 2.2 FAULT TESTING
- I C0.
METERING AND CONNECTIONS PER FIG 2. RECORDER IN MANUAL TRIGGER MODE.
APPLY FAULT BY CLOSING " FAULT" C8 AND AT THE SAME TIME (0R 0 TO 10 MILLISECONDS BEFORE) TRIGGER THE RECORDER IN " STORE" MODE.
REMOVE THE FAULT AND RECORD THE MEMORY TO PAPER.
l AFTER EACH FAULT APPLICATION CHECK THE UPS FOR DAMAGE.
' REPA!R THE UPS IF REQUIRED BEFORE PROCEEDING.
C1.
!NSTALL-JUMPER "A" T0 " FAULT" C8 WITH "8ACKUP SOURCE" C8 OPEN WITH RECORDER AT 20KHZ TIME BASE APPLY FAULT PER C0.
C 2..,
REPEAT C1 EXCEPT WITH 500HZ TIME BASE.
C 3.-
OPEN " NORMAL SOURCE" C8 AND CLOSE "8ACKUP" WITH RECORDER 20KHZ TIME BASE APPLY FAULT PER C0.
'~
C4.
REPEAT C3 EXCEPT WITH 500HZ TIME BASE.
C5.
REPEAT C1, C2, C3 & C4 WITH JUMPER "B" INSTEAD 0F "A" CONNECTED T0 " FAULT" CB.
C6.
REPEAT C1, C2, C3, & C4 WITH JUMPER "C" INSTEAD OF "A"
CONNECTED T0 " FAULT" CB.
C7.
REPEAT C1, C2, C3, & C4 WITH CONNECTIONS.T0 HOT FAULT SOURCE (UPS RUNNING.AT NO LOAD).
2.3 COMPLETE TEST
SUMMARY
SHEET FOR EACH TEST OR TEST GROUP.
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Revised Text FSAR Sections:
13.2.1 13.2.1.1 13.2.1.1.1 13.2.1.1.1.1 13.2.1.1.1.2
[
13.2.1.1.1.3 13.2.1.1.1.4 p
13.2.1.1.2 1
13.2.2 Appendix 13C Appendix 13F Appendix 13I Appendix 13J Appendix 13K (new) 630.4 630.7 630.10 r.
- - ~ --
HCGS FSAR 8/84 13.2 TRAINING 13.2.1 PLANT PERSONNEL TRAINING PROGRAM The training program for Hope Creek Generating Station (HCGS)'ils formulated to develop and maintain an organization qualified to 1FAserf-(
assume the responsibility for preoperational testing, operat on, maintenance, and technical considerations for the facility.
accomplish these objectives and to provide the necessary control of the overall plant, the following three general training programs will be implemented:
a.
Initial Plant Staff Training Programs - These programs are designed to provide competent, trained personnel in all disciplines and at all levels of plant organization.
The programs are designed to allow personnel to be placed at various points, according to their training, experience and intended position.
The training procedures are detailed in the Nuclear Department Training Manual.
b.
Requalification Training Program - A requalification program as required by 10 CFR 50.54 (1-1) will be developed to provide continuous training and upgrading of plant personnel and will meet the requirements of 10 CFR 55, Appendix A and NUREG 0737 Enclosure 1.
Use will be made of the Hope Creek specific simulator scheduled to be delivered to the facility in the summer of 1984.
Therefore, a specific requalification program will not be available until late 1984.
Upon formal acceptance of the Hope Creek specific simulator and establishment of operator shift rotation, the licensed operator requalification program will be implemented to ensure that all cold license candidates maintain a high level of knowledge and operator confidence.
The requalification program will run on an annual basis i
with all program requirements completed during the two year requalification cycle.
The requalification program will consist of three areas; pre-planned lectures, on-the-job training and requalification examinations.
The pre-planned lectures will cover fundamental review and operational proficiency.
Fundamental review training will be in those areas of heat transfer, fluid flow, thermodynamics, mitigation of accidents involving a degraded core and these subject areas delineated in 10CFR55, Appendix A.
Operational proficiency training 13.2-1 Amendment 7
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y c4 f-will involve lectures that will focus on essential qy-f f
plant operational guidelines and changes or experiences
'I in the nuclear industry.
I The on-the-job training will ensure that each licensed i
operator maintains an acceptable level of skills and l
familiarity associated with plant systems, controls and 1:
operational procedures.
This will be accomplished through reactivity manipulations, plant evolutions and ii operational reviews.
j Requalification examinations will be given to determine f
the licensed operator's knowledge of the material i
covered, areas where additional training may be l
required and operational proficiency.
These b
examinations will consist of a segmented written
],
examination and an oral examination.
1 1
i I
Personnel demonstrating a significant deficiency in a C
r given area of knowledge and proficiency may be placed into an accelerated training program.
This program
?
tructured,to upgrade knowledge is; tagill be specifically kdeficiencies. Successful completion
(
and skills iden,tified of the accel.erated training program will be evaluated H
by a weicten and/or oral examination.
Procedures V
y descritiing the co'ntent and conduct of the
?!
requalification program will be developed and will be f
a
'/
maintained in the Nuclear Department Training Procedure H
Manual.
8 d
J c.
Replacement training - These programs are designed to i
j provide qualified presonnel for the station j
j organization.
The tieneral Manager - Hope Creek y
Operations, or the designated representative, may waive i
portions of the training program for individuals based J
on their previous experience and/or qualifications.
1 J.3heTraining procedures are detailed in the Nuclear q
Department Training Manual.
3 5
The Manager - Nuclear Training is responsible for implementation y
l of this program.
Prior to implementation, each course, its j
scheduled starting date, and its duration shall be approved by a
the General Manager - Hope Creek Operations.
g
/,
The Manager - Nuclear Training will ensure that all individuals i
providing instruction are technically qualified to present the
=i material and that they have demonstrated a knowledge of
- j 13.2-2 Amendment 7 i
I
l instructional techniques as required by ANS/ ANSI 3.1-1981, l
4.4.7.2.
Individuals providing instruction to license operator l
candidates wi31 have received all appropriate training and hold or have held an SRO license or certification as required by the H.R. Denton letter of March 28, 1980, Enclosure 1, and i
ANS/ ANSI 3.1-1981, 4.4.7.2.
These individuals will take an j active part in the license operator shift W training program.
Upon completion of the cold license training program and establishment of the operator.requalification program, individuals providing specific licen'le training outlined in ANS/ ANSI 3.1-1981, 4.4.7.2.c will participate in the requalification program as specified in ANS/ ANSI 3.1-1981, 5.5.1.5.
l l
Figure 13.2-1 shows the present schedule for the various initial i
plant training program.
If significant differences or changes l
occur in those courses not yet conducted, the appropriate course i
outlines and descriptions will be revised by Amendment.
i l
13.2.1.1 Operatina Department Trainino procrams I
l l
These programs are designed for individuals who will assume the responsibility for both licensed and nonlicensed plant operating functions, as outlined in job specifications.
- a. v+A s The program is divided into the following brei: : ;rente:
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h assure the experience criteria of ANS 3.1 (1951) is met, as well as the general guidelines of NUREG-0094, additional experience will be provided by a structured-e ati @on f'"
program for all if, censed, operator candidates.4 est M MSmVik%% M4WOM iSSheWA\\A Q em 13.2.1.1.1
' Cold License Training Program This program is designed for NRC reactor operator (RO) and senior reactor operator (SRO) cold license candidates of varying backgrounds and experience.
Candidates will be factored into the program at various points, depending on their previous experience and training.
Testing and screening will be an intimate part of the overall training program.
All license candidates who are supervisors will attend the PSEEG Supervisory Skills Training Program and will meet the supervi=ory training requirements of ANSI /ANS-3.1-1981, Section 5.2.1.8 prior to core load.
I6
-fM e ttstestded eL A m that all SRO
.e ty val nt 13.2.1.1.1.1 Senior Reactor Operator Training Program The senior reactor operator (SRO) candidates will attend a training program consisting of, but not limited to, the following areas of instruction a.
Nuclear Reactor Fundamentals b.
Reactor Startup Experience c.
Advanced technical training d.
Pre-Certification system training e.
BWR Cold certification training f.
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%- NRFhIMNnj 13.2-3 Amendment 7
Hope'Craek Systems training d bwiscg N. h.hmo Exa.msO'oa 4esM$
~
Detailed course descriptions and outlines are shown in Appendices 13A, 13B, 13C, 13D, 13E, 13F and 13G3 /sr and /3 T.
l 3
7 WMew6 It is the intended M this training program that all SRO candidates here at least thirty (30) semester hours of equivalent college levelfeducation.
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- v Following the Hope Creek systems training the SRO candidates will be assigned to a shift where they will participate in the cold license operator in-plant training program described in Appendix 131..
13.2.1.1.1.2 Reactor Operator-(RO) Training Program The RO candidates will attend a training program consisting of, but not limited to, the'following:
i a.
Nuclear Reactor Fundamentals I
b.
Reactor Startup Experience c.
Pre-Certification system training d.
BWR Cold Certification training t
e.
Hope Creek system training.
4
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Detail %d course descriptions and outlines are shown in Appendices 13A, 13B, 13D, 13E,and 13G, th z om6 ta a,
l ins &T @W 3
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FollowingthfHopeCreeksystemstrainingtheROcandidateswill be assigned to a shift where they will o @ participate in 8 the cold license operator in-plant training program described in Appendix 13I.
13.2.1.1.1.3 Shift Technical Advisor Training Shift technical advisor (STA) training will meet the requirements outlined in ANSI /ANS-3.1-1981.
Training programs will consist of those areas where their prior education did not meet those requirements and will include plant specific thermodynamics, fluid flow, reactor physics, system engineering, transient and'
. accident analysis, nuclear instrumentation, process computer, plant response, and duties and' responsibilities.
l The STA training program will consist of, but is not limited to, the following areas of instruction.
a.
Nuclear reactor fundamentals 1
b.
Reactor startup experience c.
Advanced technical training d.
Pre-certification system training e.
BWR cold certification training f.
Hope Creek sy, stems training 2n-botn4 Mo.miAg Detailed course descriptions and outlines are shown in Appendices 13A, 135, 13C, 13D, 13E end 13G QoA tam.
g 13.2-4a Amendment 7
8/84 pherf 5 ~*9 Tbn reactor startup experience and BWR cold certification training may ou..i. J 0:
- ha=e individa=1= -t: e.- previously licensed.
They will hc="rr eu.wna r9WILaperational review trainina rr ;..
at an appropriate BWR simulator v6 it.; "_"cc specTfIc simulator when it becomes available.
All STA candidates will be assigned to HCGS staff where they will participate in the cold license operator in-plant training S program as described in Appendis 13I.
STA candidates will
.s eenMaue4e attend training with the SRO candidates.
It is not intended at this time to test in lieu of training as stated in ANS/ANS 3.1 1981, 5.2.1.7.
4 Proc res de ing the uct and ding cri of he ra ar nder eve ent and Ib ente int lear epartment ining y ure man 13.2.1.1.1.4 BWR Prelicense Refresher Training Breause of the long lead time required for cold license training, a Prelicensing.Refres er Course will be conducted.
This course d2 will be approximate 1y' weeks in duration and will be scheduled to end about 3 to 6 months prior to initial fuel loading.
An 6Ar M
l
[ NRC-type audit examination will be given ;t th; ;r.4 ef 2:Further training will be condu refresher training.
s identified by the audit examination.
Appendia 13J provides a detailed description of this program.
L i
i i
I 13.2-5 Amendment 7 6
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HCGS FSAR g/33 13.2.1.1.2 Nonlicensed Operator Training Program l
This program is designed to make equipment operators knowledgeable of HCGS systems, operations, and procedures.
The program will cover, but is not limited to, the following material:
a.
Hathematics Refresher b.
Physics and Basic Heat Transfer and Fluid Flow (HTFF)
Refresher Basic Power Plant Iquipment (valves, pumps, etc,),
c.
Lubrication, and Job Duties d.
NSSS e.
Electrical Systems f.
Auxiliary Systems g.
Health Physics h.
Firefighting i.
Heating Boiler z
j.
Procedures (as applicable) k.
Administrative Functions, Equipment Tagging, and Log Keeping 1.
Technical Specifications (as applicable).
It is anticipated that the classroom program,Appendia 13H)will last 12 to 14 weeks and will be followed by a period of in-plant j
training where the equipment operators will complete required l
l 13.2-6 Amendment 1
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4 13.2.1.1.3 Maintenance Department Training prog. ram l
~
l Maintenance supervisors, electricians, machinists, and boiler repair personnel will generally be selected from other operating PSE&G facilities (fossi1 and nuclear) or be direct hire, journeyman level q'Jalified.
As such, they will already have received training appropriate for their particular skill area.
Through their previous experience and selection / testing procedures these personnel will exhibit _a high degree of manual dexterity and the capability to learn and apply basic job skills in performing maintenance activities.
Maintenance personnel will receive on-the-job training during the preoperational test program by* performing maintenance activities.
Selected personnel will receive specialized vendor training on specific equipment or skills.
Personnel promoted to the
~
journeyman or supervisory level will be required to satisfactorily complete the PSE&G Advanced or Supervisory Training Program associated with'their particular skill area.
Additional training for experienced personnel will include a BWR Technology Course, appropriate quality assurance training, training on plant specific maintenance procedures, and radiation worker and general employee training, as well as other programs deemed necessary.
Procedures for these training programs will be available in the Nuclear Department Training Manual.
Personnel below the supervisory and journeyman level, as a minimum, will complete the various required apprentice level training programs as their career progresses.
These programs will also be detailed in the Nuclear Department Training Manual.
Training will be conducted by PSE&G and qualified vendor personnel.
13.2.1.1.4 Technical Department Training Program l
The objective of the Technical Department Training Program is to provide highly skilled personnel to effectively support the l
precperational testing program and plant power operations.
13.2-7 Amendment 1
Procedures for these training programs will be available in the Nuclear Department Training Manual.
13.2.1.1.4.1 Chemistry Section Training l
Supervisor and technician level personnel will be selected only citer meeting applicable experience requirements.
As such, they will generally have completed the appropriate training program cssociated with their respective job position.
Procedure for conducting these programs will be available in the Nuclear Department Training Manual.
Experienced personnel who fit that description will, as a minimum, undergo training in the following general subject areas:
a.
BWR Technology b.
Chemistry Practices and Procedures I
c.
Chemistry Equipment and Use
^
d.
Applicable Administrative Procedures Special Courses presented by the Nuclear Training e.
Center and/or vendors, as appropriate.
f.
QA Program g.
General Employee and Radiation Worker Training.
Personnel promoted to the supervisory or technician level taill be required to complete the PSEEG Chemistry Technician Advanced Course or Nuclear Supervisor Course, as appropriate to the respective job position.
Personnel below the supervisory and technician level, as a minimum, will complete the various required apprentice level training programs as their career progresses.
13.2-8 Amer.dment 1
HCG 5 FSAR 8/83 Personnel below the supervisory and technician level, as a ainimum, will complete the various required apprentice level training programs as their career progresses.
Isc personnel will receive on-the-job training during the preoperational testing program by performing their job associated tasks in support of that testing.
Training will be conducted by qualified PSE&G and vendor personnel.
1 13.2.1.1.4.3 Reactor Engineering Training Program l
Prior to core load, selected re* actor engineering personnel will have attended a vendor-offered course typically entitled " Station Nuclear Engineer".
Typical subject matter will include reactor behavior, control rods, shutdown margins, technical specifications and Fuel Warranty Operation Provisions, core flow and thermal limit calculati'ons, fuel failure and Preconditioning Interim Operating Management Recommendation and water chemistry.
- 13. 2~.1.1. 5 Radiation. Protection Department Training Program l
Supervisory and technician level personnel will be selected only afte'r meeting applicable experience requirements.
As such, they will generally have completed the appropriate training program associated with their respective job position.
Procedures for i
conducting these programs will be available in the Nuclear Department Training Manual.
Experienced personnel who fit that description will, as necessary, undergo training in the following general subject areas:
s.
BWR Technology
,b.
Radiation Protection Practices and Procedures Radiation Protection Equipment and Use
, c.
d.
Applicable Administrative Procedures 13.2-10 Amendment 1
_. _ _. _ _ _. _ _, _ _. _ _ _.. _...... _ _ _. _ _ ___.~ __ _..___ _ _ _ _ __ _ _. _
HCGS FSAR-8/83 Chemistry personnel will receive on-the-job training during the preoperational testing program by performing their job associated tasks in, support of that testing.
4 Training will be conducted by qualified PSEEG and vendor personnel.
13.2.1.1.4.2 Instrumentation and Controls Section Training l
Supervisory and technician level personnel will be selected only after meeting applicable experience requiremer.ts.
As such, they will generally have completed the appropriate training program associated with their respective job position.
Procedur'e for conducting these programs will be available in the Nuclear Department Training Manual.
Experienced personnel who fit that description will, as a minimum, undergo training in the following general subject areas:
a.
BWR Technology b.
Instrumentation and Controls Practices and Procedures c.
Instrumentation and Controls Equipment d.
Applicable Administrative Procedures Special Courses presented by the Nuclear Training e.
Center and/or vendors, as appropriate.
f.
QA Program g.
General Employee and Radiation Worker Training.
Personnel promoted to the supervisory or technician level will be required to complete the PSEEG Instrumentation and Controls (IEC)
Technician Advanced Course or Nuclear Supervisor Course, as appropriate to the respective job position.
13.2-9 Amendment 1
HCGS FSAR 8/83 Personnel below the supervisory and technician level, as a ainimum, will complete the various required apprentice level training programs as their career progresses.
IEC personnel will receive on-the-job training during the preoperational testing program by performing their job associated tasks in support of that testing.
Tra,ining will be conducted by qualified FSE&G and vendor personnel.
13.2.1.1.4.3 Reactor Engineering Training Program l
Prior to core load, selected re' actor engineering personnel will have attended a vender-offered course typically entitled " Station Nuclear Engineer".
Typical subject matter will include reactor behavior, control rods, shutdown margins, technical cpecifications and Fuel Warranty Operation Provisions, core flow cnd thermal limit calculations, fuel failure and Preconditioning Interim Operating Management Recommendation and water chemistry.
I 13.2.1.1.5 Radiation. Protection Department Training Program I
Supervisory and technician level personnel will be selected only cite'r meeting applicable experience requirements.
As such, they will generally have completed the appropriate training program Essociated with their respective job position.
Procedures for conducting these programs will be available in the Nuclear Department Training Manual.
Experienced personnel who fit that description will, as necessary, undergo training in the following general subject areas:
a.
BWR Technology
,b.
Radiation Protection Practices and Procedures j
ac.
Radiation Protection Equipment and Use l
l l
d.
Applicable Adminiscrative Procedures
[
13.2-10 Amendment 1 C'
Special. Courses presented by the Nuclear Training Center and/or vendors, as appropriate f.
0A Program g.
General Employee and Radiation Worker Training.
Personnel promoted to the supervisory or technician level will be required to complete the PSE&G Radiation Protection Technician Course or Nuclear Supervisor Course, as appropriate to the respective job position.
Personnel below the supervisory and technician level, as a minimum, will complete the various, programs as their career progresses.
Radiation Protection personnel will receive on-the-job training -
during the pre-operational testing program by performing their job associated tasks in support of that testing.
Training will be conducted by qualified PSEEG and vendor personnel.
13.2.1.1.6 General Employee Indoctrination l
All persons regularly employed at HCGS, including temporary maintenance and service personnel, who are permitted unescorted access shall be given General Employee Indoctrination.
This training covers the following areas:
a.
Site Description b.
Security System d.
Quality Assurance Program 13.2-11 Amendment 1
Radiological Health.
Personnel will be tested in the above areas to determine the effectiveness of General Employee Indoctrination.
Personnel who will routinely work in radiation and/or contaminated areas will also complete a Radiation Worker Training Program of approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 13.2.1.2 Refresher Trainina for Nonlicensed Plant Personnel A retraining program will be provided for all personnel to ensure that they remain proficient in thei particular jobs.
Retraining in specific areas is provided to th extent necessary for personnel to safely and efficiently carry out their assigned responsibilities in accordance with established policies and procedures.
This includes operating experiences, design changes, revisions to procedures, and new procedure indoctrination.
Such training may consist of vendor presentations, technical
~
training sessions, on-the-job work experience or programmed instruction.
Personnel are evaluated on an annual basis where individual needs for retraining will be identified.
13.2.1.3 General Employee Indoctrination Reaualification All persons regularly employed at HCGS, including temporary maintenance and service personnel who are permitted unescorted access, shall requalify in General Employee Indoctrination annually.
This is accomplished by attending the requalification class and obtaining a satisfactory score on an examination covering the areas mentioned in Section 13.2.1.1.6.
l Personnel trained in the Radiation Worker Training Program will requalify annually by attending the Radiation Worker Review Program of approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Satisfactory completion of that program also meets General Employee Indoctrination Requalification requirements.
4 4
13.2-12 Amendment 1
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l HCGS FSAR 8/34 13.2.1.4 Replacement Trainino for Nonlicense4, Plant Personnel Replacement training is designed to supply qualified personnel at all levels and job positions within the plant organization.
Training is carried on at all job levels to qualify that particular individual to effectively perform the required job functions.
Qualified personnel who are promoted to the next job level are placed, as rapidly as possible, into the appropriate training program.
It is the general policy of PSE&G to promote from within.
In this manner, as an individual progresses, he/she is immediately trained for the new position and capable of supporting and training personnel in the lower classifications.
Personnel who are directly hired into job positions above the entry level will meet or exceed the applicable requirements of that position.
Training programs will be developed for these personnel to familiarize them with appropriate HCGS-specific material.
Training will be conducted by qualified PSE&G and vendor personnel.
The training programs will be described in the Nuclear Department Training Manual.
13.2.1.5 Replacement Trainino for NRC Licensed Plant Personnel Training for NRC licensed replacement personnel will, as a minimum, meet the existing NRC requirements as outlined in 10 CFR 55.21,
.22,
.23, appropriate NUREGs, and the H. Denton letter of March 28, 1980 and all applicable training requirements of AES/ ANSI 3.1-1981.
These programs are described in the Nuclear Department Training Manual and are revised as regulations and job requirements change.
13.2.2 FIRE BRIGADE TRAINING PROGRAM l
Insett 4 x
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with e
Fire protee on training will be condutior, in accord g aideline of the SRP (
R,0800) S 13.2.2.I
.6, l10CF 0
Appendi an Branch echn Leal Pos' tion CMEB 9.5.
/
S cti C.3.d.
This raining wil include clas oom in tru tiop, ha
-on fire guishing lant dril i
13.2-13 Amendment 7
HCGS FSAR 8/84 T e classroom instruction will include the followin course ma erin 1:
Firefighting Plan l
Response to alarms l
2.
esponsibility of members l
3.
Rea n for fire brigad l
b.
Identificati of Fire H ards l
1.
Concept of re l
2.
Properties of maable and combustible liquids l
3.
Hazardous c mical roperties l
4.
Boiling 1 quid, expand ng vapor explosion l
c.
Products of ombustion l
1.
Prod ts of burning plastics l
2.
Pr ucts of smoke l
3.
operties of carbon monoxide l
l l
4.
Properties of contaminated smoke l
5 Effects of heat l
6.
Ventilation l
l 13.2-14 Amendment 7
..--..___._-._____.___.___.___.__._.__._______..._,l_
HCGS FSAR 8/84 Firefighting Equipment l
1.
Fire detection l
2.
Fire suppression l
e.
Type of Fires l
f.
Auxiliar Equipment l
g.
Plant Modif ations l
Actual hands-on fire exti guishing wil be conducted to provide l
-brigade members with actua fire.exti guishing and the use of
-~emergency breathing apparat s under trenuous conditions.
These practice sessions will be he d at I ast once per year for each fire brigade member.
l Plant drills will be held at sp ified intervals not to exceed 3 months for each shift to allow fi e brigade members the
. opportunity to practice as a eam nd to ensure adequate procedures and readiness.
Each drill will be prepla ed to estab sh training objactives and will be critiqued to etermine how 11 the training objectives have been me.
Performance d iciencies noted will be remedied by additional raining.
Fire drills as a min mum will assess the fire alarms effectiveness, time to assemble the fire briga e, use of the
-firefighting equi ent, firefighting strategies and the effectiveness of e brigade leader.
The Fire Briga Training program is designed to en ure that the employees assi ned to the fire brigade are capable o providing adequate manu 1 firefighting strategies to control fi es that might occur t the Hope Creek Generating Station.
The rogram will cover,p but is not limited to the following:
a.-
Indoctrination of the plant firefighting plan.
l 13.2-15 Amendment 7
.,--e
-.,-w.,-.
.-,..w.
-,.w.-.,e--..-.,--...-o.
.-.---...m
....w,mm--.--e--....,--.w i.,,-,.e-r,..,+e,-v
HCGS FSAR 3/84 Identificatior. of fire hazards.
l c.
Th roperties of the products of bustion.
l d.
Identific on and use of all refighting equipment.
l e.
The proper use o ommuni tion, lighting, ventilation, and emergency breat ng uipment.
f.
The proper method fo fig ing fires inside buildings and confined space.
g.
The direction d c'ocedination o the firefighting activities.
ire Brigade leaders nly).
~
h.
Detailed eview of firefighting strateg s and procedu s.
i.
Rev of the latest piant modifications and co esponding changes in firefighting plans.
~
Procedur describing course content, grading criteria and recordk eping are under development.
These procedures are schedu ed to be completed by January 1985.
l l
l l
13.2-16 Amendment 7 l
Z~nserb R
~
FMR.5 ec:Non
- ' '2 I" 'Y p
g w.
ws. goy enkre G HCGS FSAR 13.2.2 FIRE BRIGADE TRAINING PROGRAM Fire protection training will be conducted in accordance with the
, guidelines.of-the SRP (NUREG 0800) Section 13.2.2.II'.6,
- 10CFR50, Appendix R and Branch Technical Position CMEB 9.5.1, Section C.3.d.
This training will include classroom instruction, hands-on fire extinguishing and plant drills.
The Fire Brigade Training Program is designed to ensure that the employees assigned to the fire brigade are capable of providing adequate manual fire fighting strategies to control fires that might occur at the Hope Creek Generating Station.
The prog ram will cover, but is not limited to the following:
a.
Indoctrination of the plant fire fighting plan.
b.
Identification of fire hazards.
c.
The properties of the products of combustion.
d.
Identification and use of all fire fighting equipment.
e.
The proper use of communication, lighting, ventilation, and emergency breathing equipment.
f.
Familiarization with the layout of the plant, including access and egress routes to each area.
g.
Correct method of fighting fires, including fires in ener-gized electrical equipment, fires in cable and cable trays, hydrogen fires, fires involving flammable and combustible liquids or hazardous process chemicals, fires. resulting from construction or modifications (welding).and record file fires.
h.
The direction and coordination of the fire fighting activi-ties (fire brigade leaders only).
i.
Detailed review of fire fighting strategies and procedures, j.
Review of the latest plant modifications and corresponding changes in fire fighting plans.
The classroom instruction will include the following course material:
a.
Fire Fighting Plan 1.
Response to alarms 2.
Responsibility of members 3.
Reason for fire brigade py r
..,,w e-
-m-i w
pr-
4 b.
Identification of Fire Hazards 1.
Concept of fire 2.
Properties of flammable and combustible liquids 3.
Hazardous chemical properties 4.
Boiling liquid, expanding vapor explosion c.
Products of Combustion 1.
Products of burning plastics 2.
Products of smoke 3.
Properties of carbon monoxide 4.
Properties of contaminated smoke 5.
Effects of heat 6.
Ventilation d.
Fire Fighting Equipment 1.
Fire detection 2.
Fire suppression e.
Types of Fires f.
Auxiliary Equipment g.
Plant Modifications Actual hands-on fire extinguishing will be conducted to provide brigade members with actual fire extinguishing and the use of emergency breathing apparatus under strenuous conditions.
These practice sessions will be held at least once per year for each fire brigade member.
Plant drills will be held for each shift to allow fire brigade members the opportunity to practice as a team and to ensure ade-
_quate procedures and readiness.
Each fire Lrigade member must participate in at least two drills per year.
Each drill will include the simulated use of fire-fighting equip-ment' required to cope with the situation and type of fire select-ed for the drill.
The area and type of fire chosen for the drill will differ from those used in the previous drill so that brigade members are trained in - fighting fires in various plant areas.
The situation selected will simulate the size and arrangement of a fire that could reasonably occur in the area selected, allowing for fire development due to the time required to respond, to obta.in equipment, and organize for the fire, assuming the loss of automatic suppression capability.
At least one drill per year will be performed on a back ahift for each shift fire brigade.
At ~1 east one drill for each shift fire brigade per year will be unannounced to determine the fire fighting readiness of the plant fire brigade, brigade leader, and fire protection systems and equipment.
Personnel planning and authorizing an unannounced drill will ensure that the responding shift fire brigade members are not aware that a drill is being -planned until it is begun.
Unannounced drills will not be scheduled closer than four weeks.
Unannounced drills will be planned and critiqued by members of the management staff responsible for plant safety and fire protection.
Performance deficiencies of a fire brigade or indi-vidual fire brigade members will be remedied by scheduling addi-tional training for the brigade or members.
Unsatisfactory drill performance will be followed by a repeat drill within thirty days.
At three-year intervals, a randomly selected unannounced drill
-will be critiqued by qualified individuals independent of the licensee's staff.
A copy of the written report from such indi-viduals shall be available for NRC review.
Regularly planned meetings will be held every three months for all members to review changes to the program.
Periodic refresher training will repeat classroom instruction over a two year period.
These sessions may be concurrent with planned meetings.
Training of the plant fire - brigade will be coordinated with the local fire department so that responsibilities and duties are
' delineated in advance.
This coordination will be part of the training course and will be included in the training of the local fire. department staff.
Local fire departments will be provided training in operational precautions when fighting fires en nuclear power plant sites and will be made aware of the need for radiological protection of personnel and the special hazards associated with a nuclear power plant site.
Instruction will be provided by qualified individuals who are knowledgeable, experienced and suitably trained in fighting types i
l-of fires that could occur in the plant and using types of equip-i ment available in nuclear power plants.
Instruction will be provided for all employees once a year.
It will be repeated on an annual basis.
The instruction will be i
given on (1) the fire protection plant, (b) the evacuation routes, and (c) the procedure for reporting a fire.
Instruction will be provided for security personnel that addresses (a) entry procedures for outside fire departments, (b) crowd control for people exiting the station, and (c) procedures for reporting potential fire hazards observed when touring the facility.
l
Instruction will be provided to appropriate shift personnel that complements that given to members of the fire brigade.
Instruction will be provided to temporary employees so that they are familiar with (a) evacuation signals, (b) evacuation routes, and (c) the procedure for reporting fires.
Station personnel will participate in an annual accountability and evacuation drill.
Fire Protection Staff Training for the fire protection staff members shall include courses in:
1.
Design and maintenance of fire detection, suppression and extinguishing systems.
2.
Fire prevention techniques and procedures.
3.-
Training and manual fire-fighting techniques and procedures for plant personnel and the fire brigade.
S e
o w
SC.EDUI.E 1952 1983 1994 1985 1986 JFMAMJJASOND JFMAAJJASOWD JFMAnJJASOND JFMARJJASODO JFMA2JJASOps ACTIVITY Sanior reactor operator (ETA / Senior Shitt l
1 l
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ll e l e
{ h l h ef M g [l 21 l
l
]l f Supervisor)
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(womesperience own I i I
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)l S.nior reactor oper-etor sc pertene. nwa l
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9 g,
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eo E Equipment operator
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12 I
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17 I
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.-3 APPENDIX 13B REACTOR STARTUP EXPERIENCE
% e.\\ % W O h W C O r Presented by:
P.r p.i;; Ot:t U i errity y::ilitj f
Obiective To assign cold license applicants, with no previous nuclear experience, to a Research Reactor Training Course.cendveted-by P-rhir State ".itrreity.
This ' -
actual hands-on experience with 6. g e gives the student
..... nuclear reactor and allows the cold license applicant to obtain at least the minimum of 10 reactor startups necessary to establish cold license i
eligibility requirements of ANS 3.1, 1981.
if
Reference:
a i.
ANS/ ANSI 3.1-1981 Section 5.2.1.1 l
l l
l l
13B-1 Amendment 7
3 HCGS FSAR 8/84 APPENDIX 13C ADVANCED TECHNICAL TRAINING
% 4A or S(E.C.N-O o M C. k. e Presented by:
M:r;hi: Stete U.1 e eity Obiective To provide advanced technical training in Thermodynamics, Heat-Transfer, Fluid Dynamics, Reactor Materials, Reactor Physics and Human Behavior to senior supervisors and STA candidates.
Course Description The Advanced Technical Training Program consists of nine (9) courses.witich tet:1 29 errester :::dit heur; Of :: d::ic ir.:tructie.
A list of the courses and their content is outlined in the following pages.
References:
ANS/ ANSI 3.1-19El Section 5.2.1.6, 5.3.3 l
10 CFR 55.22 NUREG 0737 Appendix C Section 6.1.2 l
M USM h1Mr\\
pPC (C m W %
kk-l Nh UR
'h -
pde M
\\AsM kt b Q.of\\bC.Md 9%(4 0" 5Gd.tIdeML NW.
i l
l 13C-1 Amendment 7
~
HCGS FSAR 8/84 APPENDIX 13E COLD LICENSE CERTIFICATION TRAINING 7286 o-
% ele.cStA Co N h b P Q.4 o.n cQptoued Presented By:
Cener:1 Ph:, rice, C;;;;. :t th: Seege-hanna
- ir.cl:t :.
Th: fir t ? a;;ks ef ejst :: tr:ining will b-et sur facility.
Objective:
1.
To ensure that non-experienced (nuclear) personnel meet' the cold license eligibility requirements of NUREG-0094 and ANS 3.1 1981
References:
ANS/ ANSI 3.1-1981 Section 5.2.1.3.2 s
NUREG 0737 enclosure 1 10 CFR 55.23 Y e-In ob5h dcv&
fno W Loc <o 9 eseaQ PL/w arp a+ Wu s g uo/ m N &. W& fS WA N a
& Mr a a~trad~,
e 1
l l
l l
13E-1 Amendment 7
HCGS FSAR 8/83 APPENDIX 13F SS-N TRAINING PROGRAM d
Presented By:
Ir. part,, Cencr:1 E1:ctri er.d ir pert by M;;pt.is Et t: ';.. i v n s i i.yStjec.k d % & N t"
?kS 4 6 or-Objective:
To provide advanced instruction to Senior Operators and Supervisors on BWF 3pecific topics LOW g ft-GENERAL ELECTRIC 1.
BWR Chemistry - I wk.
4 2.
Nuclear Engineering - 3 wks
'N 3.
Corrosion - Materials - 1 wk.
4.
Radiological Emergencies - I wk.
5.
Abnormal Event Analysis - 1 wk.
6.
. Degraded Core Damage - 1 wk.
(
- MSU M
Materials Study - 2 wks.
'3 crcdit;}
8E Human Behavior - 2 wks.
'2 cc;ditJ l
\\
l l
13F-1 Amendment I l
-N NOTES:
College Credits for GE Course - 8 (awarded through the a.
e N.Y.S.
Regents) b.
College Cr its for MSU Course - 6 c.
Total Course Length - 12 wks.
d.
Description of Modules 7 and 8 are provided in Appendix 13C.
l.
- The STA's and SS-N's will be integrated for these. courses.
/
l
References:
s ANS/ ANSI 3.1-1981 10 CFR 55.21 and 55.22
\\
b ME ins Y $20 & & so,'If allead %he SS-s0 Ytrain tn proy am % k+ sg seneet ejeckh Comaq *nd }&
'h )
7 wau.ans:s s = av==
L h e d8 o
g% p %s gat W Aou.p.(
3 l
selecAed con cAecpe and.
l l
l 13F-2 Amendment 5 l
--..--.--..____.._..---..___.-.,--,....,.--,,,__--,,..,-,,,n,,--~,
,,,w,_.,,e----,-m,,n,
gh l
COLD LICENSE OPERATOR INPLANT TRAINING P'
Presented'by!PSE&G [ selected contractor personnel jectives: To provide cold-license candidates with a structured and documented program of plant ebrervatienbreoperational testing and work assignment participation requirement 5+sW M A,
Descript. ion: The cold license in-plant training program is designg_tgj{pp7_geg{gggg.g Q.y~.. ::Qp:nsure:: hat 20 n
- = :y e
t
__.....,.....___...,,.....uvi.4 cm a,.y e.ch candidate receives sufficient practical work experience
-necessary to gain a thorough knowledge of the plont.
In d
addit on, this program provides for a structuraf :t:?ruetie=.
program where each candidate receives an oral examination and system check out on plant systems emphasizing system operation, local control and interactions.
This in-plant training is documented in the form of individual.In-olant Trainino Guideline g
- b l
fer the RO, SRO and STA candidates.Y ihe completed in-plant training guideline will be maintained in the individuals training record.
' N5er
References:
ANS/ ANSI 3.1-1981 Section 5.2.1.2.2, 5.2.1.3, 5.2.1.4.
i l
l i
i i
l l
L 131-1 Amendment 7 l
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,_.. _. _ _ _,. _ _. _ _ _ _ _ _.. _.. _ _ _. _... _ _ _ _ _..... _ _ _. _ _ _. ~ _ _ _ _ _. _ _ _ _. _ _
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- ~ '
II.
System Knowledge c
III.
Performance Items
.T._.[_. _ __.._..__
IV.
Technical Specifications V.
Reactivity Manipulations
~
~
l SEB l
I.
Control Board Checkouts II.
Technical 3pecifications III.
Radiological Controls IV.
Plant Safety V.
Refueling VI.
Procedures VII.
Performance Requirements STh I.
Control Board Checkouts f
I II.
Plant Safety III.
Procedures IV.
Performance Requirements
APPENDIX 13J PRE-LICENSE EXAMINATION TESTING AND TRAINING l
Presented by: PSEEG or sslected contractor personnel.
l Objectives: To determine individual candidate's ability to operate the plant in a safe and competent manner and to identify areas of weakness that may be corrected prior to administration
.of the NRC license examinations.
==
Description:==
The pre-license examtnation testing and training period will consist of an intensive period of instruction and testing prior to the NRC license examinations.
The instructional phase of this program will consist of the the following:
h 8 deed) l a.
Classroom presentations on:
l 1.
Reactor theory ev 2.
Heat transferres;e4 2.
Fluid mechanics res;eJ 4.
Thermodynamics resca f4enl&%5ics fCVicd r::::dur:_ :23 ::::: tin; ;511:::;51::
l 5.
Technical Specification Wd Od*Mr Bw/we.s resica 6.
7.
Related industry events relevant to operation.
Simulator Operation [clauroom Preferofloa [y6Mfo)
[~/tJde/s) b.
1.
During normal, abnormal and emergency operations l
rocedural and o@in}
4e e.csac. under%ssdim3 f
k ysof ic5 13J-1 Amendment 7
To demonstrate the proper use of the emergency operating procedures.
c.
1 p1;nt f: rnetrrtirn
- .ip
- : t :;:r-t ir= f c-: --
1;;;l :p:::iin; ;:nrir
--! :: inn:ni 1-_::ti be emulut4t2 hf b O*"
The testing phase of this program willA :n-i t Of; (g 4 Mfk#lly 45si3pe.d WhcMonsl tv{eff and will e,ms.'s+ el; a.
A written examination to determine knowledge level of theory, operating procedures and philosophies, system constrt.ction end design and technical specification requirements.
b.
An oral examination to determir.e knowledge level of plant operation from both simulator demonstration and
~
in-plant walk through.
References:
ANS/ ANSI 3.1-1981, Section 5.2.1.5.
t 13J-2 Amendment 7
Mces F58e APPeonix.
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HCGS FSAR 8/84 OUESTION 630.4 (SECTION 13.2)
With regard to training in the use of plant systems to concrol or mitigate an accident in which the core is severely damaged, please provide the tralning programs and schedule for:
Licensed oper0 tors and senior operators a.
b.
Other plant personnel (Ref. H. R. Denton letter of March 28, 1980 and II.B.4 of NUREG-0737)
RESPONSE
Licensed Operators and Operations Personnel l
NUREG 0737,Section II.B.4 requires that training of plant personnel be conducted to teach the use of installed equipment and systems to control or mitigate accidents in which the core is severely damaged. to the H. R. Denton letter dated 3/28/80 identifies the top sthatshouldbeincludedinthg NND b
db odynami ecus;d 10 m$jeh Mfee The HCGS operator training for mitigating core damage is under "T # '
l development.
It will incorporate all areas identified in jl enclosure 3 of the 3/28/80 letter as they are applicable to a BWR:
l A.
Incore Instrumentation l
1.
Use of fixed or movable incore detectors to determine the extent of core damage and geometry changes.
2.
Methods of determining peak temperatures, extended range readings and direct readings at terminal junction.
3.
Methods of calling up incore data from plant process computer.
B.
Vital Instrumenta; ion l
1.
Instrumentation response in an accident environment; failure sequence & indication reliability.
2.
Alternate methods for measuring flows, pressures, levels and temperature.
h 630.4-1 Amendment 7 l
8/84 HCGS FSAR OUESTION 630.7 (SECTION 13.2)
Section 13.2 of the HCGS FSAR contains the training program The gegments for licensed and non-licensed operations personnel.Please cegment outlines are contained in the Appendices of 13.2.
provide the details or information for the following:
Prerequisites for personnel assigned to each program.
a.
For licensed training, which course (s) will contain the b.
use of HCGS specific procedures including; Individual Systems, Integrated Plant, Administrative, Abnormal and Emergency, Radiological Emergency Response Plan, Technical specifications, Initial Fuel Loading, Low Power and Periodic Surveillance Testing?
Please provide the applicable references (Industry 10 CFR and Regulatory Guides for c.
Standards, NUREGs, each of the segments outlined in the Appendices.
Identify those training segments which include the subject areas contained,in 10 CFR Part 55 Section 21, d.
22 and 23.
The Appendices do not contain a course description of Please provide the course e.
segments i-k of 13.2.1.1.
description or a schedule for submittal of the course q
-p descriptica.
The Appendices do not contain the details of the Please f.
observation training referenced in 13.2.1.1.
provide the course description or a schedule for submitting the observation program.
for NRC Concerning replacement training (hot licenses) candidates in Section 13.2.1.5, the FSAR must contain, g.
as a minimum, those courses or segments identifled in Section 13.2.1.1 or provide a schedule for submittal of Ref. (NUREG-0800, this program prior to fuel loading.
13.2.1.B)
Please provide information on the details of SS-N In addition, why h.
training contained in Appendix 13F.
are Senior Operators with previous experience excluded from this course as indicated in Figure 13.2-17 (sic)
(Ref. NUREG-0800, 13.2.18)
RESPONSE
Personnel assigned to the licensed and non-licensed operator training programs come with diverse backgrounds; however, a.
J Amendment 7 630.7-1 I
f j
msann%s 8/84 occh individual will r.ast tha education and experience requirem2nto_of ANS/ ANSI 3.1 - 1981 loading.
prior to initiate fuel
- N' In general training co,me from one of the following areas:'the personr
'1.
Degreed engineer 2.
Previously licensed (BWR/PWR) 3.
Navy nuclear plant operator 4.
Fossil plant operator 5..
Sales EO upgrade In-general, personnel assigned to the non-licensed trainin@ will come from one of the following areas: operator 1.
~0ualified utility /equiseent operator from Salem Generating Station 2.
Navy nuclear plant operator
-3.
Fcssil plant operator satisfactory score on a screening examination as aT i.
l a
At present, Power Operator Service Selection (POSS previously held a NRC license and for degreed p is used.
All prospective employees must participate in a
,i physiological screening process.
Personnality Inventory (NHPI)
The Minnesota Multi-Phasie is presently in use.
b.
specifications will be conducted as the procedure l
available.
become -svailable at various intervals throughout theThe training period.
T candidates are thoro,o ensure that all licensed o9erator technical specifications, training on plant specificughly familia procedures and technical specifications will be incorporated into the training programs outlined in Appendices 13 G training program Appendix 13 J, will be implemen
, 13 H &
(3) to six (6) months prior to the license examinations This training will cover all the HCGS specific operating abnormal and emergency procedures, administrative and i
f emergency response procedures, technical specifications and
(
low power and surveillance testing procedures.
i will be covered by classroom instruction, in-plant oral Training on the Hope Creek specific simulator. examinations, written e 3
630.7-2 Amendment 7
,e
- _. -. -,.. _ ~.,, - _ _ _,..
. ~.
}!
8/84 HCGS FSAR Applicable references for each of the segments outlined in c.
the appendices are shown on the appropriate cover sheet of each appendix.
d.
Trai'ing segments which include 10CFR Part 55 Section 21, 22 and 23 are identified in Appendix 13A, 13C, 13E, 13F and 13G.
' The following segments of the training program are still o.
under development:
Appendix i -
Cold license operator in-plant training Appendix J -
Pre-license examination testing and training f.
A course description for segments i and j of the training program is contained in Appendices 13 I and 13 J,d;ft e(ernb,9 respectively.A desty p of m-
. etratawe 4madw'*gewhy B K provWs a 3
Hot license training for'NRC candidates will be conducted to g.
augment the shift staffing allotment, allow for promotion or fill vacancies c'ue to reassignment.
This training will utilize & major portion of the existing cold license training program; noww.cr. certain areas may be waived based on an individual's prior experience and educational background.
Procedures describing the content and jf administrative requirements will be completed by June 1985.
h.
Appendix 13F has been revised to incorporate th.ie response.
l lJP 630.7-3 Amendment 7
-l OUESTION 630.10 (SECTION 13.2)
Please provide the training programs for all 2.anagement personnel, technical support staff, and other personnel contained in Figure 13.1-9 thecugh 13.1-13.
We believe that Figure 13.2-1 I
may be modified-t:; include the personnel and training prograr.s.
(Ref. NUREG-0800 Section 13.2.I)
RESPONSE
e Figures-13.1-9 through.13."-13 outline the organization structures:of the HCGS operations department.
The training for each department varies as does the training for the diff.erent levels of perso nel within each department.
This training is
.onducted as th d the procedures describing the
- content of th,4 ',eneed arQes anpecgrams is' contained in the Nuclear Departme Training Procedure Manual.
Figure 13.2-1 reflects the initial training of plant. staff perr,onnel; however, it is our policy to provide additional training whenever personnel performance identifies as training need.
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V ATTACHMENT 6
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