ML20094D182
| ML20094D182 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/27/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8408080290 | |
| Download: ML20094D182 (111) | |
Text
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Pubbc Servce I
O PS G Cornpany Electre 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 July 27, 1984 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:
HOPE CREEK GENERATING STATION DOCKET NO. 50-354 DRAFT SAFETY EVALUATION REPORT OPEN ITEM STATUS Attachment 1 is a current list which provides a status of the open items identified in Section 1.7 of the Draft Safety Evaluation Report (SER). Items identified as " complete" are those for which PSE&G has provided responses and no confir-mation of status has been received from the staff. We will consider these items closed unless notified otherwise. In order to permit timely resolution of items identified as
" complete" which may not be resolved to the staff's satis-faction, please provide a specific description of the issue which remains to be resolved, Attachment 2 is a current list which identifies Draft SER Sectione not yet provided.
In addition, enclosed for your review and approval (see Attachment 4) are the resolutions to those Draft SER open items listed in Attachment 3. A signed original of the required affidavit is provided to document the submittal of these DSER open item responses.
Should you have any questions or require any additional information on these open items, please contact us.
Very truly yours, b5 0 8408080290 840727 g(
PDR ADOCK 05000354 E PDR Attachments 00l The Energy People g e491214V) 1 A3
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Director.'of Nuclear T ' Reactor' Regulation 2 7/27/84 -
. C. D. - H . Wagner
-USNRC' Licensing Project Manager.
.W.:H'. Bateman.
'USNRC-Senior Resident Inspector c
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UNITED STATES OF AMERICA
_. NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-354 J-PUBLIC SERVICE ELECTRIC AND GAS COMPANY Public' Service Electric-and Gas Company hereby submits the enclosed. Hope Creek Generating Station Draft Safety Evalua-tion Report open-item responses.
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 I
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Ohomas J. M rtih l&
Vice Pres
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Engineeri g and Construction 1
Sworn to and subscribed before me, a Notary Pub ic of New Jersey, this f7 day of July 1984.
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DATE: 7/27/84
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ATTACINENT 1 DSER R. L. MITTL TO OPEN SFLTION A. SCHWENCER l l
ITEM NUMBER SUBJECT STATUS LETTER DATED 1 2.3.1 Design-basis temperatures for safety- Open !
related auxiliary systems l 2a 2.3.3 Accuracies of meteorological Cmplete 7/27/84 measurements 2b 2.3.3 Accuracies of meteorological Ccuplete 7/27/84 measurements 2c 2.3.3 Accuracies of meteorological Cmplete 7/27/84 measurements 2d 2.3.3 Accuracies of meteorological Open measurements 3a 2.3.3 Upgradirg of onsite meteorological Cmplete 7/ 27 / 84 measurements progran (III.A.2) 3b 2.3.3 Upgrading of onsite meteorological Cmplete 7/27/84 measurements program (III.A.2) 3c 2.3.3 Upgrading of onsite meteorological open measurements program (III.A.2) 4 2.4.2.2 Ponding levels Open 5a 2.4.5 Wave impact and runup on service Complete 6/1/84 Water Intake Structure 5b 2.4.5 Wave impact and runup on service Open water intake structure Sc 2.4.5 Wave impact ard runup on service water intake structure 5d 2.4.5 Wave impact ard runup on service Cmplete 6/1/84 water intake structure 6a 2.4.10 Stability of erosion protection Open structures 6b 2.4.10 stability of erosion protection Open structures l 6c 2.4.10 Stability of erosion protection Open l structures l
M P84 80/121-gs
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ATTACHMENT 1 (Cont'd)
DSER R. L. MrITL 'IO OPEN SECTICN A. SCHWENCER ITEM NUMBER SURTECT STATUS LETTER DATED 7a 2.4.11.2 Thermal aspects of ultimate heat sink Open 7b 2.4.11.2 Thermal aspects of ultimate heat sink Ccurplete 6/1/84 8 2.5.2.2 Choice of maximsn earthquake for New Open England - Piedmont Tectonic Province 9 2.5.4 Soil danping values Complete 6/1/84 10 2.5.4 Foundaticn lovel response spectra Complete 6/1/84 11 2.5.4 Soil shear moduli variation Couplete 6/1/84 12 2.5.4 Combination of soil layer properties Ccaplete 6/1/84 13 2.5.4 Iab test shear moduli values Couplete 6/1/84 14 2.5.4 Liquefaction analysis of river bottcm Ccuplete 6/1/84 sands 15 2.5.4 Tabulations of shear moduli Ccxrplete 6/1/84 16 2.5.4 Drying and wetting effect on Ccaplete 6/1/84 Vincentown 17 2.5.4 Power block settlement monitoring Couplete 6/1/84 18 2.5.4 Maxinun earth at rest pressure Ccmplete 6/1/84 coefficient 19 2.5.4 Liquefaction analysis for service Cczrplete 6/1/84 water piping 20 2.5.4 Explanaticn of observed power block Completc 6/1/84 settlement l 21 2.5.4 Service water pipe settlement records Ccrrplete 6/1/84 22 2.5.4 Cofferdam stability Ccmplete 6/1/84
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23 2.5.4 Clarification of ESAR Tables 2.5.13 Ccrrplete 6/1/84 and 2.5.14 o
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I M P84 80/12 2 - gs
ATIACHMENT 1 (Cont'd)
DSER R. L. MITIL TO CPEN SECTION A. SQiWENCER ITEM NUMBER SUELTECT STATtJS LETTER DATED 24 2.5.4 Soil depth nodels for intake Cmplete 6/1/84 structure 25 2.5.4 Intake structure soil modeling Open 26 2.5.4.4 Intake structure sliding stability Open 27 2.5.5 Slope stability C mplete 6/1/84 28a 3.4.1 Flood protection Caplete 7/27/84 28b 3.4.1 Flood protection Cmplete 7/27/84 28c 3.4.1 Flood protection Cmplete 7/27/84 28d 3.4.1 Flood protection Cmplete 7/27/84 28e 3.4.1 Flood protection Cmplete 7/27/84 28f 3.4.1 Flood protection . Open 28g 3.4.1 Flood protection Ccnolete 7/27/84 29 3.5.1.1 Internally generated missiles (outside Cmplete 7/18/84 containment) 30 3.5.1.2 Internally generated missiles (inside Closed 6/1/84 containment) (5/30/84-Aux.Sys.Mtg.)
31 3.5.1.3 Turbine missiles Cmplete 7/18/84 32 3.5.1.4 Missiles generated by natural phenmena Open 33 3.5.2 Structures, systems, and emponents to Open be protected frm externally generated missiles 34 3.6.2 Unrestrained whipping pipe inside cmplete 7/18/84 containment 35 3.6.2 ISI program for pipe welds in Ccmplete 6/29/84 break exclusion zone M P84 80/12 3 - gs
ATIACHMENT 1 (Cont'd)
- DSER R. L. M1TIL TO OPEN SECTIOJ A. SCHWENCER ITEM NUMBER SUBJECT STATUS LETTER DATED 36 3.6.2 Postulated pipe ruptures Caplete 6/29/84 37 3.6.2 Peedwater isolation check valve Open operability 38 3.6.2 Design of pipe rupture restraints Open 39 3.7.2.3 SSI analysis results using finite Open element method and elastic half-space approach for containment structure 40 3.7.2.3 SSI analysis results using finite Open element method and elastic half-space -
approach for intake structure 41 3.8.2 Steel contaireent buckling analysis Carplete 6/1/84 42 3.8.2 Steel containment ultimate capacity Caplete 6/1/84 analysis 43 3.8.2 3RV/IDCA pool dynamic loads Caplete 6/1/84 44 3.8.3 ACI 349 deviations for internal Caplete 6/1/84 structures 45 3.8.4 ACI 349 deviations for Category I Caplete 6/1/84 structures
- 46 3.8.5 ACI 349 deviations for foundations Caplete 6/1/84 47 3.8.6 Base mat response spectra Caplete 6/1/84 48 3.8.6 Rocking time histories Caplete 6/1/84 l
49 3.8.6 Gross concrete section Ccrplete 6/1/84 50 3.8.6 Vertical floor flexibility response Caplete 6/1/84 i spectra 51 3.8.6 Conpariscn of Bechtel independent Open verification results with the design-basis results l
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M P84 80/12 4 - gs
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ATIACHMENT 1 (Cont'd)
DSER R. L. MITTL 'IO OPEN SECTICN A. SCHWENCER ITEM NUMBER SUE 7ECT STATUS IErrER DATED 52 3.8.6 Ductility ratios due to pipe break open 53 3.8.6 Design of seismic Category I tanks Cmplete 6/1/84 54 3.8.6 Cmbination of vertical respcnses Caplete 6/1/84 55 3.8.6 Torsional stiffness calculation Cmplete 6/1/84 56 3.8.6 Drywell stick model develognent Complete 6/1/84 57 3.8.6 Rotational time history inputs Cmplate 6/1/84 58 3.8.6 "O" reference point for auxiliary Cmplete 6/1/84 building model 59 3.8.6 overturning mcznent of reactor Cmplete 6/1/84 building foundation mat 60 3.8.6 BSAP element size limitations Caplete 6/1/84 61 3.8.6 Seismic modeling of drywell shield Cmplete 6/1/84 wall 62 3.8.6 Dryall shield wall boundary Cmplete 6/1/84 conditions 63 3.8.6 Reactor building d me boundary Conglete 6/1/84 conditions 64 3.8.6 SSI analysis 12 Hz cutoff frequency Cmplete 6/1/84 65 3.8.6 Intake structure crane heavy load Cmplete 6/1/84 drop 66 3.8.6 Ingedance analysis for the intais Open structure 61 3.8.6 Critical loads calculaticn for Cmplete 6/1/84 l
reactor building d me 68 3.8.6 Reactor building foundation mat Cmplete 6/1/84 contact pressures M P84 80/12 5 - gs i
ATIACHMENT 1 (Cont'd)
DSE2 R. L MITIL 10 OPEN SECTICN A. SCHWENCER ITEM NUMBER SURTECT STATUS IEITER D&TED 69 3.8.6 Factors of safety against sliding and Complete f/1/84 overturning of drywell shield wall 70 3.8.6 Seismic shear force distribution in Complete 6/1/84 cylinder wall 71 3.8.6 Overturning of cylinder wall Ccuplete 6/1/84 72 3.8.6 Deep beam design of fuel pool walls Ccuplete 6/1/84 73 3.8.6 ASHSD dome :nodel load inputs Cmplete 6/1/84 74 3.8.6 Tornado depmssurization Ccoplete 6/1/84 75 3.8.6 Auxiliary building abnornal pressure Caplete 6/1/84 76 3.8.6. Tangential shear stresses in drywell Caplete 6/1/84 shield wall and the cylinder wall 77 3.8.6 Factor of safety against overturning Cmplete 6/1/84 of intake structure 78 3.8.6 Dead load calculations Caplete 6/1/84 79 3.8.6 Post-modification seismic lads for Cmplete 6/1/84 the torus i
! 80 3.8.6 Torus fluid-structure interactions Caplete 6/1/84 l 81 3.8.6 Seiraic <'asplacement of torus Cmplete 6/1/84
\
l 82 3.8.6 Review of seismic Category I tank Complete 6/1/84 design 83 3.8.6 Factors of safety for drywell Cmplete 6/1/84 buckling evaluation i
! 84 3.8.6 Ultimate capacity of containment Complete 6/1/84 (materials) 85 3.8.6 Inad ccxnbination consistency Cmpleta 6/1/84 l
! M P84 80/12 6 - gs
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e ATTACHMENT 1 (Cont'd)
DbdR R. L. MITIL TO l CPEN SECTION A. SOfWENCER ITEN' NUMBER JUBJECT STATUS IEITER DATED 86 3.9.1 Cmputer code validation open 87 3.9.1 Information on transients open 88 3.9.1 Stress analysis and elastic-plastic Cmplete 6/19/84 analysis 89 3.9.2.1 Vibration levels for NSSS piping Ccr.plete 6/29/84 systems 90 3.9.2.1 Vibration rronitoring gegram during Cmplete 7/18/84 testing 91 3.9.2.2 Piping supports and anchors Cmplete ' 6/29/84 92 3.9.2.2 Triple flued-head contairment Cmplete 6/15/84 penetrations 93 3.9.3 1 Load cmbinations and allowable Cmplete 6/29/84 stress limits 94 3.9.3.2 Design of SRVs and SRV discharge Ccaplete 6/ 29/84 piping 95 3.9.3.2 Fatigue evaluation on SRV piping Cmplete 6/15 / 84 and IECA downcomers 96 3.9.3.3 IE Information Notice 83-80 ccmplete 6/15/84 97 3.9.3.3 Buckling criteria used for emponent Cmplete 6/29/84 supports 98 3.9.3.3 Design of bolts Cmplete 6/15/84 99a 3.9.5 Stress categories and limits for Cmplete 6/15 /84 ccre support structures 99b 3.9.5 Stress categories ard limits for Cmplete 6/15/84 core support structures 100a 3.9.6- 10CER50.55a paragraph (g) Cmplete 6/29/84 M P84 80/12 7 - gs
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ATIACHMENT 1 (Cont'd)
DSER R. L. MITTL TO OPEN. SECTICN A. SCHWENCER ITEM NUMBER SUBJECT STATUS LETTER DATED 100b - 3.9.6 10CFR50.55a paragraph (g) Open 101 3.9.6 PSI and ISI programs for pumps and - Open valves 102 3.9.6 Leak testing of pressure isolation Couplete 6/29/84 valves 103al -3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103a2 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment -
3 103a3 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent I
103a4 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment 103a5 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103a6 3.10 Seismic ar.:1 dynamic qualification of Open mechanical and electrical equipnent 1
103a7 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103bl 3.10 Seismic and dynamic qualification of open mechanical and electrical equipnent 103b2 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103b3 3.10 Seismic and dynamic qualification of Open mecbanical and electrical equignent 103b4 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103b5 3.10 Seismic and dynamic qualificaticn of Open mechanical and electrical equipment i
M P64'80/12'8 - gs,
ATIACHMENT 1 (Cont'd)
DSER R. L. MITIL TO OPEN SECTION A. SOlWENCER ITEM NUMBER SUBJECT STATUS IEITER DATED 103b6 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment 103cl 3.10 Seismic and dynamic qualification of Open
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medianical ard electrical equipment 103c2 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment 103c3 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment 103c4 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment -
104 3.11 Environmental qualification of NBC Action mechanical and electrical equipment i 105 4.2 Plant-specific mechanical fracturing Cmplete 7/18/84 analysis 106 4.2 Applicability of seismic andd LOCA Cmplete 7/18/84 loading evaluation
- 107 4.2 Minimal post-irradiation fuel Cmplete 6/29/84 surveillance progra.n 108 4.2 Gadolina thermal conductivity Cmplete 6/29/84 equation i 109a 4.4.7 TMI-2 Item II.F.2 Open l
109b 4.4.7 IMI-2 Itan II.F.2 Open 110a 4.6 Functional design of reactivity Cmplete 7/ 27/ 84 control systems 110b 4.6 Functional design of reactivity Cmplete 7/27/84 control systens llla 5.2.4.3 Preservice inspection prcgram Cmplete 6/29/84 (cmponents within reactor pressure l boundary) l' l
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M P84 80/12 9 - gs
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ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL TO I SECTIm A. SOfWENCER OPEN ITEM NUMBER SUBJECT STATUS IErrER DATED
'lllb 5.2.4.3 Preservice inspection program Cmplete 6/29/84 (ccmponents within reactor pressure boundary) lllc 5.2.4.3 Preservice inspection program Cmplete 6/29/84 (wwirnts within reactor pressure boundary) ll2a 5.2.5 Reactor coolant pressure boundary - Cmplete 7/27/84 leakage detection 112b 5.2.5 Reactor coolant pressure boundary Ccmplete 7/27/84 leakage detection 112c 5.2.5 Reactor coolant pressure boundary Cmplete' 7/27/84 leakage detection 112d 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 leakage detection ll2e 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 leakage detection 113 5.3.4 GE procedure applicability Ccmplete 7/18/84 114 5.3.4 Ccmpliance with NB 2360 of the Sumer Ccuplete 7/18/84 1972 Adderda to the 1971 ASME Code 115 5.3.4 Decp weight and Charpy v-notch tests Cmplete 7/18/84 for closure flange materials 116 5.3.4 Charpy v-notch test data for base - Ccmplete 7/18/84 materials as used in shell course No.1 117 5.3.4 Cmpliance with NB 2332 of Winter 1972 Open Addenda of the ASE Code 118 5.3.4 Isad factors and neutron fluence for Open surveillance capsules M P84 80/12 10- gs l
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ATTAGMENT 1 (Cont'd)
DSER R. L. MITIL 'In CPEN SECTION A. SGWENCER ITEM NUMBER SUBJECT STATUS IEITER DATED 119 6.2 'IMI item II.E.4.1 Cm plete 6/29/84 120a 6.2 TMI Item II.E.4.2 Open 120b 6.2 'IMI Item II.E.4.2 Open 121 6.2.1.3.3 Use of NUREG-0588 Cmplete 7/27/84 122 6.2.1.3.3 Temperature profile Cmplete 7/27/ 84 123 6.2.1.4 Butterfly valve operation (post Cmplete 6/29/84 accident) 124a 6.2.1.5.1 RW shield annulus analysis Cmplete . 6/1/84 124b 6.2.1.5.1 RW shield annulus analysis Cmplete 6/1/84 124c 6.2.1.5.1 RW shield annulus analysis Cmplete 6/1/84 125 6.2.1.5.2 Design drywell head differential Complete 6/15/84 pressure 126a 6.2.1.6 Redundant position irdicators for Opn vacum breakers (and control rocm alams) 126b 6.2.1.6 Redundant position irdicators for Open vacuum breakers (and control room alarms) 1 27 6.2.1.6 Operability testing of vacum breakers Cmplete 7/18/84
! 1 28 6.2.2 Air ingestion Cmplete 7/27/84 1 29 6.2.2 Insulation irgestion Cmplete 6/1/84 130 6.2.3 Potential bypass leakage paths Cmplete 6/29/84 l 131 6.2.3 Administration of secondary contain- Cmplete 7/18/84 ment openings l M P84 80/12 11- gs
ATTACHMENT l' (Cont'd)
DSER R. L. MITIL TO CPEN SECTICN A. SCHWENCER ITEM NUMBER SUB.7ECT STATUS IEITER DATED 132 6.2.4 Containnent isolation review Canplete 6/15/84 133a 6.2.4.1 Containnent purge system Open 133b 6.2.4.1 Containnent purge system Open 133c 6.2.4.1 Containnent purge system Open 134 6.2.6 Containnent leakage testing Canplete 6/15/84 135 6.3.3 IPG and LPCI injection valve Open interlocks 136 6.3.5 Plant-specific LOCA (see Section Canplete ~ 7/18/84 15.9.13) 137a 6.4 Control roon habitability Open 137b 6.4 Control roon habitability Open 137c 6.4 Control roon habitability open 138 6.6 Preservice inspection program for Canplete 6/29/84 Class 2 and 3 cu yonents 139 6.7 MSIV leakage control system Catplete 6/29/84 140a 9.1.2 Spent fuel pool storage Canplete 7/27/84 140b 9.1.2 Spent fuel pool storage Canplete 7/27/84 140c 9.1.2 Spent fuel pool storage Canplete 7/27/84 140d 9.1.2 Spent fuel pool storage Corplete 7/27/84 141a 9.1.3 Spent fuel cooling and cleanup Open systen 141b 9.1.3 Spent fuel cooling ard clearup Open system 141c 9.1.3 Spent fuel pool cooling and cleanup Canplate 6/29/84 system M P84 80/12 12- gs
ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL 'IO OPEN SECTION A. SCHWENCER ITEM NUMBER SUBJirr STKIUS LETTER DATED 141d 9.1.3 Spent fuel pool cooling and cleanup Open system 141e 9.1.3 Spent fuel pool cooling ard cleanup Open system 141f 9.1.3 Spent fuel pool cooling and clearup Open system 141g 9.1.3 Spent fuel pool cooling and cleanup open system 142a 9.1.4 Light load handling system (related Closed 6/29/ 84 to refueling) (5/30/84 --
Aux.Sys.Mtg.)
142b 9.1.4 Light load handling system (related Closed 6/29/84 to refueling) (5/30/84-Aux.Sys.Mtg.)
143a 9.1.5 Overhead heavy load handling Open 143D 9.1.5 Overhead heavy load handling Open 144a 9.2.1 Station service water systen Open 144b 9.2.1 Station service water systen Open 144c 9.2.1 Station service water system open 145 9.2.2 ISI program and functional testirg Closed 6/15/84 of safety and turbine auxiliaries (5/30/84-coolirg systems Aux.Sys.Mtg.)
146 9.2.6 Switches and wiring associated with Closed 6/15/84 HPCI/RCIC torus suction (5/30/84-Aux.Sys.Mtg.)
147a 9.3.1 Ccmpressed air systems Open 147b 9.3.1 Cmpressed air systans open M P84 80/12 13- gs i
ATIACHMENT 1 (Cont'd)
DSER R. L. MITIL TO
'CPEN SECTION A. SCHWENCER ITEM NUMBER SUBJECT STA'IUS IEITER DATED 147c 9.3.1 cmpressed air systems Open 147d~ 9.3.1 Cmpressed air systems Open 148 9.3.2 Post-accident sampling system Open (II.B.3) 149a 9.3.3 Equipment and floor dre . ' age system Caplete 7/27/84 149b 9.3.3 Equipment and floor drainage system Cmplete 7/27/84 150 9.3.6 Primary containment instrument gas Open syst m 151a 9.4.1 Control structure ventilation system C mplete 7/27/84 151b 9.4.1 Control structure ventilation syste Cmplete 7/27/84 152 9.4.4 Radioactivity monitoring elements Closed 6/1/84 (5/30/84-Aux.Sys.Mtg.)
153 9.4.5 Engineered safety features ventila- Cmplete 7/27/84 tion system 154 9.5.1.4.a Metal roof deck construction Cmplete 6/1/84 classificiation 155 9.5.1.4.b Ongoing review of safe shutdown NRC Action capability 156 9.5.1.4.c Ongoing review of alternate shutdown NBC Action capability 157 9.5.1.4.e Cable tray protection Open 158 9.5.1.5.a Class B fire detection system Cmplete 6/15/84 159 9.5.1.5.a Primary ard secondary power supplies Cmplete 6/1/84 for fire detection system 160 9.5.1.5.b Fire water pump capacity Open M P84 80/1214- gs
ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL TO OPDi SBCIION A. SOfWENCER ITEM NUMBER SUELTECT STAT 11S IErrER DATED 161 9.5.1.5.b Fire water valve supervision Cmplete 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 Cmplete 6/1/84 164 9.5.1.6.e Remote shutdown panel ventilation Canplete 6/1/84 165 9.5.1.6.g Emergency diesel generator day tank Cmplete 6/1/84 protection l l
166 12.3.4.2 Airborne radioactivity monitor Complete 7/18/84 positioning ,
167 12.3.4.2 Ebrtable continuous air acnitors Cmplete 7/18/84 168 12.5.2 Equi [ ment, training, and grocedures Cmplete 6/29/84 for inplant iodine instrunentation 169 12.5.3 Guidance of Division B Regulatory Cmplete 7/18/84 Guides 170 13.5.2 Promdures generation package cmplete 6/29/84 subnittal 3 171 13.5.2 TMI Item I.C.1 Ccmplete 6/29/84 l 172 13.5.2 PGP Cm mitment Cmplete 6/29/84 173 13.5.2 Procedures coverity abnormal releases Complete 6/ 29/84 of radioactivity 174 13.5.2 Resolution explanation in FSAR of Cmplete 6/15/84-TMI Items I.C.7 and I.C.8 175 13.6 Physical security Open
, 176a 14.2 Initial plant test grogram Open M P84 80/12 15- gs
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i ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL TO SECIION A. SG WENCER CPEN NUMBER SUEk7ECT STATUS LEITER DATED ITEN 176b 14.2 Initial plant test program Open 176c 14.2 Initial plant test program Cmplete 7/ 27/ 84 176d 14.2 Initial plant test program Cmplete 7/27/84 176e 14.2 Initial plant test program Cmplete 7/ 27 / 84 176f 14.2 Initial plant test program open 176g 14.2 Initial plant test program open 176h 14.2 Initial plant test program Open 176i 14.2 Initial plant test program Cmplete 7/27/84 177 15.1.1 Partial feedwater heating Cmplete 7/18/84 17L 15.6.5 IECA resultirg fran spectrum of NRC Action postulated piping breaks within RCP 179 15.7.4 Radiological consequences of fuel NBC Action handling accidents 180 15.7.5 Spent fuel cask drop accidents NRC Action 181 15.9.5 TMI-2 Item II.K.3.3 cmplete 6/29/84 182 15.9.10 TMI-2 Item II.K.3.18 cm plete 6/1/84 183 18 Hope Creek DCRDR Open 184 7.2.2.1.e Failures in reactor vessel level Couplete 7/27/84 sensing lines 185 7.2.2.2 Trip system sensors ard cabling in Cmplete 6/1/84 l
turbine building j
l 186 7.2.2.3 Testability of plant protection open systems at power M P84 80/12 16- gs
4 ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL TO A. SCHWENCER CPEN SECTION SUBJECT STATUS LETTER DATED I1TM NUMBER ,
187 7.2.2.4 Lifting of leads to prform surveil- Open lance testing 7.2.2.5 Setpoint methodology Open 188 189 7.2.2.6 Isolation devices Open 190 7.2.2.7 Regulatory Guide 1.75 Ccrnplete 6/1/84 191 7.2.2.8 Scram discharge voltme Ccanplete 6/29/84 192 7.2.2.9 Reactcr mode switch Ccrnplete 6/1/84 193 7.3.2.1.10 Manual initiation of safety systems Open 194 7.3.2.2 Standard review plan deviations Ccrnplete 6/1/84 195a 7.3.2.3 Freeze-pcotection/ water filled Open instrument and sampling lines and '
cabinet temperattre control 195b 7.3.2.3 Freeze-protection / water filled Open instrunent and sampling lines and cabinet temperature control
! 19t 7.3.2.4 . Aring od ccmnon instrument taps Open 197 7.3.2.5 Microprocesscr, multiplexer and Ccznplete 6/1/84 ccrnputer systems 198 7.3.2.6 TMI Item II.K.3.18-AES actuation Open 199 7.4.2.1 IE Bulletin 79-27-toss cf ncn-class Opn IE instrumentaticn and control power l system bus during cperation 200 7.4.2.2 Renote shutdown system Ccanplete 6/1/84 201 7.4.2.3 RCIC/HPCI interactions Opn 202 7.5.2.1 tevel neasurement errces as a result Open l of environmental temprature effects t
on level instrumentation reference leg l
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! M P84 80/1217- gs
ATTACHMENT 1 (Cont'd)
DSER R. L. M1TIL TO CPEN SECTION A. SCHWENCER ITEM NUMBER SUR7ECT STATUS IEITER IATED 203 7.5.2.2 Regulatory Guide 1.97 Open 204 7.5.2.3 TMI Item II.F.1 - Accident nonitoring Open
~
205 7.5.2.4 Plant process cmputer system Ccmplete 6/1/84 206 7.6.2.1 High pressure / low pressure interlocks Cmplete 7/27/84 207 7.7.2.1 HEIBs arti consequential control system Open failures 208 7.7.2.2 Multiple control system failures open 209 7.7.2.3 Credit for non-safety related systems Cmplete - 6/1/84 in Chapter 15 of the FSAR 210 7.7.2.4 Transient analysis recording system Cmplete 6/1/84 211a 4.5.1 Control rod drive structural materials Caplete 7/27/84 211b 4.5.1 Control rod drive structural materials caolete 7/27/84
.s, 211c 4.5.1 Control rod drive structural materials Cmplete 7/27/ 84 211d 4.5.1 Control rod drive structural materials Co@lete 7/27/84 211e 4.5.1 Control rod drive structural materials Cmplete 7/27/84 212 4.5.2 Reactor internals materials Caplete 7/27/84 213 5.2.3 Reactor coolant pressure boundary Cmolete 7/27/84 material 214 5.1.1 Engineered safety features materials Cm plete 7/27/84 215 10.3.6 Main steam and feedwater system Cceplete //27/84 materials 216a 5.3.1 Reacter vessel materials Cmolete 7/27/84 M P84 80/12 18- gs
l l
l ATTACliMENT 1 (Cont'd)
DSER R. L. MITIL TO-CPEN SECTION A. SQiWENCER ITEM NUMBER -
SUILTECT STATUS LETTER DATED 216b 5.3.1 Reactor vessel materials Ccmplete 7/27/84
- 2 17 9.5.1.1 Fire protection organization Open 2 18 9.5.1.1 Fire hazards analysis Ccmplete 6/1/84 219 9.5.1.2 Fire protection administrative Open controls 220 9.5.1.3 Fire brigade ard fire brigade Open training 221 8.2.2.1 Physical separation of offsite Open transmissicn lines -
222 8.2.2.2 Design provisions for re-establish- Open ment of an offsite power sourca 4
223 8.2.2.3 Independence of offsite circuits open between the switchyard and class IE buses i 224 8.2.2.4 Ccunon failure mode between onsite open and offsite power circuits 225 8.2.3.1 Testability of autcmatic transfer of Open power fran the nonnal to preferred power source 226 8.2.2.5 Grid stability Open 227 8.2.2.6 Capacity and capability of.offsite Open
- circuits i.
228 8.3.1.l(1) Voltage drop during transient condi- Open tions 229- 8.3.1.l(2) Basis for using bus voltage versus Open actual connected load voltage in the voltage drcp analysis 230 8.3.1.1(3) Clarification of Table 8.3-11 Open M P84 80/12 19- gs
. 1 l
ATTAC19ENT 1 (Cont'd)
D6ER R. L. MITTL TO OPEN SECTIOi A. SCIDENCER ITEM NUMBER SUELTECT STATUS LETTER DATED 231 8.3.1.l(4) Undervoltage trip setpoints Open 232 8.3.1.1(5) Ioad configuration used for the Open
. voltage drop analysis 1.
233 8.3.3.4.1 Periodic system testing Open 234 8 .3 .1.3 Capacity and capability of onsite Open
, AC power supplies and.use of ad-ministrative controls to prevent overloading of the diesel generators 235 8.3.1.5 Diesel generators load acceptance Open -
test 236 8.3.1.6 Compliana with position C.6 of Open i
IG 1.9 237 8.3.1.7 Decription of the load sequencer Open
! 238 8.2.2.7 Sequencing of loads en the offsite Open power system -
t 239 8 .3 .1.8 Testing to verify 80% mininun Open voltage 240 8.3.1.9 Ccupliance with BrP-PSB-2 Open 241 8.3.1.10 toad acceptance test after prolonged Open no load operation of the diesel generator 242 8.3.2.1 Cmpliance with position 1 of Regula- Open tory Guide 1.128 243 8.3.3.1.3 Protection or qualificaticn of Class Open IE equipment frca the effects of fire suppression systems 4
244 8.3.3.3.1 Analysis ard test to denonstrate Open adequacy of less than specified separation i
M P84 80/12 20- gs
_ _ __ _ .. u ._ . __ _ _ . __ _ . . . _ . . _ _ - - - - , _ _ _ . _ . _ . _ . . , _ _
KITACHMENT 1 (Cont'd)
DSER R. L. MITTL TO OPEN SECTICH A. SCHWENCER ITEM NUMBER SUETECT STATUS LETTER DATED 245 8.3.3.3.2 The use of 18 versus 36 inches of Open separation between raceways 246 8.?.3.3.3 Specified separaticn of raceways by Open analysis and test 247 8.3.3.5.1 Capability of penetrations to with- Open stand long duration short circuits at less than maximum or worst case short circuit 248 8.3.3.5.2 Separation of penetration primary Open and backup protections
- 249 8.3.3.5.3 The use of bypassed thermal overload Open protective devices for penetration protections 250 8.3.3.5.4 Testing of fuses in accordance with Open R .G . 1.63 251 8.3.3.5.5 Fault current analysis for all Open representative penetration circuits 252 8.3.3.5.6 The use of a single breaker to provide Open penetration protection 253 8.3.3.1.4 Conmitment to protect all Class lE Open equipment frcm external hazards versus only class lE equipnent in one division 254 8.3.3.1.5 Protecticn of class lE power supplies Open frca failure of unqualified class lE loads 255 8.3.2.2 Battery capacity Open l 256 8.3.2.3 Autcmatic trip of loads to maintain Open
.aufficient battery capacity i
M P84 80/12 21- gs l
ATIACHMENT 1 (Cont'd)
DSER R. L. MITIL TO OPEN SECTICN A. SCHWENCER ITEM NUMBER SUETECT STATUS IEITER DATED 257 8.3.2.5 Justification for a 0 to 13 second Open load cycle '
258 8.3.2.6 Design and qualificaticn of DC Open system loads to operate between ministan and maximLa voltage levels 259 8.3.3.3.4 Use of an inverter as an isolation Open device 260 8.3.3.3.5 Use of a single breaker tripped by Open a IDCA signal used as an isolation device -
261 8.3.3.3.6 Autmatic transfer of loads and Open interconnection between redundant divisions TS-1 2.4.14 Closure of watertight doors to safety- Open related structures TS-2 4.4.4 Single recirculaticn loop operation Open TS-3 4.4.5 Cote flow monitoring for crud effects Ccmplete 6/1/84 1
TS-4 4.4.6 Imse parts monitoring systen Open TS-5 4.4.9 Natural circulaticn in normal Open operation TS-6 6.2.3 Secondary contairanent negative Open pressure TS-7 6.2.3 Inleakage and drawdown time .in Open secondary containment TS-8 6.2.4.1 Isakage integrity testing Open TS-9 6.3.4.2 ECCS subsystem periodic ocmponent Open testing TS-10 6.7 MSIV leakage rate l
M P84 80/12 22- gs l
_ . - - . , , . . _ _ . . . _ . - . . . . . . . _ _ . . . , - _ - . - , _ , c. - . - . . . , . _ .,-._, _ _ . . . _ _ . . - _ - -
ATTACHMENT 1 (Cont'd)
DSER R. L. MITTL TO OPEN SECTION A. SCHWENCER ITEM NUMBER SUBJECT STATUS LETTER DATED TS-Il 15.2.2 Availability, setpoints, and testing Open of turbine bypass system TS-12 15.6.4 Primary coolant activity II-l 4 .2 Fuel rod internal pressure criteria Conglete 6/1/84 If-2 4.4.' Stability analysis submitted before Open second-cycle cperation
~
l i
l l
e M P84 80/12 23- gs
- - . - _-- -m _ - . ,__.-_.- . _.y-- ..-- _ . - - , . - -- - - - - - - - _ . _ _ . _ . , . .m_. ..-.-
ATTACHMENT 2 DATE: 7/27/84 DRAFT SER SECTIONS AND DATES PROVIDED SECTION DATE SECTIOtt DATE t
3.1 3.2.1 11.4.1 ,
3.2.2 11.4.2
.5.1 11.5.1 5.2.1 11.5.2 6.5.1 13.1.l '
8.1 13.1.2 8.2.1 13.2.1 8.2.2 13.2.2 8.2.3 13.3.1 8.2.4 13.3.2 3.3.1 13.3.3 8.3.2 l'3 . 3 . 4
- 8.4.1 13.4 8.4.2 13.5.1 8.4.3 15.2.3 8.4.5 15.2.4 8.4.6 15.2.5 8.4.7 15.2.6 '
8.4.8 15.2.7 9.5.2 15.2.8 9.5.3 15.7.3 9.5.7 17.1 9.5.8 17.2 10.1 17.3 10.2 17.4 10.2.3 10.3.2 10.4.1 10.4.2 10.4.3 10.4.4 11.1.1 11.1.2 11.2.1 11.2.2 11.3.1 11.3.2 CT db MP 84 95/03 01
DATE: July 27, 1984 ATTACHMENT 3 OPEN ITEM DSER SUBJECT SECTION
- NUMBER 2a 2.3.3. Accuracies of meteorological measurements 2b 2.3.3. Accuracies or meteorological measurements 2c 2.3.3. Accuracies of meteorological measurements 3a 2.3.3 Upgrading or onsite heteoro-logical measurements program (III . A. 2. )
30 2.3.3 Upgrading of onsite meteoro-logical measurements program (III.A.2.)
28a 3.4.1 Flood Protection 28b 3.4.1 Flood Protection 28c 3.4.1 Flood Protection 28d 3.4.1 Flood Protection 28e 3.4.1 Flood Protection 28g 3.4.1 Flood Protection 110a 4.6. Functional design ot reactivity control system 110h 4.6 Functional design of reactivity control system 112a 5.2.5 Reactor coolant pressure boundary detection ll2b 5.2.5 Reactor coolant pressure boundary detection
OPEN ITEM DSER- SUBJECT SECTION NUMBER
- 5. 2. 5. Reactor coolant pressure 112c .
boundary detection 5.2.5 Reactor coolant pressure 112d boundary detection .
5.2.5 Reactor coolant pressure 112e bouncary detection 121 6.2.1.3.3 Use of NURE6-0588 122 6.2.1.3.3 Temperature profile 128 6.2.2 Air ingestion 140a 9.1.2 Spent tuel pool storage 140b 9.1.2 Spent fuel pool storage 140c 9.1.2 Spent fuel pool storage 140d 9.1.2 Spent fuel pool storage 144a 9.2.1 Station service water system [
144o 9.2.1 Station service water system 144c 9.2.1 Station service water system 149a 9.3.3 Equipment and floor drainage system 149b 9.3.3 Equipment and floor drainage system 151a 9.4.1 Control structure ventilation system 151m 9.4.1 Control structure ventilation system 176c 14.2 Initial plant test program 176d 14.2 Initial plant test program 176e 14.2 Initial plant test program 1761 14.2 Initial plant test program
OPEN ITEM DSER SUBJECT SECTION NUMBER 184 7.2.2.1.e Failures in reactor vessel level sensing lines 206 7.6.2.1 High pressure / low pressure
- interlocks 211a 4.5.1 control rod drive structural materials 211b 4.5.1 Control rod drive structural materials 211c 4.5.1 Control rod drive structural materials 211d 4.5.1 Control rod drive structural materials 211e 4.5.1 Control rod drive structural materials 212 4.5.2 Reactor internals materials 5.2.3 Reactor coolant pressure 213 boundary material 214 6.1.1 Engineered safety features material 215 10.3.6. Main steam and feed water system materials 216a 5.3.1 Reactor vessel materials 216b 5.3.1 Reactor vessel materials
-4 4
9
- .f e
- 4 ATTACEMENT 4 O
e 7
DSER Open Item No. 2a (Section 2.3.3)
Accuracies of Meteorological Measurements The applicant states that the entire onsite meteorological measurements system complies with the accuracy specifications presented in RG 1.23, "Onsite Meteoro-logical Programs." However, the applicant has not provided (as requested in
- RAI 451.10) estimates of the overall system accuracy for each parameter measured.
The types of wind speed and direction sensors and recording equipment identified by the applicant in Table 2.3-29 have been used by other applicants and licensees to meet the accuracy specifications of RG 1.23.
Response
For the information requested above, see the response to DSER Open item 3a and b.
i l
i p -, -- -- - - - - . ,- ------..- ~ , , - ,- -,-,., - n-,--.-... ,, w,. -- . - + - . - - ,
DSER Open Item No. 2c (Section 2.3.3)
The metacrological measurements program, during plant operation, will include j those parameters currently measured. Meteorological parameters are to be available for display through the radiation monitoring system central radia-tion processor (CRP), although the method of display has not been specified.
Calculations of atmospheric transport and diffusion are also to be available through the CRP, although the models and/or methodology have not been described.
Response
DSER Open item 3a and b.For the information requested 3e to above, see th l
l
DSER Open Item No. 2b (Section 2.3.3)
The applicant's method for aetarsining vertical temperature gradient is uncommon, using a matched pair of therwistors. Additional information is required from the applicant to demon-o) strate that the accuracies of noteorological esasurement comply with the system accuracy specifications presented in RG 1.23.
Response
For the information requested above, see the response to DSER Open item 3a and b.
i g ER Open Item No. 37&b (S cticn 2.3.3) tipgrading of Onsite Meteorological Measurements Program (7II.A.2)
To address the setecrological requirements for emergency preparedness planning outlined in 10 CFR 50.47 and Appendix E to 10 CFR 50, the applicant will be required to upgrade the operational meteorological seasurements program to m t the criteria in NUREG-0654, Appendix 2, " Criteria for Preparation and Evaluation of Radiological Energency Response Plans and Preparedness in Sup-port of Nuclear Power Plants." The upgrades must be in accordance with the schedule of NUREG-0737, III.A.2, " Clarification of TMI Action Plan Require-ments," or its supplements. The incorporation of current meteorological data i
into a real-time atmospheric dispersion model for dose assessment
- will also be considered as part of the upgraded capability.
Response
For the information requested above, see revised Question Response 451.6, FSAP. Section 2.3.3.3 and Table 2.3-29a, b and c:-
4 l
, - - . . .,,-,-.,,,,.,....,.,..-..--.,,.,.-,..,.,,-_-_,,,_-,m,-__ _. _ ,,,,,, _ ,,n. ,-.-n,,n.._, ,,,m _ __,n,,,,,-,v,,,.m
- DUESTION 451.6 -
Section 2.3.2 provides comparisons of meteorological data collected at the Hope Creek site with data collected at the National Weather Service station at Wilmington, Delaware to determine the representativeness of "the key meteorological parameters crucial to the safety, operation, and construction of Hope Creek Generating Station." Additional meteorological data have also been collected on Artificial Island since 1969 in support of construction and operation of the Sales Nuclear Power Plant. These data can also be compared to data for Hope Creek if j
different meteorological measurements programs are in use for 3
each Nuclear Power Plant.
a) Provide comparisons of annual wind direction frequencies at the 33-ft, 150-ft, and 300-ft for both the Salem and Hope Creek facilities for the available period of record.
Include the number of valid observations and the total l
possible observations for each period of record.
i b) Provide comparisons of annual atmospheric stability ~
- distributions (Pasquill stability classes A-G) based on the measurement of vertical temperature gradient between the 300-ft and 33-ft levels and between the 150-ft and 33-ft l
levels for both the Salem and Hope Creek facilities for the available period or record. Include the number of valid observations and the total possible observations for each period of record.
RESPONSE
- a) Annual wind direction frequencies at the 33 ft, 150 ft, and 300 ft levels observed during June 1969 to May 1971 (SGS preoperational data) are shown in Table 451.6-1. The 150 ft i
wind distribution was derived from January 1970 to May 1971 data. Annual wind direction distribution for the same three 4
levels for the period January 1977 to December 1981 are l
presented in Tables (51.6-2, 451.6-3 and 451.6-4, respectively. q 3 gg,g7 I
COMPARISONS s l I
33 feet l i
l Highest wind direction frequencies from the period 1969 to 1971The (SGS) compare favorably with those frcr 1977 to 1981 (HCGS).
SGS data shows the highest site has a bimodal distributicn.
frequency of wind directions are SE-SSE-S and W-WNW-NW. HCGS data shows the same pattern. Frequencies other than these modes are evenly distributed throughout the compass points. For all individual years, the data recovery rates are above 90 percent.
451.6-1 Amendment /'
DSER OPEN ITEM u - - . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _
INSERT A .
Data collection for the period of 1969 to 1971 was from a tower located 1400 feet north of the Hope Creek Reactor Building at a latitude of 39 degrees, 28 minutes,13 seconds north, and a longitude of 75 degrees, 32 minutes, 12 seconds west. This tower was originally located to support preoper-ational data collection for the Salem Stations. The tower was relocated to the existing location to f acilitate the construction of the Hope Creek Station and the cooling tower, O
M P84 93/04 3-dh 4
0 ih DSER OPEN ITEM ] I
~ ~ ' ' - * - - - ' ' * - - ,vg - _ _ _ , , _ __ _
HCGS FSAR 10/83 Monthly and annual joint frequency distributions of wind speed
__ and direction, based on atmospheric stability classes, are referenced in Section 2.3.2.1.1. Thc 5-year data base containing hourly site meteorological data from January 1977 to December
- 1981 was used as input in the analysis.
2.3.3.3 Operational Cata Displav l The meteorological parameters required by Regulatory Guide 1.97 will be incorporated in the data base to be included on the control roce integrated display system (CRIDs) computer. The i display of those parameters will be available as part of the i
display function along with all other related Regulatory Guide 1.97 variables. -
1
- The radiation monitoring system central radiation processor (CRP) computer will provide 15-minute average meteorological monitoring system parameters. The parameters available for display are 33-ft wind speed and wind direction, 150-ft wind speed and wind direction, 300-ft wind speed and wind direction, delta-temperature stability indicators between 300- and 33-ft and 150-4 and 33-ft, as well as precipitation, barometric pressure, solar radiation, and ambient temperature at 33 ft.
, Atmospheric transport and diffusion during normal operation will be calculated by the CRP. A method for determining atmospheric transport and diffusion throughout the plume exposure emergency planning zone during emergency conditions is being developed.
! 19scNT [6 j 2.3.4 SHORT-TERH DIFFUSION ESTIMATES 1 l ,
i 2.3.4.1 Obiective 3:
The objective is to provide conservative and realistic short-term estimates of relative concentration (1/0), at both the site boundary and the outer boundary of the low population zone (LPZ) following a hypothetical release of radioactivity from HCGS. The assessment is based on the results of atmospheric diffusion modeling'and onsite meteorological data.
A ground-level accidental radionuclide release from HCGS is analyzed at various distances. Conservative and realistic X/Q values at the exclusion area boundary (EAB) are derived for the oson oPEN IrEM jf 2.3-27 Amendment /
. / l'8
. - - -. _ - _ . ~ . .
I U s s n.~i 5 The postoperational data collection program will consist.of an enhancement to the preoperational program. . The enhancement consists of a primary and backup data acquisition system (DAS) and a {
, communication computer. A diagram of the system configuration is provided in Figure 2.3-6. A list of the system hardware components is tabulated on Table 2.3-29a. There are no changes to the meterological tower, sensors, power supply, strip chart recorders, or translator cards. The rain gauge has been changed fror. a weighing bucket to a tipping bucket which meets the NRC criteria of measuring .01 inches of precipitation. This change has been incorporated in Table 2.3-29.
The primary and backup DAS are configured with identical hardware. Each DAS consists of a Hewlett-Packard 982Ga Computer, 3497A Data I Acquisition / Control Unit, and a Dames & Moore transient protection system. Each DAS is provides with two communication ports, one as a link to the communications computer, and the other for direct
- dial-up capability. Each DAS provides for up to seven days of fifteen minute averages. The pri-l mary DAS collects data from the meterological parameters listed in Table 2.3-29. The backup i DAS collects wind speed and direction from the the 4
three tower elevations and two delta T's, as well as the backup meterological tower. The data l
acquisition system calculate a sigma theta for each of the three level wind directions.
{
. The communications computer which consists of a j DEC PDP 11/23 computer and RX02 dual disk drive.
The communications computer is configured with nine I/O ports. The I/O ports support data transfer / interrogation with the Salem Control Room ~
the Hope Creek Radiation Monitoring System via a
- meteorological system link (whien incorporates a HP9915 computer) as well as three dial up ' ports.
l The communication computer also supports a display i unit in the the Hope Creek EOF as well as communi-i cation to the primary and backup DIS.
System accuracy is presented on Tables 2.3-29b and 2.3-29c.
)
i M P84 93/04 1-dh
- j. DSER OPEN ITEM
plSsAT S 2
k*d" J The postoperational data collection program also includes an additional ineterological tower identi-fies as a backup meterological tower, consisting of a 10 meter telephone poll. The backup tower is located aproximately 500 feet south of the primary meterological monitoring tower. Backup meterolo- i gical data provides wind speed, wind direction, Backup meteorologicl and a computed sigma that'a. !
data provides wind speed and wind direction and a computer sigma theta.
N 0N M MAbEMO % reMe5 s.
- m g % < w RY:dh 5/30/84 M P84 93/04 2-dh DSER OPEN ITEM
,. , . , ,,n.,--,,---,ww.-,w-w- --- ,--,-----.,-,,-,,n, _,-,---,,--,,,--.,_,--,,-n,.-,-.,-,,..-,-,---n...
l .
i i
MCGS FSAs l Page 2 of 2 TABLE 2.3-21 (cont)
! meight Above i Toeser asse, I nst s-- - " and Characteristic Stris Beegreare I ft Sensed Parameter Bocorded Feraspet.or i climet - stodel 011, 3 cop Boterline - Anger l 33 wind speed wind speed stedel Ls25 anemometer. Threshold 0.6 mesa, distance constant <5 f t, i operating range e to 110 mph, accuracy a15 or G.15 mpsi,
- l whichever le greater j ,
C11 met - pendel 012-10 wine Esterline - Segue Wind direction wind direction wane. Threshold 8.75 mph, stedel LS2S 41 stance constant <3.3 ft, l dampia, ratio e.e .
j1 i
i
! Temperature-elfferential l T300-1 33888
! T 150"T33aa8 BESG stedel SN 110 accuracy et.5*F leastsemise M113 j
Dow point Dew point .
C11 met - stzeel 416-1 peotor- Leede S porttureep w ratere-ambient 6 Temperature speseemas aspirated temperature shield .
with C11 met 015-3 thermister saniti-point i
accoracy te.15*C' f
i climet - sendel 014-90 pressere asterline - Angus 6 Baremetric presonare Barometric pressure f
- traneencer. Range 28-32 in. sg seodel &
r " - ? 000 ^ i- --- ramm ,_ ,_~ Ester 11me - Angue 3 mainfall mainfall l
IM R.I. Mo6*L 30Z. l~e rP.y seodel A i
D.e we r A ce one y a.o t .an e
{
as 8 Temperatura taken as part of temperature dif ferential messetement T300 - T33 ses temperature taken as part of temperature differential measurement T 50r- 1 T33 (s a Patr=4 Climet 015-3 thermistor. Accuracy a0.1*C.
DSER OPEN ITEM l
see WGe /W ,=,, e Mf
HCGS FSAR TABLE 2.3-29a DATA ACQUISITION SYSTEM HARDWARE 1
MANUFACTURER MODEL QUALITY DESCRIPTION Hewlett Packard 9826A 2 Computer Hewlett Packard 98256A 2 256K-Byte Memory Expans ion Hewlett Packard 98626A 4 Serial Ports Hewlett Packard 3497 2 Data Acquisition /
Control Unit 44421A 20-Channel Analog Hewlett Packard 2 Multiplexor Hewlett Packard 44425A 2 16-Bit Status Inout Dames & Moore -- 2 Transient Protection Modules (analog, status, voltage ref e rence) -
Hewlett Packard 9915 1 Computer DEC 11/23 1 Computer DEC RXO2 1 Disk Drive DEC Vt 103BA 1 CRT DEC Bell 212A 1
5 Serial { cts Bell 202T 6 Modem Modem 1)(I 1
1 (1) Or equivalent modem 4
DSER OPEN ITEM
. . . - _ -- - ~ -.. . - -
O m
M
- At O .
au tg 21 H
- S
$ ---- - .~ __.._
L .coe .a ThaLa 2.3-396 STsTen - - smaos DELTA TeeptanfUS:
Tsnesmaguas seusote? paecte Tafseer coneousm? wtre sprao wino stascTaoss (3ee-33 tase-333 tamos aseness (3e npal flee meal toscassst (---e cas.stos) e- es catsaust ( - -*=<cas.stus) tincassi
+3 + e.te + e.te + e.te 1 e.S .et sensor + e.ts + .3e + 1.oe Trans lator + 4.21
+ e.21 1 e.21 e - - -
e.ess -
own + e.se35 + e.ee65 + e.et?
- e.es2 + e. seas + e.es26 I a.e13 - -
e e e.se e e e e -
software e other -
Total e. 34 3s e.st65 a.227 1 3.se2 + e.1e26 + e.1826 + e.113 + e.5 s 2 naalaus arter
+ e.te3 + e. net 1 **** .e1 naos s.= e.21 e.3T t.e2 3.se + e.te3 ,
sqaare arrer 1 e.S + 1.5 .et m.c. 1.23 a.s e.5 - 5.s + e.15
+ e.15 specifScatton til tastrwatation type end spectiteetion proelded en Table 2.3-29 and 2.3-29e.
8 23 The [metentaneave error for wind speed meneeremente, esametag the indletdeal compmeest errous are addittee end ineapondent t reet osa egeere error), to within the m.C. 1.23 orectitcattame for att uted speede less than 45 mph. The grror of time everagea wind speede wit! be lese them the imetentaneese root swa egeare error Ethis statement to applicable for att other parametear in this dirceeston). Therefore, for wtad speede coseldered to be mee t critical for dispersion cancelatione, the eettested error le wall within the R.C.1.23 specificettom.
DFTage n pe4 40/11 1-ge e
e .
O e
O
BCGS FSAR Table 2.3 --29c ARTIFICIAL ISIAND DIGITAL DATA ACQUISITION SYSTEM ACCURACIES The following system accuracies are based upon VEN30R accuracy specifications and the following conditions:
o L year calibration interval o 5-1/2 digits displayed on DVM o Auto Zero ON VOLTMETER ACCURACY ,
ERROR PERCENT PLUS RESOLUTION RANGE (V) 0F READING .
ERROR (MV)
.119999 -
.015 .003 1.19999 .015 .02 11.9999 .015 .L PARAMETER ERROR OAS INPUT ERROR MAXIMUM 4 DVM ENGINEERING CALCULATION DAS
' PARAMETER RANGE VOLTAGE UNITS POINT ERROR Temperature 1.19999V 0-l.0V 45*C 45*C 0.013*C Delta-Temperature 1.19999V 0-1.0V +10*C 10*C 0.0026*C Dew Point ll.9999V 0-5.0 +100
- F 100*F 0.022*F Wind Speed L.19999V 0-1.0V 0-100, mph 50 mph 0.0095 mph Wind Speed 1.19999V 0-1.0V 0-100 mph 10 mph 0.0035 mph Wind Speed 1.19999V 0-1.0V 0-100 mph . 20 mph 0.0065 mph Wind Direction 1.19999V 0-1.0V 0-540' 540' O.092*
Precipitation 1.19999V 0-1.0V 0-1" -
0.00=b Pressure 1.19999V 0-1.0V 28-32"Hg 32Hg 0.00068"Hg Solar Radiatisn 1.19999V 0-1.0V 0-ZLy/ min 2Ly/ min 0.00034Ly/ min aThe data acquisition system error is due entirely to HP-3497A instrument error. Software calculations are computed to 12 significant digits. !
I Therefore, software error is negligible.
)
bP recipitation is calculated using a step-function conversion technique with sufficient noise margin that an error of 0.00" is achievable over an entire calibration period interval.
4 e
DSER oPEN ITEM 3
HCGS 2<
0, b, c o de s.
3 DSER Open Item No. 28 (DSER Section 3.4.13
, FLOOD PROTECTION I i ,
The design of the facility 'for flood protection was reviewed in l
,: accordance with Section 3.4.1 of the Standard Review Plan (SRP) 4 NUREG-0800. An audit review of each of the areas listed in the ,
i
! " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" portion of the SRP section. Conformance with the acceptance j'
criteria formed the basis for our. evaluation of the design of the n
F facility for flood protection with respect to the applicable regulations of 10 CFR Part 50.
i In order to assure conformance with-the requirements of General ,
- Design criterion 2, " Design Bases for Protection Against Natural
! Phenomena," our review of the overall flood protection design l
included all systems and components whose failure due to flooding j could prevent safe shutdown of the plant or result in uncontrolled j, release of significant radioactivity.
The applicant has sited the plant (at elevation 22.5 feet Mean s Sea Level (MSL)) along the Delaware River near the point where the !
The design basis flood is l- river flows into the Atlantic Ocean.
the result of the probable maximum hurricane (PMH) surge with l,
- wave runup coincident with the 10% exceedance high tide. The design basis flood level for all structures is 34.8 feet MSL,
~
i' j including wave activity (refer to Section 2.4.2 of this SER).
l The design basis flood level of 34.8 feet MSL represents plant i submergence at the plant site by 12 feet 3.6 inches. Vertical i and horizontal construction foints are provided with waterstop to I elevation 32 feet MSL. ['The applicant must water-proof all safety-
' related structures and all penetrations to those structures to a higher elevation than the flood elevation of the design basis 3
~
flood (PMH).} g
[ The probable maximum flood which results in over 12.3 feet of I water onsite is due to the PMH and is greater than the flooding
- due to the probable maximum precipitation. l l'
The personnel access doors to areas where flood protection must be provided are all submarine doors which open outward, except l' doors 318 and 158. [In order to comply with the guidelines of l Regulatory Guide 1.102, " Flood Protection for Nuclear Power l
Plants", Position C1, the applicant must modify doors 31B and '
- 1 58 to be submarine doors or equivalent for these doors to open '
- outward or assume the doors are open during the design basis floodandverifythatnosafety-relatedequipmentwillbeflooded)LSRh l (The applicant has not provided information requested concerning i Regulatory Guide 1.102, Position C.2, and therefore no conclusions t
28-1 l
i '
s .
Item No. 28 (Cc...'d) ,pg -
can be made concerning compliances at this time [7 /3he applicant has not committed to providing sensors on all doors and hatches l
~in exterior. walls which are b(low the desgin basis flood elevation, i I
plus wind-generated wave ef fects to alarm in the control room when they are opened. As an alternative, the applicant may l
, provide the results of a flooding analysis with the administra-l tively controlled doors open and which shows that no safety-related !
equipment will be flooded.} 39c, (The site contains non-seismic Category I tanks. The applicant
~
has stated that the site drainage system will prevent the contents 4
i- of the failed tanks (as the result of a safe shutdown earthquake) f rom flooding the safety-related structures. The applicant has j not identified the site drainage system as safety-related, seismic j- Category I. The site drainage. system must be safety-related and i seismic Category I in order to take credit for the system af ter t
- design basis event. Similarly, the site drainage system should be i tornado and tornado missile protected if the drainage system is needed to prevent any flooding resulting from tank (s) fsilure due to a tornadic even or due to tornado generated missiles.] -JSd i
The applicant has stated that the electrical cables will continue '
to function properly even if the manholes and duct banks are
- flooded. The ability of the cables to perform the function if j they are flooded with sea water and the lorg-term effects of continued submergence in sea water is discussed in Section3 8.3 of this SER.
['In response to our concern regarding internal flood protection, i the applicant indicated that their discussion of plant features to prevent internal flooding of redundant safety-related equipment was in Section 6.1. 3.e of the FSAR. There is no Section 6.1.3.e in the FSAR.]- Age i
j [The applicant has not addressed our concern associated with the structural integrity of the safety-related structures during the design basis flood and the effects of " floating" missiles. Since the Delaware River is a navigable waterway with the refineries and naval shipyard in Philadephia, the applicant must address the l j effects of ships and boats with a draft of less than 12 feet i i hitting the walls and penetrations of safety-related structures. !
Some ships which do travel up and down the Delaware River and can j
! have a draft of less than 12 feet are the " Newport
- class LSTs i (LST-1179 series), the "DeSoto County" class LSTs (LST-ll73 series), the " Anchorage" class LSDs (LSD-36 series), submaries (especially the non-nuclear power submaries), tug boats, visiting l
j "American" ships from foreign countries, oil tankers (when they are empty), and a large host of pleasure craf t.]. g3 $ ,
i 28-2 I I
r Item No. 28 (Cont'd)
Because the applicant has not adequately addressed the sta'ff's concerns identified above, we cannot conclude compliance with General Design Criterion 2 and the guidelines of Regulatory Guides 1.102, " Flood Protection for Nuclear Power Plants," Positions C.1 and 1.59, " Design Basis Floods for Nuclear Power Plants", Positions C.1 and C.2 and Branch Technical Position ASB 3-1, " Protection Against Piping Failures in Fluid systems Outside Containment".
We will report resolution of these items in a supplement to this SER. The design of the facility for providing protection from flooding does not meet the acceptance criteria of SRP Section 3.4.1.
RESPONSE
- a. The requested information with respect to waterproofing all safety-related structures to a higher elevation than the flood elevation of the design basis flood (PMH) has been provided in response to Question 240.8.
- b. Doors 3331B and 3315B are watertight (submarine) doors and although they are installed in an unseated position (they swing inward), both doors have been designed for specified unseating pressure of 19 feet of water. To assure that these doors will not be inadvertently opened or left open, both doors are locked closed and administratively controlled during a flood event.
- c. HCGS procedure " Acts of Nature", will commit to ensure that exterior doors and hatches are closed and locked by administrative procedure under impending flood conditions.
J d. TheresponsetoFSARQuestion410.h.hasbeenrevisedtostate i that the site drainage system is not required to prevent the contents of failed tanks (as the result of a safe shutdown carth-quake) from flooding the safety-related structures.
- e. The response to NRC Question 410.9 has been revised to refer to Section 3.6.1.e instead of 6.1.3.e.
i f
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- - - , --..,.-,.._..~_w..n. . . . , _ . . . . , , . , _ , _ - _ _ . _ , , . , , _ . - - _ _ , , . , _ , , - . . .
W HCGS FSAR 10/83 QUE37191. 41_0_. 7 (SECTION 3.4.1)
For these nonseismic Category I vessels, pipes and tanks located
^
outside of buildings, discuss the effect of failure of these items and any potential flooding of safety-related structures, systems and components. Provide a similar discussion for l' nontornado. protected vessels, tanks and piping.
, BEE 19EEE i
The failure of non-Seismic Category I and non-tornado protected tanks, vessels, and major pipes located outside of buildingt (Table 410.7-1) will not adversely affect safety-related
- structures, syatoms e.nd components by flooding, as discussed
,_ below: .
t t Failure of Tanks
, The locations of tanks in the yard area are shown on Figure !
1.2-1. Failure of the condensate storage tank, located on the i south side of the power block (Table 410.7-1, Item 1), will not
- cause flooding. Any spillage due to failure of this tank will be i contained within a reinforced concrete dike designed to be <
Seismic Category I, ,aus discussed in Section 3.8.4.1.6.
- The tanks located on the north and west sides of the power clock 7 i (Table 410.7-1, Items 2 through 7) do not have Seismic Category I i i dikes around them. Failure of these tanks could cause local
- flooding. However, this flooding would not adversely affect i safety-related facilities for the following reasons:
i '
j a. The storm drainage. system in this area will drain the spillage to the Delaware River before it reaches the '
l g &fA"y power plant complex.
i ;
1 b. Seismic Category I electrical cables and duct banks i i located in the vicinity of these tanks are protected !
i aga' inst flooding,. as discussed in the response to !
! Question 410.8.
I Failure of Coolino Tower Basin Wall (Table 410.7-1, Item 8) 4 The failure of the cooling tower basin wall would not adversely I
affect safety-related structures, systems and components, as discussed below:
The operating water level within the cooling tower basin is
] elevation 102.5 feet. The slabs and walls are conservatively
! designed for 3 feet of freeboard, allowing tne water level to .
{
ri'se to' elevation 105.5 feet. The grade around the basin well is t
l* osatorms Irnt 88Q*C 410.7-1 Amendment 2 l
t
, Insert A" i
[ a. Any spillage will be conveyed to the Delaware River l ny means of overland surface runof f without adversely affecting any safety-related structures, systems or
- components by flooding. There is a clear path to the river from the building which will assure that any surface water will not enter the building. In addition, l, storm drainage is provided to facilitate conveyance of runoff to the river which will furtner minimize tne l potential for any local ponding. -
I
, I l t i
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' osum orow nun ,J Fa.<
1 I
HCGS FSAk 10/83 1
',j at elevation 104.5 which is 2 feet above the operating water level in the basin. i The worst case flooding could result from the unlikely " wash-off" e of the soil on the south side of the tower.. For this case, the run-off would be dispersed and intercepted by the storm drainage system before it could. reach the power block are The Seismic Category I duct banks located between the intake tructure and i
the power block will not be affected as they are not located in the flow path of the water.
1 Failure of Circulatina Water.P_ines (Table 410.7, Item 9)
Failure of these pipes within the yard area between the cooling tower basin and the turbine building will cause flooding of this area. Water from the damaged pipes will erode the soil cover and l flood the yard. No Seismic Category I equipment or components
- are located in this area of possible erosion. The storm drainage system would eventually drain the water to the Delaware Rivey ,
i the most severe case, all the water from the cooling tower-
! basin could drain through the damaged pipe into the yard area between the circulating water pumphouse and the turbine building.
! This could cause flooding of the lower level of the turbine i building. However, safety-related systems and components would
- not.be damaged, as discussed in the response to Question 410.115.
l l
Failure of Maior Yard Picina j
- Failure of any of the pipes identified in Table 410.7-1, Items 10 to 14, may cause local flooding. However, the intensity and volume of water discharge from any of these pipes is less than, that of the circulating water pipes discussed above and would not j cause damage to caused by fallur,any. safety-related e of these facilities. in Soil pipes is discussed erosion the response to Question 410.64.
l i
1 l
0e the ajo.4er t,0ca./d S /w c> ver- lanc/ fa +h e
! ~b eJa usa >-e R,'ver o. s dt.s e.u s .s ed (se +anks 1
l (Z hem s ,e thru 7) i l'
l'
(
ostR 075 17 8 *
! 410.7-2 Amendment 2 I
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I l
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. TABLE 410.7-1 YARD TAtBES A84D MAJOR PIPItaC f8 ION-SEISMIC) 18/03 Item Capacity B80 Tank of Pipe Description Type et Sarmado -
or Flow Locatior Containeemt Protection 1 Coe.Jensate Storage Tank See, Gee gal South of pdeser plant Sei- alc Cat. Home comples I men.aforced Conc. tse11s 2 Fire Water Tanks (2) 340,844 gal em teorth of power plant comples alone IIone 3 Asphalt Storage Tank S.See gal teorth of power plant comptes Concrete unit Ibene .
saae=ry vr!!s 4 Fuel Oil Day Tank 10,000 gal teorth of power pleat comples Reinforced Ilone j Conc. teolls 5 Chemical Treatment Tanks 2 Sodium mypochlorite 30,000 get ea 1 Sulfuric Acid teorth of power pleet comptes metaforced mana.
26, Gee get tk.rth of penser plan:. comptes Concrete Isome 2 Sodiesa 3;ypochlorite 15, Gee get ea toest of power plant ccytes tea 13a Isone 6 Sewage Treatment Plant 1 Equalisation Taak 28,848 gal Morth of power plant comptes eenried 2 Treatment Tanks Isone e,644 gal ea aborth of poseer plant comptes geseled teone 1 Treatment Taak 35,448 ga! Isorth of power plant comptes Earth Iserm IIone 7 Fuel Oil Storage Tank 1,844,830 gal teorth of posser plant comple,s Earth dike IIcee e Cooling Toiser Basia 6.Sec ees gal sIorth of power plant comples meinforced Ilone Canc. esall 9 144*3 Circulating teater Pressure 552.048 gym Between cooling tower and IIndergrassed So!! cover Pipes (2) tastbine basilding y le 48*$ stakeup hter Pressure Pipe 38,66e gym Beactor bes11 ding to cooling
. thedergreesad Soil cover toeser 11 Og 11 34*3 makeup teater Pressure Pipe 21.068 gpa Beactor has11 ding to cooling GInderground Soil cover tower .
12 44*f Bloesecase teater Gravity Pipe 15,404 gym Cooling tobr to Belaneare g Rive r GIndergseund Soit cover E+
13 36*S Deicing hter Pressure Pipe 12,844 gym Circestating water pipe to tendergressed g , natake ser.ctur.
So&& cover 8 14 126,Firev.tertoep 2.5.e gym Around piaat .emo!.s
" ih.d.6,co ad son! .o.or E
G6/3 m .
_ - _ _ _ . - , . _ , ._. . _ . , _ , , , , , - - . __ - ,, , _ , . _ _ _ _ .. _ . _ _ < _ .-,- _ . _ , , , _ _ . , , . _ ,, . . ~ .
l HCGS l DSER Open Item No.-28G (Section 3.4.1)
FLOOD PROTECTION The applicant has not provided the information requested concerning RG 1.102, position C.2, and therefore no conclusions can be made concerning compliances at this time.
P RESPONSE !
For the information requested above see response to Question 410.4.
s 1
M P84 126/04 5-mw
i HCGS DSER Open Item No. 110 A & B (Section 4.6)
FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS
-The control rod drive system was reviewed in accordance with Section 4.6 of the Standard Review Plan (SRP), NUREG-0800.'
An audit review of each of the areas listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" por-tion of the SRP section. Conformance with the acceptance criteria formed the basis for our evaluation of the control rod drive system with respect to the applicable regulations of 10 CFR 50.
The applicant has not addressed the recommendations of NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping."
The design does not utilize a CRDS return line to the reac-tor pressure vessel. In accordance with NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drives Return Line Nozzle Cracking," dated November 1980, equalizing valves are installed between the cooling water header and exhaust water header, the flow stabilizer loop is routed to the cooling water header, and both the exhaust header and flow stablizer loop are stainless steel piping.
We have reviewed the extent of conformance of the Scram Discharge Volume (SDV) design with the NRC generic study, "BWR Scram Discharge System Safety Evaluation," dated December 1, 1980. The design provides two separate SDV headers, with an integral instrumented volume (IV) at the end of each header, thus providing close hydraulic coupling.
Each IV has redundant and diverse level instrumentation (float sensing and pressure sensing) for the scram function attached directly to the IV. Vent and drain lines are com-pletely separated and contain redundant vent and drain val-ves with position indication provided in the main control room. With respect to Design Criterion 8, the applicant stated that the "SDV Piping is continuously sloped trom its high point to its low point." In order to provide a re-sponse to Design Criterion 8, the applicant must provide a description of the SDV from the beginning of the SDV to the IV drain. The description should include piping geometry (i.e., pitch, line size, orientation).
M P84 126/OS 1-mw I
2 .
DSER Open Item No. 110 A & B (Section 4.6) (Continued)
Except for Design Criterion 8, we conclude that the design of the SDV fully meets the requirements of the above referenced NRC generic SER and is therefore acceptable.
Additionally, the above-described design of the SDV satisfies LRG-II, Item 1-ASB, "BWR Scram Discharge Volume Modifications."
Based on our review, we conclude that the functional design of the reactivity control system meets the requirements of ,
General Design Criteria 23, 25, 26, 27, 28, and 29 with respect to demonstrating the ability to reliably control reactivity changes under normal operation, anticipated operational occurrences and accident conditions including single failures, and the guidelines of NUREG-0619 and is, therefore, acceptable. We cannot conclude compliance with the guidelines of NUREG-0803 and the generic document dated December 1, 1980. The functional design of the reactivity control sytem does not meet the applicable acceptance criteria of SRP 4.6. We will report resolution of these items in a supplement to this SER.
RESPONSE
a The concerns of NUREG-0803 are addressed in response to i 0410.26. .
FSAR Section 4.6.1.2.4.2(f) has been revised to include a description of the SDV piping.
I t
i i ,
M P84 126/05 2-mw t
l
, . _ _ _ -._.. .. -- _ , - _ _ _ . - _ - _ _ _ . _ _ _ . . _ . - - _ . - . . , _ _ _ , ~ . - _ , . _ _ _ . . - - -
- l Differential pressure between the reactor vessel jroom.and the cooling water header is indicated'in the main control room'. Although the drives can function without cooling water, seal life is shortened by long-term exposure to reactor temperatures. The temperature of i
' each drive is indicated and recorded, and excessive temperatures are annunciated in the main control room. i
- e. Exhaust water header - The exhaust / water header connects to each HCU and provides a low pressure plenum and discharge path for the fluid expelled from the drives during control rod insert and withdraw operations. The fluid injected into the exhaust water header during rod movements is discharged back up to .
the RPV via reverse flow through the insert exhaust directional solenoid valves of adjoining HCUs. The pressure in the exhaust water header is, therefore, j
maintained at essentially reactor pressure. To ensure i that the pressure in the exhaust water header is maintained near reactor pressure during the period of j
vessel pressurization, redundant pressure equalizing i valves connect the exhaust water header to the cooling water header.
! la i nch d ioe *W
- f. Scram discharge volume - The scram discharge volume (SDV) consists of two sets o header piping, each of i
l which connedts to one-half o the HCUs and drains into Each set la inch d'a#7 scram discharge instrument volume (SDIV).
of header piping is sized to receive and contain all i
the water discharged by one-half of the drives during a scram, independent of the SDIV. ,
i r?)c hasde! P!/ I *s *P * * + *
- lo u)as Poin 6 W* % A i in inim u on p i+cA o f '/s p e.r fea -t shown an As urt % ~/ C ,
The SDIV for each header set is directly connected to l the low point of the header piping. The large-diameter l
pipe of each SDIV thus serves as a , vertical extension *f i
of the SDV. A a pl ping e,ca,n se; tion a.t -e:h e, ko tics j He a b sV Pro v* des alt-a
- n y:e o P %s s s si v omd s b V v's a lop ed d >od h tine s w s c. m n.e m n 'Is " pe.r 9ea+ s loPt .
i
- During normal plant operation, the SDV is empty and is i
vented to the atmosphere through its open vent and drain valves. When a scram occurs, upon a signal from the safety circuit, these vent and drain valves are closed to conserve rgactor water. Redundant vent and drain valves are provided to ensure against loss of reactor coolant from the SDV following a scram. Lights in the main control room indicate the position of these valves.
! osER OPEN MEM // O 4.6-13
HCGS FSAR 12/83 QUESTION 410.26 (SECTION 4.6)
Providetheinformationrequestedinourgenericletter81-4 <
dated August 31, 1981, regarding NUREG-0803, " Generic Safety '
Evaluation Report Regarding Integrity of SWR Scram System Piping."
RESPONSI ,
HCGS is participating in the BWROG activities related to the scram discharge pipe integrity. ~The BWROG's final response to the NRC is being prepared for NRC review and approval. N
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A HCG 5 P lant . spec.;fic response wit / b e provsded 8WR06 v;4 h;n 40 day s o S NR c. a.ccep fx.ne e af the C,'x e s, e' F pas;Hcn . M cgs w:>/ im p /ement rega.ppea uieedva.I o 9 P h 4 N R C. Fe view o.nd a n y, o.r-; sing Scom 73 pp oG S d rnl H a.l s h Me end of Me.next a.(4 se ^l LC. Lf F t'Q Va l -
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DSER OPEN ITEM //C 410.26-1 Amendment 3
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NCGS DSER Osen Item No.112 (DetR Section 5.2.5)
REACTOR C00 TANT PRESSURE BOUNDARY LEAKAGE DETECTION I Provisions have not been made to monitor all of the systems con-nected, as identified in Table 1 of Section 5.2.5 of the Standard Review Plan, to the RCPS for monitoriq and alarming intersystem leakage by using radioactivity and dif cerential flow monitors.
Specifically, the applicant has not provided monitoring capability for intersystem leakage for the pafety injection system (high and low pressure systems), residual heat removal system (inlet and discharge), reactor core. isolation cooling system, and the steam side of the high pressure coolant injection system. Thus, the guidelines of Regulatory Guide 1.45, Position C.4 are not met.
Each leakage detection system has indicators and alarms either in the control room or at .the local panels.
- The monitor signals pro-
, vided to the control room are generated through the plant computer system with no unprocessed signals available to the operators and
, no procedures to direct the operators where or how to obtain the information if the control roas indications are lost. The appli-cant should provide a discussion of the capability to maintain suf ficient onsite manpower at all times to man all locaf panels 100% of the time (this is in addition to the manpower requirements discussed in Section 9.5 of this SER) when the information is not available in the control room, to provide a seismic Category I communication system between the control room and all local panels, to provide procedures to guide the personnel at the local panels, and to propose a Technical Specification requiring the manning of the local panels when the control indications are not available.
Thus, the guidelines of Regulatory Guide 1.45, Position C.7 is not met.
The applicant does not have a sump flow monitoring system, an airborne particulate radioactivity monitoring system, and a seismic Category I monitoring system and therefore does not meet the guidelines of Positions C.3 and C.6 of Regulatory Guide 1.45.
As recommended by Regulatory Guide 1.45, at least three separate detection methods should be employed and two of these methods are to be (1) sump level and flow monitoring, and ( 2) airbone parti-culate radioactivity monitoring. We will require the applicant to provide sump flow monitoring, in addition to the existing sump level monitoring stated in the FSAR, in order to meet the first part of Position C.3. The applicant has not provided an air-borne particulate radioactivity monitoring system. Not having an airborne particulate radioactivity monitoring system is accept-able provided that the applicant provides an alternate monitortng system which ueets the qualifications of the airborne particulate system. The applicant has not proposed any alternate at this t ime . In conformance with Regulatory Guide 1.45, Position C.3, the third method of detecting leakage is the monitoring of drywell cooler condensate flows. Regulatory Guide 1.45, Position C.6, requires the airborne particulate monitoring system to be seismic Category I. The applicant must provide a seismic Category I airborne radioactivity monitoring system or a seismic Category I acceptable alternate leakage monitoring system.
112-1
O
. NCGS ,
DSER Open Item No.112 (Cont'd)
) The applicant has not provided information concerning the systems testing and calibration frequency and capability during power operation of the plant in accordance with Regulatory Guide 1.45, Position C.S. The applicant has committed to specifying the .
maximum allowable identified and unidentified leakage rates as 25 gym and 5 gpm, respectively, in the technical specifications.
Thus, the guidelines of Regulatory Guide 1.45, Position C.9, are met. Until the applicant provides the information stated above on the leakage detection systems, we cannot make any con-clusions as to the acceptability of the systems. We will report resolution of this item in a supplement to this SER.
RESPONSE
For the HCGS definition of intersystem leakage, refer to Sec-tion 1.14.1.7. .
For a discussion on leak detection for the four systems noted, refer to the following sections:
- 1. Safety Injection System (high and low pressure systems) -
Section 5.2.5. 2.1 (o ) .
- 2. Residual Heat Remeval System (inlet and discharge) -
} Section 5.2.5.2.1 (o ) .
- 3. Reactor Core Isolation Cooling System - Section 5.2.5.2.1 (m) ,
- 4. High Pressure Coolant Injection System (steam side ) -
Section 5. 2. 5-2.1 ( 1) .
Section 5.2.5.2 has been revised to indicate that the drywell floor and equipment drain sump leakage rate indications are class lE and are located on main control room panel 10C604.
^
Sections 1.8.1.45 and 5.2.5.2 have been revised to address the concerns of positions C.3 and C.6 of Regulatory Guide 1.45.
Section 5.1.5.2 has been revised to identify that the drywell equipment and floor drain sump level monitoring instrumentation is seismic Category I.
! Sections 5.2.5.9 and 11.5.2.2.15 have been revised to provide information concerning testability. ,
1 e
F65(4) 112-2 l
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l ECGS FSAR 10/03
,l6(flf t'"t (f'
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)
See Section 5.2.3 and 6.1 for further discussion and Section 1.8.2 for the NSSS assessment of this Regulatory Guide.
1.8.1.45 Conformance to Reculatory Guide 1.45, Revision 0 May
' . 1973: Reactor Coolant Pressure Boundary Leakane Detection Systeeg HCGS is designed to comply with Regulatory Guide 1.45, with the exceptions, clarifications, and amplifications discussed below.
Paragraph C.3 of Regulatory Guide 1.45 requires that three methods of leak detection be provided. NCGS does not employ an airborne part.iculate radioactivity monitor due to uncertainties in detecting 1 gpa of RCPB leakage in I hour. The uncertainties that affect the reliability, sensitivity, and response times of radiation monitors, especially todine and particulate monitors, !
are discussed below.
The amount of activity becoming airborne following a 1-gpa leakage from the RCPS varies, depending upon the leak location and the coolant temperature and pressure, which affect the ,
flashing fraction and partition factor for todines and i
particulates. Thus, an airborne concentration cannot be
! correlated to a quantity of leakage without knowing the source of the leakage.
i Coolant concentrations during operation can vary by as much as several orders of magnitude within several hours. These effects i
are mainly due to spiking during power transients or changes in i the use of the reactor. water cleanup (RWCU) system. An increase in the coolant concentrations can give increased containment concentrations when no increase in unidentified leakage occurs.
l' Not ali activity is from unidentified leakage. Changes in other ,
I sources result in changes in the containment airborne '
concentrations. For example, identified leakage is piped to the drywell ec:uipment drain sump, but all sump and collection drains are ventec to the drywell atmosphere, thereby allowing particulates to escape, causing further measurement uncertainties.
The amount of activity that is detected depends upon the amount of plateout on drywell surfaces prior to reaching the detector intake. The amount of plateout is dependent on uncertain osza onn 1:za //.R 1.0-26 Amendment 2 l
BCGS FSAR 10/03 .
i
);
quantities, such as location of the leak, distance from the "
l detectors, and the pathway to the detector.
1 1 Furthermore, under normal operating conditions a radiation-free 8
- background does not esist. There is a buildup of activity i
concentration due to both identified and unidentified leakage.
At higia equilibrium activity levels, a small change in activity i level due to a small leak is hard to detect in the desired time interval ,
3 ,
i 4
Although particulate monitors are available with sensitivities covering concentrations espected in the drywell, previously '
discussed uncertainties under operating conditions coupled with '
any calibration and setpoint uncertainties make particulat.e
) monitors a less reliable method of leak detection. -
'4 ,
l i
ECGS does employ three separate and diverse leak detection methods.
i The RCPS leak detection s tem consists of '
i sal *We. cArakar I
- 4,wi qmpua j a. Mrywell floor f drain sump level monitors (lN LIE,t/ OF A SGlSMIC -
j # ;
raft 608N I AIR, PAa,71CutME pefEc10N .sy.3 REM).
j b. A drywell cooler condensate flow monitor c.
A noble gas monitor (IM Lit.W GF AN MR, f42.TicliLAYE CE1TsC10N SY.57EM'
-- /AfSSA7~~ 1) ~
'l Paragraphs C.2 and 5 require that the leakage monitors be able to detect an increase in leakage of I gym in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. !
monitor can detect concentrations as low as 10-* eC1/ce, theThe noble gas Cinimum the primary activity system concentration coolant. Nowever, espected in the drywell based on !
, an increase in 1 gym i leakage within an hour may be difficult to detect due to high .
) equilibrium and buildup of activity levels radiation.
background for noble gases (10-* to 10-* eC1/cc) !
! The noble gas monitor is capable of detecting leaks of approutmately 10 gym and does so l
' very quickly due to the high diffusion rates of the noble gases, j t 2
! The drywell floor drain sump level monitor and the drywell cooler
! c ndensate monitor can detect fluid flows of 1 gym in I hour. ;
I Nowever, fluid flow is not always a direct indication of RCPS :
- 1cakage because chamber and the drywell. of free communication between the suppression
! The drywell atmosphere is not ,
i drywell coolers.necessarily saturated due to the water vapor removal by the
- i Not water can evaporate from the torus and ' l t
e i t
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enter the drywell. The water will condense and register on the drywell cooler condensate monitor. The condensate drains into the drywell floor drain. sump and will register on the sump level monitor. Therefore, during times of suppression pool transients, such as from heat up from main steam safety / relief valve (SRV) or .
! RPCI system testing, evaporation from the suppression chamber i will obscure values of RCP5 leakage.
Pos requires that the leakage detection sys capable of per heir functions after c event that does not require plant s a detection system is capable of operatin opera is earthquake (OBE) and a DBA p level monitor is used o ulatory
.45 and 1.97 purposes. +-__
'PUsi C.6 also suggests that at least one RLra Aeax aec method s o in functional after an SSE. Thi ity does not exist in design. The the RCPB leak detection system is to monito rity of the RCPB so that if there are any changes ant can ly shut down.
Since the plant ut down after an SSE, the etection system have to remain functional after an SSE, t t% .
s Position C.7 requires that indicators and alarms for each leakage detection system should be provided in the main control room.
Procedures for converting;various indications to a common leakage equivalent should be available to the operators. The calibration of the indicators should account for needed independent variables. .
Position C.7 is further clarified by Standard Review Plan
~
Section 5.2.5, III.5 which requires that if monitoring is computerized, backup procedures should be available to the
- operator.
- lAlSE/2 7"-* A- -
drywell-sumps-and drywell air coolers leakage moni 'g systems, and level change is electronically ttted l
from level sens n' local radiati~on proc LRP) which processes these' signa in turn t a processed data for indication"~nd~ a alarms, leve alculated flow rates to the central radiation proc P) in omputer room. Data in a keyboard the CRP is avail ' o the operator on the printer erminal CRT and/or annunciated in n control l W ~
l
_ I 1.8-28 Amendment 2 c ,
- l A)S ERT~ /- -
T ka orpeu ci,c <xol< ve o&,-;
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ince r es 4he' leakage signal's are processed locally with capability
\
for I, deal readout, procedures for converting various indications to a common leakage equivalent are not provided to the operators,A n a r c(c.
Jackup procedures ;;: : 9 provided to the operatoric.;;
<a-a < 6'- ---6-ai;;; *- ~~- e
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es that leakage detection s equipped Position .
calibration during to readily permit te or operabil plant operation. This capa not provided on RCPB leak nside the containment, because detection instrumen e calibratio sting cannot be performed n ent during reactor operation.
For further discussion of the RCPB leak detection system, see Section 5.2.5.
! 1.8.1.46 Conformance to Reculatory Guide 1.46, Revision 0, May 1973: Protection Against Pipe Whip Inside Containment The criteria set forth in Regulatory Guide 1.46 are design bases for HCGS. See Section 3.6.2 for further discussion of pipe break design and Section 1.8.2 for the NSSS assessment of this Regulatory Guide.
osER OPEN ITEM //g 1.8-28a Amendment 2 i
_ _ . - _ _ _ . _ _ - - _ . . - _ _ _ . . _ . . . _ _ _ _ _ _ _ _ _ _ _ . . _ . . . . _ _ - _ . _ _ . _ _ _ _ . _ . _ _ . . _ _ _ _ _ _ . _ _ . . , _ _ _ . - _ _ ~ . . _ . _ _ . .
l
-/NSE/2i B As Ascrikc/ in 5'ec% S~ z S~z, Afsplays of lUor drain sun 1plowis drywell aguijnuf aad nos ; g-( wk;<k. a ce n o+ dapw Aw on 4x
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DSER OPEN ITEM //g
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the RWCU pump heat exchangers and the reactor recirculation pump )
seal and jacket cooling heat exchangers. The RACS sensor l
monitors radiation emanating from a continuously flowing RACS ,
! water sample which is taken at a point downstream of the RACS l pumps. \
- i High radiation in the SACS water or the RACS water indicates intersystem leakage. The affected sensor and its associated monitoring channel will activate an alarm in the main control room when the radiation exceeds a predetermined limit. No isolation trip functions are performed by these channels.
- \
These radiation channels are part of the process radiation monitoring system described in Section 11.5.
High levels in the SACS or RACS head tanks may also indicate intersystem leakages from the sources given above. High level in either head tank will activate an alarm in the main control room.
5.2.5.2 Leak Detection Instrumentation and Monitoring ,
5.2.5.2.1 Leak Detection Instrumentation and Monitoring Inside Primary Containment
- a. Floor drain sump level and flow - The normal design leakage collected in the floor drain sump includes unidentified leakage from the control rod drives (CRDs), valve flange leakage, component cooling water, service water, air cooler drains, and any leakage not connected to the equipment drain sump.
~ /A M M T C -
1 transmitter is used in the drywell floo n sumps a fed into a local microproces . level change in the will be convert low rate by the processor. Abno les ates are alarmed in the main control room on in excess of background lenk uld indicat nerease in reactor leakage from an uniden source in ex 1 gym within 1 hour.
- b. Equipment drain sump level and flow - The equipment drain sump collects only identified leakage and valve stem packing leakoff collectively. This sump receives .
DSER OPEN I m / Q 5.2-46 Amendment 2
/A)GCne7~ C -
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+ piped drainage from pump seal leakoff and reactor vessel head flange vent drainage. The equipment drain sump instrumentation is identical to the floor drain sump instrumentation.
~
- c. Drywell air cooler condensate drain flow - Condensate from the drywell air cooler is routed to the floor drain sump.
ght e in each of two drain lines frand is trapped drywell a s drains into by a closing soleno ontrolled by a local microprocess rising n the drain line is evel transmitter that s signal to sen
. Flow in anck e4 % bo du'% gars k(owA. N ci dr We ll' C.onlerS hur coo (erS fSeMSoc r kd r) is vu cui-wed (o A. Mlo2
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i BCGS FSAR 8/83 to differentiate between. identified and unidentified leakage is discussed in Sections 5.2.5.4, 5.2.5.5, and 7.6.
\
5.2.5.7 Sensitivity and Operability Tests
- Sensitivity, including sensitivity testing and response time of the leak detection system, and the criteria for shutdown if leakage limits are exceeded, is covered in Section 7.6. ,
i Testability of the leakage detection system is contained in Section 7.6.
5.2.5.8 Safety Interfaces The Balan'ce of Plant-GE Nuclear Steam Supply System (NSSS) safety interfaces for the leak detection system are the signals from the monitored balance of the plant equipment and systems that are part of the nuclear system process barrier, and associated wiring '
and cable lying outside the NSSS equipment.'
C. .
Testino and Calibration 5.2.5.9
- /A/SEf_r E 5.2.5.10 Conformance to Raoulatory Guide 1.45
! For a discussion of compliance with Regulatory Guide 1.45, see Section 1.8.1.45. .
5.2.5.11 SRP Rule Review SRP 5.2.5 acceptance criterion II.1 requires that leak detection
- system integrity must be maintained following an earthquake, as per GDC2. This is met through Regulatory Guide 1.29 positions C-1 and C-2.
f*
DSER OPEN ITEM //2 5.2-59 Amendment 1
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information about the HEPA and charcoal filter efficiency and g
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condition. J t
11.5.2.2.12 Radwaste Area Exhaust Kadiation Monitoring System The RAE RMS is located in the exhaust duct for radwaste area compartments in which there is equipment that has a possibility )
of releasing airborne radioactive materials (Refer to Figure 11.5-1). The RAE RMS is upstream of the filters and will )
be exposed to higher concentrations than the RES RMS, thus l allowing earlier detection of any problems in the radwaste areas )
of the auxiliary building. The RAE RMS has the same components and functions as the RBVSE RMS described in Section 11.5.2.2.8. l
- l 11.5.2.2.13 Gaseous Radwaste Area Exhaust Radiation Monitoring System l
\
The gaseous radwaste area exhaust (GRAE) RMS is located in the exhaust duct for the recombiner compartments (Refer to Figure 11.5-1). This allows earlier detection of airborne
' radioactivt materials than is possible by downstream monitors where the concentrations are more diluted. -The GRAE RMS has the )
same components and functions as the RBVSE RMS described in Section 11.5.2.2.8. There are no filters upstream of the location.
11.5.2.2.14 Technical Support Center Ventilation Radiation Monitoring System i
The technical support center ventilation (TSCV) RMS is located in the inlet plenum for the technical support center (Refer to Figure 11.5-1) The purpose of the TSCV RMS is to detect radioactive materials ih the inlet air. The TSCV RMS has the same components as the RBVSE RMS described in Section 11.5.2.2.8.
If the concentration exceeds the trip setpoint, an alarm at the CRP alerts the operator to manually transfer from the normal air supply to an emergency recirculation and filtration mode. ,
I l 11.5.2.2.15 Drywell Leak Detection Radiation Monitoring System The drywell leak detection (DLD) RMS monitors the gaseous radioactive materials in the drywell (Refer to Figure 11.5-3).
The design objective of this system is to monitor reactor coolant
)
DSER OPEN ITEM //j;[ 11.5-18 Amendment 1
pressure boundary (RCPB) leakage in accordance with Regulatory Guide 1.45. Conformance to Regulatory Guide 1.45 is discussed in Section 1.8. The capability to do so declines as the normal in-
' containment background of gaseous radioactive materials increases because of the accumulation from identified leaks. An air sample l
is extcacted and returned through penetrations that are isolated
- by the PCIS described in Section 7.3.1.1.5. The DLD RMS components are one inlet and one outlet stub on the east side of the drywell, penetrations, and isolation valves. There is also a shield sample chamber, a beta scintillation detector, and an LRP.
The high-high alarm indicates excessive leakage from the RCPB.
The DLD RMS is seismically qualified to operate under conditions
! during which the reactor is operated. The functional requirements and descriptions of other leak detection equipment are discussed in Sections 5.2.5 and 7.6.1.3. Provision for a grab sample is included.
- msce r F --
11.5.2.2.16 Reactor Auriliaries Cooling System Radiation Monitoring System The reactor auxiliaries cooling system (RACS) RMS monitors a sample extracted from the RACS (Refer to Figure 11.5-1). The ,
f" RACS RMS has the same components as the liquid radwaste RMS. The high-high alarm indicates leakage into the RACS from the heat l
(- exchangers that are serviced by the RACS.
l 11.5.2.2.17 Safety Auxiliaries Cooling System Radiation Monitoring System The safety auxiliaries cooling system (SACS) RMS has two
- monitors, A and B, one for each of the two SACS loops (Refer to Figure 11.5-1). The SACS RMS monitor samples extracted from the SACS. The SACS RMS has the liquid radwaste RMS.
sample chambers are part of the SACS pressuredary boun,The and.areSACS RMS seismically qualified. The high-high alarm indicates leakage into the SACS heat exchangers from the safety auxiliaries served by the safety auxiliaries cooling system.
11.5.2.2.18 Heating Steam Condensate, Waste Radiation
> Monitoring System The heating steam condensate, waste (HSCW) RMS monitors a sample of the condensate flow from the liquid waste management system (Refer to Figure 11.2-4). The high-high alarm /crip indicates
- both leakage of radioactive materials from one or both of the DSER OPEN ITEM //g 11.5-19 i
- - - - . - _ _ _ _ _ - - _ - . - - . _ __A
b- - mse2 Y t .
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drauul calibr Rou wiJ/ Le dw durixp tafue/&g outny s.
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HCGS ,
1 DSER Ooen Item No. 121 (Section 6.2.1.3.3)
USE OF NUREG-0588 For.the drywell, the limiting accident is a small-size break that does not result in reactor depressureization due to either loss of reactor coolant or automatic operation of the ECCS equipment. For this accident, the operators are aler-ted by a high'drywell pressure signal and reactor scram, and
'it is assumed that they respond with an orderly shutdown of the reactor, which takes about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (i.e., the reactor coolant system is depressurized using the main condenser while limiting the reactor cooldown rate to 100*F/hr). The applicant has, therefore, assumed that there is a blowdown of. reactor steam for the assumed 6-hour cooldown period.
Because the worst combination of primary system pressure and drywell pressure produces a maximum superheat temperature of 340*F from the escaping steam for_drywell design purposes-and for the environmental qualification of safety-related equipment located in the drywell, the applicant has assumed a maximum drywell temperature of 340*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.- We will require the applicant to comment on whether NUREG-0588 is being used for the temperature profile beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
RESPONSE ,
In place of the temperature profile outlined in NUREG-0588, the temperature profile shown in Figure 3.11-4 of Section 3-11, is used for the environmental qualification of Class 1E equipment in the drywell as indicated in Section 6.2.1.1.2.6. This figure also shows the temperature used in the drywell beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the environmental qualification of Class lE equipment.
8 M P84 126/05 3-mw l
1
, ,1
-DSER dpen Item No. 122 (DSER Section 6. 2.1. 3. 3. )
TEMPERATURE PROFILE The applicant has not provided the temperature profile to be used for environmental qualification of any safety-related equipment located in the suppression chamber. We will require that the applicant provide us with this information, and will report on this matter in a supplement to this SER.
RESPONSE
The safety-related components in the suppression chamber are the suppression pool-to-drywell vacuum breakers and Class 1E RTDs with associated cables. The RTDs and associated cables are qualified for drywell temperature. The vacuum brea'Rers contain no 1E controls which would have to be qualified for post-LOCA suppression chamber temperature.
The vacuum breakers mechanical components are included in the program for qualification of mechanical equipment in harsh
< environments which is discussed in Section 3.11.2.6. .
l 1
1 HCGS l
DSER Open Item No. 128 (Section 6.2.2)
AIR INGESTION Ingestion potential has been. extensively studied via full scale experiments, and BWR RHR suction / strainer geometries have been tested (see NUREG/CR-2772) . Experimental results show that if the Froude (Fr) number is less than 0.8 at.the intake, air ingestion is zero. We will require the appli-cant to comment on whether or not air ingestion poses a problem at HCGS.
RESPONSE
The MCGS RHR core spray, and HPCI suction strainer / piping geometries are such that the Froude number is less than O.8 for all stre.iners. Therefore, air ingestion is not a con-cern for the HCGS design. For further discussion see Section 1.14.1.12 and revised section 6.3.2.2.5.
(
l M P84 126/04 3-mw 1 l
[" I 1.14.-l.10.6 Response (LRG II/2-RSB(d))
As indicated by FSAR Section 5.4.6.2.4(f), water hammer r the RCIC system which is comparable to
.that protection provided is provided for the @CS EC injection systems.
1.14.1.11 Adecuate SRV Fluid Flow, LRG I/RSB-8 1.14.1.11.1 Issue The applicant must perform tests to show that flow through the safety relief valves is adequate to provide the necessary fluid relief required consistant with the analysis reported in Section 15.2.9 of the FSAR.
- 1.14.1.11.2 Response See response to LRG Issue No. 5, Section 1.14.1.5.
1.14.1.12 Provisions to Preclude Vortex Formation, LRG II/7-RSB 1.14 1.12.1 Issue f
To preclude vortex formation, air entrainment, and subsequent
. damage to ECCS pumps due to cavitation, it must be shown that-adequate margin exists between the minimum suppression pool level and the depth of submergence of the ECCS pump suction strainers.
This can be shown by analysis or by observations during pre-op testing that no vortex is formed.
1.14.1.12.2 Response The ECCS pump suction strainers in the HCGS suppression chamber are provided with a minimum submergence of at least 10 feet, as measured from minimum suppression pool level. This amount of
'submergenss/is exp;;ted to provide sufficient margin to preclude tormacion of vortices the 'h---ee of air ad5"eir crt and secto.3ga indicated by ducing FSAR Section "CCO 6.3.2.2.5, "
[ p;;p :; rstica "ill ha var 4 M ad A"rir.g pr;;per: tion:1 t : ting. t~,
Iv6m=Livu --
i ha.; b e.e.n o.na.ly 2.ed
- 1.14-15 Amendment 1
- DSER OPEN ITEM /E b .
l
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( -
. }
- b. Instrumentation to indicate system performance during i test operations. -
- c. M,otor-operated valves and check valves capable of l manual operation for test purposes l
- d. Shutdown cooling lines taking suction from the 1 recirculation system to permit testing of the RHR l discharge into the RPV.after, normal plant shutdown 1
- e. Drains to leak test the major system valves.
All active LPCI components are capable of individual functional testing during normal plant operation. Except as indicated below, the LPCI control system design provides automat.ic alignment from test to operating mode if system initiation is required. The exceptions are as follows:
- a. Closure of any of the motor-operated pump suction valves in the suction lines from the suppression chamber requires operator action to reopen them.
Indication of the status of these valves is provided in the main control room.
- b. Parts of the system that are bypassed or deliberately rendered inoperative are indicated automatically or manually in the main control room.
6.3.2.2.5 ECCS NPSH Margin and Vortex Formation NPSH calculations for ECCS pumps, such as the calculation in the previcus section, have shown adequate margin to ensure capability of proper pump operation under accident conditions. This capability is verified during preoperational testing. The-aL5euww cf cir aa**=inment and unr*e- f:r;; tion during ECCC pg.7,p ' .
-epos tica is el e v: ificd during prancar=*ier.21 tasting. %s,
/Ma eN y
+
6.3.2.2.6 ECCS Discha'rge Line Fill Network A requirement of the ECCS is that cooling water flow to the RPV be initiated rapidly when the system is called upon to perform DSER OPEN I m / 62 8 ' g,3_25 ;
9 ;
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- w nser$
The geometries of the RHR core spray and HPCI suction strainer and piping in the torus have been evaluated and the resulting Froude numbers are less than 0.8 for all strainers.
Tests heve snown that no air core vortices or air withdrawal are observed for BWR Marx I geometries where tne Frouce number is less than 0.8. Therefore the HCGS design avoids the formation of air core vortices and possible air ingest' ion.
+
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' bS[A, 0)0th / M / a2 O .
4
. JUL A9 *8dD 2 6 75 0 0 HCGS .
DSER Open Item No. 140 (DSER Section 9.1.2)
SPENT FUEL STORAGE Since the applicant's application for an operating license was docketed in 1983, which is af ter the. November 17, 1977 date 4
specified in the SRP, the applicant must provide the results of an analysis which shows that a failure of the liner plate as a result of an SSE will not cause any of the following (1) significant releases of radioactivity due to mechanical damage to the fuel; (2) significant loss-of-water from the pool which could uncover the fuel and lead to release of radioactivity due to heat up; (3) loss of the ability to cool the fuel due to flow blockage caused by a portion of one or more complete section of the liner plate falling on the top of the fuel rackst-(4) damage to safety-related equipment as a result of the pool leakage; and (5) uncontrolled release of significant quantities on radioactive fluids to the environs; in accordance to the Standard Review Plan. These buildings are also designed against flooding and tornado missiles (refer to Section 3.4.1 and 3.5.2 of this SER). We cannot conclude that the reqdirements of General Design Criterion 2, " Design Bases for Protection
, Against Natural Phenomena," and the guidelines of Regulato'ry i Guides 1.13, " Spent Fuel Storage Facility Design Basis,"
Position C.3, 1.29, " Seismic Design Classification," Positions C.1 and C. 2, have been met.
1
.The applicant has not provided the design details of the spent fuel storage racks, the results of an analysis of impacts onto j the racks, the bundle to bundle spacing, the design maximum 4
enrichment (weight percent of U235), a description of calculational methods used for criticality analysis (along with
, the results), a tabulation of the nominal value of K gg of the racks along with the various uncertainties and biase,s considered in the analysis, and a tabulation of the reactivity effect of each of the abnormal accident situations coasidered for our i review. Since credit is taken for gadolinia in the fuel, the applicant must provide a commitment that every fuel bundle will have a specified minimum amount of gadolinia distributed over a specified number of specific fuel pins, for the entire length of the fuel. As an alternative, the applicant can provide the results of the criticality anal *' sis without taking credit for i the gadolinia.
Thus, we cannot conclude that the requirements of General Design
!- Criteria 61, " Fuel Storage and Handling and Radioactivity Control," and 62, " Prevention of Criticality in Fuel Storage and Handling," and the guidelines of Regulatory Guide 1.13, l Positions C.1 and C.4, concerning fuel storage facility design are satisfied.
i i
140-1
- , . _ , , . . , _ . , - . . _ _ . - , . , _ . , _ _ . _ _ , , _ _ , - . _ _ _ . . . . _ . ~ . . . . . _ . . _ . , _ . . _ . . . _ _ . , . _ _
JL is '84 0 2 6 7 5 0 0 -
~
L DSER Open Item No. 140 (Cont'd) l l-Wo.cannot conclude that the spent fuel storage facility is in conformance with the requirements of General Design Criteria 2, L
l 61, and 62 as they relate to protection of the spent fuel against natural phenomena, radiation protection, and prevention of criticality and .the guidelines of Regulatory Guides 1.13, Fositions C.1, C.3, and C.4 and 1. 29, Positions C.1 and C.2, l
L relating to the facility's design basis and seismic
[ c las sif i, cation. The spent fuel storage facility does not meet the acceptance criteria of SRP Section 9.1.2. We will report resolution of this item in a sup,71ement to this SER.
( Additionally, the information provided through Amendment 3 was not sufficient for the staff to complete the evaluation of the compatibility and chemical stability of materials wetted by spent fuel pool water. To complete the review, the following
, information is requested:
(1) Identify and list all materials in the. spent fuel storage pool including the neutron poison material, rack leveling -
feet, and rack frame.
(2) F.rovide test or operating data showing that the neutron poison material will not degrade during the lifetime of the spent fuel storage pool.
I (3) Provide a description of any materials monitoring program I
for the pool. In particular, provide information on the
- frequency of inspection and type of samples used in the
! monitoring program.
j (4) Provide details of the spent fuel racks to show that no i buildup of gases will occur in the cavities containing the poison materials.
i
RESPONSE
The spent fuel pool liner plate was not designed to seismic Category I requirements because SRP 9.1. 2, Revision 2 (March 1979), which first invoked the seismic Category I ,
requirement, was not issued until after the design and procure-ment of the liner plate was complete and fabrication had begun (November 1978) . However, the liner plate was designed to act as a form for the concrete in the spent fuel pool walls. To perform this function a system of channels, wide flanges and angle stiffeners was welded to the back surfaces of the liner and connected to the o.!tside formwork with form ties. Thus, during the concrete placing operation the welds between the stiffeners and the liner were subject to the lateral pressure effects of the wet concrete. This may be considered a ' test' load in that after the concrete sets, the anchoring capability 140-2 w wv -N- w ym e -m---- g vg - h-yn-W-++-vwrw w-g-g-yggmW-sw---a--e m p o e-spe-w et -vyw y 4 7,yem=gy,e--e*6r e* yvg egya egwer-+--rar*T**-WW98-'e*mM
at -s 84 02 6 7 5 0 0 RESPONSE (Cont'd) of the stiffener system in holding the liner plate against seismic loads is at least equal to the form pressure load.- The estimated
. test load during . construction (appgoximately 300 lb/f t2) was lower This construction load than the design value of 690 lb/ft .
induced a correspondingly lower stress in the stiffener-to-liner welds.
< An analysis, performed to evaluate the effect of SSE loads on the liner, shows that the resultant stresses would be insignifi-c cant (approximately 1% of the stresses due to concrete placement) when added to the residual concrete load. SSE induced loads imposed on the floor liner by the spent fuel racks would also be insignificant, and will not cause a liner failure.
Based on the considerable design margin for form pressure load and the acceptable performance of the wall liner plate when sub-jected to this ' test' load, it is concluded that the liner plate is capable of withstanding SSE loads without any loss of function.
Thug, the design of the liner plate satisfies General Design Criteria 2, 61, and 62, Regulatory Guide 1.29, Positions C.1 and C.2, and Regulatory Guide 1.13, Positions C.1 and C.4. Refer to
< Section 9.1.2.5 for additional justification of the non-seismic Category I liner design. For additional information on the design and analysis of the liner plate, refer to Appendix 3F.
- For a discussion of the liner leakage collection system, which permits expedient liner leak detection and measurement, and
! prevents uncontrolled loss of contaminated pool water, refer to Section 9.1.2.2.2.1.
The spent fuel storage facility design meets the intent of 1
Regulatory Guide 1.13 Position C.3, as described in Section '
9.1.4.6 and 9.1.5.6.
r l The spent fuel storage rack design details have been provided in j the response to Questions 281.2 281.13, 410.39 and 410.42. The information requested in Questions 220.15 and 410.38 will be provided by September, 1984. This information will support the
. criticality review and demonstrate that the design satisfies
! General Design Criteria 61 and 62, and Regulatory Guide 1.13 positions C.1 and C.4.
The materials used in the spent fuel storage racks were included in the response to Question 281.13 (Amendment 5).
140-3 l
I
ft s 8402 67 500 i
RESPONSE (Cont'd)
Similar rack designs, with vented Boral poison in stainless steel racks, have been licensed and have proven successful. HCGS's maximum anticipated radiation exposure for the Boral is 5.12 x 1011 rads. Similar Boral specimens have been subjected to accumulated radiation doses up to 7 x 10 11 rads at the University of Michigan's Ford Ractor. These specimens were found to be structurally. sound and neutron attenuation capabilities were not degraded by irradiation.
In order to continually assure the adequacy of the poison material, test coupons are provided for a Boral surveillance program.
Forty-five coupons are installed in high radiation areas of the spent fuel pool. However, because stainless steel spent fuel racks with Boral poison material are already in use in other BWR fuel pools, a Boral surveillance program is not planned at HCGS.
If information from these lead plants indicates any problem with the Boral, a surveillance program can then be initiated.
The speat fuel rack poison cavities are vented to prevent.any buildup of gases. Response to Question 281.13 provides further information on venting.
j i
ll 4
140-4
l 1
HCGS k
lb DSER Open Item No.144 ( DSER Section 9.2.1)
- STATION SERVICE WATER SYSTEM i
The SSWS consists of two redundant piping loops from the Delaware Rivor to the plant. Each loop contains two 50% capacity ' pumps that are powered from ,a Class 1E power supply.
i The system is housed in seismic Category I and tornato-protected structures (see Section 3.5.2 of this SER) . (The applicant has
! not provided documentation to verify that the SSWS is protected from the flood water (including wave ef fects) of the design basis flood . ] Station service water (SSW) piping is buried a minimum
- of 4 f t below grade, which provides adequate protection from missiles. The system is designed to seismic . Category I, Quality Group C requirements. [Thus, we cannot conclude that the
- requirements of General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena," are satisfied.} However, the staf f can conclude that the guidelines of RG 1. 29, " Seismic Design Classification," Positions C.1 and C.2, are satisfi,ed.
The design of the SSWS ensures that system f unction is not los t assuming a single active component failure coincident with a loss of of fsite power. [However , the applicant has not demonstrated the design of the SSWS can provide suf ficient cooling for a safe shutdown af ter a non-mechanistic pipe f ailure (e vent) with the 4 loss of one SSWS pump (single active f ailure ) . Therefore , we i
cannot conclude that the requirements of General Design Criterion
]
44, " Cooling Water," are satisfied.]
q
} The SSW pumps are normally operating. The availability of the
! standby pumps is ensured by periodic functional tests and in-
, spe ctions. The system design also incorporates provisions for
! accessibility to permit inservice inspection as required. [How-ever, the applicant has not specified the frequency of the
! functional testing or inspection. Thus, we cannot conclude that the requirements of General Design Criteria 45, " Inspection of Cooling Water System" and 46, " Testing of Cooling Water System,"
are satisfied.]
[ Based on the above, we cannot conclude that the station service water system meets the requirements of General Design Criteria 2, 44, 45, 'and 46, with respe ct to protection from natural j phenomena, capability for transferring the required heat loads, inservice inspection and functional testing.] However, the staf f i
concludes that the system meets the guidelines of RG 1.29, gositions C.1 and C.2, with respect to the system's seismic classification. [We will report resolution of this item in a supplement' to this SER. The station service water system does l not meet the acceptance criteria of SRP Section 9.2.1.]
i i
144-1 l
1 HCGS . s no
RESPONSE
For information on the protection of the SSWS fr a flood water see the response to DSER Open Item No. 5.
Our response to Question 410.66 completely describes our design with regards to pipe break and loss of a service water pump.(and is not required acc Briefly stated our design does not for redundant treins of a dual-to BTP ASB 3-1, Section B.3.b . ( 3 ),
purpose moderate-energy essential system) consider non-mechanistic of a pipe breaks along with an additional single active f ailure pump.
l i
l 144-2 )
K53/1 l
l l
JR 23 '84 i.t ~t 6 0d ' i d HCGS FSAR 1/84 9.2 1.6 Tests and Inspections , ,
The system is hydrostatically tested prior to the station operation. All active components, e.g., pumps, valves, and controls, are functionally tested prior to startup and periodicall ANO FRiqu2Nay W /NSEAWC0 resrtN9 /$y/NC44/CMD thereaf ter./NL CAprgg E Vyt /G,Tgc a n cat ApfC.jp/CA YMAJ.S. ,
Inservice Inspection and functional testing of the safety-related portions of the system and components will be in accordance with the examination and testing criteria of Articles IWA, IWD, IWP and IWV of Section XI, ASME Code, 1977 Edition and addenda through Summer, 1978.
The specific examination and tests of the system and components will be listed in the Station Inservice Inspection (ISI) and Inservice pump and valve test (ISI) program Administrative .
Procedures.
9.2.1.7 Instrumentation .
Local instrumentation is provided at the equipment location for maintenance, testing, and performance evaluation.
Water levels at each station service water pump bay, and upstream of the intake structure, are monitored in the main control room.
The station service water pump discharge header is equipped with pressure transmitters that provide input to the plant computer.
Two dual element temperature sensors are located at opposite ends of the intake structure inlet. The river temperature displayed in the main control room is an average of these sensors.
9.2.2 SAFETY AND TURBINE AUXILIARIES COOLING SYSTEM The safety and turbine auxiliaries cooling system (STACS) is a closed loop cooling water system consisting of two subsystems: a safety auxiliaries cooling system (SACS) and a turbine auxiliaries cooling system (TACS).
The SACS, which has a safety-related function, is designed to provide cooling water to the engineered safety features .(EST) equipment, including the residual heat removal (RHR) heat exchanger, during normal operation, normal plant shutdown, loss of offsite power (LOP), and a loss-of-coolant accident (LOCA).
DSER OPEN ITEM j 9.2-8 1.mendment 4
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DSER Open Item No.149 (DSER Section 9.3.3)
EQUIPMENT AND FLOOR DRAINAGE SYSTEM The applicant has not provided an acceptable response to our con-cern of flooding due to a rupture of nonseismic Category I piping, vessels, or tanks, or due to theThe failure of a backflow prevention ,
ECCS compartments have seismic device in the drainage system.
Category I water level instrumentation to alarm in the control. room flooding. The on high water level in the event of drain blockage applicant has not provide 3 the basis for not considering flooding l after a safe shutdown. earthquake which results in the worse case for failure of the nonseismic Category I piping and only takeTherefore, credit seismic Category I structures, systems, and components.
we cannot conclude that the "system Design design meets the requirements of Bases for Protection Against 4 general D# sign criteri.s 2, " Environmental and Missile Basis," and heural Phenomena," and 4, " Seismic Design Classifi-the guidelines of Regulatory Guide 1.29, to the failure of the cation ," Positions C.1 and C. 2, with respect failure or in unacceptable drainage system resulting in equipment '
release of radiation due to natural phenomena, missiles, or pipe
! j 4 breaks. _. '
! ( Based on our review, we cannot conclude that adequate protection against flooding of safety-related equipment and areas, and pro-tection against the inadvertent release of potentially rad.nactive liquids to the environment through plant drainage paths is provided.
We cannot, therefore, conclude that the system meets the requice-1 ments of General Design Criteria /'2, 4, and 60, with respect to the need for protection against natural- phencuena, pipe breaks, environ-mental ef fe cts (flooding), and release of radicactive material to the environment, and the guidelines of Regulatory Guide ,1.29, Posi-tions C.1 and C.2, with respect to- seismic clasatification. The l
equipment and floor drain system We does will not meet the acceptance report-resolution of this cri-item teria of SRP Se ction 9.3.3. !
l in a supplement to this SER. '
god
RESPONSE
i j Section 9.3.3.5 has been revised in response to Questions,410.93 "
to addres s the seismic qualificacion of the check valves and the 4
plant saf., shutdown capability following a SSE which results in a failure of the nonseismic Category I components and drain lines.
f s
/
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f F66(3) 149-1 i__ . _ . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ .
- s. '
, , OUESTION 410.91 (SECTION (9.3.3) ,
. ', Demonstrate that a failure of the nonseismic Category I, . ..
- nonsafety-related (EFDS) will not compromise portion of the the equipment capability for and floor safe drain syst. , . . ' ,
shutdown
! because of failure of more than one redundant safety-related l- train due ,to flooding for the following reasons:
s.
i Failure of the EFDS to remove the flood water from an
' enclosure containing safety-related equipment. t I Consider flooding caused by a high energy pipe break, !
moderate energy pipe crack, and rupture of nonseismic Category I piping vessel or tanks; , ;
i -
- b. Backflow in the EFDS due to check valve or other l
. failure causing flooding of one safety-related
- i enclosure from equipment or piping failure outside of ;
j j
this enclosure. a"..,- .
MW- !
j I
. . ~*E [.' E 7'
RESPONSE ,
j The EFDS will not fail to remove flood water from enclosures containing safety-related equipment such that the capability to !
j
- .' / achieve, safe shutdown would be compromised. Complete blockage or failure to pass flow of the EFDS is not considered credible and ;
I .
is not part of the design basis of HCGS. Blockage of a single
" EFDS line would preclude removal of flood water from the ;
l compartment served by that line. The HCGS design provides dedicated drain lines from safety related equipment compartments l l
in the lower elevation of the plant to the sump to preclude cross l flooding from one safety-relaked r,ompartment to another.- Also, !
i to prevent flooding from one safecy-related compartment to ;
another, the walls
) waterticht. SeSen ,e.s.S.s h.as beenfloors and penetrationsi are designed to be !
j W W W e e %=- h *ge 9 m .m W 4e W se.addih6'ml i As discussed above, significant flooding due to the failure of i
) piping, equipment and instrumentation in the reactor building is , '
- not expected. However, in the event that significant quantities j
cf water are conveyed to the sumps at elevation 54 feet, backflow ,
into the ECCS compartments is prevented by the inclusion of a l check valve where the dedicated drain line from each ECCS
[ compartment terminates in the sump. Each ECCS compartment is -
provided with separate drain lines from the compartment to the sump. ;
Thus, failure of any check valve will not result in l flooding of more than 'a one ECCS compartmen* ' - - - - - - - - - " ' - l
! 5;in; ;:r'a---d "--ify the 25!11ty ;f t.% ;.'.;;% ;;1;;; te l cintri, 2 '"---* ~' pr ::rre t:;nf ery :; in:t i dfle;
! f e l l:--:in; !" ' S ; . .. ; .;; te 0 .;ti; . ? ' 0. 00 ) .
l f. MI: b r---l-* 3 M* E- '?": kg. on p.3/5 5 T.'.ie
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i t~e seni & Mdress sel$rm'.c,. pobAk,00n h%g.,. c}1sc)t Yaur sa ud t) assoedel fyliry,
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- BCGS FSAR
- A. 10/83 system (RACS) water'to keep the wastes at their normal o'perating .
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temperature of 140*F. .
l!
9.3.3.4 System Coeration The equipment and floor desinage systess' wastes are selectively collected gravit.
and drain directly to the area collection point byAfter .
pumped to the radwaste collection tanks for processing The sump pumps startby the They appropriate treatment subsystems. automatically when a preset hi ,
stop at a preset low water level.to the drywell floor sump, except for
, pump sump.
seal leakoffs, which are routed to the drywell equipment the turbine building and an alarm is annunciated, the sumpco I
a sample loop before discharge. '
1 i
- The sanitary drainage system collects liquid wastes and entrained l 4
solids discharged by plumbing fixtures, with the exception of l lavatory basins and showers in the personnel decontamination ares ;
1 and conveys them to the sewage treatment plant. i I !
l The stora drainage system collects water from precipitation on l i
enclosure roofs, arsaways, paved and unpaved surfaces, and l l
3 irrigation runoffs outside the buildings, and conveys them to the .
l Delaware River.
i '
j Low volume and oily water wastes from the emergency diesel ,
I generator and chemical regenerant waste from the makeupdj transformer dikes, etc, are collected and pumped to the waste treatment plant in the yard area. These westes are treated to a level that meets Environmental Protection Agency (EPA) and New discharge Jersey Department of Environmental Protection (NJDEP) limits before being discharged into the Delaware River.
9.3.3.5 Safety Evaluation '
The plant drainage systems have no safety-related function.' p I gatlure or sne sysces wm not, c -Troatse any safety-relar.vu j j l (system or prevent a safe shutdown of the plant.
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,J1ooding frS a postulated failure'of non-seismic, Category I
[ ,
sVstems and componentfwill not compromise the operation of any safety related a or prevent safe shutdown of the plant.
44 Most large volume non ismic category I canks and systems
- y are located in the turbine building and the radwaste area of JL the auxiliary building .with no potential for flooding areas containing equipment required for safe shugown.gafety re-Q laced cableis . in these. areas are located ataove ant potential c* ft :ne levels . .s.,
% gIn the reactor building and the control and diesel areas of
@ che auxilia ry' building potential flooding from postulated
,O failure of non-seismic Category I systems and components is ss
. contained within the compartment containingfloor the equipment.
drain sump.
The~ flooding will drain to the respective Essential equipment is located in areas not subject to l
flooding by the failure of nonseismic category I components or in compartments that are protected from flooding from ~
sources ' extended to the compartment.
In the unlikely event that a seismic event also causes the exposed drain line from the postulated flood area to leak, the fluid may drip into an area containing essential equip-ment.
The essential equipment i.s either located such that it is not subjected to the dripping or is designed to withstand the ef fects of the dripping . The dripping fluid will drain from the compartment through the floor drains.
e F67(1)
.., .~ ~ ,
(' ' )
,, OUESTION 410.93 (SECTION (9.3.3) ,
j Verify that all check valves which~ protect safety-related equipment from flooding due to backflow through the drainage 'i.r. .
. systems are seismic Category I. ,.
RESPONSE
J.li the ch d vol-wE which yi.wi.wu seiecy-relar.eu equipowni f ::
f!; ding d:: to b:'kflew thce gh the-d::in:g: ;i ;t;;; vill be eei--n--i n ; -ii f iet rete-se e,-n " cei:n uni t,2 co;gletec-b; ::== n::.
wm s.s.s. s ho.o ben.n reosed b oadne.ss h geh& g ao66n e4 %e. eke.ck-, "J as1. and th .
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- HCGS FSAR Each emergency core cooling system (ECCS) compartment is provided with a separate drain line to the reactor, building DRW sump. (
Flooding of the ECCS pump compartments in the react .
l is prevented by the use of Aadjustable normally closed check manual valves (backwatee) valve is ling installed in these lines.
provided for the floor artment drain line to present in the safety auxil ary c us+mf.f4.,
backflow. ,W system (SACS) pump c 't3 M w re. A su % j
. ciede_yel e ue .#.v w!G SMe cnWA nJv<r/M hn aa y2e) y, hed*Ag '
7
% pim pA pe porps assEach ECCS In compartment is equipped the ECCS compartment, W%T with4//watertig
/,,
prevent any spread of the flooding. & .r.fg' Seismic Category I level instrumentation installed in the mai ,
I blockage or flooding, l The drywell drain sumps and the floce drain sumos in the reactor !
enclosure are also used as a means to detect plant leakage as.. l I
discussed in Section 5.2.5. ~
- 9.3.3.6 Tests and Insoections Alt drainage piping is tested prior to its embedment in concrete. ) i Potentially radioactive drainage piping is pneumatically tes Nonradioactive oily, acid, and storm drainage i
J l
ANSI B31.1 (1973). .
. piping for 10 minutes.,
is hydrostatically tested to the equivalent of 20 e5 psigThe s to the National Standard Plumbing Code Plantat drainage a hydrostatic pressure systran of to feet of water for 15 minutes.
i operability is checked by normal use and by the instrumentation ;
j provided in the sumps and the main control room. )
l 9.3.3.7 Instrumentation Acolication
- a. Drywell equipment and floor drain sumps - A level measurement in each sump is fed to a local radiation processor that starts and stops the lead sump Thepump at a preset high and low levels, respectively.
. processor also starts the second pump and alternatesThe alars on high .
the lead pump after each pump cycle.
level in each sump is annunciated in the main control .
room.
i 9.3-34 Amendment 2 gmw 14 9
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APR 26 '84 0 2 6 3 3 3 9 HCGS I
DSER Open Item No. 151 ( DSER Se ction 9.4.1)
CONTROL STRUCTURE VENTILATION SYSTEM 1 The CRS and CREF systems take outside air from a common tornado- J missile-protected air intake. The air intake for the CERS system is also tornado missile pr ote ct ed ; however , there is no protection '
for the nonsafety-related WAS system intake. The exhaust for the CABE, WAE, CASE, and CAE systems are tornado missile pr ote ct ed .
Thus, the staf f concludes that the requirements of GDC 4, " Environ-mental and Missile Design Bases," are satistied. The air intakes have no chlorine' monitoring capability but do nave radiation monitor-ing capability. ' Signals from the radiation detectors ala rm in the control room, automatically isolate the fresh air intake from the control room HVAC system, and automatically start the CREF system to purify the fresh air. There is no automatic operation associated with the redundant CREF system train upon loss of the ope ra ti ng system. The CRS and CREF systems are designed to maintain the operability of the equipment in the control room. The control room systems are designed to maintain the control room u6 der a posit ive pr ess ure to minimize infiltration of gases intoThus, the the con-trol room except during 100% recirculation operation.
staff concludes that the requirements of GDC 19, " Control Room,"
and the guidelines of Regulatory Guide 1.78, " Assumptions for
- Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release," Positions C.3, C.7, and C.14, are satisfied. We cannot conclude that the guide-lines of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Opera to rs Against an Accidental Chlorine Release,"
Positions C.4a and C.4d are satisfied.
The CRS, CREF, and CERS systems consist of two 100% capa city trains of filters. The CREF system consists of a prefilter, a HEPA filter, i a charcoal filter, and a fan in series for the removal of radio-activity. The CRS and CERS systems consist of a prefilter, high ef ficiency filter , and a fan. There is no filtration of the ex-haust; however, it is isolated upon a high radiation signal.
Chilled water is supplied to the two 50% capacity cooling coils in each of the air handler units. The maximum ambient tempe ra tur e for which one train will maintain the proper environment is 94"F.
The applicant must demonstrate that one train of ventilation sys-tems can maintain the compartment environmental conditions within the qualification limits with an outside ambient temperature of 102"F for all design basis accidents with the loss of the redundant ventila tion systems. Based on the above, we cannot conclude that the requirements of General Design criterion 60, " Control of Releases.of Rad'oact.ve Materials to the Environment," and the i
j 151-1
arn to uw y 6 y a v v g3
- HCGS DSER Open Item No. 151 (Cont'd) guidelines of Regulatory Guides 1.52, " Design, Testing, and Mainten-ance Criteria for Atmospheric Cleanup System Air Filtration and Adsorption Units of Light Water-Cooled Nuclear Power Plants,"
Position C. 2, a nd 1.14 0, " Design, Testing and Maintenance Criteria for Normal _ Ventilation Exhaust System Air Filtration and Adsorption Units of Light Water-Cooled Nuclear Power Plants," Positions Col and C.2, are satisfied with respect to ensuring environmental limits for proper operation of plant controls under all normal conditions, including LOC A conditions.
and accidgent Based on the above, the staff concludes that the CSV systems are in conformance with the requirements of the GDC 2, 4, and 19 with res pe ct to protection against natural phenomena , tornado missile protection, and control room environmental conditions and the guidelines of RGs 1.29, Positions C.1 a nd C.2, a nd 1.78, Positions C.3, C.7, and c.14, relating to the seismic classifi-
' cation and prote ction against hazardous chemical release ..and is, therefore, acceptable. We cannot conclude that the CSV systems are in conformance with the requirements of General Design Cri-terion 60 with respect to control of radioactive releases and the guidelines of Regulatory Guide 1.52, Position C.2, 1.95, j Positions C.4.a and C.4.d, and 1.140, Positions C.1 and C.2, re-lating to the design for emergency operation, pr ote ction of pe rsonnel against a chlorine gas release, and normal operation.
We will report resolution of this item in a supplement to this SER. The HV AC systems which make up the. CSV systems do not meet the acceptance criteria of SRP Se ction 9.4.1.
RESPONSE
Evalua ti w .f accidents relating to the release of toxic chemicals includi ng chlorine is addressed in FSAR Se ction 2.2.3.1.3.
Also, per DSER Section 6.4, Page 6-3:
"With res pe ct to toxic gas protection, the staf f's evaluation in accordance with SRP Section 6.4, RGs 1.78 and 1.95 indicated ;
that there is no danger to control room personnel from toxic I chemicals, including chlorine, stored onsite or offsite, or l transported nearby (See Se ction 2. 2 3 ) . " l l
Section 9.4.1.3 has been revised to include reference to Sec-tion 2.2.3.13.
The CRS system provides cooling (with chilled water cooling coils) during normal operating conditions. The system also provides cooling, in conjunction with the CREF unit, in the event of an accident condition.
I 151-2
-_ _ ~ . - . _ _ . .-_._ .-_.._.-__-~..___.___.._--____.u .._
us 2s 8402 633 3 9 i HCGS DSER Open Item No. 151 (Cont'd)
The function'is either: ,
Au
- 1. 1000 cf m outside , pee makeup mixed with 3000 cfm of room return air diverted through the CREF unit. The balance of air is recirculated from the air cor ditioned space or,
- 2. A 100% recirculation mode, i.e., without outside air and with the use of the CREF unit.
See FS AR Se ction 9.4.1.2. 3.
Function Mode 2 is selected in the event of an accident condition.
When the outside ambient temperature condition is 102'F, 1000 cfm air is a minimal quantity (Xpproximat31 ygg.4% of the total air supply) which will increase the suppihgtemperature by less than l'F. The re fo re , this increase in temperature will not affect the operation of the plant controls due to the use of cooling coils as stated above. Since neither outside air is brought into the system nor is the control room exposed to solar load , outside a mbient temperature of 102*F has no ef fect on Function Mode 2.
O e
F64/5 151-3
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HCGS TSAR
. l
~" ' " - ' -" -- -' i s r n ...-
af er to the f ellcwing = = -' -" '- ,
j included in the design of the saf ety-related cen:rol area HV;ic )
systems (
- a. Protection from wind and tornado effects - Section 3.3
)
- b. Flood design - 3ection 3.4
- c. Missile protection - Section 3.5
- d. Protection against dynamic effects associated with the !
1 postulated rupture of piping - Sec. tion 3.6 ;
- e. Environmental design - Section 3.11
- f. Fire protection - Section 9.5.1. .
~
9.4.1p4
. bc. chasmcals - Seda s. 2.s.t.3 Tests and Inspections The CRS, CERS, CREF, and CABE systems and their components are tested in a program consisting of the following: -
1
- a. Factory and in-situ qualification tests (see ..,'~
4
. .4 Table 9.4-6) .
4
- b. Onsite preoperational testing (see Chapter 14)
- c. Onsite operational periodic testing (see Chapter 16).
Written test procedures establish minimum acceptable values for all tests. Test results are recorded as a matter of performance )
record, thus enabling early detection of faulty operating i performance.
All equipment is factory inspected and tested in accordance with the applicable equ'ipment specifications, codes, and quality j assurande requirements. Refer to Table 9.4-6 for details of inspection and testing.
9.4-15 O
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HCGS DSER Open Item 176c (Section 14.2)
INITIAL PLANT TEST P,ROGRAM Provide a response to 0640.9.
RESPONSE
The complete response to 0640.9 was provided as part of Amendment 6 to the HCGS FSAR.
i 0
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4 h
l M P84 126/07 1-dh l l
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, 7 i
HCGS DSER Open Item 176d (Section 14.2)
INITIAL PLANT TEST PROGRAM The response does not address the concerns of I&E Information Notice Number 83-17, March 31,1983. The concern is that if a time delay prevents fuel from being supplied to the diesel generator following a shutdown signal, the air supply may be exhausted before the fuel supply is reinstated. The response to this item should be modified to address these concerns.
RESPONS E
! The response to Q640.10 has been revised in Amendment 6 to
e i
M P84 126/07 2-dh
. . , _ . . - - _ ~ , , _ _ _ _ _ - . . . _ _ . _ _ _ - . . . _ _ . _ _ . - . _ _ . , . - - . . . . . _ - - , . . - . . , v._
.g I
HCGS DSER Open Item 176e (Section 14.2)
INITIAL PLANT TEST PROGRAM Provide response to 0640.11 Item 2.
l
RESPONSE
The information requested above was provided as part of Amendment 3 to the HCGS FSAR.
O-I I
p M P84 126/07 3-dh
it HCGS DSER Open Item 1761 (Section 14.2)
INITIAL PLANT TEST PROGRAM Provide . response to Q640.21 items 4, 5, and 6.
RESPONSE
I M P84 126/07 4-dh 1
., .___ - _ . . . . , _ _ . . . _ _ - - - . . - . _ . , , - - , , _ ~ , - , , , . . . . ..--,_ _.- __ _- _..,,..__, ..._, . _ _ . - - - _ . . - _ , - . . -
1 HCGS DSER Open Item No. 184 (Section 7.2.2.1)
' FAILURE-IN REACTOR VESSEL LEVEL SENSING LINES The applicant is required to submit the results of the analysis concerning failures in reactor vessel level sensing lines to the NRC for review and provide a description of the proposed modifications or justify why modifictions are not necessary.
RESPONSE
For the information requested above, see the response to Question 421.23. ,
1 M P84 126/04 2-mw
4 HCGS DSER Open Item No. 206 (Section 7.6.2.1)
HIGH PRESSURE / LOW PRESSURE INTERLOCKS The applicant was asked to discuss the design details utilized at HCGS for overpressurization protection of the low pressure ECCS. In response, the applicant provided acceptable design details for the ECCS high pressure / low pressure interlocks. However, the staff remained concerned regarding the setpoints utilized for these interlocks.
The applicant is required to provide the design basis for the selection of the setpoints utilized for ECCS high pressure / low pressure interlocks.
RESPONSE ,
Design details for the ECCS high pressure / low pressure "
interlocks are presented in the response to Questions 440.21 and 440.26 and are summarized in the response to DSER Open Item No. 135 (DSER Section 6.3.3). As these responses describe, overpressurization protection for the RHR low pressure piping is provided by the LPCI injection check valve rather than by differential pressure interlocks on the LPCI injection valves. Hence, there are no LPCI pressure interlock setpoints.
The core spray system injection (isolation) valves are
, interlocked directly with reactor pressure. A pressure indicating switch, N690 (see Figure 6.3-7) with a nominal setpoint of 461 psig and an allowable value of 441 psig, provides an opening permissive signal when the reactor pressure falls below the maximum design pressure (approximately 460 psig) for the core spray discharge piping.
'i M P84 126/04 1-mw
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HOPE CREEK -
DSER OPEN ITEMS 211a, b, d; 212, 213, 214, 215, 216a The main concern is that the applicant's alternative approaches to RGs 1.37, " Quality Assurance Requirements for Cleaning Fluid Sys-tems and Associated Components of Water-Cooled Nuclear Power Plants," ~
and 1.44, " Control of the Use of Sensitized Stainless Steel," do not provide an acceptble level of protection from intergranular stress corrosion cracking . . . For Regulatory Guide 1.44, the applicant has set high chloride content limits that exceed the recommendations of the guide. The applicant's high chloride limit of 200 ppm and the chloride limits for other materials that come in contact with austenitic stainless steels do not provide protection from concen-trations of chlorides that can occur by evaporation. The same situacion applies to the 100-ppm limit for chloride content of the final flushing water.
Cleaning and cleanliness control are not. in accordance with the recommendations of RG 1.37, " Quality Assurance Requirements for Cleaning Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants." The chloride content limits for flushing fluids are too high and are not acceptable to the staf f.
1 RESPONSE In Regulatory Guide 1.44 reference is made to Regulatory Guide 1.37 1 for the quality of water for cleaning and flushing of fluid systems.
l Regulatory Guide 1.37 further references ANSI N45.2.1.-1973 as an
- acceptable basis for complying with the pertinent quality assurance requirements of Appendix B to 10CFR Part 50.
For the NSSS and non-NSSS scope of supply the requirements speci-fied in the applicable GE and Bechtel specifications for cleanness
- of piping and equipment are in strict complitnce with Regulatory Guide 1.37 and ANSI N45.2.1-1973 regarding the water quality re-quirements of freshwater and domineralized water for rinsing and flushing purposes.
For non-metallic materials that come in contact with austenitic stainless steel, such as die lubricants, marking materials, masking tape, Cleaning solutions, etc., the GE and Bechtel specifications require that the chloride concentrations be controlled in accordance with the various relevant Regulatory cuides and ANSI standards.
Further these materials are removed and the surfaces cleaned and 1
rinsed immediately following the operation in which they are used. Since the quality of the rinse and flush water is being
t maintained there is adequate protection from concentrations of chlorides that could occur by evaporation.
FSAR Section 1.8.1.44 has been reviewed to the applicable GE and Bechtel specifications. This review resulted in the revision of Position Cl to provide clarification of several statements and the deletion of references to the use of trichlorotrifluoro-ethane (TCTFE), which is prohibited, such that this section more accurately describes the actual practice.
T I
t 5
- m. ,~ ,-..-~ _ .. v - , , - . .,..nn,c m ., . , - . - - . - -
. ', r ace controllad so na.t hcdo3an and sulfer levels ayee win na vivious .
Repl ater3Guidesov AMsr sbJanis HCGS FSAR C*#'"3 0 858 7 4 ass me:fsrials o.re ruhieved smewdifely
- Qawing de o isninuilifcIthart-used and in onj elswded-
%,u -f pd4 man a+.
1.8.1.44 Conformance to Regulatory Guide 1.44, Revision 0, May 3 1973: Control of the Use of Sensitized Stainless Steel HCGS complies with Regulatory Guide 1.44, excep,t as'noted below.
Architect-engineer-procured items and architect-engineer field work comply with Regulatory Guide 1.44, subject to exceptions or clarifications stated below that are applied to ASME B&PV Code, -
Section III equipment and piping in safety-related systems. They are not generally applied to HVAC sy.=tems or to instruments.
(in cuanbes win R*4al* te"4 4.ide t.S7and ANSI N45.2.1-M7 Position C.1 of Regulatory Guide 1.44 is complied with since contamination of austenitic stainless steel (Type 300 series) by compounds that could cause stress corrosion crackino is avoided _
during all stages of fabrication and installation y x pu fut tri or ritip toeca e LT FE) etin the r uire nts o M t Sp+ Ifica n MI C-81 2B, eanin is 1 ited ol ions at c ain t mo tha 00 p of ocid .
R sing r flu ing i done ith ter t t co ains t mo han 0 ppm f chl ides Spe al eipding t hni s ar use to sure mplet remov 1 of CTFE Were c vic or u rai 1
} [a as oc rf Nonm dsls y 4; :fs.llsc.
qn : Q> ct n:c_ in contact with austenitic stainless steel p p. ., te uor canc , pe ecs. u ma ria , m xtng ace a4 ypsk g pe etc are ont olle so at t ey co ai o' th 20 pp of hlor des or ey e re ved medt te__
(f _ 16 nn he cera on wh ch t yw e um Penetrant materials may conform to the higher contaminant levels specified in Article 6,Section V, of the ASME B&PV Code, provided.that the materials are thoroughly removed #Iihmediately arter tne Tand the sur4ce examination has been completed. Crevices and edeeinable :::: 4 84H
'roce wed ior c cne u e vi m certal conc inin mot cn 3
200 ppm fc rides All bstan s in ntac wi aus c., -
P [(at ::t-r-Vsvnatt e openm3s, am, pr'otecded imm condesda.&.
s ini. s st 1 ara ramov orio to anv leva d m6e tu Completed components are packaged M uch : ::7. at they are
- protected from the weather, dirt, wind, water spray, and any other extraneous environmental conditions that may be encountered during shipinent and subsequent site storage.
1.8-23
r !
(
In the field, austenitic stainless steel components are stored clean and dry. Components either are stored indoors, or, if outdoors, are stored off the ground and covered.with taros.
- e. a.w lent ~to eacAor- c ocl Contamination of austenitic stainless steels in the field during installation is avoided as described above. The system
- hydrostatic test and the preoperational testing and final flushing of the completed system is performed withawater that cwus.ine n;t ;;;; then ;00 ,g. v. ......____. Nonmetallic insulation composed of leachable chloride and fluoride materials that come into contact with austenitic stainless steel are held to the lowest practicable level by the inclusion of the requirements of Regulatory Guide 1.36 in the insulation purchase specifications.
Position C.1 of Regulatory Guide 1.44 is complied with sines all grades of austenitic stainless steels (Type 300 series) are required to be furnished in the solution heat-treated condition before fabrication or assembly into components or systems. The solution heat treatment varies according to the applicable ASME '
or ASTM material specification. .
Position C.3 of Regulatory Guide 1.44 covers all austenitic stainless steels furnished in the solution heat-treated condition in accordance with the material specification. During fabrication and installation, austenitic stainless steels are not permitted to be exposed to temperatures in the range of 800 to 15000F, except for welding and hot forming. Welding practices are controlled to avoid severe sensitization, and solution heat treatment in accordance with the material specification is also required.following hot forming in the temperature range of 800 to 15000F. Unless otherwise required by the material specification, the maximum length of time for cooling from the solution heet-treated temperature to below 8000F is specified in the equipmer.t specification. Corrosion testing in accordance with ASTM A 262-70, Practice A or E, may be required if the maximum length of time for cooling below 8000F is exceeded, or the solution heat-treated condition is in doubt.
No austenitic stainless steel is subjected to service temperatures in the range of 800 to 15000F, as discussed in Position C.4 of Regulatory Guide 1.44. The only exposure of austenitic stainless steels to this range of temperatures occurs on the containment hydrogen recombiner system (CHRS) and ,
subsequent to solution heat-treating during welding. Welding .
practices are controlled as discussed below. In addition, the architect-engineer-supplied austenitic stainless steel piping and
,1.8-24 s
- ~---.,--__. . _ _ _ _ _ _ _ _ _ _ _ _
9, HCGS DSER Open Item 211c (Section 4.5.1)
. CONTROL ROD DRIVE STRUCTURAL M TA ERIALS The allowed welding heat input limit of 100 kj/in for the fabri-cation of control rod drive components has been shown by General Electric to sensitize Type 304 austenitic stainless steel and accordingly is unacceptable.
RESPONSE
The welding specification controlling the fabrication of con-trol rod drive ( CRD) components at GE's Wilmington, NC manu-facturing operations has always specified a heat input limit of 50 Kj/in. The HCGS CRD components were f abricated under this specification . Section 4.5.1.2.1 has been revised to remove the reference to the description of compliance to Regulator *J Guide 1.44 in Section 4.5.2.4.4, which deals with reactor vessel internals.
4 4
4 4
0 a.- - - . . . - - - , , . - - ...-,,,-~,--..,,,_,-y, -.mm, ,,. . _ . .. _ , ~ . - - n y .- - , . - - - ,.-..g..e, , - , , .m,--.-
- a. The cylinder and spacer (cylinder, tube and flange essembly) and the retainer (collet assembly) are hard surfaced with Colmonoy 6.
- b. The foll.owing componwnts are nitrided to provide a wear resistant surface:
- 1. Piston tube (piston tube assembly)
- 2. Index tube (drive line assembly)
- 3. Collet piston and guide cap (collet assembly).
! Colmonoy hard surfacing is applied on the cylinder, spacer, and retainer by the flame spray process.
Nitriding is accomplished using a proprietary process called New Halcomizing. Components are exposed to a temperature of about 10800F for approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> during the nitriding cycle.
Colmonoy hard surfaced components have performed successfully for the past 20 years in drive mechanisms. Nitrided components have
, been used in CRDs since 1967. It is normal practice to remove some CRDs at each refueling outage. At this time, both the Colmonoy hard surfaced parts and the nitrided surfaces are accessible for visual examination. In addition, dye penetrant examinations have been performed on nitrided surfaces of the longest service drives. This inspection program is adequate to detect any incipient defects before they can become serious i
enough to cause operating problems.
l f Welding is performed in accordance with Section IX of the ASME i B&PV Code. Heat input for stainless steel welds is restricted to a maximum of 50,000 Joules per inch and an interpass temperature i of 3500F. These controls are employed to avoid severe j sensitization and comply with the intent of Regulatory i
Guide 1.44. "-- ----- ' -- --- -- - '- -------" "
rrrre:::.t f: 90;rlatery Cuid: '
!!, :: Errtier t.5.2.! ! -
4.5-4 D$dAOMN trd m a ll (
n 4
HCGS FSAR 4.5.2.4.2 'Conformance with Regulatory Guide 1.34, Control of Electroslag Weld Properties Electroslag welding is not employed for any reactor internals.
4.5.2.4.3 Conformance with Regulatory Guide 1.36, i Nonmetallic Thermal Insulation for Austenitic Stainless Steel
' For external applications, all nonmetallic insulation meets the requirements of Regulatory Guide 1.36.
' 4.5.2.4.4 Conformance with Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel All wrought austenitic stainless steel is purchased in the ~
solution heat treated condition. Heating above 8000F is i
prohibited (except for welding) unless the stainless steel is subsequently solution annealed.
1'
. carbon content in excess of 0.035% For 304 stainless carbon, purchase steelowith specifications restrict the maximum weld heat input to 140,000 Joules per inch, and the weld interpass temperature to 3500F maximum.
4 Welding is performed in accordance with Section IX of the ASME B&PV Code. These controls are employed to avoid severe i
sensitization, Guide 1.44. and comply with the intent of Regulatory j 4.5.2.4.5 Conformance with Regulatory Guide 1.71, Welder 5
j Oualification for Areas of Limited Accessibility 4
t i There are few restrictive items described welds involved in the fabrication of in this section.
the welds with most difficult access. Mock-up Mock-ups welding areis2xamined performed on' with I,
radiography or by sectioning.
4.5.2.4.6
.i Conformance with Regulatory Guide 1.37, Quality i Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled
- Nuclear Power Plants
- . Exposure to contaminants is avoided by carefully controlling all i
cleaning and processing materials that contact stainless steel i
4.5-11 ps CR I TCm 2 // C
.._ . _ - _ , - -. __._ _. _ ___ _ _ - .._ -.__..__._. - ~ ~ _ _ _
~
s _ . . ._ _.
HCGS DSER Open Item No. 211e (Section 4.5.1)
CONTROL ROD DRIVE STRUCTURAL MATERIALS The applicant should identify the materials specifications used in the control rod drive components made of ARMCO 17-4 PH, and Inconel X-750.
RESPONSE
The fingers of the collet assemblies and the coupling spuds of the drive line assemblins of the HCGS control rod drives ( CRDs )
were fabricated of Inconel X-750, which was specified by.a General Electric specification similar to ASTM A637, G688, Type 2. The collet springs of the CRDs were fabricated of Inconel X-750, which was specified by a General Electric specification similar to AMS 5699. -The piston heads of the drive line assemblies were fabricated of 17-4-PH, which was specified by a General Electric equivalent to ASTM A564, Type G630 with a 11000F age hardening.
l
j 1
l l
HCGS DSER Open Item No. 216b (Section 5.3.1)
REACTOR VESSEL MATERIALS The reactor vessel studs and fasteners satisfy some of the recom-
- mandations of RG 1.65, " Materials and Inspections for Reactor vessel Closure Studs." The FSAR does not-discuss the nondestruc-tive examinations of the stud bolts and nuts.
RESPONSE
The main closure studs, nuts, and washers for the reactor vessel are ultrasonically examined in accordance with Paragraph N-322 of Section III of the ASME B&PV Code and additional GE requirements.
Magnetic particle inspections of the surfaces of the main closure studs, nuts and washers, are conducted in accordance with Para-graph N-626 of Section III of the ASME B&PV Code.
.