ML20094P277
| ML20094P277 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/15/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8408170103 | |
| Download: ML20094P277 (578) | |
Text
Put$c Service Electnc and Gas Company 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 August 15, 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 ha. 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 ramains to be resolved.
Attachment 2 is a current list which identifies Draft SER Sections not yet provided.
In addition, enclosed for your review and approval (see Attachment 4) are the resolutions to the Draft SER open items, and PSAR question responses listed in Attachment 3.
Attachments 5 and 6 contain the responses retransmitted on August 3 and 10, 1984, respectively. A signed original of the required affidavit is provided to document the submittal of these items.
Should you have any questions or require any additional information on these open items, please contact us.
Very truly yours,
/
The Energy People E
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' Director of Nuclear Reactor Regulation 2 8/15/84 C_.D. H. Wagner USNRC! Licensing Project. Manager W. H. Bateman USNRC Senior Resident Inspector
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UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION j DOCKET NO. 50-354 PUBLIC. SERVICE ELECTRIC AND GAS COMPANY Public Service Electric and Gas Company hereby submits the enclosed Hope Creek Generating Station Draf t Safety Evalua-tion Report open item responses and FSAR Question 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 h
By: .N/4 A1
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Thomas J. arfin~
Vice Pre dent -
Enginee ng and Construction Sworn to and subscribed before me, aNotaryPupic of New Jersey, this /5- day of August 1984.
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DAVID K. BURO NOTARYPUBUC 0F NEW JERSEY My Comm. Lapires 102388 GJ02
DATE8 8/15/84 ATTAOMENT 1 DSER R. L. MITTL TO OPEN SECTION A. SCHWENCER
.ITkM NLMBER SUETECT STATUS IEI"IER DATED 1 2.3.1 Design-basis - tenperatures for safety- Cmplete 8/15/84 related auxiliary systems 2a. 2.3.3 Accuracies of meteorological Cmplete 8/15/84 measurements (Rev. 1) 2b 2.3.3 , Accuracies of meteorological Cm plete 8/15/84
~ measurements (Rev. 1) 2c 2.3.3 Accuracies of meteorological Cmplete 84 5 /84 measurements ' (Rev. 2)
L
- 2d 2 3.3 Accuracies of meteorological Complete 84 5 /84 measurements (Rev. 2)
I 3a 2.3.3 Upgrading of onsite meteorological Ocmplete 8/15/84 measurements program (III.A.2) (Rev. 2) 3b 2.3.3 Upgrading of onsite meteorological Cm plete 8/15/R4 measurements progran (III.A.2) (Rev. 2)
] 3c 2.3.3 Upgrading of onsite meteorological Open measurements program (III.A.2) 4 2.4.2.2 Ponding levels Conplete 8/03/84 Sa 2.4.5 Wave inpact and runup on service Ca plete 6/01/84 Water Intake Structure Sb 2.4.5 Wave impact and runup on service Open
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water intake structure Sc 2.4.5 Wave inpact and runup m servi Ca plete 7/27/84 water intake structure
! 5d 2.4.5 Wave inpact and runup m service Cmplete 6/01/84 water intake structure 6a' 2.4.10 Stability of erosim protection Open 2 structures t i, 6b 2.4.10 Stability of erosion protection Open structures 6c 2.4.10 Stability of erosien protectim Cmplete 8/03/84 structures M 164 80/12 1-gs l
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ATTACINENT 1 (Cont'd)
DSER R. L. MITTL TC OPEN SECTION A. SOMENCER ITEM NUMBER SUR7ECT STATUS LETTER DATED 7a 2.4.11.2 Thermal aspects of ultimate heat sink Cmplete 8/3/84 7b 2.4.11.2 Thermal aspects of ultimate heat sink Caplete 8/3/84 8 2.5.2.2 Choice of neximian earthquake for New Complete 8/15/84 England - Piechnont Tectonic Province 9 2.5.4 Soil danping values Cmplete 6/1/84 10 2.5.4 Foundation level response spectra Conplete 6/1/84 11 2.5.4 Soil shear moduli variation Cmplete 6/1/84 12 2.5.4 cmbination of soil layer properties Couplete 6/1/84 13 2.5.4 Iab test shear moduli values Cmplete 6/1/84 14 2.5.4 Liquefaction analysis of river bott m Conplete 6/1/84 sands 15 2.5.4 Tabulations of shear moduli Caplete 6/1/84 16 2.5.4 Drying and wetting effect cn Caplete 6/1/84 Vincentown 17 2.5.4 Power block settleent monitoring Conplete 6/1/84 18 2.5.4 Maximtsu earth at rest pressure Conplete 6/1/84 coefficient 19 2.5.4 Liquefaction analysis for service Cmplete 6/1/84 water piping 20 2.5.4 Explanation of observed power block Cmplete 6/1/84 settlement 21 2.5.4 Service water pipe settlement records Cmplete 6/1/84 22 2.5.4 Cofferdam stability Cmplete 6/1/84 23 2.5.4 Clarification of FSAR Tables 2.5.13 cmplete 6/1/84 )
and 2.5.14 l
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! M P84 80/12 2 - gs t
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7 ATTACININT 1 (Cont'd)
DSER R. L. MITIL 'IO OPEN SECTIOi A. SCHWENCER I'IDt MJMBER SUBJECT STA'IUS IEI'rER DATED 24 2.5.4 Soil depth models for intake Conplete 6/1/84 structure 25 2.5.4 Intake structure soil modeling Conplete 8/10/84 26 2.5.4.4 Intake structure sliding stability open 27 2.5.5 Slope stability Conplete 6/1/84 28a 3.4.1 Flood protection Complete 7/27/84 28b 3.4.1 Flood protection Conplete 7/27/84 28c 3.4.1 Flood protection Conplete 7/27/84 28d 3.4.1 Flood protection Couplete 7/27/84 28e 3.4.1 Flood protection Conplete 7/27/84 28f 3.4.1 Flood protection Conplete 7/27/84 28g 3.4.1 Flood protection Conplete 7/27/84 29 3.5.1.1 Internally generated missiles (outside Conplete 8/3/84 containment) (Rev. 1) 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 Couplete 7/18/84 32 3.5.1.4 Missiles generated bf natural phenomena Conplete 7/27/84 33 3.5.2 Structures, systems, and conponents to Conplete 7/27/84 be protected from externally generated missiles 34 3.6.2 Unrestrained whipping pipe inside Conplete 7/18/84 containment 35 3.6.2 ISI program for pipe welds in Couplete 6/29/84 break exclusion zone l
M P84 80/12 3 - gs ,
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T ATTAOMENT 1 (Cont'd)
DSER R. L. MITIL TO OPEN SECTICN A. SCHWENCER IT1!M NLMBER SUBJECT STATUS IEITER DATED 36 3.6.2 Postulated pipe ruptures Cmplete 6/29/84 37 3.6.2 Feedwater 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 Cmplete 8/3/84 elanent method and elastic half-space approach for containment structure 40 3.7.2.3 SSI analysis results using finite Conplete 8/3/84 element method and elastic half-space approach for intake structum 41 3.8.2 Steel containment buckling analysis Cmplete 6/1/84 42 3.8.2 Steel containment ultimate capacity Couplete 6/1/84 analysis 43 3.8.2 SRV/LOCA pool dynamic loads Cmplete 6/1/84 44 3.8.3 ACI 349 deviations for internal 02nplete 6/1/84 structures 45 3.8.4 ACI 349 deviations for Category I Cmplete 6/1/84 structures e 46 3.8.5 ACI 349 deviations for foundations Qmplete 6/1/84 47 3.8.6 Base mat response spectra Caplete 8/10/84 Rev.
48 3.8.6 Rocking time histories Cmplete 6/1/84 49 3.8.6 Gross concrete section Caplete 6/1/84 50 3.8.6 Vertical floor flexibility response Ccmplete 6/1/84
...!' spectra 51 3.8.6 Capariscn of Bechtel independent Ca plete 8/10/84 f verification results with the design- (Rev. 1) basis results 1
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M P84 80/12 4 - gs
i ATTACIMENT 1 (Cont'd)
DSER R. L. MITTL 'IO OPEN SECTICN A. SCHWENCER ITEM NUMBER SUBJECT STA'IUS IETTER DATED 52 3.8.6 Ductility ratios due to pipe break Conplete 8/3/84 53 3.8.6 Design of seismic Category I tanks Conplete 6/1/84 54 3.8.6 Combination of vertical responses Conplete 8/10/84 Rev<
- > 55 3.8.6 'Ibrsional stiffness calculation couplete 6/1/84
- 56 3.8.6 Drywell stick model developnent Conplete 6/1/84 57 3.8.6 Rotational time history inputs Conplete 6/1/84 58 3.8.6 "O" reference point for auxiliary Conplete 6/1/84 building model 59 3.8.6 Overturning moment of reactor Conplete 6/1/84 building foundation mat 60 3.8.6 BSAP element size limitations Conplete 6/1/84 61 3.8.6 Seisnic nodeling of drywell shield Cocplete 6/1/84 wall 62 3.8.6 Drywell shield wall boundary Conpletc 6/1/84 conditions 63 3.8.6 Reactor building dome boundary Conplete 6/1/84 conditions 64 3.8.6 SSI analysis 12 Hz cutoff frequency Conplete 6/1/84 65 3.8.6 Intake structure crane heavy load Couplete 6/1/84 drop
- 66 3.8.6 Inpedance analysis for the intake Conplete 8/10/84 Revc structure 67 3.8.6 Critical loads calculation for Conpleto 6/1/84 reactor building dome
[ 68 3.8.6 Reactor building foundation ant Conpletc 6/1/84 contact pressures i
i M P84 80/12 5 - gs i
ATTACINENT 1 (Cont'd)
DSER R. L. MITTL IC OPEN SECTICN A. SCHWENCER ITEN NUMBER SUBJEX.T STATUS IEITER DATED 69 3.8.6 Factors of safety against sliding and Cmplete 6/1/84 overturning of drywell shield wall 70 3.8.6 Seismic shear force distrhtion in Cm plete 6/1/84 cylinder wall g 71 3.8.6 Overturning of cylinder wall Cmplete 6/1/84 72 3.8.6 Deep beam design of fuel pool walls Caplete 6/1/84 73 3.8.6 ASHSD dame nodel load inputs Cmplete 6/1/84 74 3.8.6 Tornado depressurization Cmplete 6/1/84 75 3.8.6 Auxiliary building abnornal pmssure Cmplete 6/1/84 76 3.8.6 Tangential shear stresses in drywell Cmplete 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 Imd calculations Ca pleto 6/1/84 79 3.8.6 Post-modification seismic loads for Caplete 6/1/84 the torus 80 3.8.6 Torus fluid-structure interactions Cmplete 6/1/84 81 3.8.6 Seismic displacement of torus Complete 6/1/84 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 84 3.8.6 Ultimate capacity of containment Cmplete 6/1/84 (materials) 85 3.8.6 toad cambination consistency Cmplete 6/1/84 M P84 80/12 6 gs 1
l ATTAOMNr 1 (Cont'd)
DSER R. L. MITIL 10 OPEN SECTICN A. SCHWENCER l ITEM NUMBER SUE 7BCr STA1US LETTER DMTD i i
86 3.9.1 Caputer code validation Open 87 3.9.1 Information m transients Open 88 3.9.1 Stress analysis and elastic-plastic Caplete 6/29/84 analysis 89 3.9.2.1 Vibration levels for NSSS piping Caplete 6/29/84 systems
. 90 3.9.2.1 Vibration monitoring program during Cmplete 7/18/84 testing 91 3.9.2.2 Piping supports and anchors Caplete 6/29/84 92 3.9.2.2 Triple flued-head containment Ca plete 6/15/84 penetrations 93 3.9.3.1 Ioad cabinations and allowable Caplete 6/29/84 stress limits 94 3.9.3.2 IMsign of SWs and SW discharge Caplete 6/29/84 Pi ping 95 3.9.3.2 Fatigue evaluation on SW piping Caplete 6/15/84 and IDCA downconers 96 3.9.3.3 IE Information Notice 83-80 Capletc 6/15/84 97 3.9.3.3 Buckling criteria used for conponent Complete 6/29/84 supports 4
98 3.9.3.3 Design of bolts Complete 6/15/84 99a 3.9.5 Stress categories and limits for Caplete 6/15/84 corw support structures 99b 3.9.5 Stress categories and limits for Caplete 6/15/84 core support structures i
", 100a 3.9.6 10CFR50.55a paragraph (g) Caplete 6/29/84 M P84 80/12 7 - gs
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_ATTAOMENT 1 (Cont'd)
DSER R. L. MITTL 'Id OPEN SECTION A. SCHWENCER ITEM NUMBER SUEUBCr STA'!US LETTER IWrED 100b 3.9.6 10CFR50.55a paragraph (g) Open 101 3.S.6 PSI and ISI programs for punpa and Open valves 102 3.9.6 IAak testing of pressure isolation Conglete 6/29/84 valves 103a1 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 103a3 3.10 Seismic and dynamic qualification of Open
. mechanical and electrical equipment 103a4 3.10 Seimnic 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 and dynamic qualification of Open
, mechanical and electrical equipment 103a7 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103b1 3.10 Seismic and dynamic qualificaticn of Open mechanical and electrical equipment 103b2 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103b3 3.10 Seismic and dynamic qualification of Open
. mechanical and electrical equipment 103b4 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103b5 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment M P84 80/12 8 - ga s
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ATTACEMENT 1 (Cont'd)
DSER R. L. MITTL TO OPEN SECTICN t A. SCHWENCER ( '
ITEM NUMBER SUBJECT STATUS LETTER DPLTED 103b6 3.10 Seismic and dynamic qualification of open mechanical and electrical equipnent 103c1 3.10 Seismic and dynamic qualification of open m chanical and electrical equipnent 103c2 3.10 Seismic'and dp duic qualification of open mechanical and electrical equipnent 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 equipnent 104 3.11 Environmental qualification of NRC Action mechanical and electrical equipment 105 4.2 Plant-specific mechanical fracturing Cmplete 7/18/84 analysis 106 4 .2 Applicability of seismic andd LOCA Complete 7/18/84 loading evaluation 107 4.2 Minimal post-irradiation fuel Cmplete 6/29/84 surveillance progran 108 4.2 Gadolina thermal conductivity Cm plete 6/29/84 equation 109a 4.4.7 TMI-2 Item II.F.2 Open 109b 4.4.7 TMI-2 Item II.F.2 Open 110a 4.6 Functional design of reactivity Cmplete 7/27/84 control systens 110b 4.6 Functional design of rea'ctivity Cm plete 7/27/84 control systems 111a 5.2.4.3 Preservice inspection program Cmplete 6/29/84 (conponents within reactor pressure boundary)
M P84 80/12 9 - gs.
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ATTACHMENT 1 (Cont'd)
DSER R.L.MITILh OPEN SECTICN A. SCHWENCER ITEM NUMBER SUELTECT STA'IUS LETTER DATED lllb 5.2.4.3 Preservice inspection program Cmplete 6/29/84 (cmponents within reactor pressure boundary) lllc 5.2.4.3 Preservice inspection program Cmplete 6/29/84 (caponents within reactor pressure boundary) ll2a 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 leakage detection ll2b 5.2.5 Reac'_or coolant pressure boundary Caplete 7/27/84 leakage detection ll2c 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 leakage detection ll2d 5.2.5 Reactor coolant pressure boundary Complete 7/27/84 leakage detection 112e 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 leakage detection 113 5.3.4 GE procedure applicability Cmplete 7/18/84 114 5.3.4 Compliance with NB 2360 of the Sumer Cmplete 7/18/84 1972 Addenda to the 1971 ASME Code 115 5.3.4 Drop weight and Charpy v-notch tests Cmplete 7/18/84 for closure flange materials 116 5.3.4 Charpy v-notch tost data for base Cmplete 7/18/84 materials as used in shell course No.1 117 5.3.4 Capliance with NB 2332 of Winter 1972 Open Addenda of the ASME Code 118 5.3.4 Lead factors and neutron fluence for Open surveillance capsules l
l' ll M P84 80/12 10- gs
ATTACHMENT 1 (Cont'd)
DSER R. L. MITTL TO OPEN SECTICN A. SCHWENCER ITEM NUMBER SUBJECT STATUS LETTER DATED 119 6.2 TMI item II.E.4.1 Cmplete 6/29/84 120a 6.2 IMI 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 cperation (post Cmplete 6/29/84 accident) 124a 6.2.1.5.1 RPV shield annulus analysis Complete 6/1/84 124b 6.2.1.5.1 RPV shield annulus analysis Cmplete 6/1/84 124c 6.2.1.5.1 RPV shield annulus analysis Cmplete f'l/84 125 6.2.1.5.2 Design drywell head differential Cmplete v/15/84 pressure 126a 6.2.1.6 Redundant position indicators for Open vacum breakers (and control rom alams) 126b 6.2.1.6 Redundant position indicators for Open vacum breakers (and control room alarms) 127 6.2.1.6 Operability testing of vacuum breakers Ccraplete 7/18/84 128 6.2.2 Air ingestion Cmplete 7/27/84 129 6.2.2 Insulation ingestion Cmplete 6/1/84 130 6.2.3 Potential bypass leakage paths Cmplete 6/29/84 131 6.2.3 Administration of secondary contain- Cmplete 7/18/84 ment openings M P84 80/12 11- gs
1 ATTACINENT 1 (Cont'd)
DSER R. L. MITTL T OPEN SECTICN A. SCHWENCER ITEM NUMBER SUR7ECT STATUS IEITER IRTED 132 6.2.4 Containment isolation review Ccmplete 6/15/84 133a 6.2.4.1 Contairiment purge system Open 133b 6.2.4.1 Containment purge system Open 133c 6.2.4.1 Contairnent purge system Open 134 6.2.6 Containment leakage testing Cartplete 6/15/84 135 6.3.3 LPCS and LPCI injection valve Open interlocks 136 6.3.5 Plant-specific IDCA (see Section Cmplete 7/18/84 15.9.13) 137a 6.4 Control rom habitability Open 137b 6.4 Control rom habitability Open 137c 6.4 Control rom habitability Open 138 6.6 Preservice inspection program for Caplete 6/29/84 Class 2 and 3 ca ponents 139 6.7 MSIV leakage control system Couplete 6/29/84 140a 9.1.2 Spent fuel pool storage Cmplete 8/15/84 (Rev. 1) 140b 9.1.2 Spent fuel pool storage Cmplete 8/15/84 (Rev. 1) 140c 9.1.2 Spent fuel pool storage Cmplete 8/15/84 (Rev. 1) 140d 9 .1.2 Spent fuel pool storage Complete 8/15/84
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(Rev. 1) 141a 9.1.3 Spent fuel oooling and cleanup Cmplete 8/1/84 system 141b 9.1.3 Spent fuel cooling and cleanup Ccmplete 8/1/84 systen 141c 9.1.3 Spent fuel pool oooling and cleanup Cmplete 8/1/84 system i
M P84 80/12 12- gs
ATTACINENT 1 (Cont'd)
DSER R. L. MITTL 'IO OPEN SECTION. A. SCHWENCER I'ITM NLMBER SUli7ECT STA'IUS IEITER DATED 141d 9.1.3 Spent fuel pool cooling and cleanup Cmplete 8/1/84 )
systa 141e 9.1.3 Spent fuel pool cooling and cleanup Cmplete 8/1/84 system 141f 9.1.3 Spent fuel pool cooling and cleanup Cmplete 8/1/84 syste 141g 9.1.3 Spent fuel pool cooling and cleanup Cmplete 8/1/84 systen 142a 9.1.4 Light load handling system (related Camplete 8/15/84 to refueling) (Rev. 1) 142b 9.1.4 Light load handling syst m (related Ccmplete 8/15/84 to refueling) (Rev. 1) 143a 9.1.5 overhead heavy load handling Open 143b 9.1.5 Overhead heavy load handling Open 144a 9.2.1 Station service water system Cmplete 8/15/84 (Rev. 1) 144b 9.2.1 Statico service water syste Complete 8/15/84 (Rev. 1) 144c 9.2.1 Station service water system Cmplete 8/15/84 (Rev. 1) 145 9.2.2 ISI progran and functional testing Closed 6/15/84 of safety and turbine auxiliaries (5/30/84-cooling 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 Coupressed air systems Ccmplete 8/3/84 (Rev 1) i 147b 9.3.1 Ompressed air systems Cmplete 8/3/84 (Rev 1)
M P84 80/12 13- gs
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l 1
ATTAONENT 1 (Cont'd)
DSER R. L. MITIL 10 OPEN SECTION A. SOMENCER ITEM NtNBER SURTECT STATUS LETTER IRTED 147c 9.3.1 Cmpressed air systems Cmplete 8/3/84 (Rev 1) 147d 9.3.1 cmpressed air systems Complete 8/3/84 (Rev 1) 148 9.3.2 Post-accident sanpling system Open (II.B.3) 149a 9.3.3 Equignent and floor drainage system Cmplete 7/27/84 149b 9.3.3 Equipnent and floor drainage system Cmplete 7/27/84 150 9.3.6 Primary containment instrument gas Cmplete 8/3/84 system (Rev. 1) 151a 9.4.1 Control structure ventilation system Cmplete 7/27/84 151b 9.4.1 Control structure ventilation system 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 8/1/84 tion system (Rev 1).
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 4 159 9.5.1.5.a Primary and secondary power supplies Cmplete 6/1/84 for fire detection system 160 9.5.1.5.b Fire water punp capacity Cmplete 8/13/84 l
l M P84 80/1214 g;
ATTACINENT 1 (Cont'd)
DSER R. L. MITIL 'IO OPEN SECTION A. SOMENCER ITEM NLMBER SURTECT STA'IUS LETTER DATED l 1
161 9.5.1.5.b Fire water valve supervision Cmplete 6/1/84 162 9.5.1.5.c Deluge valves Cm plete 6/1/84 163 9.5.1.5.c Manual hose station pipe sizing Caplete 6/1/84 164 9.5.1.6.e Remote shutdown panel ventilation Cm plete 6/1/84 165 9.5.1.6.g Emergency diesel generator day tank Cmplete 6/1/84 protection 166 12.3.4.2 Airborne radioactivity monitor Cmplete 7/18/84 positioning 167 12.3.4.2 Portable continuous air nonitors Cmplete 7/18/84 168 12.5.2 Equipment, training, and procedures Cmplete 6/29/84 for inplant iodine instrumentation 169 12.5.3 Guidance of Division B Regulatory Cmplete 7/18/84 Guides 170 13.5.2 Procedures generation package Cmplete 6/29/84 sutznittal 171 13.5.2 'IMI Item I.C.1 Cmplete 6/29/84 172 13.5.2 PGP Cmmitment Cmplete 6/29/84 173 13.5.2 Procedures covering abnormal releases Cmplete 6/29/84 of radioactivity 174 13.5.2 Resolution explanation in FSAR of Cmplete 6/15/84
'IMI Items I.C.7 and I.C.8 175 13.6 Physical security Open 176a 14.2 Initial plant test program Cmplete 8/13/84 i
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M P84 80/12 15- gs
ATTACMENT 1 (Cont'd)
DSER R. L. MITTL 10 OPEN SECTION A. SCHWENCER ITB4 NLMBER SUBJ W STATUS LETTER DATED 176b 14.2 Initial plant test progrm Ctmplete 8/13/84 176c- 14 .2 Initial plant test program Cmplete 7/27/84 176d 14.2 Initial plant test program Cmplete 8/13/84 Rev.
176e 14.2 Initial plant test progrart Cmplete 7/27/84 176f 14.2 Initial plant test program Cmplete 8/13/84 176g 14.2 Initial plant test program Open 176h 14.2 Initial plant test program Cmplete 8/13/84 176i 14.2 Initial plant test program Caplete 7/27/84 177 15.1.1 Partial feedwater heating Caplete 7/18/84 178 15.6.5 IDCA resulting frcm spectrum of NRC Action postulated piping breaks within RCP 179 15.7.4 Radiological consequences of fuel NRC Action handling accidents 180 15.7.5 Spent fuel cask drop accidents NRC Action 181 15.9.5 'IMI-2 Iten II.K.3.3 Cmplete 6/29/84 4
182 15.9.10 1MI-2 Item II.K.3.18 Caplete 6/1/84 183 18 Hope Creek DCRDR Caplete 8/15/84 184 7.2.2.1.e Failures in reactor vessel level Otrnplete 8/1/84 sensing lines (Rev 1) 185 7.2.2.2 Trip systen sensors and cabling in Cmplete 6/1/84 turbine building 186 7.2.2.3 Testability of plant protection Cmplete 8/13/84 Rev.
, systas at power M P84 80/12 16- gs t
i e - ~ "' ""
ATTACIMENT 1 (Cont'd)
DSER R. L. MITTL 10 OPEN SECTI W A. SCHWENCER ITEM NLMBER SUBJECT STATUS LETTER DATED 187 7.2.2.4 Litting of leads to perform surveil- Omplete 8/3/84 lance testing 188 7.2.2.5 Setpoint methocblogy Cmplete 8/1/84 189 7.2.2.6 Isolation devices 02nplete 8/1/84 190 7.2.2.7 Regulatory Guide 1.75 Conplete 6/1/84 191 7.2.2.8 Scran discharge volume Canplete 6/29/84 192 7.2.2.9 Reactor mode switch Ccmplete 8/15/84 (Rev. 1) 193 7.3.2.1.10 Manual initiation of safety systems Cmplete 8/1/84 194 7.3.2.2 Standard review plan deviations Ccriplete 8/1/84 (Rev 1) 195a 7.3.2.3 Freeze-protection / water filled Complete 8/1/84 instrument and sanpling lines and cabinet temperature control 195b 7.3.2.3 Freeze-protection / water filled Complete 8/1/84 instrument and sanpling lines and cabinet tenperature ccntrol
. 196 7.3.2.4 Sharing of carmon instrument taps Couplete 8/1/84 197 7.3.2.5 Microprocessor, nultiplexer and Canplete 8/1/84 cmputer systems (Rev 1)
-198 7.3.2.6 1MI Item II.K.3.18-ADS actuation Open 199 7.4.2.1 IE Bulletin 79-27-Ioss of non-class Canplete 8/1/84 IE instrumentaticn and control power systen bus during operation 200 7.4.2.2 Renote shutdown systen Canplete 8/15/84 (Rev 1) i 201 7.4.2.3 RCIC/HPCI interactions Conplete 8/3/84 i
- l. 202 7.5.2.1 level measurenent errors as a result Complete 8/3/84 l of envircrnnental tarperature effects l On level instrunentation reference leg M P84 80/12 17- gs
ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL It OPEN SECTICN A. SCHWENCER ITEM NUMBER SUBJECT STATUS IEITER DPJED 203 7.5.2.2 Regulatory Guide 1.97 Cmplete 8/3/84 204 7.5.2.3 TMI Item II.F.1 - Accident nonitoring Cmplete 8/1/84 205 7.5.2.4 Plant process cmputer system Cmplete 6/1/84 206 7.6.2.1 High pressure / low pressure interlocks Cmplete 7/27/84 207 7.7.2.1 HELBs and consequential control system Cmplete 8/1/84 failures 208 7.7.2.2 Multiple control systs failures Cmplete 8/1/84 209 7.7.2.3 L W.it for non-safety related systems Complete 8/1/84 in Chapter 15 of the FSAR (Rev 1) 210 7.7.2.4 Transient analysis recording system Cmplete 7/27/84 211a 4.5.1 Control rod drive structural materials Cmplete 7/27/84 211b 4.5.1 Control rod drive structural materials Cmplete 7/27/84 211c 4.5.1 Control rod drive structural materials Cmplete 7/27/84 211d 4.5.1 Control red drive structural materials Cmplete 7/27/84 211e 4.5.1 Control rod drive structural materials Cmplete 7/27/84 212 4.5.2 Reactor internals materials Cmplete 7/27/84 l 213 5.2.3 Reactor coolant pressure boundary Cmplete 7/27/84 material
. 214 6.1.1 Engineered safety features materials Complete 7/27/84 215 10.3.6 Main steam and feedwater syst m Complete 7/27/84 materials 216a 5.3.1 Reactor vessel materials Cmplete 7/27/84 M P84 80/12 18- gs i
ATTACEMENT 1 (Cont'd)
DSER R. L. MITTL 10 OPEN SECTIm A. SCHWENCER I'1TM NLMBER SUBJECT STATUS IETIER DATED 216b 5.3.1 Reactor vessel materials Cmplete 7/27/84 217 9.5.1.1 Fire protection organization Cmplete 8/15/84 218 9.5.1.1 Fire hazards analysis Cmplete 6/1/84 219 9.5.1.2 Fire protection administrative Cmplete 8/15/84 controls 220 9.5.1.3 Fire brigade and fire brigade Cm plete 8/15/84 training 221 8.2.2.1 Physical separation of offsite Cmplete 8/1/84 transmission lines 222 8.2.2.2 Design provisions for re-establish- Cm plete 8/1/84 ment of an offsite power source 223 8.2.2.3 Independence of offsite circuits Cmplete 8/1/84 between the switchyard and class IE buses 224 8.2.2.4 Cm mon failure mode between onsite Cm plete 8/1/84 and offsite power circuits 225 8.2.3.1 Testability of autmatic transfer of Cm plete 8/1/84 power frm the normal to preferred power source 226 8.2.2.5 Grid stability Cm plete 8/13/84 Rev.
227 8.2.2.6 Capacity and capability of offsite Couplete 8/1/84 circuits 228 8.3.1.l(1) Voltage drop during transient mndi- Cdnplete 8/1/84 tions 229 8.3.1.1(2) Basis for using bus voltage versus Complete 8/1/84 actual connected load voltage in the voltage drop analysis 230 8.3.1.1(3) Clarificatim of Table 8.3-11 Cmplete 8/1/84 i
l l M P84 80/1219- gs I
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l ATTAONENT 1 (Cont'd)
DSER R. L. MITIL TO l OPEN SECTI N A. SCHWENCER '
ITEM NtNBER SUBJECT STATUS IEITER DATED 231 8.3.1.1(4) Undervoltage trip setpoints Cmplete 8/1/84 232 8.3.1.1(5) Ioad configuration used for the Cmplete - 8/1/84 voltage drop analysis 233 8.3.3.4.1 Periodic systen testing Couplete 8/1/84 234 8.3.1.3 Capacity and capability of onsite Complete 8/1/84 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 Omplete 8/1/84 test 236 8.3.1.6 Compliance with position C.6 of Conplete 8/1/84 10 1.9 237 8.3.1.7 Decription of the load sequencer 0:Inplete 8/1/84 238 8.2.2.7 Sequencing of loads on the offsite Cmplete 8/1/84 power systen 239 8.3.1.8 Testing to verify 80% mininun Couplete 8/15/84 voltage t
240 8.3.1.9 Ompliance with BTP-PSB-2 Canplete 8/1/84 241- 8.3.1.10 Load acceptance test af ter prolonged Cmplete 8/1/84 no load operaticn of the diesel generator 242 8.3.2.1 Chnpliance with position 1 of Regula- Otmplete 8/1/84 tory Guide 1.128 243 8.3.3.1.3 Protection or qualification of Class Ccmplete 8/1/84 lE equipnent fran the effects of fire suppressicn systems 244 8.3.3.3.1 Analysis and test to demonstrate Cmplete 8/1/84 adequacy of less than specified separation 4
4 l
M P84 80/12 20- gs I
l
ATTAOMENT 1 (Cont'd)
DSER R. L. MITTL 'IO OPEN SECTION A. SGWENCER ITEM NtMBER SUR7ECT STA'IUS LEITER DATED 245 8.3.3.3.2 The use of 18 versus 36 inches of Cmplete 8/15/84 separation between raceways (Rev. 1) 246 8.3.3.3.3 Specified separation of raceways by Cmplete 8/1/84 analysis and test 247 8.3.3.5.1 Capability of penetrations to with- Cmplete 8/1/84 stand long duration short circuits at less than maximum or worst case short circuit 248 8.3.3.5.2 Separation of penetration primary Cmplete 8/1/84 and backup protections 249 8.3.3.5.3 The use of bypassed thermal overload Cmplete 8/1/84 protective devices for penetration ptatections 250 8.3.3.5.4 Testing of fuses in accordance with Cmplete 8/1/84 R.G. 1.63 251 8.3.3.5.5 Fault current analysis for all Cmplete 8/1/84 representative penetration circuits 252 8.3.3.5.6 The use of a single breaker to provide Cmplete 8/1/84 penetration protection 253 8.3.3.1.4 Cmmitment to protect all Class IE Cmplete 8/1/84 equipment from external hazards versus only class lE equignent in one division 254 8.3.3.1.5 Protection of class lE power supplies Cmplete 8/1/84 frm failure of unqualified class 1E loads j 255 8.3.2.2 Battery capacity Cmplete 8/1/84 256 8.3.2.3 Autmatic trip of loads to maintain Cmplete 8/13/84 sufficient battery capacity I i
'i I
l M P84 80/12 21- gs l
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ATIAONENT 1 (Cont'd)
DSER R. L. MITTL 10 OPEN SECTION A. SOMENCER ITEM NLMBER SUBJECT STATUS LETTER DATED 257 8.3.2.5 Justification for a 0 to 13 second Omplete 8/1/84 load cycle 258 8.3.2.6 Design and qualification of DC Cmplete 8/1/84 systen loads to operate between mininun and maximan voltage levels 259 8.3.3.3.4 Use of an inverter as an isolaticn Cm plete 8/1/84 ,
device 260 8.3.3.3.5 Use of a single breaker tripped by Cmplete 8/1/84 a IOCA signal used as an isolation device 261 8.3.3.3.6 Autcmatic transfer of loads and Cmplete 8/1/84 interconnection between redundant divisions 262 ll.4.2.d Solid waste control program Open 263 11.4.2.e Fire protection for solid radwaste Cmplete 8/13/84 storage area 264 6.2.5 Sources of oxygen Open 265 6.8.1.4 ESF Filter Testing Cmplete 8/13/84 266 6.8.1.4 Field leak tests Cm plete 8/13/84 267 6.4.1 Control rocm toxic chemical Cmplete 8/13/84 detectors 268 Air filtration unit drains Open 269 5.2.2 Code cases N-242 and N-242-1 Open 270 5.2.2 Code case N-252 Open TS-1 2.4.14 Closure of watertight doors to safety- Open related structures 4
M P84 80/12 22- gs
1 ATTACIMENT 1 (Cont'd) 1 DSER R. L. MITTL 'ID OPEN SECTION A. S0iWENCER ITEM NLMBER SUILTECT STATUS IEITER IRTED TS-2 4.4.4 Single recirculation loop operation Open TS-3 4.4.5 Core flow monitoring for crud effects Ccmolete 6/1/84 TS-4 4.4.6 Inose parts monitoring system Open TS-5 4.4.9 Natural circulation in normal Open operation.
TS-6 6.2.3 Secondary containment negative Open pressure TS-7 6.2.3 Inleakage and drawdown time in Open secondary containment TS-8 6.2.4.1 Leakage integrity testing Open TS-9 6.3.4.2 ECCS subsystem periodic ccuponent Open testing TS-10 6.7 MSIV leakage rate TS-ll 15.2.2 Availability, setpoints, and testing Open of turbine bypass system TS-12 15.6.4 Primary coolant activity LC-1 4.2 Fuel rod internal pressure criteria Ccunplete 6/1/84 LC-2 4.4.4 Stability analysis subnitted before Open second-cycle operation M P84 80/12 23- gs
l ATTACHMENT 2 DATE: 8/15/84 DRAFT SER SECTIONS AND DATES PROVIDED SECTION DATE SECTION DATE 3.1 3.2.1 11.4.1 See Note 1 3.2.2 11.4.2 See Note 1 5.1 11.5.1 See Note 1 5.2.1 11.5.2 See Note 1 6.5.1 See Note 1 13.1.1 See Note 4 8.1 See Note 2 13.1.2 See Note 4 8.2.1 See Note 2 13.2.1 See Note 4 8.2.2 See Note 2 13.2.2 See Note 4 8.2.3 See Note 2 13.3.1 See Note 4 8.2.4 See Note 2 13.3.2 See Note 4 8.3.1 See Note 2 13.3.3 See Note 4 8.3.2 See Note 2 13.3.4 See Note 4 8.4.1 See Note 2 13.4 See Note 4 8.4.2 See Note 2 13.5.1 See Note 4
, 8.4.3 See Note 2 15.2.3 8.4.5 See Note 2 15.2.4 8.4.6 See Note 2 15.2.5 8.4.7 See Note 2 15.2.6 8.4.8 See Note 2 15.2.7 9.5.2 See Note 3 15.2.8 9.5.3 See Note 3 15.7.3 See Note 1 9.5.7 See Note 3 17.1 8/3/84 9.5.8 See Note 3 17.2 8/3/84 10.1 See Note 3 17.3 8/3/84 10.2 See Note 3 17.4 8/3/84 10.2.3 See Note 3 10.3.2 See Note 3 10.4.1 See Note 3 10.4.2 See Note 3 10 .4 .3 See Note 3 10.4.4 See Note 3 11.1.1 See Note 1 Notes:
11.1.2 See Note 1 11.2.1 See Note 1 1. Open items provided in 11.2.2 See Note 1 letter dated July 24, 1984 11.3.1 See Note 1 (Schwencer to Mittl) 11.3.2 See Note 1
- 2. Open items provided in June 6, 1984 meeting i 3. Open items provided in April 17-18, 1984 meeting
- 4. Open items provided in May 2, 1984 meting MP 84 95/03 01 2
i i
f OPEN ITEM DSER SECTION SUBJECT
.192 7.2.2.9 Reactor made switch 200 7.4.2.2 Remote shutdown system 217 9.5.1.1 Fire protection organization 219 9.5.1.2 Fire protection administrative controls 220 9.5.1.3 Fire brigade and fire brigade training 239 8.3.1.8 Testing to verify 80% minimum voltage 245 8.3.3.3.2 The use of 18 versus 36 inches of separation between raceways.
1 4
Y i
4
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, ,, - .- - - , - - -. ~ . , . , . ... . . ~ . ,. ~.
DATE: 8/15/84 ATTACHMENT 3 OPEN ITEM DSER SECTION SUBJECT 1 2.3.1 Design basis temperature for safety-related auxilliary systems and components.
2a 2.3.3 Accuracies of meteorological measurements.
2b 2.3.3 Accuracies of meteorological measurements.
2c 2.3.3 Accuracies of meteorological measurements.
2d 2.3.3 Accuracies of meteorological measurements.
3a 2.3.3 Upgrading of onsite meteorological measurements program.
3b 2.3.3 Upgrading of onsite meteorological measurements program.
8 2.5.2.2 Choice of maximum earthquake for New England-Piedmont Tectonic Province.
140a 9.1.2 Spent fuel 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.
142a 9.1.4 Light load handling system.
142b 9.1.4 Light load handling system.
144a 9.2.1 Station. service water system.
144b 9.2.1 Station service water system.
144c 9.2.1 Station service water system.
183 18 Hope Creek DCRDR
Attachment 3 (cont'd)
QUESTION NUMBER FSAR SECTION 210.20 3.6.2 210.50 3.9.3 230.8 2.5.2.2.3.~4 430.88 (Rev. 1) 9.5.4 i 430.94 (Rev. 1) 9.5.4 5
430.132 9.5.7 i
4 4
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i I
i Li
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l 1
4 ATTACHMENT 4 N
I 4
1
nm 26 '8402 6 3 339 HCGS
- e DSER Open Item No.1 ( DSER Section 2.3.1)
DESIGN-BASIS TEMPERATURES FOR SAFETY-RELATED AUXILIARY SYSTEMS AND COMPONENTS
, The applicant has considered treme outdoor temperatures of t
34.4*C (94*F) and -20.6*C in the design of the HVAC systems for all safety-rela buildings. The bases for the selection of these temperatures were the 14 probability of occurrence (summer) and 994 (winter) probability of occurrence l, values from the distributions ' presented by ASHRAE. , The appli-i cant states that in a typical summer, the design temperature
> would be equalled or exceeded 14 of the time (about 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> ) .
i Similarly, during a typical winter, a tenperature lower than the design temperature would be expected 14 of the time (about 1 .' 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />). The applicant has stated that the diesel generator i air intake, service water valves, main steam isolation valves, and impulse lines are in Seismic Category I structures maintained within acceptable environmental conditions by safety-
!, , related HVAC systems. The applicant also indicates that the
! exhaust pipe for the diesel generator is protected by a small hood to prevent blockage. Extreme temperatures of 41.7'C ,
(107 'F) and -26. l *C (-15 'F) have been reported at Wilmington.
Temperatures in excess of 32.2*C (90*F) are expected at
. Wilmington about 19 days each year. Temperatures of -17.8'C (O'F) or lower are expected less than about 1 day each year at l Wilmington. Conditions with return periods of 100 years are l normally considered for design of safety-related auxiliary
! systems and components. The 100-year return period extreme temperatures in thw Hope Creek area are approximately 42.2*C j (108 'F) and -29.9'C (-20
- F ) . Further justification on the adequacy of the extreme temperatures considered by the applicant for the design of safety-related auxiliary systems and components
- is required.
I i Large-scale episodes of atmospheric stagnation occur in the Ii region. About 31 atmospheric stagnation cases totaling about i 128 days were reported in the area in the priod 1936-1970. Two of these cases lasted 7 days or more.
i As discussed above, the staf f has reviewed available information j
! relative to the regional meteorological conditions of importance
!, to the safety design and siting of this plant in accordance i! with the criteria contained in SRP Section 2.3.1. Based on lI this review, the staf f concludes that, with the exception of j design basis temperatures for auxiliary systems and components, j the applicant. has identified and considered appropriate regional t meteorological conditions in the design and siting of this j plant, and, therefore, meets requirements of 10 CFR 100.10 and GDC 2. The design-basis tornado characteristics selected by
- the applicant conform to the position set forth in RG 1.76 and therefore, meet the requirement of GDC 4 to d6termine an acceptable design basis tornado for missile generation.
- h 1-1
ut 26 '8402 63 33 9 l
RESPONSE
The 100-year return period extreme temperatures are considered overly conservative for HCGS safety-related system designs.
Section 2.3.2.1.2.1 compares onsite temperatures with those of Wilmington. Onsite summer maximum temperatures are slightly lower [2.2*C (4*F)] while winter minimums are slightly higher
[2.8'C ( 5 'F)] than at Wilmington. , . j; Outside d; sign tanperatures for safety-related NVAC xiliary
!. systems and components [34.4 *C (94 *F) and -20.6 *C ( are considered long tena, steady state values. They are conserva- I tively assumed to occur continuously during summer and winter, ! while extreme maximum and minimum temperatures are transient values of short duration. A. review of onsite diurnal temperature variations listed in Table 2.3-11 reveals hourly mean temperature ranges vary from
- 3.8'C (6.8*F) in December to 6. 4 *C (11.5'F) in May. Variations
' in mean values conservatively represent actual daily temperature variations. The daily range in - temperature normally used for
- HVAC system designs at Wilmington is ll.l *C (20 *F) as indicated l in ASHRAE Handbook of Fundamentals - 1977. The maximum and minimum diurnal temperatures occur between 2 to 4 p.m. and 6 to 8 a.m. respectively. On any given day it is concluded that
- the daily maximum and minimum temperature will occur for a
{ period of no more than 6 hours with two transition periods between the maximum and minimum values of 6 hours each.
- Outside air temp atures af fect HVAC system designs in three N(
ways:
- a. Building envelope heat gains or losses are a function of l inside and outside temperature dif ferences and building
! construction. 4 l b. Heat gain or loss due to infiltration air entering the butilding at other than space design temperatures. I c. Heat gain or loss due to conditioning outside ventilation X air to system design temp f tures. Building envelope heat gains or losses due to 100-year return l t period extreme maximum or minimum outside temperatures will
/C have litti; :: no effect on safety-related systems. A significant amount of the HCGS buildings are constructed below grade thus protected from outside air temperatures. Exposed walls and . roof s for safety-related structures are constructed of thick - concrete which has large heat capacity. These massive surfaces i act as " thermal flywheels' as they absorb heat from or liberate heat to the outside air as its temperature fluctuates. Therefore, it can be concluded that internal space temperatures will not be af fected by transient conduction heat gains or losses through building external surf aces.
1-2
~!
. ..- - - - - - - - - - . .... = _ -.-. -_ --
J . Am 26 8402 6 333 9 RESPONSE -(Cont'd) Infiltration air flow rates are a function of building construc-tion with _ respect to openings and cracks in exterior surfaces. l HCGS is designed as a low leakage structure with minimal outside j openings. In relation to the building volume, infiltration is insignificant. Therefore, space interior temperature is unaf fected by changes in infiltration loads. A y Extreme temperatures will Xf fect outside air ventilation loads. A cross reference to safety-related HVAC systems was provided
- in response "e" to Question 451.5. Of these systems, the following are not af fected by outside ventilation loads:
control area battery exhaust; filtration, recirculation ventila- ,l tion; equipment area cooling, SDG room recirculation; and ' diesel area safety-related battery room exhaust. These systems l are unaf fected because they either recirculate or exhaust 1004 l of internal room air. Summer cooling and winter heating conditions for af fected systems are discussed below. 1 HVAC cooling equipment is designed to of fset all system internal
- i. heat gains including: equipment , lighting, piping, electric l cable, etc. , in addition to external building envelope and 1 outside ventilation loads. External loads are a small fraction of the total cooling system load. Safety-related equipment is '
!. located in areas with minimal or no exposed surf aces, and for ) ! the " thermal flywheel" reason cited above, building envelope 3 heat gains are negligible. The following safe ty-related systems
! mix outside air with space return air and condition this mixture ,
i using heating and/or cooling coils: Control Room Supply, Control Room Emergency Filter System, Control Equipment Room
! Supply, Switchgear Room Cooling, and Diesel Area Class lE Panel
- Room Supply. The percentage of outside air entering these j systems is less than 20% of the total system supply flow. An
! increase in outside air temperature from 34.4 *C ( 94 *F) to 42.2*C (108'F) would result in no more than a 1.6*C (2.8'F) increase in the mixed air temperature entering the cooling coils. Des ign margin in the cooling coils will result in less than a 1.6*C (2.8'F) increase in the supply air temperature. Thus, the t resulting space temperatures would increase less than this amount. Transient temperature increases of this magnitude and duration would not adversely af fect the operation of safety-related equipment.
The service water intake structure ventilation system relies on 100% outside air for heat removal in the intake structure. Intake structure space temperatures would increase proportionally with increases in outside air temperatures. The continuous
; duty service temperature for certain intake structure equipment l will be exceeded for a short transient period when the maximum 1 100-year return period outside temperature is considered.
Higher toom ambient temperature would not cause a prompt failure
, of any safety-related equipment in this area. .Rather, higher room temperature would tend to shorten the long-term life of the equipment. Because room ambient temperatures will be at or i below the outside design temperature 99% of the time, thermal agi.ng would be insignificant.
i i lf 1-3 ,
APR 26 '8402 6 3 3 3 9 RESPONSE (Cont'd) Y j The 100-year return period extreme temperature sited -29.9*C (-20*F) would have no adverse ef fect on safety-related auxiliary systems or ccuponents. HVAC heating equipment is designed to of f set the entire building heat loss assuming no credit for equipment, lighting or other internal heat gains. This condition occurs only during plant shutdown for ref ueling or maintenance. Thus, with the exception - of some internal inline duct heaters, heating equipment is not l s a fe ty-re lated . The minimum building inside design temperature l for most areas is 4.4*C (40*F). However, heating equipment has a significant reserve capacity since it has been selected with the capability to provide an inside design temperature of 15.6*C ) (60*F). Credit for equipment, lighting, and other heat gains ' and the reserve capacity of the heating equipment would more than offset the increased heating load. There fore , the building minimum design temperature can be maintained during this transient extreme condition. 9 i 8 1 l t K51/2-17 1-4
Re<1 DSER Open Item No. 2a (Section 2.3.3) Accuracies of Meteorological Measurements The applicant states that the satire 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 revised Question Response 451.6, FSAR Section 2.3.3.3 and Table 2.3-29a, b and c. l
BCGS FSAR 12/83 00ESTION 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 l d;termine the representativeness of "the key meteorological l 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 ; Plcnt. These data can also be compared to data for Hope Creek if different meteorological measurements programs are in use for 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 possible observations for each period of record, 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 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
c) 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 wind distribution was derived from January 1970 to May 1971 data. Annual wind direction distribution for the same three - levels for the period January 1977 to December 1981 are - presented in Tables 451.6-2, 451.6-3 and 451.6-4, g espectively. ; l l COMPARISONS l 33 feet Highest wind direction frequencies from the period 1969 to 1971The (SGS) compare favorably with those from 1977 to 1981 (HCGS). cite has a bimodal distribution. SGS data shows the highest frequency of wind directions are SE-SSE-S and W-WNW-NW. BCGS data shows the same pattern. Frequencies other than these modes For all i l cre evenly distributed throughout the compass points. individual years, the data recovery rates are above 90 percent. s
- _-_- m a x.mun__ A 451.6-1 Amendment [
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 facilitate the construction of the Hope Creek Station and the' cooling tower.
M P84 93/04 3-dh DSER OPEN ITEM Od < :L . - - . . _ . .-. - .
BCGS FSAR 10/83 4 Monthly and annual joint frequency distributions of wind speed cnd direction, based on atmospheric stability classes, are rcierenced in Section 2.3.2.1.1. The 5-year data base containing
' i hnvely site meteorological data from January 1977 to December !
1981 was used as input in the analysis. l 2.3.3.3 Operational Data Display
~ l l l
The meteorological parameters required by Regulatory Guide 1.97 . will be incorporated in the data base to be included on the control room integrated display system (CRIDs) computer. The display of those parameters will be available as part of the
-display function along with all other related Regulatory Guide 1.97 variables.
The radiation monitoring system central radiation processor (CRP) l c: puter 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- I temperature stability indicators between 300- and 33-ft and 150- i and 33-ft, as well as precipitation, barometric pressure, solar j radiation, and ambient temperature at 33 ft. I
; Atrospheric transport and diffusion during normal operation will i
bo calculated by the CRP. A method for determining atmospheric 1' transport and diffusion throughout the plume exposure emergency planning zone during emergency conditions is being developed. TncNT {5 , 1 2.3.4 SHORT-TERM DIFFUSION ESTIMATES l l , 2.3.4.1 Obiective ; l Th3 objective is to provide conservative and realistic short-tera i cotimates of relative concentration (X/Q), at both the site boundary and the outer boundary of the low population zone (LPZ) following a hypothetical release of radioactivity from HCGS. The ccssssment is based on the results of atmospheric diffusion ,. modaling and onsite meteorological data. A ground-level accidental radionuclide release from HCGS is cnolyzed at various distances. Conservative and realistic X/Q values at the exclusion area boundary (EAB) are derived for the DSER OPEN ITEM dQ 2.3-27 Amendment _____i_______-___..__ Mf
1poset"1 b 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 from 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 9826a Computer, 3497A Data 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 fif teen minute averages. The pri-mary DAS collects data from the meterological parameters listed in Table 2.3-29. The backup DAS collects wind speed and direction from the the three tower elevations and two delta T's, as well as the backup meterological tower. The data acquisition system calculate a sigma theta for each of the three level wind directions. I The communications computer which consists of a 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 CO11 col Room ~
the Hope Creek Radiation Monitoring 9(sU n via a meteorological system link (which int ',ro ates a HP9915 computer) as well as three d %) _a ports. j The communication computer also supporto a display unit in the the Hope Creek EOF as well as communi-cation to the primary and backup DAS. i System accuracy is presented on Tables 2.3-29b and 2.3-29c. 4 s e DSER OPEN ITEM [Q 4
$ y i
\
I b passar S 2
- h. M ) i The postoperational data collection program also includes an additional meterological tower identi-fies as a backup meterological tower, consisting The backup tower is of a 10 meter telephone poll.
i located aproximately 500 feet south of the primary j ' meterological monitoring tower. Backup meterolo-gical data provides wind speed, wind direction, Backup meteorologi and a computed sigma theta. data provides wind speed and wind direction and a computer sigma theta. h C.R.9 dQi NAbEclo % rv4M t-u 4 w.a un s. w,s a RY:dh 5/30/84 M P84 9 3/04 2-dh l 1 1 l l DSER OPEN ITEM 4
BCGS FSAR TABLE 2.3-29 (cont) page 2 of 3 Reaght these sumer asse, Ynatrumment and Characteristic stria Be m eere et Amanes merameter Recorded Farasyster C11 met - sm> del 011, 3 cup Soter11mo = Angen 33 estad speed wine spee4 Threshold 0.6 spes, Model 842S en-te r. distance constant <5 ft, operating range O to 110 mph, accuracy aIE or 0.15 mph, tettichever le greater tRied directies C11amot - Dudel 012-10 wia4 Boterline = Se p 881a4 digestion Itseel 183S vane. Threshold 0.75 mph, distance constant <3.3 f t, Samping ratio 0.4 , Temperatasse=418ferecht141 T300*1 33888 T150"T33sa3 Den point
- EG6G stodel SM 110 accuracy *0.5*F teostaentes M113 gens pe&at climet - eedel 016-1 sootor- Emede S gest keep 7 ,- ^___ " 7
- Temperatuse gr==he aspirated temperature shield ,
with clim.t 015-3 thermistor gaelt&-Geist t I accuracy mo.15'c Rosenetric parecesse C11 met - Model 014-90 pressere meter 11em = Ra p i 4
- Sar m as prosesse transducer. Range 2e-32 La. asg geoept &
j ! Ester 11me = Sagen amawana Salafall ^^ : - ' 050 ; i- --; r&TE guTW
- 3' lW RfI moost 3oz, y .rp.w Itodel &
D-c ae f A cc atae v a.o t . oes } { Gal 9emperatene taken as part of temperatese differmetial smeasurement T3 co - T333 - 438 gneparatase taken as part of temperature differeattaa maamerament Ti so - T 3 j 438 poisse C11 met 019=3 tenermister. Accaracy 30.1*C. j _. l DSER OPEN ITEM jQ 4
- Amov.womm.r Jr
I HCGS FSAR TABLE 2.3-29a ' DATA ACQUISITION SYSTEM HARDWARE l MANUFACTURER MODEL QUALITY DESCRIPTION Hewlett Packard 9826A 2 Computer Hewlett Packard 98256A 2 256K-Byte Memory Expans ion Hcwlett Packard 98626A 4 Serial Ports Hewlett Packard 3497 2 Data Acquisition / Control Unit 20-Channel Analog Hewlett Packard 44421A 2 Multiplexor 44425A 16-Bit Status Input Hcwlett Packard 2 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 10 3BA 1 CRT DEC 1 Serial ({grts ' Bell 212A 5 Modem Bell 202T 6 Modem (1) (1) Or equivalent modem l l DSER OPDi ITD4 cO
O
.3 M
N O . _ . . _ . . , _ , ,,
'O t4 21 M
NCGe FSAh thele 2.3-29t> SYSTEM ftEASUREMENT ERMON DELTA TEMPEhATtic t: W1HD SPEED WINO DIRECTIOel (300-313 4150-n3i TEMPEhATURE DatfpOINT PSSCIF1TatEOS CWWeseENT
*-a (tempel (30 stral (100 Isrul ( * *=?al (DECRrts CEt.s s p:il torcMErs cELsIUsl ( ~ --seCELStus) I w"* l senser + 0.15 +, .30 + 1.40 +3 + 0.10 . U.lu 0.10 + 0.5 .01 Translater + e.21 + e.21 + e.21 0 - - -
e.est - SWIt + 0.0835
- 4.8465
+ 8.017 + 0.092 + 0.0026 + u.uult i u.013 - -
0 0 0.00 0 o 0 0 - Seitsare 8 gewr - - - - - - \ 8.3635 0.5165 1.227 + 3.092 + 0.1026 1 o. a'u 26 1 u.113
- 4.548 .81 Total sees taus 4
ersor 1 .tol t.- .1
- . 21 ..n 1.0, 3.00 1 0.1., 1 ....
l re .re.r a. 5.0 + 0.15 + 0.45 1 0.5 + 1.5 .01 O.G. 1.23 8.5 0.5 - _ sees 18Scation (1) Isotrisasetettee type and' specification provided en Table 2.3-29 and 2.3-29a. I (2) The lastantaneous error for wtad speed measuremente, assuming the individual component errors are addi tive and independent t root sua equare erserl, le ' uithin the R.G.1.23 specificatiese for all wind speeds less than 45 mph. The error us time averagal wind speeds will be less then the instantaneene ! rest esse egeare error (this statement le ep11 cable for all ether parameters in thne dLacuselon). Therefore. for wind speeds coasteeret to be et l gritiset far disposoien calculattene, the estimated error is well within the m.G. 1.2J specification. j RFT a go C3 704 80/11 1-ge
~
j . l 1
- i i
1 4
~
BCGS FSAR Table 2.3 - 29c ARTIFICIAL ISIAND DIGITAL DATA ACQUISITION SYSTEM ACCURAC'IES The following system accuracies are based upon VEN00R accuracy specifications and the following conditions: o 1 year calibration interval o 5-1/2 digits displayed on OVM o Auto Zero ON VOLTMETER ACCURACY . ERROR PERCENT PLUS RE50L'JTION R.UlGE (V) :F READING ERROR '"V)
.ll9999 .015 .003
- 1.19999 .015 .02 11.9999 .015 .1 PARAMETER ERROR DAS INPUT ERROR MAXIMUMa DVM ENGINEERING CALCULATION DAS PARAMETER RANGE VOLTAGE UNITS POINT ERROR Temperature 1.19999V 0-1.0V +45*C 45*C 0.013*C Delta-Temperature 1.19999V 0-1.0V +10*C 10*C 0.0026*C Dew Point 11.9999V 0-5.0 +100*F 100*F 0.021*F Wind Speed 1.l9999V 0-1.0V 0-100 mph 50 mph 0.0095 mph l Wind Speed 1.19999V 0-1.0V 0-100 mph 10 mph 0.0035 mph l Wind Speed 1.19999V 0-1.0V 0-100 mph . 30 mph 0.0065 mph Wind Direction 1.19999V 0-1.0V '0-540* .
540* 0.091* Precipitation 1.19999V 0-1.0V 0-l* - 0.00=b Pressure 1.19999V 0-1.0V 28-31"Hg 32Hg 0.00068"Hg Solar Radiation 1.19999V 0-1.0V 0-2Ly/ min 2Ly/ min 0.00034Ly/ min 4The data acquisition system error is due entirely to HP-3497A instrument l error. Software calculations are computed to 12 significant digits. 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. i .. .. DSER OPEN ITEM Nd
- I
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iI i Q,1 DSER OPEN ITEM [G _w,- - ,-n . . , , - - , n m.
DSER Open Item No. 2b (Section 2.3.3) The applicant's method for detemining vertical temperature gradient is uncommon, using a matched pair of therwistors. Additional information is required from the applicant to demon-y strate that the accuracies of meteorological seasurement comply with the system accuracy specifications presented in RG 1.23.
Response
For the information requested above, see the response to DSER Open item Ja.
- . - . . . - -- _ -.,s_,w v'-w-- -- -e-w ---e.----- - - - - - - ---w eme- n,ier-wwr+--w-,w--y evr--r+ww,ww w ee-.W,-,e--ev-
2 81/ 4 i and 0 l DSEROpenItemNo.2c((Section2.3.3) i; l The meteorological measurements program, during plant operation, will include those parameters currently measured. Meteorological parameters are to be available for display through the radiation monitoring system central radia-tion processor (CRP), altheagh 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
For the information requested above, DSER Open item .2a see the response to 4 e 1 l 1 l
?CV h DSER Open Item No. 3asb (Section 2.3.3)
Upgrading of Onsite Meteorological Measurements Program (III.A.2) To address the meteorological 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 measurements program to meet the criteria in NUREG-0654, Appendix 2. " Criteria for Preparation and Evaluation of Radiological Emergency 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
- into a real-tism atmospheric dispersion model for dose assessments will also be considered L. part of the upgraded capability.
Response
j-h e r e spon s c. T o t h.'s item u.r ill b e. pro vo d ed b y b e c.cm h er, / 9 84 )* er th e CommIdm en t m a ele. in the m/ pac // 3 0,198V /eMer fo A. Sc4 wcncer
)C, o n R. t. . M.* HA.
I l
i 2).scR OPEA I TEM NO & ( 3 e c.f o n ?. 5 2 A )
- c. Ho t et of maxim um EnRracunkt foR Nw EastnNo -
P'EbMdNT TEC-rows c. PRovin ce l As noted in NRC question Q230.6, the staff and the applicant are presently assessing the choice of maximum earthquake for the New England-Piedmont tec-tonic province. The applicant argues that the 1982 New Brunswick earthquakes are associated with a conjugate shear fracture system and that focal mecha-nisms are consistent with this interpretation. Also the applicant argues the 1982 events are not out of character with the region's moderate level of sais-cicity. Since comparable situations are not known in the HCGS area, the appli-cant finds it is overly conservative to assume the 1982 New Brunswick earthquake can occur in the New England-Piedmont province near the HCGS site for purposes of evaluating the HCGS seismic design. ,The staff is reviewing the applicant's arguments and evaluating the possibility that a magnitude 5 3/4 event similar to the 1982 New Brunswick earthquake could occur in other parts of the New l England-Piedmont tectonic province. Because this issue has not been resolved, a final determination of the seismic design input adequacy can not be made at ; this time. Response -
)
AofdHicn a./ in for ,v, aron o n % /.s He r>, has been ProJo Ned in the re.,pon s e s +o QuesHons ;3o.W and a 30. 9. l l I l I _ - - _ - - _ _ - - - , _ - _ , _ - . _ _ _ . , _ _ _ , - _ ~ _ . _ _ _ _ _ . _ . _ , _ _ _ , -
9 Question 230.8: Question 230.6 addressed the January 9,1982 magnitude 5-3/4 New Brunswick, Canada earthquake and the effect of the New Brunswick earthquake on the Hope Creek site. The revised FSAR discusses the 1982 New Brunswick earthquake sequence and states "the 1982 Miramichi earthquake sequence was not out of character with the region's previous earthquake history" (FSAR, p. 2.5-8'0). Is the New - Brunswick region more seismically active than the Hope Creek site areat Compare seismic activity rates and recurrence models derived l for alternative size regions around both of these areas.- (See, for 4 example, the Shearon Harris SER, NUREG-1038 or the Millstone Meeting Summary contained in a February 1,1984 letter from W. - Counsil to B. Youngblood). Response: The seismic activity rates and recurrence models for several alterna-tive sized regions surrounding the Hope Creek site and the epicentral region of the January 1982 Mirimichi, New Brunswiuc earthquake have been compared. This comparison was done in the following manner: I
- 1) earthquakes listed in Table 2.5-1 of the FSAR for which no i mangitudes were available (i.e. historical events prior to about 1950) were converted from Intensity to magnitude using Nuttli and Herrmann's (1978) relationship m b = 1.75 + 0.5 I, (1)
- 2) other seismicity data (i.e. events listed in Earth Physics Branch Files or North Eastern U.S. Seismic Network Bulletins) where magnitude scales other than m were used to characterize b
seismic events, were considered to be numerically equivalent for purposes of this analysis.
- 3) a recurrence model was constructed for both the Hope Creek region and the area surrounding the New Brunswick magnitude l 5.7 event of the forms l DSER OPEN ITEM [
k__.___.____.-._..__.-_- - --
~ ~
!^ Log N(k M) = a - bM" 1
- and normalized to 104M1 2 for direct comparison of earthquake density in a .5 x5 area centered about both the Hope Creek 4 site and the New Brunswick event (see Figure 1 and Table 1).
- 4) a qualitative comparison of a 1 x1 area about both areas was also made to reflect the relative numbers of instrumentally recorded earthquakes of varying magnitudes over equivalent recording periods at both sites (see Table 2).
Discussion: A 5'x5 area surrounding the Hope Creek site and the Mirimichi, New Brunswick magnitude 5.7 earthquake epicenter was selected as , being broad enough to represent seismicity on a regional level. Events were compiled from Table 2.5-1 of the ESAR for Hope Creek I and from data files of the Earth Physics Branch, Dominion Observa-tory, for the Mirimichi region (Adams,1984, personal communica-tion), not including the Jan~uary 9,1982 Miri~michi event or its aftershocks. Table i provides a summary of the recurrence para-meters for both areas and the recurrence curves are shown on s Figure 1. Assuming that the seismicity is uniformly distributed in both 5 x5 i regions, the probability that an event as large as M= 5.5 occurring within a 16km radius about each site would be: l Hope Creek - 1 in 41,600 yrs i Mirimicht, New Brunswick - 1 in 20,200 yrs The fact that the population density in central New Brunswick is low and that earthquakes have only been routinely recorded in that region
^
i for the last decade or so may reflect an even greater difference in . seismicity between Mirimicht and the Hope Creek site. . As a further comparison between the two areas, a 1 x1' area, ! centered on the Hope Creek site and the Mirimichi 1982 epicenter, osEa oPEN ITEM [ 4
- _ . _ _ . _ , . . . _ _ _ . _ _ , _ - . _ _ , , . , - . . _ . . .,_..__,___,._,._m._ , . - , _ . . _ _ . . _ , _ ~ , , _ _ . . , _ _ _ , _ . , , _ _ , _ _ _
n. I 1. was selected and earthquakes occurring within both the areas were i- determined. As earthquakes have been instrumentally recorded since ! about 1930 in eastern Canada, this date was used as the low cut off in both data sets available. Table 2 shows the comparison; for the Hope Creek areas, there have been a total of 8 events since 1930-2 events between magnitudes 2.0-2.9, 2 events between magnitude 3.0-3.4, 3 events between magnitudes 3.5-3.9 and one event between magnitude 4.0 and - 4.4. The 1 x1" region s,urrounding the,1982 Mirimichi epicenter, on the other hand, shows a greater number of events (25) than Hope Creek, not including the 1982 Magnitude 5.7 event and aftershocks. There have been recorded (since 1930) 14 events of magnitude 2.0-2.9; 4 between magnitudes 3.0-3.4; 5 events between 3.5-3.9; 1 earthquake between magnitude 4.0 and 4.4 and 1 event between magnitude 4.5 and 4.9. It is clearly shown that the Hope Creek site is within an area of significantly lower seismicity than for the Mirimichi, New. Brunswick Region.
.e .
9 DSER oPEN ITEM { 3
TABLE 1 RECURRENCE PARAMETERS 2 N/104mi 2 R @ on Area (mi ) 14.0 t 4.5 2 5.0 25.5 "b" 4 Hope Creek (5 x50) 9.2 x 10 - 0.03 0.01 0.002 - 0.67 l 4 New Brunswick (5"x5 ) 8.1 x 10 0.02 0.01 0.003 0.0007 0.85 (w/o Mirimichi Events) TABLE 2 SEISMIC EVENTS WITHIN A 1 x1 AREA
. CENTERED ABOUT THE HOPE CREEK SITE AND THE MIRIMICHI MAONITUDE 51 EPICENTER _
Hope New Creek
~
Brunswick Magnitude (1975 11580) (1930-1981) 2.0 - 2.9 2 14 3.0 - 3.4 2 4 3.5 - 3.9 3 5 4.0 - 4.4 1 1 4.5 '4.9 0 1 TOTAL 8 25 l 1 DSER OPEN ITEM 7
C 10-1
~
o . 10-2
)(
cv E w - o L 2 2 ~ 3 > 2 Mi rimichi New Brunswick 0 m 5 x5 Area Lc g N(2M)= 1.5-0.8 5 M
-3 F ope Cree k 50x50 A es ~
Log N(2M) = 0.66-0.67M - e 10~4 3 4 5 6 7 Magnitude DSER OPEN ITEM [ Recurrence Curves
I O Ouestion 230.9 In Question 230.6 the staff recommends developing a site specifle response spectrum with foundation conditions similar to the Hope Creek site. FSAR Section 2,5.2.6 compares a site specific spectrum developed oy Lawrence Livermore National Labora-
. tories (LLNL) to the Hope Creek design spectrum.
Compare shear wave velocity profiles for the re-cording sites examined by LLNL with the profile for the Hope Creek site. If any of the recording sites are not similar to the Hope Creek site, recalculate the site specific response spectrum without those - records.
Response
Discussions with Lawrence Livermore National Laboratories (LLNL) have revealed that the recording sites (where records were obtained by LLNL for developing the site-specific spectrum that was used to compare with the Hope Creek design spectrum) do not have shear wave profiles available. Therefore, no comparisons can be made to the shear wave profile at Hope Creek, nor can any recalculation of the site specific response spectrum be accomplished. DSER OPEN ITEM p j
,' nCoS a ao 840268 64 0
- DSER Open Item No. 140 (DSER Section 9.1.2) i ~
SPENT FUEL STORAGE Since the applicant's applidation for an' operating license was docketed in 1983, which is af ter the Novembgr 17, 1977 date specified in the SRP, the applicant must provide the results L of an analysis which shows that a failure of the liner plate as
- a result of an SSE will not cause any of the followings (1) significant releases of radioactivity due to mechanical i damage to the fuel (2) significant loss-of-water from the pool which could uncover the fuel and lead to release of radioactivity 1 due to heat upt (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 leakager 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 requirements 2 of General Design Criterion 2, " Design Bases for Protection l Against Natural Phenomena," and the guidelines of Regulatory j 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.
- The applicant has not provided the design details of the spent i fuel storage racks, the results of an analysis of impacts onto i the racks, the bundle to bundle spacing, the design maximum
- enrichment (weight percent of U235), a description of calculational methods used for criticality analysis (along with the results), a tabulation of the nomin11 value of K gg of the racks along with the various uncertainties and biase,s considered 4 in the analysis, and a tabulation of the reactivity effect of i each of the abnormal accident situations considered for our 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 l of the fuel. As an alternative, the applicant can provide the results of the criticality analysis without taking credit for
! 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, Positions C.1 and C.4, concerning fuel storage facility design
- are satisfied.
e
- 140-1
"o -
- m. ao 8402 68 6 4 0 DSER Open Item No. 140 (Cont'd)
. We cannot conclude that the spent fuel storage facility is in
! conformance with the requirements of General Design Criteria 2, l 61, and 62 as they relate to protection of the spent fuel l against natural phenomena, radiation protection, and prevention of criticality ant. the guidelines of Regulatory Guides 1.13, Positions C.1, C.3, and C.4 and 1.29, Positibns C.1 and C.2, relating to the iacility's design basis and seismic classification. 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 supplement 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) Provide test or operating data showing that the neutron poison material will not degrade during the lifetime of I the spent fuel storage pool. C - (3) Provide a description of any materials monitoring program for the pool. In particular, provide information on the f requency of inspection and type of samples used in the monitoring program. (4) Provide details of the spent fuel racks to show that no buildup of gases will occur in the cavities containing the poison materials. i RESPONSE 4 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 outside 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
. , , I
RESPONSE (Cont'd) "\ of the stiffener system in holding the liner plate against seismic loads is at least equal to the form pressure load. Th estimated
' test load during construction (appgoximately 300 lb/f tg) was lower than the design value of 690 lb/ft . This construction load induced a correspondingly lower stress in the stif fener-to-liner welds. .s' An analysis, performed to evaluate the ef fect of SSE loads on the liner, shows that the resultant stresses would be insignifi-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. Thus, the design of the liner plate satisfies General Design Criteria 2, 61, and 6 2, 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 Regulatory Guide 1.13 Position C. 3, as described in Section 9.1.4.6 and 9.1.5.6. . The spent fuel storage rack design details have been provided in the response to Questions 81.2, 281.13, 410. 3 9 and 410. 4 2. 9he inf crz:t ion ::;2c;;;d in 0"==?i;;; 2:^.1 ...a 410.30 111 be-
- pcevided ti Tr;trrt;r, 1^0". This information wwtL supportsthe Selsm
- C and criticality reviewsand demonstratesthat 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. 1 l 140-3
.. *e RESPONSE (Cont'd) -
". Similar rack designs, with vented Boral poison in stainless steel l racks, have been licensed and have proven successful. HCGS's l maximum anticipated radiation exposure for the Boral is 5.12 x 10 d rads. Similar Boral specimens i toaccumulatedradiationdosesypto7x.10ggvebeensubjected A rads at the 1
University of Michigan's Ford Sctor. These, specimens were found to be structurally sound and neutron attenuation capabilities were not degraded by irradiation.) g l In order to continually assure the dequacy of the poison material, test coupons are provided for a Bora surveillance program. Forty-five coupons are installed i igh 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 4
fuelpoo1{,aBoralsurveillanceprogramisnotplannedatHCGS. such as a r%+ tee llo ud Brown 's Fe r>% . If infor--tier frer there leed pi;nt; infierter :ni p;;bler. yith th; Scr:1, Otrerill:::: ;;;;;;; ::: th;n M initie;ed. v
- l The spent fuel rack poison cavities are vented to prevent any buildup of gases. Response to Question 281.13 ' *~
provides further ] information on v,enting. ,,, . t i P.sCG w;// developf a .. '.'s p ry ra m -to m o n & the som. Gor a l s w ve-o'//a nc e progrw a $ ed %e r F'e ern b M on h'c.e.J/o o ,- Browds Ferry b y Marc.h , /193~ l %e respera e -/* guesWon ast.sy has s een re voied +o r e Fleet %;s esspon se 1 4 ( )
.' l I' i 140-4 i
i , HCGS JSAR 6/84 - 3.8.4.8.3 Spent Fuel Rack Design l [ Acceptance Criterion II.4.f requires that the spent fuel racks be j designed in compliance with Appendix D of SkP'A.8.4, which I requires that construction materials should conform to Section III, Subsection NF of the ASME Code.
> /NSERT C=
The spent fuel racks are constructed of ASTM A-240 and ASTM A-564 - stainless steel. The A-240 and A-564 material specifications are identical to the ASME SA-240 and SA-564 material specifications. All rack steel is supplied with certified material test reports. The rack materials are procured under a Q.A. Program that is - intended to comply with:
- a. 10CFR50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants".
- b. ANSI /ASME M45.2, " Quality Assurance Program- I-Requirements for Nuclear Facilities", and
- c. ANSI /ASME NOA-1, " Quality Assurance Program Requirements for Nuclear Power Plants".
3.8.5 FOUNDATIONS Foundations for all Seismic Category I structures and the turbine building and the administration facility, which are non-Seismic Category I structures, are described in this section. 3.8.5.1 Descriotion of the Foundations The configuration of the foundation mats for the various structures is shown on Figure 3.8-37. Reinforced concrete mat foundations are provided for all structures. Except for the station service water system (SSWS) intake structure, the mats rest either on the Vincentown Formation or on engineered structural backfill placed on the Vincentown Formation. The mat and the lean concrete leveling i l 1 3.S-48b Amendment 6 osan oPEN ITEM /Y .
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. 1 HCGS FSAR 12/83 CHAPTER 9 l
tem l - Table No. Title 9.1-1 Fuel Pool Cooling and Cleanup System and Torus Water Cleanup System Design Parameters 9.1-2 Fuel Pool Cooling and Cleanup System He'at Removal Capacity and Makeup Requirements 9.1-3 Fuel Pool Cooling and Cleanup System and Torus Water Cleanup System Failure Modes and Effects . Analysis 9.1-4 Tools and Servicing Equipment 9.1-5 Fuel Servicing Equipment 9.1-6 Reactor Vessel Servicing Equipment . 9.1-7 In-Vessel Servicing Equipment 9.1.8 Refueling and Storage Equipment 9.1-9 Under Reactor Vessel Servicing Equipment and Tools 9.1-10 Overhead Heavy Load Handling System Data Summary 9.1-11 Reactor Building Polar Crane Data 9.1-12 OHLHS Loads Over Safety-Related Equipment 9.1-13 Reactor Building Polar Crane Design Comparison With NUREG 0554, Single Failure Proof Cranes for Nuclear Power Plants 9.1-14 Hope Creek Polar Crane Special Lifting Devices and Slings 9.1-15 Refueling Floor Heavy Load Height Restriction l y 9.1-16 Not Used 9.1-17 Spent Fuel Pool Liner Drain Lines l g j 9.1-18 Decay Heat and Evaporation Rates for Loss of Spent
! Fuel Pool Cooling
, 4.I -l4 spd Fuel Rack uib'cality Artalysis I'npt Wefe,s. 8 1,1-10 criWedh-vil fy Analy.ris Resdfs Amendment 3 9eI ~ N $pec[al &rt fe t&tel Sftnf Fuel RedinpaF faameWI _ _'et F UiHealdy Anaip'or_ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ . _ _
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, HCGS FSAR 8/84 CHAPTER 9 FIGURES Fioure No. Title 's.
9.1-1 New Fuel Rack Arrangement 9.1-2 General Arrangement of Spent Fuel Storage Pool 9.1-3 A Typical Spent Fuel Rack 9.1-4 Spent Fuel Rack Arrangement in Fuel Pool lCMyy0 9.1-5 Fuel Pool.Coolipg and Torus Water Cleanup, P&ID - 9.1-6 Fuel Pool Filter Demineralizer, P&ID 9.1-7 Fuel Preparation Machine Shown Installed in Fuel Pool 9.1-8 New Fuel Inspection Stand . 9.1-9 Channel Bolt Wrench 9.1-10 Channel Handling Tool 9.1-11 Fuel Pool Sipper 9.1-12 Channel Gauging Fixture 9.1-13 Fuel Grapple 9.1-14 General Purpose Grapple 9.1-15 Fuel Inspection Fixture 9.1-16 Refueling Outage Flow Diagram 9.1-17 Plan View of Refueling Floor During Refueling 9.1-18 Simplified Section of New Fuel Handling Facilities (Section X-X, Figure 9.1-17) Wa l-M R Special Sped Fuel Raefc
'l, V 10 SPEWT FttEL Rack CRITtCStTY CrEayTRY DSER OPDi ITD4 / Yt) 9-311 Amendment 7 p,* ._ .. _ _ _ _ _ ___
1 ,I HCGS FSAR g/34 1
- 1. Normal storage conditions exist when the fuel storage racks are located in the pool and are
; covered with about 25 feet of water for radiation shielding, and with the maximug number of fuel assemblies or bundles in their Gesign storage position.
- 2. An abnormal storage condition may result fr bMe, accidental dropping of ad' W ' fuel , or from damage caused by the horizontal movement of fuel i handling equipment without first disengaging the fuel from the hoisting equipment.
{ b. It is assumed that the storage array is infinite in all
; directions. Since no credit is taken for leakage, the ;
values reported as effective neutron multiplication !
- factors are in reality infinite neutron multiplication l l' factors. The biases between the calculated results and :
experimental results and the uncertainty involved in ' the calculations, as well as other uncertainties, are ! taken into account as part of the calculational i procedure to ensure that the specified K eft limits are met. i I
- c. The racks are designed to protect the fuel assemblies
- from physical damage caused by impact from fuel
; assemblies. The rack design would prevent the release of radioactive materials in excess of 10 CFR 20 and - ; 10 CFR 100 allowances under normal and abnormal storage conditions.
I l d. The racks are constructed in accordance with the OA i requirements of 10 CFR 50, Appendix B. i i e.- The spent fuel storage racks are constructed in accordance with Seismic Category I requirements. The applicable code for the design of racks is ASME Section III, Subsection NF. 1 1
- f. Spent fuel storage space is provided in the fuel storage pool to accommodate 5.3 core loads of fuel assemblies.
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HCGS FSAR 10/83 9.1.2.2.2.2 High Density Spent Fuel Storage Racks HighdensityspentfuelstoragArack,sinthefuelpoolstore spent fuel transferred from the reactor vess' elk. These are top-entry racks. - The spent fuel storage racks are of freestanding design and are not attached to either the fuel pool wall or the fuel pool liner plate. The racks are constructed of stainless steel, and the details of rack construction will be provided prior to fuel load. See A ppenclix 9A fos- a descriphvn of' h e deseg n, a nalysis and con s trac hvn of tb kc)h dc.nej .syc.d h el s+orag e ra c ks . l l 4 i l
/ 9.1-10b Amendment 2 DSER OPEN ITEM /5 C
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- 4. ,
- 1. The mesimum stress in the fully loaded rack in a faulted condition will be,provided prior to fuel load.
N. i j. The spent fuel storage racks also have the caN bility
; of storing control rod guide tubes, control rods, and defective fuel containers. When the spent fuel is stored in the spaces provided for storing the above the i K,gg does not escoed 0.95.
1
- k. Several design features reduce the possibility of heavy objects dropping into the fuel pool. The main and -
auxilian hoists of the reactor building tolar crane i are single-failure proof. In addition, tto main hoist i is physically prevented from traveling in the truncated l i segment shown on Figure 9.1-31 by mechanical stops on i
- the girders of the polar crane. The crane design is l discussed in Section 9.1.5. The removable guardrail i and the four-inch curb around the refueling cavities i further limit the possibility of heavy objects dropping ll into the fuel pool.
l'
- 1. The fuel storage pool has water shielding for the
! stored spent fuel. Liquid level sensors are installed i to detect a low pool water level. Makeup water is ,
i available to ensure that the fuel will not be uncovered ! should a leak occur. l i : r
- m. Since the fuel racks are made of noncombustible material and are stored underwater, there is no ;
! potential fire hazard. The large water volume also ; i protects the spent fuel storage racks from potential l l pipe breaks and associated jet impingement loads. k q. I .1.%3 DVSEY 9.1.2.4 Spent Fuel Rack Inservice Inspection i ad 1nserv'ipe 'ihsppetTop prog'kam is71n eff)ct the$ughodT tbd711 - l
/cf She' racks-to ensuite thatnthe'quilit'y (f thn pMaongddedks
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HCGS FSAR t d. u he fo o on m. 9.1.2.4.1 Test Coupon Description and Installation Details of test coupon description and installation will be ' provided prior to fuel load. 9.1.2.5 SRP Rule Review In SRP Section 9.1.2, Acceptance Criterion II.1 requires conformance to ANS 57.2, Paragraph 5.1.1, which states that the spent fuel storage facility, including its equipment and safety-related structures, shall be designed to Seismic Category I requirements. The spent fuel liner plates are non-Seismic Category I and are not considered safety-related. These liner plates are welded to Seismic Category I embeds in the pool walls. Their primary functions are to minimize pool leakage and facilitate j decontamination of the pool walls. Since they are essentially nonload-bearing, they will not adversely affect the structural integrity of the fuel pool and the spent fuel storage racks, and therefore do not have to comply with Seismic Category I requirements. Any pool wall attachments will always be affixed to the wall embeds. Acceptance Criterion II.6, ANS 57.2, Paragraph 5.4.1 states that at least one radiation monitor with audible alarm should be installed on'the fuel handling machine.. At HCGS, permanent radiation monitors scanning the entire refueling floor are mounted on the reactor building walls. These monitors indicate and actuate audible alarms locally and in the control room. In addition, portable health physics instrumentation will be installed on the fuel handling platform whenever the refueling machine is used over the spent fuel pool and the reactor core. The radiation monitoring system, including the portable platform mounted health physics instrumentation, is considered to be adequate for protection of personnel in the reactor building during all phases of station operation. DSER OPEN ITEM. / tl$ y e i
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INSERT A gaf6I#fIi HCGS FSAR 9.1.2.3.3 CRITICALITY ANALYSIS AND RESULTS The criticality analysis was performed using the input parameters contained in Table 9.1-19. Figure 9.1-20 shows the reference 7 n tp kr escenably,cr ticality analysisge the seev f/w 6 gg aalfti geometry g jtory the forta simuler d The criticality analysis is based onNo new fuel with a nominal, credit is taken for the flat U-235 enrichment of 3.4 w/o. ie 4 h*gf .<burnable f fuel poison fuel rods which may be present in the fuel assemblies._a The analysis uses Utility Associates International's 4;;2M TUAI's) dif f usion theoryThe model, CHEETAH-B/CORC-BLADE /PDQ7 as the analysis includes the various criticality main working model. safety-related aspects of the rack design, including various sensitivity calculations. The Monte Carlo transport model, AMPX/ KENO -IV, is used as the verification model to verify the reactivity of the nominal rack design. UAI performed similar criticality analyses for Limerick and (ql.- Susquehanna.4 The an h is includes all the normal, abnormal, and d ' accident conditions described in Section 9.1.2.3.1. Table 9.1-20 summarizes the nominal value of K effectiveThe of the racks under normal, abnormal, and accident conditions. various uncertainties and biases considered in the analysis are also included. -. _. _ _ _ _ _
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tlCGS FSAR 1 CALCULATIONAL MODELS This section presents a description of the calculational models and l the basic assumptions used in this ' criticality eqalysis. l l 1 i j The Working Model i The criticality analysis for the Hope Creek BWR spent fuel racks employs the CHEETAH-8/CORC-BLADE /PDQ-7 model as the basic engineering tool. CHEETAH-B W is UAI's BWR lattice code based on the original LEOPARD code and uses a modified ENDF/B-II . cross section ' library. CORC-BLADE enerates equivalent diffusion theory cross sections for the control blade. The PDQ-7 rogram is the well-known, few-group spatial diffusion theory code widely used by the industry. The CHEETAH-B/CORC-BLADE /PDQ-7 model, which is also a part of the LEAHS (Lifetime Evaluation and Analysis of Heterogeneous Systems) nuclear analysis series of Control Data Corporation, has been extensively tested 4 through benchmarking calculations of measured criticals as well as thmugh core physics calculations for several operating power reactors. A zero current boundary condition was applied to the four sides of the unit reference storage rack cavity .
"to produce an infinite array effect. The two-dimensional, PDQ-7 calculations were made for four neutron energy groups, two mesh intervals per fuel pin, a flat U-235 enrichment description and a zero axial buckling to simulate infinite fuel length.
' The Verification Model The verification calculation employs the KENO-I PX del.
The basic neutron cross section data comes from the master libraries of AMPX - a 123 group GAM-THERMOS neutron library prepared from ENDF/B version II data. The NITAWL module of the A M X program is DSER OPEN ITEM / [d
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.L eek tt f y aof 16 FSAlt JCGS used to perform a Nordheim integral treatment of the U-238 1 resonances accounting for the self-shielding effect. The working. library produced by the NITAWL/Al(X module retains 1 the 123 group energy structum and is used directly by KENO-IV. l In the KENO-IV calculation, the spent fuel rack geometry including each fuel and water rod cell is represented discretely.
To simulate the arrangement of a large number of storage rack units, and for a non-leakage condition in the axial directions, a specular reflective condition is applied to all six sides . of the reference case storage rack cavity '";- - Y L4L Basic Assumptions To ensure that the analysis follows a conservative approach and confonns to the generei guidel,ines of criticality safety analysis in Reference W the calculations are perfonned with the following assumptions:T.1-10
- 1. A flat 3.4 w/o distribution in an 8x8 bundle, with U-234 neglected
- 2. Fresh fuel, no burnable poison
- 3. Minor structural members replaced by water, i.e., spacer grids
- 4. Fresh water
- 5. Fuel is channeled.
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REFERENCE CASE CALCULATIONS Physical Parameters and the Basic Storace Rack Cavity Geometry The reference storage rack cavity 'J:,.. [as a pitch of 6.308" + 0.030". The stainless steel canister has a nominal inside clearance of 6.080 to acconnodate 8x8 fuel assembly channeled in 0.080" thick Zircaloy-4. Plates of the neutron absorber material Boral, consisting of 8 C 4 in an aluminum matrix core and clad with 1 an aluminum sheath, are fastened to the outside of the canister. , The Boral plate- has a nominal total thickness of 95 mils and a minimum 2 B-10 density of 0.028 g/cm . Table #contains the values of the 9 I-m input parameters used in the analys.is. 1 The rack must acconnodate both channeled and unchanneled fuel. l Studies reveal that the channeled fuel in the rack is more reactive than the unchanneled fuel. Taking the conservative approach, the study here involves channeled fuel (except in the accident condition where the dropped fuel is unchanneled in order to permit the closest contact between the dropped fuel assembly and the rack). Two small, but non-conservative changes were made to the reference case in order to facilitate modeling. First, the boral width was set at 4.48" instead of 4.465". Second, the stainless steel flanges used in welding the outer wrapper to the inner can were deleted. An adjustment was made using PDQ to account for these differences.
# Results of the Reference Case Calculations 1,1-11 9'./-
Using the input data from Table fand Figure (except as noted above), the K,ff values of de mfennce case at 684 wem cal-culated for the calculational model described in Section 2.0." The msults are: O 9" DSER OPEN ITEM
Ensett N f ase F eF I& dCCr5 FSAR PDQ-7 KENO-IV k,ff, reference calculation 0.9229 O.9306 +_ 0.0042 95% confidence interval 0.9222.- 0.9390 I i 9 i l l t DSER OPEN ITEM /yd
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asENE SENSITIVITY AND TOLERANCE REACTIVITY CALCULATIONS 4 Temperature Effect
- s, Using the reference storage rack cavity geometry, the temperature of the fuel and pool water was varied. In addition to the nominal 68'F, 40',F and 212*F were studied and the results of the CHEETAH-B/
. CORC-BLADE /PDQ-7 runs are given on . Table As shown, reactivity decreases continuously as temperature increases
--- -~ ~ ~ ~ ~ ~ ~ ~
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7 40'C g Void Effect' The effect of boiling (assuming equal voids inside and outs;de of the rack) was studied by varying the voids from 0% to 20% at a temperature of 212*F with the reference geometry. TheCH,EgAH-B/ CORC-BLADE /PDQ-7 results are shown in NgummagnumE Tablep. As* indicated, k,ff decreases con'tinuously as the void fraction increases. 6 Pitch Sensitivity The rack design permits the storage cavity pitch to diffar from the 5.308" nominal value by +_0.030". The pitch sensitivity calculations of this analysis show the reactivity effect of these tolerance components as well as the reacEivity pitch
~
sensitivit, by expanding the caiculational range from -0.060" to
+.030" at .030" intervals. The results, which are pisE455mm l . ci I- w t PfgummutuuNe tabulated in Table E Fndic&te 'that in the neighbor-hoo'd of the nominal pitch, the pitch reactivNy coefficient is about .15%ak per .030" pitch change.
l DSER oPEN ITEM /
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Insed N f* ' "" HCGS FSRR
@ Effect of Boron The Boral Plates which separate two'adhcent fuel assemblies have a nominal thickness of .095" (consisting of an 73 mil core ,
and 11 mil aluminum sheaths) a nominal width of 4.465" and an overall length of 11 feet 3 inches. The minimum B-10 loading
~
2 _ 1s 0.028 g/cm .,_
.---"~
(a) Boron Width Tolerance . The effect of reducing the Boral width was examined. The PDQ-7 calculation for the reference case con-figuration with the Boral width reduced by 0.0625" yielded k ,= 0.92641. Hence, the reactivity increases due to the -0.0625" tolerance on Boral width is Ak = +0.0029. . _ _ , (b) Boron Density 2 The boron density was maintained at .028 g/cm for all calculations. This areal density is the minimum
~ ~
density allowed by manufacturing design specificationi.. , _ _ . . (c) Boral Core Thickness Variation The sensitivity to the Boral core thickness was de-tennined by calculations in which the thickness varied from 61 mils to 80 mils (the aluminum sheaths were
~
varied. within.tpieranTe to obtai ~ e worst ~ case cirq_ thickness). ; The results, tabulated in Table show a continuous increase in reactivity as the core thickness increases. This is due to the fact that the areal density i_s_. held constant, so an increase in thickness reduces volumetric density and. 6 a smaff degree, the boral effectivene f DSER oPEN ITEM /fO M
T nget- k pqe 9 of I4 ((C-&S FSR W asMP- Dimensional and Positional Tolerances l The total Ak bias for dimensional'anaspositional tolerance calculated from five separate contributions: l (1) Pitch Reduction . (ii) Boral Width Reduction . (iii) Inter-Cavity Spacing Reduction Off-center Loading (iv) (v) Boral Thickness Increase The effect of reducing the center-to-(1) Pitch Reduction. center spacing of the rack cavities is obtained from ti Table 9.I-1p; Ak3 = 0.0015. arul {$ The Ak bias due to reducing the (ii) Boral Width Reduction _. Boral width by its tolerance. 0.0625" is obtained from e 0.0029. femuutelt45 M and is ak2 Td/t T./-26 (iii) Inter-Cavity Spacing Reduction. Any seismic effect th may reduce the separation distance between a can be detennined from the pitch sensitivity studged 4tueiemm898,' Bringing two adjacent cavities closer O.048" results in the canisters toghing andSince a reactiv increase ok j = 0.0023 (from Table 9esumsdydupe). this reduction is the maximum reduction of pitch po l in this design, this effect will not be added to ite but will replace it. l The free space existing between a Off-center Loading._ (iv) properly center fuel assembly and the top casti an assembly to be loaded off-center in a cavity. It wa
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'.' shown that this condition causes no adverse reactivity effect since ,the resulting k,ff for off-centered loading is less than that for properly centered assemblies.
N (v) Boral Thickness Increase. The worst case boral core thickness reactivity effect calculated due to manufacturing tolerance
~
stackup (.080") is 'obtained from indisakj=.000' The above positive Ak contributions are statistically combined to give the total Ak bias for mechanical and seismic uncertainties. ak = /(ak)2j +- (ak2) +(ak)5 = 0.0037 e l 1 l DSER OPEN ITEM [kh W-t _
Insert nye IQop 16 dc(rS. FSkl{ ca.2.3.0 . y SPECIAL CASES
# Grappler Drop Accident ,
N, The accident considered is the inadvertent drop of the assembly grappler used in lifting assemblies within the spent fuel pool. In this accident, the grappler is dropped in such a way that assemblies in adjacent rack cavities are displaced such that they are resting in an off-center loading arrangement.
~
The reactivity effect for this off-center arrangement'was , dis-
,. cussed ~in Sectibn @ (
9e l 1.3.3. 2(i V) (al, Assembly Drop Accident (a) Single Assembly Dropped on Top of Rack. No adverse reactivity effect is expected from dropping a fuel assembly on top of a fully loaded storage rack during fuel' handling because of the large water thickness (-14 inches) existing between the top of the assemblies already inside the cavities and the dropped assembly resting on top of the rack. Moreover the P0Q-7 model assumes an infinite fuel length in the axial direction. (b) Single Assembly Next to Rack. The dropping of an assembly outside the rack is a possible event because of the un-obstmeted water area existing between the periphery of the storage racks and the side walls of the pool. 1 . A conservative analysis to evaluate this situation is illustrated in Figure E. An asser41y, presumed to be 9.l-20 7 DSER oPEN ITEM f% l 1 .
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I/ CGS FSAR dropped during handling, lodges paralled to an assembly in the outer cavity with no Boral slab separating the two assemblies. Thedroppedassen$s1y is unchanneled to perinit the closest contact with the rac The dimensions used are those of the refemnce case. This arrangement of the dropped fuel assembly with a 31/2 x 3 finite fuel racg is reflected onthreesidesasindicatedinFigure(t$'e~ fourth'sideis a zero flux boundary. The k,ff result for this case was 0.9128. The result for the same geometry without the ' l dropped fuel was 0.9064 giving an increase of reactivity of l Ak = 0.0063 for the above dropped assembly configuration. tamwee 7 Assembly Moving Between Two Storace Racks The rack structural design does not allow sufficient room to fit a fuel assembly between any two of the high density spent fuel racks. Therefore, the movement of assemblies
. between racks is precluded.
e 9 DSER oPEN ITEM / N
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,l,1,3,3,44 New Fuel Storage in the Spent Fuel Racks The feasibility of storage of fresh fuel in the high density spent fuel racks was analyzed. Storage of new fue,1 in the mist, partly j flooded, and dry conditions are addressed below. )
l l l l DSER oPEN ITEM /YO W
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. 25% Mist Condition The storage of new fuel of unifonn 3.4'w/o U-235 enrichment in the high density spent fuel rack in a 25%
aqueous mist environment was analyzed with the KENO model 'i '- ";_r: .T.,The resulting k,, and 95% confidence interval are shown below: K
= 951 Confidence Interval 1
25% Mist .6375+.0054
.6267 .6483 Dry Condition UAI experience in the analysis of poison.ed rack critical 1t)! ,
. indicates that the fully flooded rack configuration is the most reactive with reacitivy decreasing with a decreast in moderator density. The 25% mist condition analy:is confirms this as shown below. For this reason a dry condition analysis was not perfonned since it too will be less reactive than the flooded condition.
Moderator Density K. 95% Confidence Interva' 3 Reference Case: 1.00 g/cm .93061 0042 .9223 .9389 3 . 25% Mist Condition:0.25 g/cm .63751 0054 .6267 .6483 l
@ Partly Flooded Condition
' The totally flooded condition as analyzed in the reference - case is,more reactive than that of the partly flooded condi tion . cf' l.2. 3. 3 F M Special Spent Fuel Rack Storage
' , ' 3,5x6 non-borated special rack is to be installed in the "g$ :E spent fuel pool . Storage of control rods, control rod guide tubes and defective fuel is provided for by this special rack. This rack was analyzed for storage of ruptured fuel as shown in Figure CSpecial rack input parameters are sumarized in Tabid ct. H 1 DSER oPEN ITEM /YO
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Insect & f*ft 'Y # 16 ' ' 1 tlc 6S FSA R The storage of ruptured fuel is a more reactive evaluation than that of control rods or contml md guide tubes. - j M Storace of Ruptured Fuel in the Fully Flooded Special Rack The storage of ruptured fuel assemblies within defective fuel ' storage containers inserted into l the special 5 - ! :: i::::dackwasanalyzed using the CHEETAH-B/PDQ-7 diffusion theory model. The case was analyzed as an infinite array in ' order i to simulate storage of .. :: : :# ruptured fuel ! l assemblies in the special rack. The resulting ; K,ff for this case was .6589. Considering that this K,ff accounts for no radial or axial leakage. the reactivity for the storage of fuel in the special rack is well beiow the design limit K,ff of .95. Storage of undamaged fuel within the special rack is less reactive than storage of damaged fuel . This is due to the fact that in the ruptured fuel case, the defective fuel storage container displaces water. For this reason, the storage of undamaged fuel w'as not analyzed. l DSER oPEN ITEM /f/h M - _ - - - =_
Lusent- /t f ctye t > ettb
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SUMMARY
AND CONCLUSION The final result as calculated by both the working model (CHEETAH-B/CORCBLADE/ PDQ-7) and the verification model (AMPX/ KENO-IV) 19 sumarized in this section and compared to the NRC regulation k,ff limit of 0.950. The
" Reference Case" referred to in this report uses the nominal dimensions given. in FigureWd. Table ithout the dimensional and material tolerances included.
g Results $f theTra~n'sioit Monte Carlo (AMPX/ KENO iV) Verification Calculations and the Calculational Bias
- 9. l.2 3.3. (
k,ff, Reference Case N 0 9306 t 0.0M2 - Benchmark bias, ak -0.001
.9296 t 0.0042 95% Confidence Interval k,ff . 0.9212 = 0.9380 The bias of the KENO-IV vs. measurement is based on criticality experiments perfonned with fixed neutron poisons . These experi-ments were chosen because they approach the fuel storage rack configur-ation in that they used fixed poison plates between fuel rod clusters.
The result of th benchmark calculations was that the KENO-IV results were 0.001Ak above the measured value. This demonstrates a negative bias of 0.001ak. l g Summary of Results k,ff, adjusted (KEN 06) 0.9296 t 0.0042 Dimensional and Positional Tolerance, ak (PDQ, M ) 0.0037 PDQ correction for non-conservative i assumptions in the reference case, ak (PDQ h ) 0.0006 Dropped Assembly, ak (PDQ,JEEEsuWEED) 0.0063 . DSER oPEN ITEM /f/O
]g gg&V A ffL (S of Ib IlCG3 FSAR Final K,ff 0.9402 1 0.0042 95% Confidence Interva.1 0.9318 - 0.9486
. Design Limit, k,ff - .
0.950 The final k,ff value (0.9486) includes all the design specification tolerances, the postulation of a dropped fuel assenbly, the model bias, and the 95% confidence interval from the KENO calculations. However, the negative reactivity effect (- 0.5% Ak) due to the pmsence of U-234 and the parasitic structure materials (i.e., spacer grids) in each assen61y was not included. , 9 DSER OPEN ITEM ///O
1 h i 5 . . t , N, . 3 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 g spent fuel pool. However, because stainless steel spent fuel i
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. l
\
a b of DSER OPEN ITEM /f/O 1 i l
l l l HCGS PSAR gates, and any other noncoutine heavy loads that must be carried over the spent fuel pool. , l 9.
1.6 REFERENCES
9.1-1 C. L. Martin, Lattice Physics Method, NEDO-20913, General Electric, June 1975. 9.1-2 AISC Manual of Steel Construction 9.1-3 AGMA Gear Classification Manual 9.1-4 Aluminum Construction Manual, Aluminum Association 9.1-5 AWS D1.1, Structural Welding 9.1-6 NEMA MG-1, Motor and General Standards 9.1-7 National Electric Code 9.1-8 OSHA 1910.179 9.1-9 OSHA, Vol 37, No. 202, Part 191 ON T(-(S /ES T JV zto -ic(7g i u )es7n ObjedJe> rue! D'^ Ljfd Wafer Reach SpentFacdbes Storage . Rossien o i SU"Muy RePetof- 1
$U (AAT 84-42.
Ah<cleae Cn%/;g g (U f" Ale %ntFuel Rack'
- e? Nfe G'eek Gemfy %fien ,
I i l
~
XCG6 PSA R
' TABLE 4.l-l4 &ye ingx
(. _
- ! CitlTIC4LlTV REFERENCE CASE SPENT FUEL RACK INPUT PARADETERS
'I FUEL ASSDSLY.(8x8) - s l FUEL DATA l - l Pellet OD 0.410" j Clad OD 0.483" l Clad Thickness 0.032" Clad Material Ir-2 . Fuel Rod Pitch 0.640" Active Fuel Length 150." U-235 Enrichment 3.4 w/o Effective (Stacked) Density 96.55 Theoretical WATER R00 DIATA (2 per Assembly) (, l Water Rod OD 0.591 " Water Rod Thickness 0.030" Water Rod Material Zr-2 CHANNEL DATA Channel Inside Dimension 5.27" Channel Thickness .080" Channel Material Zr- 4 ( - DSER OPEN ITEM ,,yy l
.. - - - - - - - - - - - - - - - - - l
? CGS .FSAR .
. 4.t-M Pge zoF2
( , TABLE (continued) BORAL PLATE DATA - Total Thickness O.095" + .005"
.010 Width 4'.465" 1 0625" Length 135"1 0.25" l
Sheath Thickness 0.011" 1 001" l Sheath Material Aluminum (1100 series) Com Thickness 0.073" (nominal;s range:.061" to .08C Com Material Boral B-10 Density 0.028 g/cm2 CAVITY DATA , ( Can Inside Dimension
- 6.080" (nominal)
Can Thickness -- -- 0.090" (nominal) , Outer Wrapper Inside Width Dimension 4.562" 1 020" Outer Wrapper OutsideWidth Dimension 5.36" (nominal) Outer Wrapper Inside Thicknesi Dimenif6n- 0.101" (nominal) l Outer Wrapper Material Thickness 0.024" (nominal) Cavity Material Stainless Steel Rack Cavity Pitch
- 6.308" 1 0.030" e
/
DSER OPEN ITEM /[b e
h' CGS FSAR (' TABLE h T ,[ ' d O f"I'l
- b )
CRITICAUtV .
,SLM4ARY OF 4-GROUP PDQ-7 RESULTS M REFERENCE CASE k,ff AS A FUNCTION OF TEWERATURE AND VOIDS REFERENCE GEOMETRY - FIGURE p { / 2 0 1
K Tamperature 'F 5 Voids eff 40 0 0.9265 , 0 0.9235 l 68 I 212 0 0.9028 212 10 0.8845 212 20 0.8630 ( t ( DSER OPEN ITEM /YO ! '~- r_ _1__r _
i (- . TABLE V f4f'e.2 of- J 5 l-A# N k,ff AS A FUNCTION OF PITCH VARIATION PITCH K (INCHES) . eff (INCHES)
+ .030 6.338 .9220 l
Base 6.308 .9235
.030 6.278 .9250
.060 6.248 .
.9264 l
l 1 ( DSER OPEN ITEM / pd
.- l TABLE M F#/#
- q. I-20
. s k,ff AS A FUNCTION OF BORAL CORE THIClotESS - AREAL DENSITY CONSTANT Thickness of Boral Core
! Inches off 0.061 0.9233 -
(base) 0.073 0.9235 i 0.080 0.9236 e 4 ( DSER OPEN ITEM /YO i n 9 _. , e ., , - - . , - - - - -
KC&S FSAR (~ . TABLE V T.[-l[ NSPECIAL NON-POIS0NED SPENT FUEL RACK INPUT PARAMETERS -- FOR TifE C.RITICAlITY AuktVsts '
~
~~ --~~"~~
FUEL dSSDSLY (8x8) , i FUEL DATA Pellet 00 0.41 0" Clad 00 0.483" Clad Thickness 0.032" , Clad Material Zr-2 Fuel Rod Pitch 0.640" Active Fuel Length 150." U-235 Enrichraent 3.4 w/o . Effective (Stacked) Density 96.55 Theoretical WATER ROD DATA (2 per Assembly) ( Water Rod OD 0.591" Water Rod Thickness 0.030" Water Rod Material Zr-2 CHANNEL CATA Channel Inside Dimension 5.27" Channel Thickness .080" Channel Material Zr-4 CAVITY DATA Can Inside Dimension - 11.50 + .00"
.06" Can Thickness 0.165 (nominal)
Cavity Haterial Stainless Steel Rack Cavity Pitch 11.665" + .020"
.000"
(
- DSER OPEN ITEM [%
1
~
6; - l 9 l 4 TOP VIE W
~ .
ventA .?P- - v TYP
- Bo ra.( -- m caaem -
*5 use Ty'PICAL CRW[5TER vl/I TH BCRRL AWD WRAPPERS o -
Canister -
/
/
8
. ~;% i r
,M%
Oder t%ger i A7 Vent rye t y on a a e d
%dI :
., % o 5 .
( ,
.~ k '
} i .,2 o l Il j
=
s.L '! l l _ s, N '
; a S
a sN
< d h$$
i m e vent Tff s J I ' N HOPE CREEK g I GENERATING STATION g
- FINAL SAFETY ANALYSIS REPORT E
8 .
!a o
"' ~ -
e A TYPICAL SPENT FUEL RACK FIGURE s.t.s pp y ,f q..
l I l CEFECTIVE FUEL
- I C00 TROL ROO CUCE TUSE l STCR AGE CONTAINER C.E.0108 DSS 19 . l' O.E. o781E720
! * - CONTROL ROD 11.006 11.80 -
j TYP. 1 G.E. e 814E740 INSIDE OPENING ,. ,
- o I .
! I ELEV.100.25 MAX,
- -- -- - r i7s.2s min.
p i i
\. l l l ;
I i i l l l l L - - _- l L,~_- - rmp- - _ r -
~
' a 8 g
j t i -
- E LEV.12.00
' 8
.. 0 J LB . . 2
- DIA. - . .
f_... l
- -~l E LEV.10.00 4
s u s 1 x - cs 1 re- s cr= .,. s , EjV. 8
\ / , i f
i l l f f
' ELEV. 4.75 T K'
~ 3.625-DIA. k ,,, . \
E LEV. 0.00 SECTION A-A SHOWN 4 MIN. HEIGHT
+-- 5 EQ. SP.*
@ 11665= 58 33 5l[ +l+ +
_F3' g hF+
++
+P kk 1 .665= 6 66 =
j QI+ @+d ; -
' ^ '
a.fs.2 -
!= 71.16 g SECTION 0"0 BOTTOM VIEW SCALE: NONE l $ SCALE: NONE N
HOPE CREEK GENERATING STATION g NOTES: FINAL SAFETY ANALYSIS REPORT
- 1. REGULAR FUEL SUNDLES COULD ALSO 8E STORED 0+
IN THIS MACK ON A TEMPORARY BAS 15 IF NECESSARY. p[C o ATGiCATSPENT FUEL RACK
- 2. 5 a 6 SPECIAL R ACK FOR STORAGE OF CONTROL ROD G.E. *814E749 AND DEFECTIVE FUEL 3TORAGE CONTAINER G.E. e761E720 AND CONTROL FIGURE 9.1- $
h(L ROD GUIDE TUSE G.E. e 10505619. j( i::".5 0 M " .AME46984588 M I w - - , - , , . - . . - - -- - -- -- - - - , , - , , _ _ . . - - - - - - . , - - - - . - , . - - - . - - - . . . . - - . _ , - . . . _ ,
' MGMABul&, SPECIAL ; .:;- -PE.7, RACK
( .
+.020 -
= n.as. 000m 1
f / // // // 5'///// ,
, -7.716 ID 8.03 OD Rack b
/ Water iIi i
a c (.165
. l 533044 l
/ /
5.12 SQ% 1 l/
- v. y i
/ l
/ l\' A0d
/ i
/ s
/ I
/ ~ s
~
@ s g Zire ~,,,
l ll r - Mater ' O n Channe) % . t
/ I $
; l(.000.)
/
/uel r
/ Rod /
/ . .
n . e , l
!/ ~
'/ '
n 5.27 SQ / fective / Fuel Storage
/
i j ,
/ /gg1 Container (55304)
/
i 11.50 +.00,_ , , se
/ 4 ,
s, !
/////// // ////// / / < ;. l C .
1 ! i ! ( . DSER OPEN ITEM /t/C ,, h SfECIAL SPE8 F sheefze Fi9are 9.(-l9 .,.
, IIOPE CREEK SPENT FUEL RACK IEFERDICE CASE GEWETRY
*< % *NW r OD O OJ O O 'O ~O use 0000 0000 0000 OO 00 I" Lr OOOOl 0OO O , t
, s"
[& ,Qg 3 I
- s. s os1.ow V ;
^ ^ ~ ' ' ~
,0 0 0 0l O O O OA
.O O O O O OOO 00 00 0000 0000 , , 0 0 00 -
d d ;
=
; 6.308!.030 NGTE: . _
- l. All Der 71v13'35 , 17t, xATtniAt Me Tnthes.
1 WATER ROD CELL IIRC + IIRC I cf,1-19 @ ( fhe 2 FUEL ROD CELL g 3 CHANNEL ZIRC 8 I 4 INNER CAfl 55304L T hcSc I N tr1f. 5 BORAL WRE+R 00TER WRAPPER 55304L 6 GENER TI TION 1 FINAL SAFETY ANALYSIS REPORT SPEWT* FUEL R ACk DSEE OPER ITEM /yo cgITtceL1 y GE ETRy
/
J nGuR .. .tto S4=f 1.f 2
~
fWFP 91W e dBNP9 Eft 9t?Y SPENT FUEL STORAr.E(DROPPED ASSDSLY ACCIDENT Mk l l Zero Current Boundary ff/j,i n
/
g
/W 't_
k e
.M 7,,**PPe,W .-g;,
l
, i , ,
- i. L l af y;m 4
I [ !. ./ h .Zero
' /_ ; ._
$b' /j
{
% .4, .
i \' l // ll .
. Current Boundary I s p' .l t ? .iammme Zero i
s g . Flux l,,MJw,Nrsamiin , Boundary t auss ' _ 8.0"Q.12'
,,j ,
If
'q .} 'f l .' il k: '{,..- [
//
3 -l6, l. n . i(71$'-{i
,//; *
- j.;' (.,
t i
,l ! -=== ,
3
's l Drg/ 'd . ~.~ w t. - I
' o Assembly ' f '
,_ / /h . . $ '
f/.i 5., k.
- Zero Curmnt Boundary Boral *Zire Slab Channel,py,j, -
. Assembly s Inner Can DSER OPEN ITEM' / g h9h Y ~10 Shee b1h
1
-- . - _ . . . _ A4 f E N DI X . 't 8 - - - . . .-. - - -- - . _ . . - - -. 1 1
DESl6Aly A A1 A LN s( S . AND eqN sig ucT( ort .__ of +n e n e s C.J__5 PENT _Fu Et. . S To K A GR.Acic.5 E . _ . . _ _ A O I SC0PE - .. . - -.. .-- - -- . - - - - - . .--. _. Th 6_ a.pgencicx _ cle scribes &c. dest 3n, a n a l y s a a w c4 :
._co ns huc hyn. of de s p e 1- L e t ra.eks._ _ _ _ . . . . _ _ . _..-- _ ....
l 'f d .1 _ D esc a.t e tt ons oF SPEMT Fu EL foot. AND RACKS __. SecWon 9.1.'t.<2. c o nwns a cte s c r cp h o n o f +be s7 e t. (w e l skrue facc h tacladany Ae h,gh &eascy sgeni fu r I s +orage racks . The syeat fue l racks are of free s% ding desyn and are n o t a H ac heci lo e r&er +h e fuel pool na // or +/> e fuel pool line r pla te . Figu res I. 2-lo and I. 2- 5 2 show the spen t h<el pool in relahon lo o +her p/sd s h-ac f a re.s . Fcgures 9. I. 5 and 9.t.4 show dehcis of +he spen f fuel ro.c k S . .. ._. . Tks spent fuel racks are elescgned % unhshd as \ (askla. kcl drop of a fue I bundle. Sec kon 9. /. 5 con hrins 1 DSER OPEN ITEM /(/O l 9d-I
- 1
_. _ a_ de s crop _f,yn af de oveehead heavy l aad kdliy
. . .s_y.sHus. ..for.. the reae.lvv buildaty pile. crane . i>tcludt)ty_. ..
. . . _ h.y a res sh owoty . .loa d. pan.s for. 4 e ...c ra n e .._ _ .._
. . . . _ .. .. . . . . . . . . . - . . _ . ... . .-. _ .. .-..l
_ _ %. b A P plt C A6 L E Co DES, S TAN Dant OS ANO S TECIFt C ATtoN $
. . . .... . .. . . .. . - . . .= . .-...- .-
All par ts o f de sped 6tc.l rac(cs , cyce4,F de a.djushyty _.
. . . _ screws cn +h e fee t o f .each modale and ne_ pe rs o n
_. nsafe r ca.1, are ma.de from ASTtvl A240, Tyge so4L.
. . _ _ S_ kinless 5f e e l ., The 4djushwg screms_ are ytade fro,n___. _
ASTM A % 4. , Type /,30 s htcn l e ss s he 1. Bora.l a_ de garson nta f e rca.f . _ . _. . _ . _ . . . . . . . Desrgn, b b ric a hwn and ins t a llahtn o f th e sge.t t fu e l racks are Sa bsec kwn NF fe r S ormed based upon ce u. r e m e.sb e f Re fe r e,1c c 9a-I for C la 33 3 co rag on e .t i-s u.y p o r ts . .
. __$ 4 S E L $ fn t C AND i m PA CT LOADS _ _ _ _ _ , , _ _ , _ , _
. - - . . . - . . - - . _ . . . . . . . . . . . . - . . . . . - . = . - . ..
. . . a.f Slcer .1esfanss - sge.c h a. kr th e..- 'Pt E"'I PocI sis.b.. . -
..Eloo r < cs pon s e. s p e ck a re. de v el a y ccL From . .
you ncL. . l l
** - - **-a -- " " * '
- e. e.- - . . - -
_. ._. 38 '2
-- - . J
. T
.-.resyone syae%._wkreh c. amp l y wnb de regut,emen15
-.oi Refuitkry Guides l.(ao_aord. 6 6). , Accelera hvn _
-- - . kn e. his krtes .are.. developed... for_.L>o _ horce n /a/____ _ __
..dIYec h>rt$ .and_. 9ne _ yer_h'eal $fte c/tpn___lrgin __}kc }loe r _.
cespnse - pec+ra. s . These ihree kee hakrces_.are _ _.. . imposed simu Ikneously. The. pea k .reifuses fro >ri. ea ch do're chun are_. com hined. by sgua re rov f of.. A e.-_.
. .- ._. _ s a m of. Ae sgua res. i> a ecordarece with Regula ky . . - .._
- . . . Guede. I.92. _. . .... ..._. . _ _ __.-.
._ imy a.e f Ioads _.due.I.Jo.. Jue I ra 1+lan.y..are . ea icwt a fed .
useng methods described .in Sec hun 96 - G. /mpact loads a re consedered for loc a.l a s we tl as overa rt e f fec h on the rack desqn. v 98 .5 l. 0 A D S AND loa D com & /N A TIONS l - L oads and load com brna + eon.s are in ejreemen f with _ Ta ble I af Reference 98- 2, Thermal e /feeis are included by using decreased ma ferta./ properfres a / +he appleca ble lempera hire le ve / . Sin ce the caeks a re _ free s+an ding , &e re are no +bermal siresses. . . DM OM 1m jyg * ~ ~ ~ ~ ~ ~ ~ ~ ~ - ~ ~ ~ id - 3 .
98 6 DestC,N MD A N A L.4 SLS PROC E DURES . . - _ . _ _ . _
- ! . _. r " ' i" J _ 0 s W M .._ f o y k t p f r a m . . . _ . - . .
. Each k.e 1. rtck._cs. rdca.kud as A 3D Gwde eleme+f _
mode (( Fc$u.re .98 .I .s k ows _a. . Rye -can cskr gerhwn .. of a re.ck. Th e . c a.rtc.> 1e rs_.qad_ .be am. 3 red .yI4.fea re
. . . . mo de(ed wi%. . g l a.le e (em e_n ls, The _ yerrm e fer bar. ,
u keck seca. res .h e ca n cs.4s es_a.L &c 4ep , a.sd . % e . s h4 Gening br rs ...fo r &<. g ted_-pla le a re wdeled wik
. beam e le m en ts. The %.cn s+ain le ss 3+ eel egger co^ hincy ..Se. neu.bo.n absorker and Ae ..s %'n less s kel panels u sed do. close _ o f f fbe_ a.i fe erta le . .cav /fres a re . . . .
_. not modeleci 6k neir.. ma sses (re hse lud ed . The fue i a s s ea t re s a re atodel ed o s beam e le m ewis . Figa.r e 96-2 shows a dou.ble rac k racde l in .s c h e ms.ht-forn. 31) in 4 e r fm.c c e l e mc tf5 are used fo re pre 3 e a f he fu.e i- b - c a n c3 k r .clea ra nc e as well as de rack. h,- rnck ga.p. These mon icn ear e le me4 h re graduc e forces due bfue( ra.+t-(cn.g M p a ss (.ble. . mc k- h- rack. in+erachvo_,_.. .. 3. D._g _eg ._e (e m e nts _.. w ifh ma k real...__. progerkes .b4seJ on ne_. inlez fac.e free hvn. .co e f fee ve~ts . are u.s col da s nunta4e _ske cor nee _sy(or b'y Geet . .
._ .. .uAcek wq . s iede or..Ii4 L. d L A e yav L 4.lcor. k o
... 0.8-4._ . _ . . . . _ _ . . . . . . . .
_b ounday value.s a f in e.htn .cael_kccent (o.1._a_x.d
.-. 0. 8) _. a re._u.s ed i'n order _k-.rden.k}Ae-at mos]_
.condebtns for sledeh.y aad kr_marmum_reachxx s -.
4<..syyor+ fe e f. l 3+ru clara. l. da mpmy..coe i kes:en b .oL 2 pec ceaL.for ooE . . . . .
.. . and. 4 per cent. ht .515. A re._ u s ed, _excey fAa f imfs c f __.. _.
_ . dampen 3 of.._lo .per._ cent _ o f cttheal is_usedJer d e gap -
.__ elem en b si.hce . impsef drsstysles su bs fan kn./ .a, noun h af..
ener3y. tarn 20.feef _af sabarergence, sloslung effech .. . __ are neyluf tbl.e and. Herefore are neyleeled. F/urd dampig e f fee h are sIso .neylected. To sihtulate Hee rnenrersten , eI(ech, a lI de in ternd wa ler en tray,ed uiMen de ta ch en velope a added k & e k ortron tal ,<as.s . Th e ex ferna.f wa le e ie tween adfa cent ra ch.s a dc hydrod yneostic cophy elenreat shouin in Fijare wdeled as}M 2 - A pau me b ec. amou.ab of. shdy , ta hrch conscders ve.gl
..keI A.A..sig c I _ m k_.,.it_ coa 2xchcl.+o deleem he_whech.
._of 4Le. b lio w 3 eoade6mn> s k. auld ._h e _ s.onsedeac i L um%; u 4Lc se us m re s(case of 44e rac hs . .
! __ ._ . . re.ch _ e . _ , . , _ _ , , . _ . _ . _ , . , , _ , _ , _ _ = . _ _ mct... - e s . w ._..__ _ _ __ os ,op,,27,, j fg __ - __ _ _ _ _ _ _ ____ 9 4 - *$
- i; e .
. ' M.ck be- W rap.. b(d _...
. c ,c k .htt ..
For 4ke 74., k t( (e4ded co3.4c h.'oe, _ ecce#+ve c . _ . .
& Ae L.e l on eae sede 04 tLe . >-sek cs, cou.stde re -
98.'l s uncru tt*<. Acc e PrAsc e enTret A . _ . _ _ . . _ . . __ .A\ l % bte s 6-e 5 se5 are w 4 p_e M u~4 T4 4 le .. l . . e( R e Ce,e u e. 96 '2. ._ S he55 le-els b- bem e (ea <4 c, y (g % tk ne_.fe ui w ~ 4 _tE_.Apge J x _. xvir ..-6.... . _.. . Re (eveue 9A-I , _ Sk cs_ (e-e (s 6r y la.fe _ e (e u A . e u & mie s 4r . plade d. s he (I +p e syy gvh _ .. smcc g4ress ke td> .% dese coyme~h c- bea>ust-Tw +Le %.A Av.og e endchw n , Iocal y e emu e~+ de Gruk pa ss,% ceguA e a gerws scble yvowdrel +ka.t a < u 9/CrA[k G bT5663 dC Aci e)cCCed values g e , mc ++d & l.eset D s e rnce Wch ad de res 14iq de kre4tw
.. doe s wt gerwt +Le .Ae I .codprabn K,pp.Jo . exceed
.... - . . . O . R *i . - . . - - . . - - . - . . -
=_
@ .. _ ..,e.p. .. ..*==.N ..D..
.a,,
. . - . e -h..m h. N b "
DSER OPEN ITEM ///C
.. . . S. A .T. . b. _. .._ ..
9*
L
. .- l l . e. t A , 8_ _. . _ _M. A v e a i Ats , ~
a u M. tty couTeet wo srectAc. _ _._ l l CO M G Y RM.e~t t OM "(EC dNI Q l L ES S l Made tt ( a. <.s.-.d e s c ri b ed_in Seehvn_9A 3 . L alNy_
.contro I yrace du res for. .rnaterials,.Jo.bricakan an d.
design confro I ..and ve riHca+ ion _ comply with AAIS I
.._N 4 9._% . . C o n v en ko n al ..so ns huc h'o n ._ ine +bo ds a re_
used. A,, de sc ri6 ed i'n Sec h'an . 9. I. 2. 2e 2. 2 , ayy ro xania le y l __ _ 2L per cen+ o f . S e k kl sp ea t.Ju e / s krs; e cya cdy_ _
. . . .wilI he ymided by rscks aisk fled pritt k enef,al pis st t epera hun. The remainoig rs c/cs uit/ be ins k tied is ler. The ini h a (ty ins h2 (lec/ rs e t<> are g en erally l o ca led a t ne nor& end oI de sye,t t (n e l pea l. Th e r e k re, +ke addehenal racles es n be msk(ted
}a kr withoaf be<>tg frsaspor/ed o<er exis h>ty racks n h u*c h con lain spent fuel.
_ .9 8 9 _ . R EF E R ENC ES --- _ . , 9.6- I ASME 6 oiler and fressare desseI Code , Seebmn RI., . D iurs cun I j l 9 8 0 E ch'b'on , Sum mer l4 81 Addenda.
^
DSER OPEN ITEM j/4j/6 1
]l . e l i i 1
.M . -4.8 - 2 N. .R. c.... N. ot .R E. G . 0 8 0 0 ,
- S R P SecNon 3 . 7. 4, __.. .
Ay(e ndr x D , Re v . 'o, Ju Ip i A e i . . ..
. . _ . . - M . .... ..- . . - - - 6_ ._. . . . . . .. ..-. M. . . . .. .. . .. .. . . . . . .. .
w - e .. e emes. a w. m M- asmame . e
- _ w- +**.em ee .
.6 fu. 66..* . **l# b .e a .
EM h w M gui.mmem. uip. m. m ea6 e e g h 6 st..- mp a ..ee e@ se e. ee y . h e. e+ .e a h ee = 6 W *M hwe 6 e g @ em h p y g eag.g4m. .p, g m
.Me M N eh . . '"'4mm. ** mh .ew,, . . &
W -. w W.= h 6 . = 4 . M M. #4 6 WPd. . ge pump w a 6 h &em 6de 6 -' NgSe e .e .. GGNMe" m M m.h ae . . 6. S- 6e hh weM h ae see n .e. m.. e e. Ohhem me . -m6 m e *4- -- =m3 .. # 4- . -
.l.
- 6. 6-. a muet,.ee m 6. g. - . . . . - G .me em a e MshW .h' s um.M M dM- m.esta wa ee m m.m a 6.e eM -Mw e-e . 8 gh ae.e em um m -h. se em.- h . . 4m>e mem.mb a .. m DSER OPEN ITEM fg
. .- 9 A - 6. . . . _ . . . . . . . . . -
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BCGS FSAR 4g, egg, QUESTION 22 h15 (SECTION 3.8.4) Provide sketches of the mathematical models used in the design of s' pent fuel racks. Describe in detail, the methods of analysis by wtich seismic and other loads are applied ta sthe racks and the pool.
- RESpopSE
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3.f.4. F. 3 Se c h'ons -3. 2.4.4.1 and 9. /. 2. 2. 2. 2 hue been re.vtsed and ues/cd Appendix 98 has .been a dded lo yrnede fhe r Mormahen. l I l DSER OPEN ITEM //d m 220.15-1 e l . r
.I HCGS FSAR 4/34 OUESTION 281.13 (SECTION 9.1.2)
Identify the materials, including the neutron absorbing material (poison), used in the fabrication of the high density spent fuel storage racks and all other structural components % ced by the l pool water. Indicate how the poison-contaigi.9 cavities are ' vented.
RESPONSE
All parts of the spent fuel racks, except the adjusting screws in the feet of each module and the poison material, are made from ASTM A240, Type 304L, stainless steel. The adjusting screws are made from ASTM A564, Type 630 stainless steeb. Boral is the poison materia 1. (Wlth H IICO hed titxhneuit. l{ed twenf Sca{e )),
. tentWe4. -
Thin (0.024 inch thick) outer canister sheets hold the Boral tighltly against the 0.090 inch thick inner canister walls. The outer canisters are spot welded to the inner canisters along the bottom and both vertical sides of the outer canister. The top edge of each outer canister is seam welded to the inner canister. iii. c;p betua.. ti.. sp;t w;1d; govi9)he poison vente are
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l l i DSER OPEN ITEM /YO 281.13-1 Amendment 5
- - - - - - - - - - , . - - - - - . - , - y-- , , ,.w,, -
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- ECGS FSAR 8/84 l OUESTION 410.38 (SECTION 9.1.2)
Insufficient information is provided for review of the criticality gf the spent fuel pool. Ihe design bases are-acceptable with respect to criticality. Ihe (nformation required for the review is promised for later. Euch information should include the following: a_ . Sufficient structural detail to permit an independent calculation of the criticality of the racks.
- b. A description of the calculational methods used along with the results gf the verification of the methods. Ihis may be by reference to documents previously submitted by the organizations doing the analysis.
l c. A tabulation of the nominal value of k effective of the racks along w_ith the various uncertainties and biases
~ ; considered in the analysis.
- d. A tabulation of the reactivity effect of each of the abnormal 1 accident) situations considered.
RESPONSE
Ouffluieni infer.:tien for review cf the criticality of the sp C . i Tuel peel, incitidiTig that linied :beve 3till be-ava44able by ' Septemtree t9847&nd wt11 De addedwectinn 919 o Sechen cl. ( . z . 3.3 hts teen reeme/ +c inchele tl c i, 6 a.h!?n < qa ce fel a (ke I i I 1
" " " " /Yd 410.38-1 Amendment 7 l 1
- . . - _ _ . . _ _ . - . . . . - - _ _ _ - . - . . _ , . . - - , , - , _ . . . , _ . . . . .,____._-__..-.....-_-.-,e., , , _ , - - . . . . . - - , _ _ - . ..,_.,,w.._......-.__.,..,.-m,
I l HCGS FS"AR 10/83 OUESTION 281.14 (SECTION 9.1.2) Provide details of the materials monitoring program for the spent fuel pool, including type of samples used and frequency of inspection.
RESPONSE
Details of the materials monitoring program for the spent fuel pool, including type of samples used and frequency of inspection, will be provided in June 1984. d el eg Zn orderh conbinuall,Y AS3ure _lbe Adtguacy.- af.Ae poisan ma+eriaI., +es+ coup.ns are .- provided for .a . Baral surveillance yragram. Forfy - five fesh coupons are ins /alled in hyb ........ . radiafien . areas af the xpenE fue/paal. However,.. van **8' 4ecauses rfainless sfeel spenf fuel racks wiB... Boral poison maleri I are already in use in-aber2vn spen + fuel s+orage peels, .rueh 0 0 rionHe.e/lo o n el B row n Ferny 's , a Boral .s.sreeillance. p rog ram can e t, e n b e. in Mied. f3C44 w, ll de v elop e a proya m 40 men Nor t.he do ral .w ri d ila n c e pr o.g em of e.isher p e,m :, gos pc e,,, or Bro wn ' s f~c e"y by piarc A ,19ss-l l !
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'281.14-1 Amendment 2
R e y .L DSER OPEN ITEM 142a AND b (SECTION 9.1.4) LIGHT LOAD HANDLING SYSTEM (Related To Refueling) Redundant interlocks and limit switches have not been provided to prevent accidental collision with pool walls. The applicant must provide these redundant interlock and limit switches or provide the results of an analysis which shows that the effects of a fuel bundle colliding with the pool wall is bounded by the fuel handling accident analysis in Chapter 15 of the FSAR. Based on the above, 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, Position C.3 with respect to prevention of unacceptable radioactivity releases and criticality accidents are satisfied.
RESPONSE
Strict administrative and procedural controls will assure that a collision of a fuel bundle with the pool wall.will not occur. An analysis has shown that a postulated, i accidental fuel-bundle collision with the pool wall cannot result in more mechanical damage to the bundle hardware or the fuel or result in more fission product release than could result from the postulated drop of a fuel bundle over the reactor core. The analysis of the fuel drop accident, described in Section 15.7.4, shows that the resulting fission product release would be within the guidelines of 10CFR100 - Since the consequences of the wall collision accident would be bounded by those of the fuel drop
- accident, re dundant interlocks and limit switches are not required. FSAR see.fion 9. /. y. 3 7 an d th e t-espon t e io g ue sHon Y/o Gt ha ve been t-e vis ed ye reJ/c c.t Ws re spo"' C-MP84 95 08 01-az
_. _ .. ~ _ _ - _ - - - . . . _ . _ . _ . - ._ - . ', HCGS FSAR ! . / 9.1.4.3.5 Reactor Vessel Servicing Equipment ( i Section 9.1.4.2.5 includes discussions of the safety aspects of some of the reactor vessel servicing equipment. Failure of any of the vessel servicing equipment listed in Table 9.1-6, except the dryer and separator sling, head strongback, or service ' ! platform, poses no hazard greater than the effects of the refueling acciden,t. . Safety evaluation of the dryer and' separator sling, head strongback, and service platform is provided in Section 9.1.5. 9.1.4.3.6 In-Vessel Servicing Equipment 1 i Section 9.1.4.2.6 includes discussions of the safety aspects.of I the in-vessel servicing equipment. Failure of any of the in-vessel servicing equipment listed in Table 9.1-4 poses no hazard greater than the effects of the refueling accident. 9.1.4.3.7 Refueling Equipment
)
- Section 9.1.4.2.7 includes discussions of the safety aspects of
! the refueling equipment. The most severe failure of the refueling equipment would be a failure of the refueling platform and its associated fuel grapple, which would cause a fuel assembly to be dropped onto the reactor core. Thi.s fuel handling accident is analyzed in Chapter 15. The refueling platform is designed to prevent it from toppling
- into the nools durino a safe shutdown earthauake (CCF). ,
i Redundant safety interlocks are provided, as well as limit ; switches to prevent accidentally running the grapple into the ,
. noel walls.1 Tne grappie usea Ior muwi movemens as on ene ena of
~a telescoping mast. At full retraction of the mast, the grapple l
_ . is 7 feet below the wat4r surface, so there is no chance of raising a fuel assembly to the point where it is inadequately shielded by water. The grapple is hoisted by redundant wire rope cables inside of the mast and is lowered by gravity. A digital readout is displayed to the operator, showing the exact coordinates of the grapple over the core. The mast is suspended and gimbaled from the trolley, near its top, so that the mast can be swung about the axis of platform
- 3. b62 .
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', O BCGS FSAR 10/83
! OUESTION 410.61 (SECTION 9.1.4) Verify that there are redundant interlocks and limit switches to prevent accidental collision with pool walls or other structures or components on all light load handling devices. ,
)
1
- l
+%. MY
! A H4 11ght load handling (LLB) devices M ave redundant interlocks and limit switches. However, the effects of an LLH device e accidental collision in the sp:nt fuel p;1: d .i it e-i_ch # would be Me! W =ith theredund: effects oft the int::_ fuel:h: i less than handling accident described in Section 15.7.4. i
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The M LLR devices that could be involved in accidental pool hoists,
- wall collisions are the two the two fuel pool ib cranes, and the channel handling j refueling platform auxiliary /oom.
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i - l l l i i DSER OPEN ITEM Nk 410.61-1 Amendment 2
a .s'840268981 HCGS k 8 V i"* DSER Open Item No.144 ( DSER Section 9.2.1) STATION SERVICE WATER SYSTEM , The SSWS consists of two redundant piping loops from' dh,e ' Delaware River to the plant. Each loop contains two 504 capacity pumps that are ' powered f rom a Class lE power supply. 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 staff can conclude that the guidelines of RG 1. 29, " Seismic Design Classification," Positions C.1 and C. 2, are satisfied. The design of the SSWS ensures that system f unction is not lost 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 failure (event) with the loss of one SSWS pump (single active failure). Therefore, we cannot conclude that the requirements of General Design Criterion 44, " Cooling Water," are satisfied.) The SSW pumps are normally operating. The availability of the standby pumps is ensured by periodic functional tests and in-spections. The systen design also incorporates provisions fior 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 respect to protection from natural I phenomena, capability for transferring the required heat loads, 4 inservice inspection and functional testing.] However, the staff concludes that the systen meets the guidelines of RG 1.29, Positions 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 not meet the acceptance criteria of SRP Section 9.2.1.] 144-1 l
f Es.s740268381 HCGS
RESPONSE
For information on the protection of the SSWS frami flood water see the response to DSER Open Itant No. 5. l Our response to Question 410.66 cmpletely describes oJr design with regards to pipe break and loss of a service water pump. Briefly stated our design does not (and is not required according to BTP ASB 3-1, Section B. 3.b . ( 3 ) , for redundant trains of a dual-purpose moderate-energy essential system) consider non-mechanistic pipe breaks along with an additional single active failure of a Pump. The frequency of the camponent functional testing will be in accordance with the criteria of . IWP and IWV of Section XI of the i ASME Code. See FSAR Section 3.9.6 for the specific edition and addenda. FSAR Section 9.2.1 has been revised to state that the level and frequency of inservice testing is included in FSAR Chapter 16, Technical Specifications. s K53/1 144-2 i
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, NCGS FSAR .
1/84 9.2.I'.6 Tests and Inspections The system is hydrostatically tested' prior to the station , > operation. All active components, e.g., pumps, valves, and l controls, are functionally tested prior to startup and ! periodically thereaf ter. LE vfl AN resryNg As /NcL4/Oso /N cMAprsa M,O FREqd1NC)' TscwocAL of /NSKAME.:%eraja1CA M A45. l Inservice Inspection and functional testing of the safety-related portions of the system and components will be in accordance with the esamination 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 i 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. : t 9.2.2 SAFETY AND TURBINE AUXILIARIES COOLING SYSTEM The safety and turbine turiliaries 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 offsit.e power (LOP), and a loss-of-coolant accident (LOCA). DSsR OPsN ITsM / 9.2-8 Amendment 4
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.DSER OPEN' ITEM No '183 (Section 18)
HOPE CREEK DCRDR P'rovide SPDS safety analysis
~
RESPONSE
As discussed in our April 2, 1984 letter to the NRC, the Hope Creek SPDS will be based on the graphic displays
-developed by the BWR owners group. Preliminary displays have been. prepared,sbut they must still go through an operator evaluation program. As discussed in the referenced letter we are presently committed to a September 1, 1984 submittal'to the NRC. The display evaluation program is due to be completed in September 1984, and the final report is due in December 1984. We will not be able to make a submittal until December 1984 based on the above program delays.
e
April 2, 1984 l i Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
, 7320 Worfolk Avenee Bethesda, Maryland 20014 Atts: Mr. Albert Schwencer, Chief Licensing Branch 2 ,
Division of Licensing j Centlemens j HOPE CREEE GENERATING STATION DOCERT No. 50-354 GENERIC LETTER 42-33: SAFETY PARAMETER DISPLAY SYSTEM Pursuant to the commitsent made in Attachment A to our l 1etter of April 14, 1993 (R. L. Mitti, PSE4C, to A. l Schwencer, NRC), the following is an update of the status of ; the Safety Parameter Display Systes (SPDS). ! In Generic Letter 82-33, the MRC identified the SPDS as a ; priority safety improvement, deserving precedence over control room design review, emergency operating procedures development, and other activities referenced in NUREC 0737, Supplement 1. Utilities were advised to implement the SPDS, and then perform post-installation analyses to verify para-meter inputs, display layouts, and operator comprehension all after the fact. The bases for these analyses were to be function and task analyses using the revised Emergency Procedure Gelde11nes (EPCs)/ Emergency Operating Procedures (E0Ps). The protest for proceeding with the SPDS first was that EPCs were unavailable and that undue delay in SPDS implementation sight result from a procedures-based approach. Now, how-l ! ever, EPCs are available for DWRs in essentially final form, and utilities can proceed to develop proceduree-based ,
)
1 1 : DSER OPEN ITEM /83
Mr. Albert Schwencer 2 4/2/84 displays (or extend existing display sets to support procedures) without any substantial time penalty. In short, l
- top-down* SPDS design and implementation for BWRs is no l longer inconsistent with the priority SPDS designation.
l Section 4.lf of NUREG 0737, Supplement 1, stipulates that the SPDS provide information to plant operators about reactivity control, reactor core cooling and beat removal, i reactor coolant system integrity, radioactivity control, and i containment conditions. The EPGs (and ultimately EOPs) address these functions and conditions symptomatically, so ! that displays developed from the EPG/EOPs will provide the ! required information. i A functional and task analyses will be performed on the EPGs which will provide an overall display concept. This display concept will include a listing of plant parameters. i The function and task analyses also support the requirements i in Sections 5.lb (ii) and 4.2a of NUREG-0737, Supplement 1. In fact, although the results of these analyses will be the i basis for display development, they may also be used to 3 support control room design reviews and develop training ' ! requirements. Rather than employing the analyses to verify 1 SPDS designs, this project lays the groundwork for develop-ing these designs from the analyses. In other words, func-
- tion and task analyses will be used as design input rather
- than post-implementation review criteria for the SPDS.
Since the EPGs address the functions and conditions specia f fled in Section 4.lf in NUREG-0737, Supplement 1, and since the function and task analyses identify the information requirements of the cperators, the SPDS display development methodology developed under this project binds Section 4.lf { 1 with Sections 4.2a and 5.lb (ii). A direct consequence is a j set of SPDS displays which fully support the procedures information requirements, which, in turn, encompass the basic functions and conditions specified in Section 4.lf. Hence, the initial HCG 8 SPDS parameters were selected so j that the graphic displays being developed by the BWR Owner's
- Group, based on Revision 3 of the Emergency Procedure i
Guidelines, can be implemented. The BCGS plant specific DSER OPEN ITEM /t3
.~-- J s 1 l
Mr. Albert Schwencer 3 4/2/84 displays and associated function and task analyses based on the EPGs and EOPs will be provided to the NRC for review by September 1, 1984. Very truly yours, i i R. L. Mi tl General Manager - Nuclear Assurance and Regulation - 22A JESadb C D. H. Wagner USNRC Licensing Project Manager Mr. W. H. Bateman
, USNRC Senior Resident Inspector BC General Manager - Engineering , General Manager - !! ope Creek Operations ! Project Manager - Hope Creek Manager - Licensing and Analysis Chief Project Engineer - Hope Creek
. Chief Controls and Electrical Engineer i J. P. Whooley l Associate General Solicitor l B. C. Markowitz - Bechtel Power Corporation
! Conner a Wetterhahn J. Ashley - Bethesda Office CARMS OB 04 01/03-D e
DSER OPEN ITEM /[3 1 l
Re v 1-HCGS DSER Open Item No. 192 (DSER Section 7.2.2.9) REACTOR MODE SWITCH We require the applicant to augment the response concerning the reactor mode switch to indicate that HCGS has responded to IE Notice '83-42 and that the In modified mode addition, we switch requirewill thebe installed prior-to fuel load. applicant td clarify the response regarding the use of operational procedures as ' the primary method of controlling rod movement during refueling.
~
RESPONSE
The response to Question 421.26 addresses the concerns about the reactor mode switch installed at HCGS.
, t i
L ! s 1 h I MP84 56/11 6-db
l NCGS FSAR 4/04 00EST10N 421.26 (SECTIONS 7.2, 7.3, 7.7) , l Mode switch contact and mode switch operating mechanism i malfunctions have caused inadvertent protective actions. Similar malfunctions could have rendered redundant channels of protective functions inoperable. IE Information Notice 83-42 provided notification of potentially significant events concerning mode switch malfunctions. Section 7.2.1.1.11 of the FSAR indicates
- that the Further reactor discussion mode switch can bethe regarding used to initiate reactor mode aswitch reactor l scram.
capability to bypass and enable protective functions and provide ! rod withdrawal interlocks and refueling equipment interlocks is not provided in Section 7.2. Provide a detailed discussion on how the mode switch is incorporated into the overall design, supplemented with detailed drawings and schematics. Please i include the following: i
- 1. Identification of the reactor protection system, rod block, 1 refueling interlock and other functions important-to-safety that are dependent on proper mode switch contact operation.
! 2. Identification of the analyzed transients and accidents l where credit is taken for the operation of any function identified in 1 above. l 3. The surveillance actions necessary to positively verify mode l l switch contact positions, detect mode switch contact failures and detect mode switch operating mechanism failures for each function identified in 1 above. RESPONSE {~g The reactor mode switch curr tly installed at HCGS is of the l l type having the potential for h as described in IE This switch will be replaced prior to i Information Notice 83-42. i fuel load with a modified switch having an identical contact configuration and wiring scheme. Jn hus..medL An assessment A of the system impact of postulated mode switch misoperations f or the presently installed' " HCGS mode switchges -
-r: rid:e ir th: ::;;;t * * --- --- ' f ?::te. t:ft
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"Mpro;vides an TJu m
#~~ evaluation of the impact of postulated mode switch misoperations on the analyses described in Chapter 15. It identifies normal "RUN7
', SHUTDOWN, switch cpptact ppsitigns for gach mode of operation REFUEL, and STARTUP - and summarizes the system consequences should one or more pairs of contacts misoperate.
All of the identified misoperations of contacts are detectable by annunciation, instrumentation checks, or surveillance tests. !
-?;... :;;;;.e. .ill L; ;;fified by 2; : !!!! te id::tify th: <.
i . osER OPEN ITEM /9 N 421.26-1 Amendment 5 l l __t_._
NCGS FSAR 4/34 Ph=p*ar 5 ::;idgni e,.ely;;; th;t b;;;f eede ;;itchs misopec t5 ::.t 1 Assessment of the effects of Mode Switch Misoperations. - 1 1 I. lNode Switch Contacts 1-2, 9-10, 17-18, and 25-26 (Division 1 through 4 respectively). The contacts listed above are normally closed when the mode switch is in the *RUN" position and are open in all other mode switch positions (i.e., " SHUTDOWN", " REFUEL" and "STARTUP"). A. If the contacts were to open in thd "RUN" mode, then:
- 1. Scram or half scrams would result as K15 relays are deenergized.
- 2. The IRM functions would be enabled and the APRMs, would revert to setdown trip-level scrams or half scrams if the plant is at greater than 12% to 15%
of rated power or IRMs are upscale.
- 3. A rod block would be annunciated because the RPS l signalssto the reactor manual control system would incorrectly infer that the mode switch is in the
" SHUTDOWN" position.
B. If the contacts were to close in the *STARTUP",
" REFUEL", or " SHUTDOWN" modes, the effects would be similar for all'three modes of operation, i.e.,
In all three modes, the IRM acram function would
~
1. be bypassed. This would not be immediately detectable but would be detected at the time of the weekly channel functional tests because half scrams would not result. Weekly channel functional tests will cover IRMs in "STARTUP", <
" REFUEL" and hot and cold " SHUTDOWN".
- 2. ., Assuming two or more contact pairs fail, the 1
" SHUTDOWN" scram would be redundantly bypassed in the "STARTUP" and " REFUEL" modes with no consequence although this would not be immediately detectable. However, if the contacts were closed prior to going to " SHUTDOWN", either no scram or only a half scram could result when the mode switch is placed in " SHUTDOWN", assuming that the contacts remain closed during and after the transfer to " SHUTDOWN".
i 421.26-2 Amendment 5 l DSER OPEN ITEM
/Q :
HCGS FSAR 4/34
- 3. Ir. all three modes (" SHUTDOWN", ' REFUEL", and "STARTUP") the APRM setdown scram function would not be in effect, APRM setpoints would be raised to their high setpoint level (approximately 120%
of the NB rated power). Again, this condition would not be immediately detectable, but it could be detected at the time of the weekly channel functional tests.
- 4. In all three modes (" SHUTDOWN", " REFUEL" and "STARTUP") there would be an unannunciated capability or permission via the reactor manual control system to move more than one control rod according to the rod pattern control definition.
This would be a redundant permission in the "STARTUP" mode. In the " REFUEL" mode, the operator would not be directly aware of this capability unless he attempted h-to' withdraw more than one rod. 'ha -- -- - Eter ^^ =5e-e ;;;1d ;r:t bly prar'rf: :::h ::ti:7..t The operator ,rilliverify this rod block by awn attempt {Eo withdraw a second rod af ter the first control rod is withdrawn. In the " SHUTDOWN" mode, the operator should be aware of such permission as the normal _ annunciation of"erld the rod withdrawal 804r block would)be presentfer 5:1f :::::::::::d b; :t th:t rereltic; e e--- ti-- == +h- cere c
-fr= It:: J.; e k;;. t C.
Conclusions:
- i. ~ for switch closures where the switch is
" mode, postulated switch f s
the "ST would result in ast a hal . condition at S NJEzr A % the time of the swit re. This would alert
's conclusion the operator PS failure.
ailures only of mode swite -10, addre I and 26-26 contacts.
- 2. The potential consequences related to bypassed IRM trips and high-setpoint APRM trips, while in the "STARTUP" mode, are not determined in this assessment. These are reported in a separate evaluation.
II. Mode Switch Contacts 3-4, 11-12, 19-20 and 27-28 (Divisions 1 through 4, respectively). These contacts are normally closed when the mode switch is in the "STARTUP" position and are open in all other mode switch positions (i.e., " SHUTDOWN", " REFUEL", and "RUN".) 421.26-3 Amendment 5 DSER OPDi ITD4 /pg
INSERT A
- 1. Addressing only the multiple failures of the mode switch contacts l 1-2, 9-10,17-18, and 25-26, the items cf principal concern are:
- a. The unannounced bypass of the IRM scram function in the *
- "STARTUP " " REFUEL " and "$HUTDOWN" modes.
- b. The potential failure to scram by positioning the mode switch to " SHUTDOWN."
- c. The unannounced bypass of the APRM setdown scram function in the "STARTUP." " REFUEL," and " SHUTDOWN" modes.
- d. The unannounced permission to move more than one control rod in the " REFUEL" mode.
, e. The annunciated (via the removal of the signal) removal of the normal " SHUTDOWN" mode rod-withdrawal block. 1 i DSER OPEN ITEM /fA t.
HCGS FSAR 4/34 , preceding, the time delays relays K16A-D, would also be energized such that after 10 seconds, the second set of contacts in the shutdown scram bypass circuitry would clcse. At that time, the
" SHUTDOWN" scram function would be redundantly l .
bypassed (i.e., already bypassed by the "RUN, l
"STARTUP" or " REFUEL" modes). When this redundant l ,
bypass occurs, the operator would be alerted since the annunciators would indicate "RPS Mode Switch Shutdown Scram Bypass".
- 4. If the switch were in the " REFUEL" mode, a redundant mode-switch permissive signal to allow bypass of the high water level trip function on the scram discharge instrument volume would occur with no consequence. However, if the permissive signal does exist in the "RUN" or "STARTUP" mode, the bypass switches for the SDV high water level trip must have been incorrectly placed in the bypass position. This switch placement is abnormal for these modes. Such bypass is annunciated and would initiate a rod withdrawal block at the same time.
C.
Conclusion:
Addressing only the failure modes of the mode-switch contacts 7-8, 15-16, 23-24 and 31-32, the items of principal concern are the annunciated bypasses of the MSIV closure scram function and steam-line low pressure isolation function when the switch is in the "RUN" mode. Ten seconds later the operator would receive an additional input that something is wrong with the RPS when the annunciation system indicates that the "RPS mode switch shutdown scram bypass" is in effect. A44cWe4=f V. Summary and Conclusions e[ A b M S N h e rf M N ^ A. All failure modes for the mode switch contacts where contacts open that should be closed would result in scrams or half scram depending on the number of contacts that are open. At the same time, for conditions of operation where steam line pressure is low, isolations of the main steam lines would occur. B. In the "STARTUP", " REFUEL", and " SHUTDOWN" positions of ! the mode switch, contact closures of the contacts that should be open, (i.e., contacts 1-2, 9-10, 17-18 and/or 25-26) would result in a bypass of the IRM scram function in one or more of the RPS channels and also would result in raising the setpoint of the normally DSER OPEN ITD4 / fk
NCGS FSAR 4/34 setdown APRM high fluz scram function from 15% to list in one or more of the RPS/NMS channels. C. In item B above, although the mode switch failure, (i.e., contacts closing), would not be immediately apparent to the plant operator, the failure would be detected during the weekly IRM and APRM channel functional tests. If these te were performed prior _, to the power increase and afte ransfers4affEhe mode switch to the "STARTUP" positio , then the IRM channel functional tests would detect Oh: the failures because preg:::dt W ir: : no half scram would result. th t th : = chst;;;"t _encific tica r e ir::::t .;i!! i fe--ti:n:1 L i e.,0 the J. = ch;;;;I !...ctienei t;;t b;'~ aarfarrr' rithin M 5:::: prier te etertep, if it h:2 %
.at s_--- y -fa - ' in t h. nr vi a. . m.v.n days. e_--p y s
-er r:ill:::: .,.la " ::quir:d f:: th: 02-- rh:::hy th:,
"** STEMY" :nditi;; ie seint in:' 'ar lang nari~* %
ef tin:. J.dditenel infe--etica rill L p :vid:d i= A.no 1 m vn ihe ewde e.; itch. - D. In the "RUN" position of the mode switch, contact closures of the contacts that should be open, (i.e., contact closures of all contacts other than 1-2, 9-10, 17-18 and 25-26) would result in the bypass of'one or more RPS trip channels related to the MSIV closure scram functions and would also result in the bypass of one or more NSSSS trip channels related to the steam-line low pressure isolation function. Concurrent with the incorrect mode switch contact closures, there would be annunciations that one or more of the RPS MSIV If closure scram trip channels have been bypassed. l contacts 7-8, 15-16, 23-24 and/or 31-32 were to close, then the operator would receive additional information that something is wrong approximately 10 seconds af ter the contacts close with the annunciation that the,re is a "RPS mode switch shutdown scram bypass". Addi$n::: 4nfers ti:n .ill M previde' in 2:n 19 M = th: ose 4 (
,,..:-...,-.2.
-. v . . . . . - g E. Closure of several sets of contacts can bypass the
" SHUTDOWN" mode scram functiog If the contacts remain closed during and after transfer of the mode switch to the " SHUTDOWN" position, such closed contacts would not allow a scram to occuzh. This That is, only a half scram or sw. Ms. fact would be immediately d* -
no scram would result. :D.ux; Jds l
)) a,pparent to the pperator.7/4,q&
{ '^^^^^ : Aech4 w & 2ie m Rw-&p(nata #eMN y F. M &*"**
- FUEL" mode, closure of the mode switch!f4 In the RE 4, 25-26 and/or 27-28 would negate the contacgs normal rg ygl 1-2,y mo de restriction of the "all rods in first, bove 6nly one rod at a time" and would allow the
"" " #" /fA 421.26-10 Amendment 5
HCGS FSAR 4/84 movement of any number of rods within the rod pattern definition existing at the time. This fact would not be apparent to the operator if contacts 3-4 and/or 27-28 were closed.) G. In the " SHUTDOWN" mode, closure of the mode switch contacts 1-2, 3-4, 5-6, 25-26. 27-28 arid/ar 29-30 would-f remove the normalh od withdrawal block (restriction el A.]
' associated with this mode. This fact would be apparent to the operator because the window for the normal rod withdrawal block annunciator would be extinguished, and its change of state would alert the operator.
paceN kl== cal spe=fic.R=e d Dh * *** '"l N'I*. ' 5 *d b a
- 3:* r % k
- 4. Also, h 4e ad bl=.k LwMa.t 4 , W* +-n N@fer av m w . % ,.h ,
d , . a6 _ [NSERT Bf NEW DSER OPEN ITEM /pg 421.26-11 Amendment 5 1 _ _ _ _ _ , , _ _ _ _ _ , , _ _ _ _ __
e
, ns INSERT B
- 2. Evaluation of the Impact of the Effects of Mode Switch Misoperation '
on the Chapter 15 Analyses. The potential impacts of the effects of mode switch misoperation on the analyses of transients and accWnts presented in Chapter 15
- were evaluated. The focus was on certain specific events because of- previously expressed NRC concerns with those events or because the events might impact the limiting transients. These specific events were c assified into two groups according to the consequences of mode switch misoperation.
I. The Evaluation of Group 1 Events The events in Group 1 include:
- a. The abnomal startup of an idle recirculation loop.
- b. The failure of the recirculation flow controller with ircreasing flow.
- c. A rod drop accident.
These are events for which the concern is related to the i bypass of the scram function of the intemediate range monitor (IRM) while the mode switch is in the "STARTUP," " REFUEL," or
" SHUTDOWN" positions. This would also raise the scram ' setpoint of the average power-range monitor (APRM) from the 15% "startup" value to the 118% "run" value, which corresponds to the analytical limit of 121% used for the' analyses of Chapter 15 transients and accidents.
None of the Chapter 15 analyses of the events in Group I takes credit for either the IRM scram function or the APRM scran function with the setpoint setdown to the 15-to-25 level. Events a and b of Group I were analyzed from a "RUN"-mode power condition since the Chapter 15 analyses are initiated from about 55% power and 38% core flew. In the "RUN" mode, the IRM trips are bypassed and the APRM flux scram-setpoint is approximately 118% (121% analytical limit). The rod drop accident analysis was initiated from 0% power, (501 rod i density); consequently, the mode switch would be in the "STARTUP" position. No impact would result from the misoperation of the mode switch in the " REFUEL" or "$HUTDOWN" modes. A. For the analysis of the abnomal recirculation-loop startup transient, no creditustaken for the flow reference in the scram for high neutron flux. The high neutron flux setpoint of 1211 was used. The Analysis of this event was initiated from a power level significantly in excess of where recirculation-loop startups would DSER OPEN ITEM hd l
i nomall{RUN" in the mode.eriginate and levels, At lower power corresponding the to the mode consequences of the event would be less severe; the impact of the mode switch misoperation' consequently,is on the analys of this event is of no significant consequence. , The initiation of an abnomal recirculation-loop startup i i transient when the mode switch is in the "STARTUP" position would also be of no consequence since operating procedures would require the initial power level to be less than 155. The resulting power increase probably would not cause a scram. If the resulting power level were in excess of technical specification requirements related to power, pressure, and core flow, the operator would take corrective action in accordance with those requirements. B. The Chapter 15 analyses of the recirculation flow-controller failure with increasing flow were i initiated from a 57% power and 40% core flow conditions, with a 1215 flux scram terminating the power excursion. Similar events originating from the startup power range of 0 to 15% power would be of lesser consequence. Also, l at this low power level, normal operating procedures I would infer minimum pump speed with individual loop operation. These operating conditions would lessen the effect of a single-loop flow increase and would preclude the event of flow control failing with increasing flow on both loops. C. The analysis in Chapter 15 of the rod drop event only takes credit for the 121% APRM trip and takes no credit for the IRM scram function. The event, as analyzed from the 0% power level, is terminated by the Doppler effect and is of significance only below about 2 to 31 power. At high power levels, the rod drop would be less of a problem because of the influence of the resulting steam voids in the core on the local high reactivity. II. The Evaluation of Group 2 Events The events in Group 2 include:
- a. The inadvertent closure of the main steam isolation valve.
- b. The loss of an auxiliary power transformer. l l
- c. The break of a main steam line outside the i containment.
- d. The failure in the open position of the steam l pressura controller.
i DSER opgy
-~
i . These are events for which the concern is either the byoass of the main steamline isolation function due to low steanljne pressure by the nucl:ar steaa s:pply shutoff system (NS ) in the "RUN" mode or the less of the position scram function of ! the HSIVs in the "RUN" mode. Only the isolation function that - l should result whenever the turbine-inlet steamline pressure drops below the (analysis) setpoint level of approximately 825 - psjg is of conc 6rn. No other isolation functions of the NS are impacted by the potential mode switch misoperations. A. The analysis of the MSIV closure event in Chapter 15 does take credit for the scram initiated from limit switches of the MSIV while the mode switch is in the "RUN" mode. potential mode switch misoperation could cause this scram function to be bypassed while the mode switch is in the ;
"RUN" position. However, this bypass would be ,
annunciated in the control room. The operating i procedures would require corrective action since the ' technical specification requirement that all four channels for the MSIV-closure trip function be operable in the "RUN" mode would be violated. Depending upon the number of inoperable channels, the affected channels and at least one trip system of the reactor protection system [ haw f g, yg I (RPS) would have to be placed in the tripped condition i desgn sE M N within one hour. If both RPS trip systems were affected, i the plant would have to be placed in the "STARTUP" i twag3g (gg condition within 6 hours. I888 8b
$scfda $.[h*1 i B. The consequences of the n auxiliary power, as analyzed in sesed We d b 6 Chapter 15, are also not affected by any mode switch k saw, consegwecasy "U f"M"4.IEfi T R f M E E li 5 12.
x: urn: u r rur-cr.x , . ss k tw of sR
?
27:grr :1- rwr 1c- y g tenuMou. - 3 37 ~ '"""~ "" " " ' ~ ~ - -
%e dera d C. The analysis of the main steamline break outside of the l MM*n tsafrel I containment does not take credit either for the low steamline isolation signal that would probably result O s ,satam e ,a g fron low steamline pressure or for the scram from MSIV g g gg closure. In this analysis, the event is initiated at the Level 3 scram to start out with a minimum inventory. At Maded af Ya'sc. cbout 0.5 seconds into the event the isolation,is assumed to be initiated because of high steamline flow. Although SM N8 *f s this is not addressed in the analysis, a level-8 yygg hd F '
high-water turbine trip would be expected due to sudden depressurization. musapermTun Failure of the steam pressure controller in the open [ pi(g g Q j ggD. position would result in a level-8 high-water turbine trip, which would initiate a scram and a recirculation pump trip. Since the depressurization would be limited to the capacity of the turbine bypass, the low-pressure isolation would be delayed to beyond a minute. Since_an anrunciation in the control room would have alertetMo yn d6' the bypass of the isolation function, the operator would be prepared to actuate MSIV closure manually should this event occur. ( DSER OPEN ITEM / f J., l
.e III. Conclusion cf,the Event Evaluations Conclusions from these evaluations are that all misoperations '
of the mode switch are detectable by one or more of the following means. A. The operator would be imediately aware of a problem because of the annunci ion of bypasses that should not exist for the given po ion of the mode switch. All mode x switch misoperations t t might impact the severity of consequences of transients and accidents analyzed in Chapter 15 are in this category. Hence, the probability , of a transient occuring before the operator takes l corrective action would be extremely low. B. The operator would be innediately aware of a problem in the RPS because of scrams or half scrams, which are also annunciated. C. The remaining modes of mode switch misoperation would be detected during the weekly channel functional tests of the NMS channel inputs to the RPS. If these tests were performed prior to the power increase and after the transfer of the mode switch to the "STARTUP" position, the IRM channel functional tests would detect the failures because no half scram would result.
- 3. Surveillance Actions Necessary to Detect Mode Switch' Misoperation The proposed technical specification requirement is that the IRM l
channel functional test and the APRM channel functional test be perfonned within 24 hours prior to startup, if it has not been perfonned in the previous seven days. Also, weekly surveillance should be required when the " HOT STANDBY" condition has been maintained for long periods of time. f DSER OPEN ITEM / f./.
l ReV 1 l
. HCGS , DSER Open Item No. 200 ( DSER Se ction 7.4.2.2)
REMOTE SHUTDOWN SYSTEM I l The applicant is required to confirm that the HCGS design meets the staf f's guidance for remote shutdown capability.
RESPONSE
The response to Question 421.38 identifies how the HCGS remote shutdown systems design meets the NRC staf f's guidances fo r remote shutdown capability.psdd s ecd /o n 7. V. /. V. 3. $~ A a s h e e ri r- e s i s e d p e ,- a , s e 3 3 i o n ., win ne a c c._ o n r a ,, c 2% 19 8V. e i I I \ l =
. .. --. w.,ww i
r HCGS FSAR 4/84
~
. t
~
7.4.1.4.3.5 RSS Redundancy and Diversity l The instrument'ation and controls on the RSP are redundant to one train of safe shutdown systems in the main control room. Operation of the transfer switches on the RSP to the emergency position isolates only this train of safe shutdown systems from the main control room. Sufficient instrumentation and controls redundant to those at the RSP are available at the switchgear panels (shown on Figure 1.2-36) and the control equipment room (shown on Figure 1.2-35) such that no postulated single failure at the RSP can prevent achieving and maintaining (verifying) the reactor plant in a safe shutdown condition when the main control room is uninhabitable. No jumpering, rewiring, or disconnection of circuits will be required to accomplish this. The capability for local manual or remote-manual operation of valves which must be repositioned to achieve cold shutdo'wn has been provided. This ensures cold shutdown capability in the event of a failure to a valve remote control circuit in accordance with the requirements of GDC 19. Instrumentation and controls redundant to that provided at the P are listed and described in Table 7.4-3. f
-/ AIM k -
7.4.1.4.3.6 RSS Actuated Devices l Operation of the RSP transfer switches to the emergency position will initiate signals to close or open certain valves to facilitate safe shutdown. These valves and their after-transfer cositions are identified on Table 7.4-2. ECCS jockey pumps DR-228 (RHR loop B) and BP-228 (RCIC) are l signalled to run on RSP transfer. Operation of the RSS equipment identified in Table 7.4-3 as being redundant to the equipment provided on the RSP is indepedendent of the RSP transfer switches. DSER OPEN ITEM MN l 7.4-12 Amendment 5
. . _ . , . ,..__.____,v_ --
JLt -6 '84 0 2 6 7 4 5 0 i 1 INSERT- 4-I e In the event that the RSP is lost, the design provides for separate equipment independent of that in the RSP. This equipment is presented in Table 7.4-3 and all is designed in accordance with Class IE requirements. l e DSER OPEN ITEM mob l
I DSER OPEN ITEM 217 (Section 9.5.1.1) Fire Protection Program Requirements The applicant has committed to conform with the guidelines in BTP CMEB 9.5-1, Section C.1.a; however, he has not provided sufficient information for the staff to independently verify compliance with its guidelines. The staff will require that the applicant provide descriptions of how the fire protection organization, responsibility assignments, and personnel qualification comply with the guidelines in BTP CMEB 9.5-1, Item C.l.a.
RESPONSE
FSAR Section 9.5.1.5.1. has been revised to provide the information requested above. l
! HCGS FSAR 1/84 , ' l ! I 9.5.1.4.2.10 Fire Barrier Penetration Seals l l Fire barrier penetrations are inspected periodically to ascertain penetration integrity. 9.5.1.5 Personnel Qualification and Trainino
, de l e+ e._
9.5.1.5.1 Design and Installation Qualification Bechtel Power Corporation is responsible for the preparation of the fire hazards analysis, as well as the plant design and selection of equipment. Completed fire protection systems have l been inspected and tested by qualified personnel under the direction of Bechtel. I n s e r-t(it s w - j 9.5.1.5.2 Fire Brigade Organization, Training, and Equipment A minimum Fire Brigade of five (5) men will be available per shift. The Fire Brigade will be composed of dedicated personnel available to fight fires in both Salem or Hope Creek Generating Stations. At least three (3) of the brigade members will be trained in Hope Creek Safety Systems. The Fire Brigade Training Program will include instruction in the Station Fire-Fighting Plans, identification of hazards within the station, location and use of station fire fighting equipment, proper communications and the proper method for fighting fires
! inside buildings and confined spaces.
The Fire Brigade will be provided with personal protective equipment, emergency communications equipment, and portable fire fighting equipment required to fight those fites that may occur within the station. osza oPEN ITEM o7/ 7 9.5-45 Amendment 4 e -:
' M o 9.5.1.5.1 . Fire Protection Organization b Personnel Qualifi-cations
- The Vice President - Nuclear has ultimate res'ponsibility over the Hope Creek Fire Protection Program.
The General Manager - Nuclear Services shall establish overall policy relating to the fire protection program and shall coordinate overall fire protection support from Nuclear. Support, Hope Creek Operations and Quality Assur-ance. , ! a. The Manager - Nuclear Site Protection will formulate specific fire protection policies and will develop, implement ,ano check compliance with the fire protec-tion program. He . is supported by a staff which is knowledgeable in nuclear plant fire protection.
- b. Proto-Power Management Corporation will assist in the j develorment of the Hope Creek Fire Protection Pro- i gram. Proto-Power maintains on its staff Fire Protec-tion Engineers who meet the requirements of member of
! the Society of Fire Protection Engineers. The fire protection program at Hope Creek will be designed to implement a defense-in-depth concept towards fire pro-tection to all safety-related areas.
- a. Reporting to the Manager - Nuclear Site Protection is .
a decicated group of fire department personnel trained in fire fighting, fire protection systems and fire protection administrative procedures. The fire cepartment will have the primary responsibility of , ensuring the readiness of fire protection systems and equipment. The fire department will also be respon-i sible for ensuring the proper control of combustibles i and ignition sources; In the event of a fire, the fire department will be fully trained and capable of performing fire fighting duties.
- b. The fire brigade members will be required to satisfac-torily complete a physical examination for performing ;
strenuous activity and the fire brigade training pro- l gram which includes testing of fire protection equip- l ment and systems. 1 g The Manager - Nuclear Training will be responsible for g implementing the fire protection training program for g operating. plant personnel and the fire department. g a. The General Manager - Nuclear Support will evaluate l l m design changes anc plant modifications for impact on l the Fire Hazards Analysis. ! , k0 l l
- b. The initial design of fire protection systems and the E Fire Bazards Analysis was performed by Bechtel Power 3 Corporation under the direction of fire protection engineers who meet the requirements of member of the -
Society of Fire Protection Engineers. _,~ . _ _ _ _ -__ __ _ ,_ _ _ _ ____ _ _ __ _ _ _______ _ _ _
M 1 il
- . 'DSER OPEN ITEM No. 219 (Section 9.5.1.2)
. Administrative Controls The administrative controls for fire protection consist of the fire protection program and organization, the fire brigade training, the controls over combustibles and ignition sources, the prefire plans and procedures for fighting fires, and quality. assurance. The applicant has committed to meet' staff Lguidelines in Section C.2 of BTP CMEB 9.5-1; however the applicant has not provided sufficient information for the j
sta'ff to independently verify compliance with its guidelines.
, The staff will require that the applicant provide details-showing compliance with the guidelines in BTP CMEB 9.5-1,
- Item C.2 regarding administrative controls.
i i
RESPONSE
i FSAR Section 9.5.1.5.3 has been revised to provide the information requested above. i-e I i
- , ,- . . . , . . . - . . . ~ . , _ , - . . . .. . _ - ,.,. - _... - ,_.. ,.-... - .._, . , , , . . . - - , - . . . . _ . _ _ . _ . .
HCGS FSAR 4/84 9.5.1.5.3 Administrative Controls _uliert y Administrative controls will be developed for the purposes o controlling combustibles, ignition sources, and hot work authorization. L - ot ei e+e_ 9.5.1.6 SRP Rule Review All differences and clarifications discussed below are < differences and clarifications to BTP CMEB 9.5-1, Revision 2, dated July 1981. 9.5.1.6.1 Paragraph C.1.c.(2) Paragraph C.1.c.(2) requires that a single active failure or crack in a moderate-energy line in the fire suppression system should not impair both the primary and backup fire suppression capability. At HCGS, automatic and manual sprinkler / spray systems headers are connected to in-plant loops that are fed from the main underground fire protection water piping or yard loop by two separate lines. Since the in-plant loops are fed by two separate supplies, they are considered an extension of the main underground yard loop. Except for the radwaste/ service area, automatic and manual sprinkler / spray systems and hose stations serving a single area, including safety-related areas, have takeoffs from the in-plant loop, separated by sectional control valves normally locked open. The header arrangement is such that by manual positioning of the sectional valves, no single piping failure can impair both the primary and backup fire protection provided for a single area. Failure of the fire water piping in the radwaste area will not affect safe shutdown of the plant. HCGS is designed for , separation of redundant safe shutdown trains to meet the ' requirements of Appendix R to 10CFR50 to the extent noted in Appendix 9A. I 9.5.1.6.2 Paragraph C.5.a.(1) l
"" U" I" 2/ 9.5-46 Amendment 5
i Xn 3 e.rY
- 1. Administrative controls will be implemented at Hope Creek for the purposes of controlling combustible and hazardous materials, controlling ignition sources, controlling fire protection system impairments and testing fire protection equipment ano systems.
- 2. Procedures will be established for governing actions to be taken in the event of a fire for general employees, con-trol room personnel, fire brigace members and other requirac support groups. Pretire plans will be written for all safety-related areas.
' Administrative Procedures will be written and available for review by Orch 1, 1985. Procedures will be imple-mencea for all areas where new fuel is to be storec prior to receiving fuel on site. The fire protection program
.will be implemented in its entirety prior to fuel loac.
DSER OPEN ITEM ((
DSER OPEN ITEM No. 220 (Section 9.5.1.3) Fire Brigade and Fire Brigade Training The applicant has not provided a description of the plant fire brigade, including equipment and training to verify the guidelines contained in BTP CMEB 9.5-1, Item C.3. The staff will require that the applicant provide details showing compliance with the guidelines in BTP CMEB 9.5-1, Item C.3, in the establishment and training of the fire brigade. Fire brigade training is evaluated in Section 13.2.2 of this report.
RESPONSE
FSAR Section 9.5.1.5.2 has been revised to provide the information requested above.
A HCGS FSAR 1/84 9.5.1.4.2.10 Fire Barrier Penetration Seals Fire barrier penetrations are inspected periodically to ascertain penetration integrity. i 9.5.1.5 Personnel Qualification and Trainino 9.5.1.5.1 Design and Installation Qualification Bechtel Power Corporation is responsible for the preparation of the fire hazards analysis, as well as the plant design and selection of equipment. Completed fire protection systems have been inspected and tested by qualified personnel under the direction of Bechtel. 9.5.1.5.2 Fire Brigade Organization; Training, and Equipment , inimum Fire Brigade of five (5) men will be available p l shi The Fire Brigade will be composed of dedicate rsonnel t availab o fight fires in both. Salem or Hope Cr Generating Stations. east three (3) of the brigade ers will be { trained in Hope k Safety Systems. The Fire Brigade Training P will include instruction in the Station Fire-Fighting Plans, - ification of hazards within the station, location and use stati fire fighting equipment, proper communications the proper hod for fighting fires inside buildings an onfined spaces. The Fire Br ade will be provided with personal pr etive equipmen , emergency communications equipment, and por ble fire figh g equipment required to fight those fires that ma ccur wit n the station. J I n s er1b . ) DSER OPEN ITEM A/O 9.5-45 Amendment 4
I
% ndtr$
- 1. The Hope Creek Fire Brigede will be comprised of de@icated personnel from an eight man trained fire fighting team capable of responding to a fire in either Hope Creek or Salsa nuclear plants. At least three of the eight man team will be sufficiently knowledgeable in Hope Creek safety systems to uncerstand the effects of fire and fire suppressants on safe shutdown capability.
- 2. The fire brigade will be. provided with personal protective clothing, emergency communication equipment, portable lights ano portable ventilation qquipment. Portable extinguishers will be available throughout the plant. At least ten self-containec breathing units with a minimum of 30 minute rating will be available for fire brigade per-sonnel'. Aoditional self-contained breathing units will be available for control room personnel. At least two extra
' air bottles per unit will be located on site. .In acci-tion, a six hour supply of reserve air will be provided by means of a breathing compressor and storec air available to Salem or Hope Creek stations.
- 3. The fire brigade training program will ensure that the capability to fight potential fires is established and maintainea. The program will consist of verbal ciassroom training, f tre fighting practice and fire drills. Details of the training program are incluced in Section 13.2.2.
l DSER OPEN ITEM dSO L - . _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _
DSER Open Item No. 239 (DSER Section 8.3.1.8) TESTING TO VERIFY 80 PERCENT MINIMUM VOLTAGE section 8.3.1.1.5 of the FSAR indicates that safety related motors are designed with the capability of accelerating the l driven equipment to its rated speed with 80 percent of motor nameplate voltage applied at the motor terminals. The designs of each diesel generator unit is such that at no time during the loading sequence does the voltage decrease to less than 75 percent of nominal. By Amendment 4 to the FSAR, the appli-i cant, in response to a request for information, indicated that the only time the voltage dips below 80 percent is upon closure of the diesel generator breaker. The dip below 80 percent lasts for approximately 0.1 seconds and is caused by excitation in-rush of the unit-substations trans formers, the applicant further indicated that the Hope Creek design is such that no motor will i l be started or operated at voltages less than 80 percent of nominal voltage at the motor terminals, and that this design capability will be verified by test of the diesel generator. Verification testing for all modes of plant operation versus only LOCA loading of the diesel generator will be pursued with the applicant. i
RESPONSE
I The t-c.spo,s.te h j u c 5Non V30 .a c ha s b een r e vis e c/ a t s+e.o/ ce ho ve. , , fa p ro v oele %e in [cema Won re i F70( 8)
HCGS FSAR 4/84 OUESTION 430.20 (SECTION 8.3.1) l Section 8.3.1.1.5 of the FSAR indicates that safety-related motors are designed with the capability of accelerating the driven equipment to its rated speed with 80 percent of motor nameplate voltage applied at the motor terminals. The designs of each diesel generator unit is such that at no time during the loading sequence does the voltage decrease to less than 75
- percent of nominal. Justify operating or starting motors with 75 versus 80 percent of nominal voltage at the motor terminals.
RESPONSE
The only time the voltage dips below 80 percent is when the unit-substation transformers are energized upon closure of the SDG circuit breakers. This voltage dip is due to the excitation inrush of the unit-substation transformers. The first motor The inrush lasts load applied is the approximately six cycles. RHR motor, after closure of the SDG circuit breaker. The circuit breaker for the RHR motor has a closing permissive from the bus undervoltage relays. With the current setting of.these relays (set to dropout at 70 percent and pickup at approximately 78 percent), the RHR motor circuit breaker will close when permitted. It takes 4.5 cycles for the RHR motor circuit breaker to close. During this interval the generator has recovered its voltage in excess of 90 percent. This will be verified during the preoperational tests described in Section 14.2.12.1 qpR]P7 9 l During the rest of the load sequencing cycle, if the voltage ! drops below 90 percent, then the voltage is restored to 90 percent of normal within 60 percent of the time interval between the starting of two consecutive loads. No motors will be started or operated at voltages less than 80 percent of nominal voltaae at the motor terminals i t d g pr a e p e l DSER OPEN ITEM /3k 430.20-1 Amendment 5
. 1 i
i l
. HCGS FSAR 1/84 3
4-kV loads on the bus and tripping of incoming feeder breakers to the bus, connecting the diesel generators to the bus after reaching rated voltage at rated frequency, and finally the timed i sequential restart of normal loads.
- 3. Provide a simulated LOCA signal with normal power available and test ECCS integrated response by injecting to the vessel beginning with normal i system lineup.
- 4. Demonstrate the independence and correct load .
assignments for each standby bus. Disconnect the
- Division B, C, and D ac and de power sources and provide simultaneous loss of offsite power and LOCA signals to the Division A bus. Verify load shedding and tripping of incoming feeder breaker to the bus, connecting the diesel generator to the bus after reaching rated voltage and frequency, i
and the timed sequential restart of loads. l Operate the-Division A loads until stable conditions are achieved. Repeat the test for Divisions B, C, and D.
- 5. Demonstrate a simulated LOCA signal simultaneously with a loss of offsite power. Verify diesel i generators load shedding and sequencing and ,
integrated ECCS respon.se. Verlf ve/ ad dess a*#82 . nw'ar termins/s arc. not /en than ! f* **f"henieres{ vol y e b rin3
- mm.
The battery chargers for each standby de bus are to remain disconnected for the duration of the test. The test is continued for a period of 4 hours. Battery voltage and current is measured to check thats i (a) The de system loads remain within the design load profile values for each respective de bus. (b) The de voltage remains greater than or equal to the minimum design value under conditions of steady state and transient loading. DSER OPEN ITEM ,2 3 f 14.2-111 Amendment 4
NEV'I DsBR @n Ites No. 245 (DsBR Section 8.3.3.3.2) f ACEWAY5 l Tus oss Or 18 vamsus 36 INCBBS OF SEFARAT of the FSAR it in is the In stated sections 1 8.1.75 and that separation 8.1.4.14.3.1 between redundant cablel trays oom, l ically i e cable spreading area, control equipment roce,, re ay r and main control room are separated by 18 inches i vert as opposed to the recommended 384-1974. 36 inches of separat on required by IEEE Standard j FSAR, indicated that this by Amendment 4 to th9 h The applicant, 18 inches of separation was approved by the The staf staff'sf during t
, preliminary design review of the Bope for this Creek itenplant.
states i preliminary safety evaluation report that: , l ~
"The applicant claims these separation type cabling distances will be are adequate because a high grade specified and results of extensive testing show that no cable degradation or flame propagation occurs h upperwhen j the lower tray, separated by 12 inches from t e
- i tray, is exposed to a gas flame for 15 minutes. ;
d ion The results of these tests, that demonstrate no the degra tray ex- at licant. to cables located in the trays 12 inches abovep
RESPONSE
to Question 430.5'. Section 8.1.4.14.3.1 and the resp:r.se for have been revised to provi.de additional justification l the separation distance. i I l e t
> DB/en -
. r7o(8) ...--~r.
Mmm
l fev. l NCGS FSAR , 1/34
,, OUESTION 430.51 (SECTION B.3.1 and 3.3.2) ,
! In Sections 1.B.1.75 and 8.1.4.14.3.1 of the FSAR you* state that separation between radundant cable trays in the cable spreading area, control equipment room, relay room, and main control room
. are separated by 18 inches vertically as opposed to the recommended 36 inches of separation required by IEIZ Standard 384-1974. Provide analysis substantiated by test that demonstrate the adequacy of 13 inches of separation.
l RESPONSE , i The HCGS PSAR was approved with It inch vertical separation between redundant cable trays. ? S:tt.;; di;;:: den refer.-ee-0;; tic 5.!.5. A J Opy Vh A.-b.C#I cepsrf %a$ $ tdSbaabtaU
'tb C _//$ C O h N Y - k'CCblCO.f tC$a/~tLk/00___haS ' heCn mbmiAeo! und.ee s epaea.ke__c oxer (
/ htk< + k.n.w.
fS<, b A. ScAwencee . /VAC < s
. la)e/ eAnwt Jr. MfY a s kc XL.$.kY b C.C- / 0 /2 W* l 'h --lY*]* l& /~Q YW $E -O n &pi s ba_S_cd_ao &A l s Th s f ~+a de.mdc $dra'f C- -
Ue Adejus.Ljf_ o h 12 i c4 e s 3cfza M+/ch 1 1 DSER OPEN ITEM gj y, / 430.51-1 Amendment 4 ;_ f -
.c
~ ~ ~~
6 .i
~
S .1. d . ,4 . 3.1 Cabl spreadi Area, ntrol Equipment Roong and S "- Nain Control Room . . :. Mefgtn.::c,a - - -
. ine. AC had F}&S The cable spreading area, control equipment roomb:i;il::n, and main control room do not contain high energy equipment such as i
switchgear, transformer, rotating equipment, or potential sources of missiles or pipe whip, and are not used for storing flammable materials. power supply circuits are limited to those serving These 208/120-V power these areas and their instrument systems. Conduits containing redundant cables are installed in conduits. cables are separated by a minimum of 1 ipch. Conduit couplings, clamps, locknuts, bushings, etc, shall mot be considered For conduits in determining the required separation distances. carrying redundant neutron monitoring cables, bones also shall not be considered in determining the required separation. ; Redundant cable trays'are separated by at The least 18 inches for configurations, vertically and 12- inches horizontally. ^- not be separated by which the redundant - ^';f._ ,; canwill eitner be analyzed or tested to-distances specified above , demonstrate the compliancekwith the intent of Regulatory Guide 1.75. Separation distance requirements between Class 1E and non-l Class 1E raceways are the /same as for the separation among redundant channels. Q p 4MM AT A Strict administrative control of operations and maintenance activities is developed to control and limit the introduction of potential hazards into these areas. 3.1.4.14.3.2 Limited Hazard Areas Limited hazard areas are the general plant areas from which potential nonelectrical hazards such as missiles, pipe whip, and exposure fires are excluded. The hazards in this area are limited to failures or faults internal to the electrical equipment or cables. ,These areas include elevations 77, 102, 124, ~130, and 137 feet in the auxiliary building Minimum wing areas and separation in I elevation 87 feet in the radwaste area. t.hese C;;p.i_._]~. press is as follows:
- a. Conduits containing redundant cables are separated by a minimum of 1 inch, unless consideration of haz'ards indicates greater separation is required. Conduit couplings, clamps, locknuts, bushings, etc, shall not be considered in determining the required separation distances. For conduits carrying redundant neutron DSER OPEN ITEM f//[ /
8.1-21 Amendment 4 -
~ -'
hstA1 A TD Ll*4' 14.% L_ ' 3 w-tg M me!'e3 a ,.k/me vee 4te4( skpe=fic, ee 1b 4*et g, 4 rays , y4e ut.As =:aimem *e4 eat sepe=4;** disfance is ig, ; .nes, the fettsasin aantysis pesvides +Ae.ios4tifcafi*a for
.s+anet *
, '~! !
4sa sessee sepnen4les
' *t.- ****Y N* O**
#:sgt este, are 4teme re4seda * *st meet as deasa***f'I h **f8
! tes+ 3Fai4ied in 2 ass. s t 5-897 4 l gagge 4est refer +s art on 4{te and at*It<W hr **dit. 1"i f *4 A k a,. ns, i=Jiented in 4se *>**c pra1rar n , ki$ ***rjj *W o
- (rem 444 p.4eaftal searees of .,.ssiles or rire ek.r are aset netAs. Pense eirsut4s in +ke area.s nee installed in condai4 s -4k s barriers; the ~<s: man yeta.+;<( .4 4se p aer- cireai+s ga<lify is timt4e<l de EOS/ISD **tts A C. er ISS ve~gis dc. , There are n p eer e.46tes of At 3&se p feaftst servi.9 ,p;7.,.4 /. y4e ,re4s ,
+A<+ <
- c. sks caste +rny 4est repr4 priormal 4.e- salem skomel Stre in a saste +rsy tec<tel st*erdirectly edle setene another 4 ray 4rny mer elegrade 4ne 4il ne+ prep 3<4e 4. 4he urr and cables in ske oppe e<ble 4 ray . The +ess eenfigura4 ten calles ,ere. represenfefive of 44e Nee,5 design tt sad insfadefia.,
ina4 ver+te<( eyeer+ +6m4 %e +ss+ e.enfiguretten used o. seferation. Seenase %e Setem 4est demens+ rated st insk vertical sepenison *aas adequele , 4ke Mct,$se)refr dis +ance is joskilled. 74 sniem fesf ryert, '
"8ms Sr Calle S y sk Destyn Awer Groed,'y ja., j<en s wd,,,; ffe/undw
' sfen s c),tes/ yt/, is, tr77, separk cwer [le+te,- Sn,,, Lt.n;4+('pg n' g g.Sg ,cer., 1 un, dhl thy,r ir, irry) . s 2 .e
) ,
I DSER OPEN ITEM M e(/,[ 1
* '~ ._ . . . . _ . _ . . . . . . _ . . _ _ . . .
6/34 BCGS FSAR OUESTION 210.20 (SECTION 3.6.2) , Provide the basis for assuring that the feedwater isolation check valves can perform their function.followin,g a postulated pipe break of the feedwater line outside containment.
RESPONSE
l Per;r _::
-= =-*-- ?lS_?? ' re**-- * ' # j The feedwater line break outside the containment will be . simulated using the RELAP5 computer code. A check valve model which has been developed specifi'cally for calculations of this nature will be used for obtaining the valve dynamics.
From this thermal hydraulic analysis, the peak pressures - opstream and downstream of the valve disc as well as the maximum disc angular speed will be obtained. As part of this analysis, a sensitivity analysis will be performed to determine the break location and feedwater check valve selection that yields the most conservative stress results. N The stress analysis for'feedwater swing check valve will be performed for fluid transient loads induced by the pipe break event. Elastic and/or i astic analysis will be per-formed to determine the primar stress intensities and/or For the pressure retaining strains boundariesat such critical as valvelocationsbody [. valve disk, the calculated primary stress intensities shall not exceed Level D, Service Limit (3.0 S$ For other valve parts such as seat ring, hinge, Q actuator shaf t, structural integrity will be evaluated based on strain criteria determined by material propertf N res=l42 of W,'s and ys ss m ' I h f evvh A Ary WWa mbse, E t 6 +, l l l 1 i l 210.20-1 Amendment 6
'l> l l
[~ HCGS FSAR 6/84 QUESTION 210.50 (SECTION 3.9.3 ) , (.' Provide the buckling criteria used for the design of the RPV
~
skirt and stabilizer.
RESPONSE
The support skirt was designed to canply with the requirements of the ASME B&PV Code for Class 1, pressure-retaining portions of ao on T.ne upport ski is limi eo w the_RPV. J he c ressive duce a yi d stress di ided by sus or t load t t would
.125, the afety f tor for lted cond , ions.
An nalysis o the buck
- ng of the upport sk t under a aulte d cond ion estab 'shed tha the desig eets the riteria o Paragr h F-1370 ) of the 'ME B&PV Co , Se ctio III. The e riteri include a aulted co ition limi of 0.67 'me s the c 'tical ekling st ngth of t support s 'rt at te ra t ure . 4 The itica buckling rength is e allowed tress ca ulated by me t ds pr cribed by ar agr ap h 370(c) an A$endix VII o_fj -
t he Code .f A gnerie bur. 4/5 shdyass mdtadsd uig 4lse busy oy %. Lamarnek k i ud 2 cg nd.Ae I suppd skirfs, wkie,k kaw 4ke small.d c k of 6,eknas 4a dks. 'iks sU enaniesd g 4le sidd kekly u.mlecaseief c=pressi.4, keep sbss, and 4ransverse sliear (see RJe.ee so.so-t) ad
,L.w.d 4kJ ud.c k of b 4 pec 3 of lo.A & crifical 6=kI,-5 sks is mek grea.lec em Se gisld thess..
wee #is YsUg .sl. owed #d ineladic dbilih IWdsyi Ilus did S 9'raaaa a sar,3 u.ea.m isu w.oc& A.,a &
' ~
- a. comgressive, load Ndwould (4ob.s o. 9i eld thus 33,1,;446gI.listo a pr,ir4J
/ wtM b,udy% 4 provide k.for & A.4.bs J -fa.$/c.b. ,,. ecc f a skows L loads on 4ke 1464s sy pd did sisil,,a .k;g,&. 4%ses Ac.sses ed. e. cad 4we-LJs J 4e J h-p.ak, crilii.i lidl, t/nus A /.=b pderAsa % g4~.gsgk G-Inole) d See1,en1E of L A ME 64H Cede.
liske eso.so-1. q<esa, k uD H.su,%, 'L&4 of Shf4 I . s % \ 6 - Pa d M , boekl9 of Cacui Pla.4es ud si, alls," McA Teausseo.l Alek. 37i33,IJa k.[Dd,1957. 210.50-1 Amendment 6
>/, -
Question 230.8: Question 230.6 addressed the January 9,1982 magnitude 5-3/4 New Brunswick, Canada earthquake and the effect of the New Brunswick earthquake on the Hope Creek site. The revised FSAR discusses the 1982 New Brunswick earthquake sequence and states "the 1982 Miramicht earthquake sequence was not out of character with the
. region's previous earthquake histo'ry" (FSAR, p. 2.5-8'0). Is the New Brunswick region more seismically active than the Hope Creek site area 7 Compare seismic activity rates and recurrence models derived l
for alternative size regions around both of these areas. (See, for l example, the Shearon Harris SER, NUREG-1038 or the Millstone Meeting Summary contained in a February 1,1984 letter from W. Counsil to B. Youngblood). Response: The seismic activity rates and recurrence models for several alterna-tive sized regions surrounding the Hope Creek site and the epicentral region of the January 1982 Mirimichi, New Brunswick earthquake have been compared. This comparison was done in the following manner:
- 1) earthquakes listed in Table 2.5-1 of the FSAR for which no mangitudes were available (i.e. historical events prior to about 1950) were converted from Intensity to magnitude using Nuttli and Herrmann's (1978) relationship m b = 1.75 + 0.5 I g (1)
- 2) other sei3micity data (i.e. events listed in Earth Physics Branch l
Files or North Eastern U.S. Seismic Network Bulletins) where magnitude scales other than m were used to characterize b seismic events, were considered to be numerically equivalent for purposes of this analysis. l 3) a recurrence model was constructed for both the Hope Creek region and the area surrounding the New Brunswick magnitude 5.7 event of the form: k 1
~
Log N(k M) = a - bM-and normalized to 104 M12 for direct comparison of earthquake density in a 50x5 area centered about both the Hope Creek site and the New Brunswick event (see Figure 1 and Table 1). 0
- 4) a qualitative comparison of a 1 x1 area about both areas was also made to reflect the relative numbers of instrumentally recorded earthquakes of varying magnitudes over equivalent recording periods at both sites (see Table 2).
Discussion: A 5 0x5 area surrounding the Hope Creek site and the Mirimichi, New Brunswick magnitude 5.7 earthquake epicenter was selectsd as being broad enough to represent seismicity on a regional level. Events were compiled from Table 2.5-1 of the FSAR for Hope Creek and from data files of the Earth Physics Branch, Dominion Observa-tory, for the Mirimichi region (Adams,1984, personal coinmunica-l i tion), not including the Jan~uary 9 1982 Mirlinichi event or its aftershocks. Table 1 provides a summary of the recurrence para-meters for both areas and the recurrence curves are shown on Figure 1. 0 0 Assuming that the seismicity is uniformly distributed in both 5 x5 regions, the probability that an event as large as M= 5.5 occurring within a 16km radius about each site would be: Hope Creek - 1 in 41,600 yrs Mirimichi, New Brunswick - 1 in 20,200 yrs The fact that the population density in central New Brunswick is low and that earthquakes have only been routinely recorded in that region for the last decade or so may reflect an even greater difference in seismicity between Mirimichi and the Hope Creek site. As a further comparison between the two areas, a 1 x1 e ea, centered on the Hope Creek site and the Mirimichi 1982 epicenter, 2
- mNv-r1+mit-*e-i=t, ereye. g
O
- was selected and earthquakes occurring within both the areas were determined. As earthquakes have been instrumentally recorded since l
about 1930 in eastern Canada, this date was used as the low cut off in both data sets available. Table 2 shows the comparison; for the Hope Creek areas, there have been a total of 8 events since 1930-2 events between magnitudes 2.0-2.9, 2 events between magnitude 3.0-3.4, 3 events between magnitudes 3.5-3.9 and one event between magnitude 4.0 and 4.4. The 1 x1 region s_urrounding the,1982 Mirimichi epicenter, on the other hand, shows a greater number of events (25) than Hope Creek, not including the 1982 Magnitude 5.7 event and aftershocks. There have been recorded (since 1930) 14 events of magnitude 2.0-2.9; 4 between magnitudes 3.0-3.4; 5 events between 3.5-3.9; 1 earthquake between magnitude 4.0 and 4.4 and 1 event between magnitude 4.5 and 4.9. It is clearly shown that the Hope Creek site is within an area of l l significantly lower seismicity than for the Mirimichi, New Brunswick Region. l - l 1 3 b
9 o TABLE 1 RECURRENCE PARAMETERS ! 2 N/104mi 2 R @ on Ama (mi ) t 4.0 44.5 2 5.0 25.5 "! - 4 Hope Creek (5'x5") 9.2 x 10 - 0.03 0.01 0.002 - 0.1 ' 4 New Brunswick (5 x5 ) 8.1 x 10 0.02 0.01 0.003 0.0007 0.E
.. (w/o Mirimichi Events) i TABLE 2 SEISMIC EVENTS WITHIN A 1 x1' AREA CENTERED ABOUT THE HOPE CREEK SITE AND THE MIRIMICHI i MAGNITUDE 5 7 EPICENTER .
Hope New Creek Brunswick Magnitude (1M80) (1930-1981) 2.0 - 2.9 2 14 , 3.0 - 3.4 2 4 3.5 - 3.9 3 5 4.0 - 4.4 1 1 4.5 - 4.9 0 1 TOTAL 8 25 i 4k
o l l 10~1
~
o . i 10-2
- 1 .
( . ct -
,E -
o t - t . 2 2 - 3 . Mirimichi New Brunswick 5*x 50 Area Lc g N(aM)= 1.5-0.8 5 M
~
l- ope Cree k 50 x50Acea ~
~
Log N(2M) = 0.66-0.67M - 10-4 l 3 4 5 6 7 Magnitude t S l Recurrence Curves Figure 1
k eV. [
. aCGS FSAR Ued gossrton 430.:: <sscTlow s.s.4)
Provide additional justification to support your statement in Section 9.5.4.3 that sufficient additicnal fuel can be delivered to the plant site by truck, or barge. In your discussion include sources where diesel quality fuel oil is available and distances travelled from the source to the plant. Also discuss how fuel oil will be delivered onsite under extremely unfavorable environmental conditions. (SRP 9.5.4, Part I) RESPO4SE . by diesel generator fuel oil storage tank fill connecti are in Section 9.5.4.2.6. The total capac he SDG fuel oil tanks and day tanks is ont for seven days of SDG operation e rated fu indicated in Section 8.3 for a DBA and in this period, additional fuel can be delivered plant a truck or barge. The supply depot i ed about 44 miles from nt in Pensauk . . Under extremely unfavorable enviro tons, deliveries would be made by truck. (INSG A.T 'A ' ) l l D f - .. a s/4 , 430.88-1 Amendment 4
4 INSERT $ TO 410 f#
<R Site flooding (i.e. flooding above plant grade elevation) is a highly unlikely event. The highest historical high water was 97.5 feet (PS Datum), recorded November 1950, 4 feet below plant grade. As an estuarine, site flooding is primarily a result of the effects of tide combined with severe storms. The tidal cycle being approximately 12 hours in duration would reasonably be expected to contribute to site or local flooding for only a few hours. This would affoudtheopportunitytorefuelthefueloilstoraget[anks within a few hours of any scheduled re, fueling.
Severe site flooding to the design flood level is due to the PMH as defined in Regulatory Guide 1.59. Precise track position and forward speed (27 knots) as well as other assumptions are necessary to develop the flood levels calculated for the design basis event. A description of the analysis is presented in Section 2.4.5. A forward speed of 27 knots would cause the hurricane to move over 300 miles past the site in 10 hours. The maximum winds are assumed to extend 39 nautical miles. The forward travel speed is a critical parameter in the calculation, as this is what causes the large volume of water to be first forced into the Delaware and then carried up the estuary past the site. Even in the event that the storm should stall, flood water will tend to drain out the bay as the forcing function is no longer available to push water into the bay. There would also be a further reduction of flood waters due to the tidal change. It would be unrealistic as to expect site flooding to persist for more than 6 hours. Upon continuous operation of the diesel generators for any 5 day period, a new fuel oil shipment will be delivered. Conservatively assuming that flooding persisted for a full day, this will ensure that the 7 day supply will never be exceeded. IP "> wwer.t e RSC: vw MP84 112 07 1-vw sti
3Seg "g" to 43o.ffe i he 5 ( <*& he fodJe d. -M
} Q o f .Q.\t w s's 1. s q k s w9
% d ave.: 3 n < - ,( rya e sge.3 coaa%ns -
i Tu<\ %gf\ke &ren=a4 b.S et 4 14 CG.5 Crouse. E.st s < Wee e
.Ne. m /k. ,Osta m e. 4o A f6
$ bs 0 \ Co. .
tz mL - Se am , neua lerse) ...__. ..
.Trs. % 01 ca.
80 ms Tredoef N w 1*esej ... . . Rec., Pa+.e.t 0 . ca. Rem #g , b=sgb ..e, so %
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Fut,l
. ASC o ,
s
/
40 mi. ./ An<Aa.in , %nyt >an k, Nf-I n - - - ~.. . . .
,- -. _ - ne - _
0 0 e
,e
- e ee ,
. p , .Traa.so s.A & %< On t pq kn r*G*uM JE:, o, 1 5 e '.\ < mJ..s of & sh, .
9 + l gg 6 = eG e 0 e i l - I
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- . _ . _ _ _ _ . . _ . _ - _ _ ~ _ _ _ . . _ . . . _ . . . . _ _ . _ . _____ _ - , _ _ ~ _ . .
( 0E W. l 6/84 BCGS FSAR
-QUeETION 430.94 (SECTION 9.5.4) .-
__In Section 9.5.4.2.6 of the FSAR you state that the emergency flood protected truck fill connection for the fuel oil storage l tanks is located inside the auxiliary In Section building 3.4, at Elevation Table 3.4-1 you i 102 feet (plant grade level). state that the design flood elevation for the D/G building is 120.4 feet with the still water height at 113.8 feet. Provide the following: .
- a. Describe or provide adequate drawings to show the i
location of the emergency fuel oil storage tank fill connect 1on.
- b. Assuming the emergency fill connection must be used to Describe how fuel refill the fuel oil storage tanks.
oil will be delivered to the site during flood conditions and describe the procedures that will be used in refilling the storage tanks during flood The procedures conditions and non-flood conditions. should include fuel hose routing and fire watcher.
- c. Describe how flood water is prevented from entering the building during refueling operations. (SRP 9.5.4,
' Parts I, II & III)
- eFAPONSE i'
The diesel fuel oil emergency fill line is located in the auxiliary building at floor elevation 102 feet-0 Theinches and diesel emergency a l l center line elevation of 106 feet-6 inches. i fuel oil fill connection is located in an area which is flood ! protected by the auxiliary building main service entry doors. The location of the diesel fuel oil connection is shown on i Figure 430.94-1, reference Figure 1.2-35 for location relative to watertight door. l
,fd ' ~1Yn~k ed i ed r f- se ertfit l o at g fwIl
/ en a e n.j . i. e 3'<
f
, at Ant, ipe ha jIo all ing(fron(occurir at g~- -- 154 wi4 pgevent refu ystem operating procedures (SOP) wilh accress'the refilling of the storage tanks from any fill connection and include proper fuel hoseAbnormal routing and the establishment opecating procedure, of aoffire acts l
watch, when necessary. nature, shall provide direction to the operator as to whichIhese SOP is to be used, dependent upon environmental conditions. /Eferto procedures will be in place by Januaqy 1985. se ero emese/m y3o. es Ar in t*** *e perm Pl" . 'M ftv3a;j de /Nu3 <. & NC GS OI"'/9 Leakage through '4he door seals is removed by drainage systems in the building. The flood doors are capable of withstanding the flood height as described in Section 3.4 and Table 3.4-1, i
I - g!- 1 HCGS FSAR 1/84 '
, OUESTION 430 132 (SECTION 9.5.7)
. In Section 9.5.4 you state that diesel fuel oil is available from I l local distribution sources, but you have not discussed the
! availability of lube oil. Identify the sources where diesel quality lube oil will be available and the distances required to be travelled from the source (s) to the plant. Also discuss how the lube oil wilf be delivered onsite under extremely unfavorable environmental conditions. (SRP 9.5.7, Parts II & III)
RESPONSE
Section 9.5.4 has been revised to indicate that diesel fuel oil and lube oil are available from load distribution sources. The lube oil vendor has not been selected yet, but it is expected to be one of several possible vendors within 50 miles of the site. Since the lube oil makeup tank is refilled from an outside l connection on the west wall of the Auxiliary Building at the - 105 foot elevation, local flooding could temporarily affect delivery. However, the engines have a minimum of.7 days operating supply of lube oil : d ::::; ::y lub: ril re;10 i:Prent
- r 5: ' f!:d t tt: :::: tir: ::::;:::y ft:1 til : ;1 erie'---t eseuse, (n ta) l
')%
430.1,32-1 Amendment 4
,,. - - - - - - - - - . - , , - , . - . . - ----- - , , - - - - - - - - . . . , . . - , , - , . , . - - - - . - - - - - . - - . - - - . ._-,-.-,v,.-.---,-----%. - - - -
( .
. ~
A I' 4*0' N' INSERT f 9ISite flooding (i.e. flooding above plant grade elevation ) is a highly unlikely event. The highest historical high water was 97.5. feet (PS Datum), recorded November 1950, 4 feet below plant grade. As an estuarine, site flooding is primarily a result of the effects of tide combined with severe storms. The tidal cycle being approximately 12 hours in duration would reasonably be expected to contribute to site or local flooding for only a fe ours. This would oil N t[anks afford the opportunity to refuel the3 within a few hours of any scheduled refueling. #"" T" Severe site flooding to the design flood level is due to the PMH as defined in Regulatory Guide 1.59. Precise track position and forward speed (27 knots) as well as other assumptions are necessary to develop the flood levels calculated for the design basis event. A description of the analysis is presented in Section 2.4.5. A forward speed of 27 knots would cause the hurricane to move over 300 miles past the site in 10 hours. The maximum winds are assumed to I extend 39 nautical miles. The forward travel speed is a critical parameter in the calculation, as this is what l causes the large volume of water to be first forced into the l Delaware and then carried up the estuary past the site. Even in the event that the storm should stall, flood water will tend to drain out the bay as the forcing function is no longer available to push water into the bay. There would also be a further reduction of flood waters due to the tidal change. It would be unrealistic as to expect site flooding to persist for more than 6 hours. of the diesel generators for any 5 day Upon continuous period, operation a new dash /L /L oil shipment will be delivered. Conservatively assuming that flooding persisted for a full day, this will ensur that the 7 day supply will never be exceeded. RSC: vw MP84 112 07 1-vw l
- l d
~
l l
i l k p ATTACHMENT 5 1
' 1 DSf9 f,,)ome, t.e s e t.s dPE N r1TtA No 4 (D49- S it M J a' 4. :-
)
^
The appitcant states that safety-related buildings have been designed with - I either roof parapets and scuppers or no curbing at all so that if internal roof drains became clogged, the precipitation ace:mulation would overflow before the basic roof loading would be exceeded or roof hatches flooded. Because there w'as Tasufficient inforsation availaole to enable the staff to reach the same conclusion, the appifcant has been asked to provide additional infonnation and detailed analysis of the roof drainage systas including the ponding levels on roofs of safety-related structures. Until the additional information and analysts are available, the staff cannot conclude that the plant meets the
. requirements of GDC 2 with respect to the effect of local intense precipitation
' on roofs. _p c .. . u , : l, fl. : , - e ..- . - yc.-%la Tu 7L. ,v..y . . . -( . >
;4 0..u% 2 Vo . l3
I Eu p ^^ .
- QUESTION 240,13 (Seetion 2. 4. 2.3 ) ,
\
Provide your de tailed analysis of the PMP ponding levels on i.s ' roofs of safe ty-related structures requested in 0240. 7 De tail s should identify and provide information on the roof area of each sub-drainage area for each safe ty-related structure; the size, nurr.ser and distance above roof (elevation) of the inve r t of each u: upper (overflow drain) for each drainage system, and the elevi : ion of the curb of each roof hatch within each roof drainage area' system. Also provide details used to conclude that the ponding resulting from PMP does not of fact sa fe ty - related facilities., . RES PONS E Section~ 3.4.1.1 has been revised to respond to this que stion. e e 4 l l i DSER OPEN ITEM f t 4
~ - . . ..- s. .
. . HCGS FSAR 4/84
.m -
\- b. Waterstops provided in exterior wall construction joints and seismic separation jolnts below flood level
. c. A minimum number of openings in exterior walls and
! slabs below flood level (these openings are designed to prevent intrusion of flood water.)
- d. Water-pressure-tight doors installed in exterior walls below flood level
- 10 0* e. Exposed equipment hatches installed above flood level:
C <C those below flood level installed behind exterior walls 4 designed to prevent intrusion of water 27 b- f. Continuous waterproofing systems applied to the eg underside of base slabs and on exterior walls to grade,
,53 as discussed below.
Except for the intake structure, the HCGS safety-relate'd 2J b~
.( structures are provided with roof drainage systems capable of handling a maximum rainfall rate of 4 inches per hour for a 4
1
- uJ 20 neriod of 20 minutes s/ In t un11xely ev c that tnp roof dra) <,1 J I; ("become oggea sacundant erflow dra are prov W d ~
~
uJ 3 appro matel ' inches ve the ma oof deat47' evation celled ar , which h "' parape The y% , exc t for , e plant r f dra ge syste isposes w er through/ yard d nage l
. stem To preely pending e signific ely great rainfall I inte ities segM ts of the arapets are moved whe e necessary, j
. The intake structure roof is designed without parapets or other continuous obstructions and is sloped to shed the water.
Accordingly, no significant pending will occur. I l To prevent seepage into any Seismic Category I structure all roof : openings are watertight and provided with either metal sleeves or i concrete curbs of sufficient height to exceed any possible i pending levels. ; As an additional margin of safety, all Seismic Category I roofs I are designed to withstand a loading of 150 lb/fta, which is greater I than the loading'resulting from the maximum pending on the roofs. Doors and penetrations in exterior walls of the auxiliary and ' reactor buildings are protected against water inflow up to elevation 127 feet for parts of the south exterior walls and up to elevation 121 feet of other exterior walls. Interior drains j from the radwaste areas are independently piped to the liquid I weste disposal system and are not connected to the yard drainage j system. Wall penetrations above elevations 121 feet and 127 feet i l l 1 3.4-2 Amendment 5 l 1 DSER OP1DI ITD( y
--,_,..,,-.,es.-.,
,- ,e- -,-,..w,w,m .---ew_, -yy-.ww.y- -,.-.,_-.__...-,_.._m---.-,,,,- . - - = - . -
. x f
INSERT A The roof
- drainage E(stem consists of roof drains and 6-inch -
J diameter scuppers located 6 inches above the roof drain \- elevations. Supplementing the roof drain system is a ceries of openings in the parapets of the roof s of the buildings. The 6-hour , l oc al , all-se a so n PMP wa s u sed to siz e the se ope ning s. The PMP, which is 27. 5 inches, is distributed into 5-minute increments su:h that the maximum amounts for durations of 1 hour, 30 minutes, 15 minutes and 5 minutes are 18.1, 13.7, 9.5 and 6 inches respectively. Roof elev a tions, sub-dr ainage areas, and the dimension of parapet openings are shown in Table 3.4-3. A schematic of the roof drainage is shown on Figure
- 3. 4 -4. .
The routing of the PHP assume s no losse s, the roof drain system to be non-f unctional , and the ponding is allowed up to the , limiting elevation of the top of the curb of each roof hatch within each roof drainage area system. Prior to the PHP, an initial level of ponding at the invert elevation of the parapet openings is assumed (invert elevation is 6 inches above the roof drain elevation). A - see rating curve for each rectangular parape t opening was derived using the equation Q = CLHl .5
- whe re :
~
Q is the discharge in cubic fee t pe r second C is the discharge coef ficient (3 0) L i s the le ng th in fe e t of the parape t opening H is the head in feet of water above the i nve r t of the parape t cpening The flow capacity of the 8-inc h di ame te r o pe ni ngs i s de r ive d using the following short culvert 4quations: Inlet control flow for unsubmerged inlets : H = H. +k (1. 273 )" 5 y 2 Inlet control flow for submerged inle ts: , H =h ) 5 Th + ki( 2 where H is the total head above the invert of the opening in feet DSER OPDt ITD( ([ I
1
- ' e -
Rectanau/or arm,/ o,miurk,es idors mn4'
/ red
._ n's n bw_ag'-utshe.'._ujeir -Ecr 4p.sheoM._wd/W .
..._ dur. face-- .eAlla'b4ns 4.t/se)._zVie _-}cp_.o.dRe_.o,pwhg,--.
fa- om.1uAmay;d nur dihin3 +1w
& . F.or. co..a..d..d.i.b..ns a. sken. .fhe op' s..frednt.
u.xrkr .swkc -
..<leVaNh .u)a.s. htkhr_fho's. -hke. fop ol %. qcm
'e.. . ..O!. . eiftcc. gfitahe.M Lo 5.1. aSeb .... . . . . .. -. ... _ .
. . . _ . _ . -._-- _ -.-.... 0 _ =. C . 8 ~1/ 3.3. H ' _ .._
I .Luitert*-. .... _ .. . .. ...._.... - . . . . . . . .. . . . . . . . . l .. ---. Q is_. fhd4Chdyt... 1. citbs...ktk )yr secqx{_
.. ...-_ C. tk. fj1eAccl1sige_ ae.ffeteaf feAen.es s.s.) . - -
._ - ..b /E ?.i1< t2ZE'i% 6l N4f .op.eHL5f _J.5_ rftMre kt'lb.
l . . . . _ -
- A if fde brad' medsure/Ew &e.zzirkz(a&_.&
.. _ +4.E.jpMIN IN b!l
- -.3_is._E1c-acceleM A 'S ..ofgwwk n:z.fedde~c W .
t ...--.. -- --.-- . . - . . . _ . . . - - . . . . . . .. ... ._ . ... ... . Ci Q.
. =. . . - - . . _ . . - . . . . . . . . . . ._
N m
.. C . . - - . . - . . , -
)
_ . - , 1
t . . - . - _ - - . . - .. Insert A (cont'd)
H e is the specific energy 0 is the discharge in cubic feet per second D is the opening diameter in feet .
k,m,pandk1 are the inlet control perf.ormance
- coef ficients. The experimentally de termined value s for .
. a square edged entrance are s ,
k = 0.0098 - m = 2. 0
= 0.67 kl = 0.0645 Since the limiting water depths are greater than the ponding levels resulting from the PMP (as shown in Table 3. 4-3 ), the ponding levels do not ef fect safe ty-related f acilities.
I s . . \ I i Ds n OPEN ITM y
- - , ,, .,. ,,.,,.-e --,--w, , , -,----, -- ,, y- n-------- --. - -..- -n,--s -- - - - , - -------,.--,--.=-y e
. p[
* [- ' '
HCGS t%AR , E TAB 3. 4-3 ' Il ' Nazimum Ponding Depths on Roof s
! g
- of Safe ty-Relat&d Structures l g for local 6 Hour PHP ,
.4 Limi ting j I Water Max. Water Width of Depth Over . Depth Over Number of Width of T ,
8-i nch 8-i nch Parape t Roof Drain Roof Drain ; i Mi n. Roof H igh' Sl ot Opening Elevation Elevation - Roof Elevation Sub-Dr ai nage Diame ter (in.) (in.)
' No. (2) (ft) Area (ft2) Openings _ (ft) (ft)
'1 *
- 2. 5 - 12.0 11.5 I 159 2720 -
2 -
- 28.8 18. 0 2 137 2570 2 - - 15.0 13.6 i 1
3 172 1530 28.8 16.1 4 153 1930 1 .
- 50 '12.0 11,9 -
4 5 155.25 3700 - l - 25 13.0 12.6 6 172 38850 - 18420 - - 35 10.0 9. 8 7 198 ~ 3.0 - 12.O 11.7 8 155.25 3490 -
- 2. 5 - 19.0 18. 1 ,
9 158.33 7380 - , . 0.83
- 18.O !' 15.8 10 172 5220 1 i
- 2 - - , 18.0 17.5 11 124 5030 l ;
17.6 14 18.0 i 12 132 3,3500 l 1 l ! Notest The invert elevation of openings and the crest elevation of slots and parape t i_ f 1. ! ope o8 ys are 6 inches above the roof drain elevation. l
- 2. See Figure 3. 4-4. .
e FSAR H/11 ,
- ~
,' h h w:
r O i' 8 m
=
2 k. e I 8! E s u. ('
- G. '
@ O 1f
,,r= ~ ~ ~ ~ ~T I I
, I
,l l ,
! ,t 4 f
, Oa L_____ y e
8__ h 2 8 * HOPECREEK it -* GENERATING STATION
$*
- FINAL SAFETY ANALYSIS REPCAT l , O" __
5 O ROOF DRAINAGE UNDER
$ i PMP CONDITIONS E 9
. - rlaung 2.4.s I
- . . ~ ... . . .
.. DL? (L D .~ . . TLs... t)c. &C D S EL 'S te +.m 2 Vu 1 fonO gn G- LEvn t.S Based on its review and analysis as described above, the staff has requested the applicant to provide additional infomation on roof ponding levels -
due to intense local precipitation (PMP), flood protection for the service water intake structure and power blocki and flood protection structures adjacent to the intake structure. Until the appifcant provides the adcitional infoma-tion, the staff cannot conclude that the plant meets the requirements of GDC 2 with respect to flooding. The staff also cannot conclude that the plant meets the hydrologic criteria of GDC 44 with respect to the themal aspects of the UHS. k &3pcr% ct h V;,. ,;...a, a N Yi.:
**40.13
.$$-~~ u k e: pen h' 5 O }"
D J. ~ SUO I2 Q ges b.eu , it eskd.c.l>.I % b5 = R- Co*~ 9
.&w N0 Y.
e
, , . , - - - - , , , , , - ~ . ,-,---,w-, ------n.-- - ---- - -
. T HCGS , j 1
^*
DSER Open Item No. 7 (DSER Section' 2.4.1'l.2)
~
THERMAL ASPECTS OF ULTIMATE HEAT SINK
- j. s 1 The applicant has analyzed the ability of the service cooling water supply to withstand the effect of such severe natural .
phenomena as ice blockage, flooding, low water, and thermal - aspects of UHS. As indicated in Section 2.4.7, the effects of ice blockage would not obstruct the flow to safety-related
- ' pumps. Thus the staff concludes that the intake structure and essential service cooling water flow is adequately protected
,, against ice effects. As indicated in Section 2.4.5, the ability
- of the service water intake structure to withstand the effects of PMH surge flooding and associated wave runup and overtopping
- remains an open item.
- The applicant reported that the minimum historical low water level at the Reedy Point, Delaware, t'ide station is -8.6 ft asl. _
The applicant's analysis of the maximum setdown considered the PMH wind speed of 85 mph (the overland PMH wind speed for
.- the direction resulting in maximum setdown) to be blowing down the estuary coincident with 10% exceedance low spring astro-nomical tide of -3.9 ft asl and the associated trough of the
- 6.0 ft maximum wind wave. The resultant low water level would
(~' be -13. 0 f t mal. The applicant has stated that -13.0 f t asl is ! the design basis minimum low water level for service water pumps. Based on its independent analysis, the staff concurs that -13.0 ft mal is an appropriate design basis minimum low water level. The applicant has not identified the maximum intake temperature that will allow the plant to safely shut down under normal and emergency conditions as discussed in Regulatory Guide 1.27 nor the ability of the Delaware River to
- supply water below this temperature. Until this information is-available, the staff cannot conclude that the plant meets GDC 44 with respect to the thermal aspects of UHS.
l Based upon the evaluation described above, we conclude the ! hydrologic characteristics of the Ultimate Heat Sink meet the requirements of 10 CFR Part 100 and 10 CFR Part 100, Appendix A. As indicated above, certain aspects related to flooding level i for the service water intake structure are unresolved. There- i f ore, the staf f cannot conclude that the Ultimate Heat Sink System meets the requirements of General Design Criterion 2 with respect to hydrologic characteristics. In addition, the staff cannot conclude that the Ultimate Heat Sink meets the requirements of GDC 44 with respect to thermal aspects of the heat transfer system. 7-1 l l
y _ . _ - . e b
RESPONSE
For information on the ability of the service water intake l structure to withstand the ef fects of PMH surge flooding and associated wave runup and overtopping, see the response to DSER Open Item Number 5. The maximum intake temperature that will allow the plant to safely shut down under normal and emergency conditions is dis-cussed in the response to FSAR Question 240.15. L 1 e e e O O e 7-2 DSER OPEN ITEM 7 , i
+
I
/ .
~
.. G u e=mor) .*40.tr ( s.ufa 2.9. } l . G)
.antify the maximum tamperature of the intake water that wi !
allow th p I anh to safely sht down under normal and emergency ' conditions and discuss the ability of the Ultimata Heat Stak to ! supply service cooling wate'r below this maxisms intake temperature. , k-:,:e,s: e. . E.:
.u 9. 2. c. 2. L .bu . . :.J' A &a..:4
% >: .. 2 ;~ .
I l l 4 i DSER OPEN ITEM y l l l
y
. (
JUL .is e40 26 8 0 5 7
~
I I i j
; 9.2.5.2
- system Description j The URS is the Delaware River, which provides the source of cooling water to the SACS heat exchangers through the asWS, as shown on Figure 9.2-1. The SACS, in turn, provides domineralized cooling water in a closed loop to the ESF components. The water from the SSWS is discharged into the CWS to provide makeup for that systest. :
Details of the safety-related and nonsafety-related systems and heat load dissipation are discussed in the following sections: ,
- a. SSWS and intake structure - Section 9.2.1
.Q b. Circulating water and cooling tower - Section 10.4.5 i
i c. SACS - Section 9.2.2. I
)
A discussion of Delaware River water temperatures is provided in the Hope Creek Generating Station Operating License Stage-
- Environmental Report.
~
W 9.2.6 CONDENSATE AND REFUELING WATER STORAGE AND TRANSTER
" "' SYSTEM y N#" N .
9.2.6.1 Desien Bases
? ' The condensate and refueling water storage and transfer system has no safety-related function, except for that of supplying 4 condensate to the suction line of the high pressure coolant 7 injection (NPCI) and the reactor core isolation cooling (RCIC) g pumps. The system is designed to perfor:n the following .] functions:
E 5' !: a. Supply water to fill the reactor well, the 4 z
~
dryer / separator storage pool via the reactor well, and S E the spent fuel cask storage , pool during refueling x operations, and provide storage for this water when E refuellag is completed n a 1 _ _ _ ? __ ? - ? L _ _ _ _ _ _ _ _ _ . ___ _ _ _ _ _
e ee W ,e . me eww e $M W ep.e l g
. ... . .d m. w w w es e I*
h 8 l 4 . - . I .. . .. . .. . .. {
, i ,
e . .
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,, d [o8 [ Md E U N d h4 [ M ( ;
E - Ukf $ l sh
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/sAlar mad M eAsvW 4dvr b o<,ad uAex dmeAdras.te/
l CaJA s 7s on/J tarro/ s9a./ s4Veds9c,C / w spx'r s,as9 red 77mpel/) l'a4r 4.s h' seed 43 90.J *K Wa Mxssesum . O ram n.enraxe sea.e emsees eey edu70awJ' s.s , s+d Grx #m c e a w a s r s a a p as s heries,.i A,ss/ ode.s e , lT A3 d**/e$d7Eb T"ds1 T 5 d v'E xs4dp f As /dd. s,dsf 7dA ' MM/&&47*u A c t,Je L r 4 der.f ts** 2ref 414ed YJ'/s4 meumuws. f I l l 4 e t I oszR OPEN IN M f . . . . I
\ _. . . . _
3 t I I - \- - ___----.-_____ _ _ __ _ . _ _ _ . . _ _ _ _ _ __ _. __ _ _ . _ _ _ _ . _ . _ _ _ _ _ _ _ _ , _ , , _ _ , . . . ,_ ._
NCGS DSER Open Item No. 29 (Section 3.5.1.1) .. .. I INTERNALLY GENERATED MISSILES (OUTSIDE CONTAINMENT) With respect to rotating equipment, the applicant has stated that the pumps and fans were manufactured to the same industry standards as Palo Verde and therefore the results
- of the Palo Verde's analysis for internally generated -
missiles is applicable to Bope Creek. In order to rely upon the analysis performed by Palo Verde, the applicant must verify that every rotating component (pumps, fans , motors , and turbines, except the main turbine-generator) is designed and constructed to exac.tly the same coder, and standards (including addenda and editions) , to be of the same manufac-turer, size, and materials as the analyzed components at Palo Verde. Palo Verde relied mainly upon compartmentaliza-tion as the means to protect the redundant equipment. For each component where compartmentalization was relied upon at Palo Verde, the applicant must verify the identical l components at Hope Creek provided with comparable compart- ~ a menta112ation. Similarly, the applicant must verify the use of barriers, For T' separation and orientation as was used by Palo Verde. every component which is not identical with Palo Verde, the 4 applicant must provide a discussion of the analysis which verifies that the casing would be capable of retaining the internally generated missile or that the missile would not strike safety-related components or generate a secondary missile. Unless the applicant either verifies comformance with the Palo Verde design (as outlined abcVe) or provides the results of an analysis which shows that the casings will contain the internally generated missiles, the appli-cant must provide protection by any one or a combination of compartmentalization, barriers, separation, orientation, and equipment design. Safety-related systems mu'st be verified to be physically separated from nonsafety-related systems and components of safety-related systems are physically separated from their redundant compartments. MP 84 112 15 01-bp
.s
' Based on the above, we canhot conclude that the design is in
,J . -
conformance with the requirements of General Design Criterion 4 as it relates to protection against internally generated missiles until the applicant provides an acceptable discussion concerning rotating components as potential sources of internally generated missiles. We cannot determine that the design of the facility for providng protection from internally generated missiles meets the applicable acceptance criteria of SRP Section 3.5.1.1. , We will report resolution of chis item in a supplement to - this SER.
RESPONSE
FSAR Section 3.5.1.1 has been revised to include the results of an analysis of the internally generated rotational missiles outside containment. d i MP 84 112 15 02-bp oszs cars In n #9
._ ~
i 12/83 ([' HCGS TSAR CHAPTER 3 TABLES Table No. Title 3.2-1 HCGS Classification of Structures, Systems, and
- Components 3.2-2 Code Requirements for Components and Quality Groups for GE-Supplied Components 3.2-3 Code Requirements for Components and Quality Groups for Public Service Electric and Gas /Bechtel-Procured Components 3.3-1 Design Wind Loads on. Seismic Category I Structures 3.3-2 Tornado-Protected Structures, Systems, and
, Components 3.4-1 Flood Levels at Safety-Related Structures v ,
-3.4-2 Outside Wall / Slab Openings and Penetrations Located Below Design Flood Level 3.5-1 Internally Generated Missiles Cufs/d'e. A/ mary Cont 6:rio 3.5-2 Target Parameters 3.5-3 Missile Characteristics 3.5-4 Ejection Point Coordinates 3.5-5 Turbine Barrier Data 3.5-6 Target Barrier Data ,
3.5-7 Computed Probabilities
-3.5-8 Summary Number of Operations 3.5-9 Crash Rates Per Mile and Effective Impact Area by Category of Aircraft 3.5-10 Aircraft Crash Density by Location / Route / Altitude
, 3.5-11 Probability Summary oszz onx Irzu oPf 3-1x Amendment 3 e
i _ _ _ _ _ _ . .
?
.**'*'N='
-e,,
12/83
-1
- HCGS FSAR u.
- CHAPTER 3 l
l TABLES (cont) Table No. Title . 3.5-12 Tornado Missiles * ' y 3.f-/3 fadsenally Ma.ded Janu%f Misso'les eds de ?n~f M 3.6-1 High Energy Fluid System Piping 3.6-2 Main Steam System Piping Stress Levels and Pipe Break Data (Portion Inside Primary Containment) 3.6-3 Main Steam System Piping Stress Levels and Pipe Break Data (Portion Outside Primary Containment) 3.6-4 Blowdown Time-Histories for High Energy Pipe Breaks Outside Primary Containment , 3.6-5 Pressure-Temperature Transient Analysis Results for High Energy Pipe Breaks Outside Primary Containment 3.6-6 Recirculation System Piping Stress Levels and Pipe Break Data 3.6-7 Recirculatf.on System Blowdown Time-History 3.6-8 Feedwater System Piping Stress Levels and Pipe Break Data (Portion Inside Primary Containment) 3.6-9 Feedwater System Piping Stress Levels and Pipe Break Data (Portion Outside Primary Containment) 3.6-10 RWCU System Piping Stress Levels and Pipe Break Data (Portion Inside Primary Containment) 3.6-11 RWCU System Piping Stress Levels and Pipe Break l Data (Portion 04tside Primary Containment) ! 3.6-12 HPCI System ?iping Stress Levels and Pipe Break ' Data (Portier Inside Primary Containment) 3.6-13 HPCI. System Piping Stress Levels and Pipe Break Data (Portion Outside Primary Containment) ! 3.6-14 RCIC System Piping Stress Levels and Pipe Break Data (Portion Inside Primary Centainment) DSEE CEG ITDt d? 9 3-x Amendment 3
- - - , . - ,. -. ._ ,,.-,-.n. - - , , - - - - _.._ _ _ . ___ . _. ,- _ - . - - - . _ _ _ , . .__ _. _ . .__ - . -
HCGS FSAR 10/83 3.5 MISSILE PROTECTION . The Seismic Category I and safety-related structures, equipment, and sy' stems are protected from postulated missiles through basic plant arrangement so that a missile does not cause the failure.of systems that are required for safe shutdown or whose failure could result in a significant release of radioactivity. Where it - is impossible to provide protection through plant layout, 3 suitable physical barriers are provided to shield the critical
' system or component from credible missiles. Redundant safety-related Seismic Category I components are arranged so that a single missile cannot simultaneously damage a critical system component and its backup system.
- l .
A tabulation of safety-related structures, systems, and components, their locations, seismic category, quality group classification, and the applicable FSAR sections is given in Table 3.2-1. - General arrangement drawings are included -as
- Figures 1.2-2 and 1.2-41.
i {} 3.5.1 MISSILE SELECTION AND DESCRIPTION 3.5.1.1 Internally Generated Missiles (Outside Primary Containment) and 3 5-13 The systenc located outside the primary containment hav been examined to identify and classify potential missiles. hose systems and missiles are listed in Tables 3.5-1 undant systems are normally located in different areas of the plant or separated by missile-proof walls so that a single missile can not damage both systems.
. g -_ ;c, ru9 M [ e residual heat removal (RHR) and core 1
spray pumps, are located.in separate missile-proof compartments W are not considered a potential missile source or hazard to, __ other systems. od */>.;i ;,,,p .ny.s ye . rid.s ed
"' * **** h s+ructare eA e.i a. fee =
Refer to Section 3.5.3 for barrier design procedure. There are three general sources of postulated missiles: [
- a. Rotating component failure
-oszz orza rrza a79 3.5-1 Amendment 2
1 l - BCGS FSAR 10/83 l Pressurized component failure l b. j
- c. Gravitationally generated missiles.
3.5.1.1.1 Rotating Component Failure Missiles pro ko.ble. Catastrophic failure of rotating equipment having synchronous motors, e.g., pumps, fans, and compressors, that could lead to the generation of missiles is not considered ;;;dibi;.* Massive and rapid f ailure of these components is improbable because of the conservative design, material characteristics, inspections, and quality control during fabrication and erection. Also, the rotational speed is limited to the design speed of *the motor, thereby precluding component failures due to runaway speeds. ; ] Similarly, it is concluded that the high pressure coolant-injection (HPCI) and reactor core isolation cooling (RCIC) pumps and turbines cannot generate credible missiles. These pumps are not in continuous use, but are periodically tested and otherwise operate only in the unlikely event of a postulatedOverspeed accident.
'- They are classified as moderate energy systems.
tripping devices ensure that the turbines do not reach custaway speed, where failure leading to the ejection of a missile could take place. Oth;; ;;t; tin; ;sei;;;;t 2;;; ;;t ;; :titut: : rirril: "???rd 5;;;;;; ef it; r-li si;; es,i/e; the ;laelihwed that it;
.;..;.1..; . r ;est; ..12 ,;;;t;;t; i;.; h;;;i;;.
IB;2 e N 1 x r 3.5.1.1.2 Pressurized Component Failure Missiles l The following are potential internal missiles fece pressurized equipaent:
- a. Valve bonnets
- b. Valve stems
- c. Temperature detectors
- d. Nuts and bolts DsER N N [ 3.3,7 Amendment 2
c "~s
'~ INSERT 1 A tabulation of missiles generated by postulated failures of
( ' rotating components, their sources and characteristics, and a safety evaluation are provided in Table 3.5-13. The evaluation identified one instance where a postulated I missile, which could penetrate through the flexible connection of a vane-axial fan, could have the potential to l in the room. In order to damage safe-shutdown equipment prevent the postulated missile from damaging safety-shutdown equipment, a missile shield has been added to the design to withstand the impact of the postulated fan blade missile. . The formulas used to predict the penetration resulting from missile impact are provided in Reference 3.5-4. The penetration and perforation formulas assume that the missile strikes the target normal to the surface, and the axis ofThe the missile is assumed parallel to the line of flight. rotating components is assumed to fail at 120 percent overspeed. These assumptions result in a conservative., estimate of local damage to the target. l es
- e
' MP 84 112 15 03-bp asER OPEN I5M [f
. ~ - - - - - - - _ . . , _ _ _ _ _ _ - _ _ _ _ _ . _ _ _ _ _ _ _
~
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\ \ '
NCGS FSkB 4 TknLE 3.5-1 pggssuAltfb COMMU'Y Page 1 et 2 INTEaunLLY GEusm&T Ess5Las osers/DE (.aWT,8 #NM8 , Protect 6en Evaluation
- M(gg{Jg pggg h ton codeosaa 3131g3__ f HE Sectiog c +
NPCI 4.3 Test connectica c i' Startup 81ange c Pressure indicator (PI-3883) . j Dratae c .' Ca0 hydraulic 4.6.1 e j. Pressure indicatore (PI-Bete, 4013 4, 9l ; Pressure indicatore (PE-B421, PE-Wets, c. l PI-9416, PI-9412, PE-8447, PE-De tt, PI-8806) Test inatcators (TI-telt, TE-4414. TE-Netti c Test connections c , a vent stine flange c Test connectione c Main eteen S.1 c > Temperature stemente (TE-N644) Pressure inetcators (PP-3632 A B, C, Dl c , ! Temperature elemente (TE-N057 A, S, C, D, El c Main steam S. 8 e sealing Pressure transmitter (PT-50348 o i bline flange or Y-etrainer Test connection . c Temperature element (TE-H060) , c i Test connection a ! Feedwater S.1 S. 4. 8 Bline flange e pWcu c Temperature sensore/etomente {TE-MG47, TE-WG19, TE-He lS, TE-peet, TS-169, TS- 576, TS-242 4, B) Pressure transmitter GPT-pees) c Pressure point (PP-3876 A, Sg FP-3875 A, B3 c PP-39 86 A, B3 PP-3917 A, el . Pressure indicatore (PI-3377 A, pg FI-peegg a pucu S.4.s j ^ PI-84043 PI-80083 POfS-3981 A, B3 P058-3988 A, tl - Pressure switches (PSL-He t), Psm-Ne tti c Flou elemente (PE-3986 A. Si c 1 DSER OPEN ITEM
f n DSER OPEN ITEM Nh HCGS FSAR 0/04 , I 1 f . TABLE 3.5-13 Page 1 of a l IlffEltNALLY GENER ATED I40TATING COMPONENT MISSII.ES OUTSIDE CONTAINNENT l l . < 1 CAlfDIATED , EIBSILE SOUSCE. ge]ggi l e g!tACIggigT!cs MAX. STEEI. CASING IDEN1I- OF I4 CATION VEIDCITY DIA. WEIGHT PERF. DEP118 THICENESS REMARES IKMIDM B11111:5 ___ _JtTLEL II%L 18.:92L (18 1 _11)! l Fan Blade containment peactor 199.0 1.25 3.7 0.211 0.1406 Fan blade may the surround 1r.g penetrate Pre-purge Bldg concrete fan caelag. well for
- Cleanup Fan E1. 1628 the fan Le 12" thick. The calcu-lated depth of fan blade penetra-10V-200 tion into the cancrete well to i scentri- 1.43". Therefore, mise 11e has no fugal Fan) ef fect on plart safe shutdown cap-ability. Therefore protection le l
not needed. , ran flade Diesel Aux Bldg 116.0 1.24 4.05 0.1066 0.0701 Perforation of fan casing may Generator SDG Area occur. Due to the orientation of Wing Area El. 178' the fans, the psetulated f an blade Exhaust Fan mise 11e will not damage any este ' shutdown equipment in the room. 1A, B-V414 %erefore, protection to not
- # Centri- needed. (s) fugal Fan) !
! Fan Blade control Aux aldg 105.0 0.969 0.6 14 0.034 0.0101 Casing perforation will not occurg Area control however, tan blade may exi*. through 1 Exhaust Fan Area the flexible connector on the fan
- E1. 155' discharge. There le no safe shut-1A, B-V02 .
ohutdown equipment in the room. (Cen t r i-
- fugal Fan) ,
Fan Blade FRVS Recir. Deactor 248.0 1.4 5.42 0.310 0.1406 Perforation of the fan casing or Fan Bldg floathis connector may occur. However, due to the orientation of 1A thru F-V213 E1. 1128, 1628, and the fans, only ceiling and floor may be hit. The calculated depth h l Centri- 1788 of the f an blaoe penetration on the concrete le 3.61 . Since there are fugal Fan) no mate shutdown equipment cd impacted, protection to not needed.(9 h i Q 10 0) 00 C) Amendment 7 CJ1 0)
l- ,, _ l (~ (,
!' DSER OPEN ITEM g .
ICGS FSAR 0/04 * ? 9 TABLE 3.5-13 (Cont) Page 2 of a 3 c:tsat t.E sounCE CALCUI.ATED ggtss_Is.E CuAp4Crenistres aent. STEER. IDautt- or incATson vEnoCITY DIA. WEIGHT PEpr. DEPT 16 CAsina FICAfl93 NISSIII THICENESS DEMARES jfTfGL jl!!zl j[:!LsL (In.1 *
,_(I N . ) .
j Fan Blade FRVS Vent Reactor 144.0 . Fan Bldq 1.02 1.99 0. 100 0.1406 Casing perforation will not occurg ' 21. 145' hoerever, tan blade may exit thromgh 1A, 3-v206 the flexible connector on the fan i l Centri- discha rge. The calculated depth of fugal Fant the fan blade penetration it to the y concrete le 1.130*. Due to the ' orientation of the fan, only the catlling and floor could be hit. gherefore needed.(i), protection is not Faz tiede Control Aus 31dg 197 0.772 0.764 0.115 0.1406 moom Emerg. Control Casing perforation will not accorg ' Filter Fan Area however, tan blade may esit through the fleutble connector en the fan E1. 155' 4 1A, 3-V400 discharge. 7t'e calculated depth of (Centri- the blade penetration inco the i concrete le 1.098 ; fugal Fan) There is no ' ) safe ehetdown equipment in the s ] room. Werefore protection te not 1 needed. . 4 ' Fca Blade Battery Aus Bldq 01 0.046 0.23 Boom , SDG Area 0.014 0.0625
- Casing perforation will not occurs ,
i A Eshaust ran E1. 1638 however, fan blade may eult through , the flemible connector on the fan { IA thru D- discharge. The calculated depth ot i I V406 the fan blade penetration ire the e (Cent r i- concrete le 0.0068 Due to orien- + ! fugal Fan) tation of the fan, safe shetdown ,
- equipment will not be impacted and i protection to not needed.(t) l Ftn Blade Control Area Aux B1de) 14J Control 0.034 0.206 0.029 0.0625 Casing perforation will not occurg h Battery Area however, f an blade may exit through 8 i
Eahaust ran E1. 170' the flexible connector on the fan i discharge. There are condesite that of i 1A, B-9410 belong to A, C, and D channele in A (Cen t ri- the room that may be needed for C) fugal Fan) safe shetdown. However, the con- N dette are thicker than the calce- g lated maninun eteel perforation depth (0.029 l, therefore, protec- . .g . tion to not needed.tlM G US G Amendment 7 l l l
,7 h ,,.
) : ;
DSER OPEN ITEM Mf 9/e4 acGS rSAn . TAlif.E 3.5-13 (Cont) Page 3 of 0 l* CALCUBATED MAX. STEEL CASING MISSILE SOURCE MISSitE CHARACTERISTICS THICENESS PBNARES I4 CATION 55~l.OCITY DIA. WEIGHT PERF. DEPTH IDEN18- CF trw.i _ g w.) , IICA119H HISSILE - _HTl8L_ EU:L 1;ast_ 0.846 0.23 0.014 0.0625 Casing perforation will not occurs , Fan flade Battery Aus midg 88 honeever, f an blade may exit through Boom SDG Area the flexible connector en the fan Enhaust Fan E1. 178' discharge. There are condelte that belong to A, C, and D channels in , 1A, B-V416 the room that may be needed f or ICentri- eafe ehetdown. However, the con-fugal Fan) dette are thicker than the calcu-lated maximum steel perforation depth (0.014*), therefore, protec-tion to not nendedf (Q(s) 0.984 0.792 0.027 0.0781 Casing perforation will not occorg Fan plade Aus aldq Aus alog 78.5 however, tan blade may emit through Rattery SDG Area the flexible connector on the fan Enhaust Fan E1. 1788
- discharge. There are conduite that 1 belong to A, C, and D channels in 2 1A, 3-V417 the roon that may be needed for (Centri- eafe shutdoun. Bowever, the con-fugal Fan) duf te are thicker than the calcu-lated maximian eteel perforation depth (0.0278), therefore, protec-tion is not needed(s)(s) 8.341 0.25 Perforation of fan casing may ,
Fan Elade Control Aux Bldg 235 1.68 8.8 occorg however, the fan to inalde a i Equirment SDG Area filter housing that is 3/16" thick. e l Supply Fan E1. 1788 The calculated steel perforation ! af ter the fan blade penetration ] 1A, B-VH-407 through the fan casing to 0.176*.
,1 Centri- Therefore t.he f an blade will not fugal Fang
- exitfrohhefilterhousing. h
# (J
- 8. 37 J.16 0.115 0.1075 Filter housting perforation will not o Fan Blade Diesel Aus Bldg 149 i (filter occur. ,
Generator SDG Area housing c0 I Panel El. 16 38 " l Supply Unit thicknese) C l l ran FC Cy) 14, 2-VH-400 00 (Centri- g L f ugal Fan) 07 CD Amendment 7
(J
~
DSER OPEN ITEM J2 f ' HCGS FSAR dfM Page 4 of e TABLE 3.5-13 (Cont) . CALCUIATED MISSILE SOURCE MSSIIE CHARACTERTSTICS MAX. STEEL CASING WEIGHT PERF. DEF115 THICKNESS REMARKS IDENTI- OF LOCATION VELOCITY DIA. E sq11s _jtnSL 11!!:L It.ast #1w.s t r u.s FImicli
~
0.094 0.le75 Filter homeing perforation will not ran slade suitchgear Aus slag 157 3.31 a.09 (filter occur. , Roos Unit SDG Area housing coolere E1. 16 3' thicknese) 14, 3-VH-401 , d dCentri-
- fugal Fan) i Fan Blade control Aus Bldg 174 1.45 4.867 0. 170 0.1875 casing perforation will not occur.
Also, tne f an le inside a filter Roos Supply SDG Area homelag. Unit El. 178' i i l 1A, B-VM-403 , ICentri-fugal Fan) t Fan Blade control Aux Bldg 280 1.37 0.153 0.069 0.1875 Casing perforation will not occur. - nouever, the fan blade may exit . l Area smoke Control through the auction side flezahle Vent ran Area ] E1. 1788 connector. There le no safe shut-l down equipment within the room. 4 10-V408 Therefore, protection le not l (vano-Aulal needed. I Fan) , Fan Elade D!esel Area Aux Bldq 281 1.72 0.902 8.092 0.1875 Casing perforation w111 not occur. ; l Houever, the fan blade could exit - , Euhaust Fan SDG Area through the auction side flexible El. 178' 1A, 5-V411 connector. A 1/48 thick et 1 i barrier to provided to enc 1 the 1 (Vane-Au tal section fleuthle connector.( l Fan) 1 Aux Bldg 260 3.33 23.9 0.383 0.25 Fan blade will penetrate throtelh Fan Elade Diesel the fan casing. However, there at
- Generator SOG Area no safe shutdoun equipment in the u poom pecir. El. 77' room. %erefore protection le not o Fan needed. co 1A thru H- g v412 O (Vane- Anlal N
' ran) G
- CO G
1 U G Amendment 7 i
.M,w
( q . . . . . . _ DSER OPEN ITEM gf , Neca rSAn , me Page 6 of e TABLE 3.5-13 (Cont) . CAICURATED sOueCr t!IDGI!?_ct!aESCTEatstics NAE. stEEI. CAsIMa NIssII.E DEMARES . I4 CATION VELOCITY DIA. WEIGHT PERF. DEP118 TMICENE8s IDEN11- OF (TN.) (IN.) ITCA1Ioll afISSIt s _1EI(SL j[Nd, jbBsL 8.49 0.22 0.25 caetag perforation will not accer. Fan glade ' Intake Intake 250 2.72 re se a wtre screen on ene Structure Supply Fan Structure E1. 1228 section of the fan cooler which WM i will prevent a fan blade from 1A thre leaving the cooler at an obliqueg
. angle.
D-v503 (Van-Aulal Fan) 8.49 0.22 0.25 Casing perforation will not accer. A Intake Intake 250 2.72 Fan elade Structure Structure to to no zzet1 Die Connector on he auction or the discharge side Euhaust Fan El. 122' lhle.4 1A thru 4 D-V504 . ' (Vane-Aulal
- l 3
Fan) 138 1.368 0.746 0.04 0.1015 Casing perforation will not occur. Fan Elade Traveling Intake The intake damper and vane guide Screen Structure on the section of the fan prevento Motor Room a fan blade from esiting in that Fan direction and the wane guide on the ' discharge of the fan prevente a f an OA, a-v55a blade from leaving the fan housing (Van-Aulal on the discharge direction. There-i Fan) fore, protection to not needed. peactor 98.s 16. 1 1816. 0.267 0.625 No caelng perforation. Impeller SACS Pumpe Bldg , E1. 102' r" 121.6 5.3 46.4 0.136 0.59 No casing perforation. Impeller Fuel Fool peactor L3 Cooling Bldq C3 Pues El. 1628 1 oi 0.0629 No casing per.foration. o. i' Impeller ECCS peactor 93.0 2.56 a.35 0.43 Jockey Bldg C Pump El. 54' g No castng perforation. O ' Imrela,r focus peactor 119.9 5.3 44.6 0.132 0.59 CD
- Bldq l
Mater G Cleanup El. 548 g. Pump C1: Amendment 7
e S/e4 - . seCGS FSAR DSEH OPEt3 ITEM gf : page 7 of e TAat.E 3.5-13 (cont) , cA14U1ATED . t!IssItz__c_ganactinig;Les Max. stEEt. cas m BEnARES nissstE sounct WEIGHT PESP. DEPTil THIcENESS i IDEN13- I.0cATIOct VEtOCITV talA. OF _n7Lst_ nt! L 3.sen ein. Jr u.i FIcMIgif missIts -
~
79.75 0.104 4.63 300 casing perforation.
- Impeller Ct:111ed Aus Bidq 82.8 5.97 68ater Pump control ?
Area E1. 155' It.79 0.068 9.39 100 casing perforation. D/G 1E Aus 8149 94.5 3.04 Impeller ! Panet Diesel clailled Area 't leater Pump E1. 17ee 0.77 Ito casing perforation. BACS Pump peactor 79.6 6.24 83.66 0.096 Impeller Didq E1. 778 1, 0.0596 0.51 Iso casing perforation. Impeller Service Intake '67.3 3.7a 26.3 Wter Structure Rooster E1. 798-8" Pump 15.2 1215.5 0.314 8.75 peo casing perforation. Impeller Service Intake 97.G kater Pump . Structure El. 9 3' i 0.219 1.125 too casing perforation. peactor 158.1 4.79 48.2 Impeller stecu pecir, utdg Pusp E1. I32' 31.8 0.053 e.5 300 casing perforation. 7 mpeller 36eCU Reactor 62.4 4.31 Precoa:, pidg , Pump E1. 145' 0.0429' O.801 3e0 casing perforation. i Beactor 57.6 4'09
. 25.6 '
1 Impeller 364cu isoldup Bldq
- u. i45'
. p Pu.p Reactor 70.4 4.04 15.9 8.0423 S.43 No casing perforation. Q Impeller 966C0 l
Backuash Bldg of E1. 132' % j Pump I e.675 too casing perforation. C3 Reactor 120.6 3.98 23.4 9 . 10 9 N Impeller CBD P ap aldq g El. 77' CO G 1 3 1 G a Amendment 7 9 i
DSER OPEN ITEN jl27 IICG13 FSAR S/84 . . Page 8 of a Tant.r i.s-ti (cont) l , cAirui.ATEo MItcsta sousCE gggastr.e c!!gAct! Lag;! TICS _ MAX. STEEL CASING IDE211- 0F IACATION VElaCITY DIA. WF.IGIIT PERF. DEPTN THICENESS PENARES ,
=
US&1195 MISS1I E iPT/SL jjNA 11.BsL _ _ {IN.] Irpeller Service Reactor 64.1. 3.98 13.94 0.0347 0.5 300 casing perforation kater Ridg Dewater E1. 54' Pump IIpe11Gr DCIC Pump Reactor 168.8 4.04 29.3 0.204 0.5 No casing perforation. Bldg El. 54' Impeller BPCI meactor 169.9 3.17 55.a2 0.339 0.625 too casing perforation. 4 Booster Bldg ! Pump E1. 54' , Icpeller NPCI Nain peactor 224 2.65 18.2 0.330 0.687 No casing perforation. fump Bldg '
! E1. 54' f HPCI Reactor 4Lates)* -("atslA "_9 -@ --ii i.. ; C M4M C $
i turbine Bldg DJlA 41a e4 WlA p3lA yJjA El. 54 (Later)a. qs g ,,gg A I * (t.aterf .(bateeld-(LaIA 9 DCIC 1urbine Reactor slag talA PalA -tN)^- ealA i E1. 54' b " aus o.ow F3 erl Sete 3,,, ntro 181.4 2.5 m.M -*A
- 1 E1. 548 .ta. ,
Ndciwd%5 Tf d'ensssala Shoesid u
"[he c.ondd.Ls have a rrnavirnus1r1 (41) volsJeric of 5 3 EdrConf- . o
- t. h e em ble is im)..-o ba b k Mp h h the condu.t. , crimring of
- f, Erire
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kG c.)LJ UtrlLt d ih o N G y',ll lee. codi%ca (d M c. M' n' Q *iede met [o.. s ve Luaho.s a. fJ1 sask de.an. , G UI G Amendmeest 7
~
JUI. 30 '84 0 2 6 8 6 5 8
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& G e poss'Me. Ah 4<Mne, robe 4 Mbc.
. bum + specel Qnot soh.cdical ateo.,rn. rhecelere_.,
prek.chon G needed. . G Tnaax+. e The. p3 Adtact miss'it.e_, 6 x w c- sdn c? % e_. l hv,v>hC4 k . TO wtht)./ c & td.di pu-[-ev- o e Oc q ,md
- - m pkde o.nd. pereJeoto #2. s" $vr6 -?%.e. sucreundik3 ccnceeAe s
- atekcc . Thes.e. e n.c p a s ac gench e{ sfQ mab wa h %
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DSER OPEN ITEM f2}
.. JUI. 30 '84 0 2 6 8 6 5 6 t l
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_ e e e e DSER OPEN ITIM j t
=
l ,
__...c. . HCGS FSAR 1/84 OUESTION 410.11 (SECTION 3.5.1) l The FSAR states that f ans are not considered as credible missile sources. Recently (Palo Verde, 1982) a fan at a nuclear facility generated a missile which penetrated the fan housing and damaged , a safety-related structure. Provide a discussion of the effects l of fan blades as a missile source and the means used to prevent damage of safety-related equipmen*, for each fan. RESPONSE 'I)c/cdc.,
" Y ., ...__.
_ _7 , _ m _ m _ . ... ..
. __m co er through-fan-housing missiles that would damage s y-relate ructures to be credible. The condition tha isted at Palo Verde lved workmanship deficiencies as t ade locknut torque and bla angle did not meet the su ter's specification. As sult, the blade ex enced fatigue failure and was ultimate ropelled e of the fan housing at an angle that penetrated the f 4 le nections of the fan ~and impinged the containment liner . HCGS has conducted a '
survey of vane-axial and c .ifugal . s in safety-related areas employing flexible con ors. We identt one instance where a postulated missil rough the flexible con tion of a vane- {': axial fan may h the potential to damage saf e-s 'down
' equipment i e room. In order to prevent the postu l sissile om damaging saf e-shutdown equipment, a missile Id I has en added to the design to withstand the impact of the
,-_tul:ted ri::il:.
INSERT l Section 3.5 has been revised to provide the results of an analysis which shows that internally generated rotating component missiles have no adverse effect on plant safe shutdown capability. oszz arts Irsx 8f 410.11-1 Amendment 4
~ .<
O HCGS FSAR 10/83 ( l 00ESTION 410.12 (SECTION 3.5.1) I The FSAR states that rotating equipment which is~not specifically identified does "unliklihood" thatnota constitute missile would a missile penetratehazard thebecause casing. ofProvide the the results of a quantitative analysis to verify this conclusion. l
-- ~
RESPONSE
rastA r x -
~~ d.ai ouy y=y w. ! =s a w wIa= = LI6=..
Mydeaihillig,on 3.5.1, will fail at HCGS and generate remote. ch a missi has suf:. ent energy to penetrate a component easing iles A review of analyses of internally generated missiles from performed for Pa rde verified that postu mp impeller or blade) typically do pumps and fans (e.g., enet the component casings. not have sufficient energy The formulae used by Palo Verd redict.the penetration resulting from missile is are pr ed in Reference 3.5-4. s and fans which are desi and constructed l Since HCGS uses the same recognized industry e and in accordane h standar those installed at Palo Verde, results of e analyses conducted for Palo Verde are indicative o {.1 ri
.... . p. . ;;;r.t . -
i ZNSERT Section 3.5 has been revised to provide the results of an analysis which shows that internally generated ' rotating component missiles have no adverse effect on plant safe
- shutdown capability.
l l l osza oFzN ITEM c77 - 410.12-1 Amendment 2
r , 1 JLt 30 'e4 0 2 68 6 5 6 HCGS FSAR 8/84
" OUESTION 410.13 (SECTION 3.5.1)
Provide a discussion of an analysis for each rotating component which verifies that the casing would be capable of retaining an internally generated missile. For each rotating component whose casing cannot retain the internally generated missile, verify that no secondary missiles will be generated from any internally generated missile.
RESPONSE
Section 3.5 has been revised to provide the results of an , analysis which shows that internally generated rotating component missiles have no adverse effect on plant safe shutdown capability. Secondy m*b cu-e nat. e.ondda n d web. M
" O M1 h% Mg CVVZ.
- N )
&C CVLf CM , M pdub O ry % 449 a W 6% %
1 I i DSER OPEN ITEM S 410.13-1 Amendment 7
.,_,__ ___ , -,.,--,--,_. ,.,_ _-.,__ _, ,-- , ,_. , -.____ .._.,., .,-- . _ _ _ _ _ - , _ _ - -- m
- 1. -
- l. .
i HCGS l (. DSER Open Item No. 3i (DSERSection3'/,1,$$) Ei
- 1. .
33I analysts results using finite element method and elastic half-space approach for contatament structure Atheresultsofthefiniteelementsoil-structureinteractionanalysisandthe impedance soil-structure interaction analysis of the containment structure. Ass PCv13f- . for the ins 0
- Ab'*A l-efgested odoVC f See th e Pc span.t e -fo h.scs open Meni c /. '
t.- i j
i l ,
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. l
- i HCGS l ,
s#
-k' OSER Open Item No. YO (DSER Section 17.T / 3 )
l i' S5I analysis results using finite element method and elastic half-space approach for intake structure Ressen se_ For the in forma Son ngu e s fed o.bo ve, s e e the tespanse +o bs EA opes ,*dem 64. t
HCGS
.. OSER Open Item No. II (DSER Section I I+ b )
Comparison of Bechtel independent verification results with the design-basis results.
RESPONSE
fees,, the 7his itkon correspds to Nem A./ 3 f NR c sMe.+ ural /6eclech,, co./ ma anny a fan u.sey so,191Y. A compwison o C Be e.Ade/ inofepes das d yes,'hea +iss resa/fs wi%-h the dash basis resu.as is a. Ha e s es' O
~ , . - - - - , -- --r- -- , , ,-- , , , , ,.,.n , - . , , - - -.,.,--,--._,-,,g-, - . - - ,,g, ------,
-. . ~~
\ Revision 1 i
- s . Meeting Date: January 10, 1984 7/2/84 l
~
Question Not A.-13 - 1 m Question: Provide ccuparison of Bechtel Independent Verification ! Results with the Design Basis Results. Responses As described in Amendment 1 of the FSAR (Section 3.7.2.4), three independent seismic soil-structure interaction The analyses design basis are performed for the major plant structures. l: analyses are performed using the finite element method by EDS Nuclear, Inc. (presently known as Impe11 Corporation). Independent finite element soil-structure interaction analyses i are sesequently performed by Bechtel to verify the design basis i analyses. In addition, in accordance with the reqdirements of ! the Standard Review Plan, Section 3.7.2 (NUREG 0800), impedance r approach (the half-space) soil-structure interaction analyses are performed by Bechtel. The analytical method utilized for the impedance approach seismic soil-structure interaction analyses of power block structures and service water intake structure is given in FSAR Section 3.7.2.1. Figure A-13-1 summarizes the division of responsibilities for the seismic analyses. Figures A-13-2 to A-13-37 show the caparison of the response spectra (2% damping) obtained fra the above three seismic soti-structure interaction analyses. Discussions of these m comparisons are as follows:
~
Power Block Structures
- , I. Caparison of Design basis and Independent Finite Element Verification Response Spectra Bechtel's independent soil-structure interaction analyses are performed using the computer code FLUSH. The results of l
1 independent finite element analyses are in reasonable agree-ment with those of the design basis analyses. As can be seen frem Figures A-13-2 through A-13-37, the horizontal response spectra obtained from the independent finite elanent analyses are generally enveloped by those obtained ' from the design basis analyses except for the frequency The vertical response spectra range lower than 2 Hz. showed some exceedances at the frequency range .of 18 Hz. These exceedances are listed in Table A-13-1. The effects of these exceedances are evaluated for the cambined responses in three directions using the SRSS { N I approach and ccupared with the design basis results. Table h A-13-2 provides these ccziparisons. In all cases, these
- variations are judged to be minor and can be accamodated z within the design margin. In areas where multimodal~ analy-5 sis is performed, the ef fects of these variations will be f ur ther reduced . It has been concluded that the variations
!o between these two analyses are within the accuracy of analyses and can be accamodated within the design margin.
> a l G5/48-1 i _ _ _ _ . . _ , _ . _ _ _ . _ . _ . _ _ . . . _ _ _ __ L .- ,_ _ .____ _ _ _ - _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ . -_
Response to NRC Audit Page 2 , l e
- k. II. Caparison of Design Basis and Impedance Approach Response
- spectra l The. peak spectral accelerations obtained frem the impedance approach analyses are generally lower than those obtained from the design basis analyses. However, these response spectra are not empletely enveloped by those obtained from the design basis analysis, especially in the frequency range between 1.0 and 3. 5 Hz. Also, there are some local exceed-ances in the higher frequency range, as shown in Figures A-13-2 through A-13-37.
As discussed during the NRC Structural Design Audit, dated January 10, 1984, sampling studies have been performed to . confirm the adequacy of the plant design. Table A-13-3 ) describes the criteria used in selection of the sanples for this study. The results of sampling studies are as I!ollows:
- 1. Structures All major reinforced concrete shear walls at the base of j the reactor building have been evaluated for seismic forces and moments obtained fra the impedance approach analyses. The actual shear stresses resulting from the
- s. impedance approach analyses were evaluated and found to be lower than the design basis stresses. Table A-13-4 provides the caparision of shear stresses at El. 54'-0. '
Tables A-13-5a and A-13-5b show the comparision of impedance approach and design basis moments for OBE and SSE cases respectively. The impedance approach momencs exceeds the design basis moments at a few wall locations as indentified on Tables A-13-5a and A-13-5b. These walls were reevaluated and the resulting moments were found to be less than the allowables. The auxiliary building seismic forces and moments obtained from the impedance approach analysis are less than the design basis shears and moments. Therefore, no further evaluation of the auxiliary building structure is neces-sary. I Based on the above, it is concluded that the as-built power block structures can acccounodate the loads obtained fra the impedance approach analysis.
; 2. Equipment
( , ( 5 The ef fects of the impedance approach response spectra
. U was evaluated on 26 types of equipment. The selected z items are located in the areas where the impedance approac E
O E G5/48-2 . E
m ponco to sus muuts - * ' Fage 3 - n 2. (Cont'd) - o C spectra were found to have. higher spectral accelerations than those of the design basis response spectra. Each equipment was evaluated in accordance with the procedure i described in Table A-13-3, and the results of the evalua-r tion are summarized in Table A-13-6. In all cases, the as-built equipment designs were found acceptable.
- 3. Cable Tray and HVAC S'apports i
- a. Cable Tray Support Approximately 200 supports were evaluated. Iri all cafes, the existing designs wre determined to be ac optable.
- b. HVAC Supports
- Over 200 supports were evaluated. In all cases, it was found that the design basis spectral acceleration exceeded the impedance approach spectral acceleration i for the support fr equencies. There fore , the HVAC
, . supports were considered acceptable.
- 4. Pipino ahd Pipe Supports O A total of 10 representative piping system calculations were selected out of 64 calculations af fected by the impedance approach analysis results. The selection of these calculations was based on the criteria given in Table A-13-3.
The objective of performing detailed dynamic seismic analysis of the sample calculation was to demonstrate tha although the design basis curve did not envelop the impedance curves in the low frequency range , such devia-tion do not have any af fect on the adequacy of existing piping analysis and support design. In other words , the stresses and loads generated using the imA dance response spectra curve as input are still within the ASME Section III code allowable for pipe and pipe support design. The methodology used for evaluation was to subject the selected existing mathematical models of piping systems to the impedance approach response spectra and to capare the resulting pipe stresses with the ASME Section III
- code allowables for pipe and pipe support design. The reactions at equipment nozzles were capared with vendor' design allowables. All pipe supports were evaluated for adequacy under the revised loads.
* ** OPEN z m
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l Response to letC Audit . Page 4 ' In all cases, tho pipe stresses were found to be Also, within the code allowables as shown in Table A-13-7. as illustrated in Table A-13-7, the equipmernt nozzle The existing pipe support allowables were also met. designs were also found adequate for the new loads and met the ASME Section III code subsection NF allowables. This is 111ust:sted in Table A-13-8. Intake Structure
~ "See responses to questions A-14 and A-16, meeting date January 11, 1984. !
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m . N Table A-13-1 (Cont'd) O Camperleon of Design Basis and Independent ! Q Finite Element verification Response Spectre _
;I '
Locatione r: spectral Aeoeleration
$ Va riations Figure l Ites (Note 2)
N PAa11 dine Eey Design Earthquake Elevation Earthquake Direction (Note 1) No. No. Design Baste! Bechtel FLUEN (e) (W) H-S 1.7 Ms A-13-21 12 8.34 S.42 , REACTOR 102 OnE l 54 onE E-W 4.3 us A-13-2 3 13 9.54 e.67 , q 201 On E - E-W 1.8 Ms A-13-2 5 14 0.30- 0.55 i OnE vertical 22.0 Hz A-13-27 15 1.20 1.42 102 > l < 201 Ons vertical 90.0 Ms A-13-28 16 1.68 1.45 1 ADEILIApr 54 OnE H-S 4.9 Ms A-13-29 17 1.15 1.48 I 54 ons E-W 4.4 Hs A-13-32 18 8.75 0.05 l i 54 OnE Vertical 22.0 Hs A-11-35 19 1. 17 1.26 i 102 Ons Vertical 18.0 Ms A-13-37 20 1.47 1.54 4 ! 17e ons Vertical 18.0 Hz A-13-37 21 1.e4 1.95 i I HofES: 1. This column identifies those locations there the results of the 1 independent analyste exceed those of the design heele analyele. i 2. For vertical earthquake direction, spectral acceleration includes i . a the ef fect of gravity load ( 1.0 g). l c-5/4a }, .
- - . - - . . .2._ m Tnbin A-13-2
- SRSS Spectral Acceleration Comparison between Desien Basis and Finite Element verification Analysis b,
Item SRSS Snectral Acceleration Coenarison( g) (Note 1) No. (A) (3) (s-A)/A Desien Basis Bechtel-FLUSE Difference (4 ) 1 1.97 1.75 -11 _ 2.24 2.20 -2 2 1.33 1.78 16 3 1.39 1.72 24 4 2.23 2.49 12 5 2 86 2 68 -6 6 2.34 2.32 -1 7 2.56 2.48 -3 8 4.27 3.44 -19 9 1.87 1.93 4 10 L 11 1.73 1.93 11 1.41 1.38 -2 13 13 2.02 1.66 -18 1.52 1.50 -1 14
- ~ -
1.21 1.43 18 15 1.71 1.86 9 16 17 2.24 2.07 -8 I 18 2.23 1.94 -13 l l 1 . 19 1.27 7 19 20 1.36 1.99 7 1.51 1.56 3 21
. ECTE: 1. The SRSS spectral acceleration values include the effect of gravity loads (1.0 g)
DSER OPC1 ITEM ff l . 1 l i
Y TABLE A-13-3 PROCEDURES FOR EVALUATION OF_ STRUCTURES, EQUIPMENT & COMPONENTS ,
^
USING IMPEDANCE ANALYSIS RESULTS k.3 , INTRODUCTION , The results of the impedance analysis are used to assess the existing design of the HCGS structures, equipment and , components. A sampling approach is used. The procedure for this evaluation is as follows: A. STRUCTURES: Since the maximum shear and axial forces and the maximum overturning moments occur at the base of the structures, and the design margins for the upper elevations are greater than
- those of the base, the ef fects of these loads at the base of each structure are evaluated.
B. EQUIPMENT: The impedance analysis spectra in general are not completely enveloped by the design basis spectra in the following areas i) 1.0 to 3.5 Hz range throughout the reactor and auxiliary buildings ii. ) 6 to 15 Hz range in the reactor building at elevation (,_ 102 ft and below. iii.) 6 to 15 Hz in the auxiliary building at elevation 54 f t. Since typical equipment frequencies are not found in the range of 1.0 to 3.5 Hz, the iten (i) above does not need any f urther evaluation.. Items (ii) and (iii) are reconcile as follows: Review the significant frequencies of approxime tely 30% of all equipment selected at randem and located in the areas where spectral variations were noted.
. If the significant equipment frequencies f all in the range where the dif ference in the spectra exist, additional eval-uation is necessary. No f urther evaluation is necessary if the significant frequencies are outside the frequency range in question.
. The evaluation is performed either by comparing the test response spectra of the equipment with the impedance spectr (if the equipment is qualified by testing) or ccuparing the actual-to-allowable stress ratios with the spectrum exceed-ance ratics.
. If the above evaluation shows the equipment may not be qualified for the impedance spectra, detailed evaluation consisting of analysis and/or testing is performed.
DSER OPEN I"Di d~/ 6
As a result of evaluation, if equipment requires modifications, the sam, le size for this evaluation. is q expanded as required. Qi C. CABLE TRAY AND NVAC SUPPORTS . l Cable tray and HVAC supports do not have frequencies in the range of 1.0 to 3.5 Hz. Therefore any dif ferences between the two spectra in this frequency range do not' require 'any evalua-tion. The ef fects of -the spectrum exceedances at frequency range - between 6 and 15 Hz are evaluated for approximately 200 cable i tray and HVAC supports. These supports are selected at random Dut are located at the lower elevation (Reactor Building El. 54 to 102 f t. , Auxiliary Building El. 54 f t. ) where the spectrum dif ferences exist. If the results of evaluation i'ndicate need for modifications to any support, the sample size for this
- evaluation is expanded as required.
D. PIPING AND PIPE SUPPORTS In genefal, impedance curves resulted in significant reductions in response spectrum peak accelerations as ccmpared to those of the design basis curves. However, freque ncy shif ts were observe in some curves, particularly in the low frequency ranges. To evaluate the effects of the frequency shif t, a ' biased" sample of affected piping systems is reanalyzed and reevaluated. The sample is selected as follows:
- Individual impedance curves for various elevations and structure are superimposed on their correspnding design basis curves to identify those impedance curves weich are not enveloped by desit Dasis curves. Those impedance curves are then superimposed on the design basis " enveloped" response spectra usec for various piping system design calculations. If the design basis envelop.
response spectra curves af facting a calculation did not totally envelop all the corresponding impedance curves, that particular calculation is then identified as "af fected" and a candidate for sampling . A "bi& sed" sample of the "af facted" calculations was selected I which emphasized the following important piping parameters : l l 1. Stress levels in the existin;; pipe stress -calculations. I' Samples included systems with high stress levels, l' 2. Dif ference in "g" level (Ag) between impedance and design basis curves in the af facted frequency zones. Sample selec-to include curves showing significant dif ferences.
- 3. High equipment nozzle loads in existing calculation.
- 4. Relative location of piping system in the plant in an atten to include response of all structures in the sample selecte-DSER OPEN ITEM f/
~
1
. l i
I l \ The number of calculations included in the sample is: ('3 Total No. No. of Calca No. of Calcs No. of Calcs Building of 0-cales_ Reviewed arrece.a in the sample _ 32 23 3
- l. Drywell 32 34 5 Reactor 213 213 124 7 2 Auxiliary 124 Results of the analysis including su'pport loads are compared against the design basis values for acceptability.
4 O l l l l l l l v e 1 I DSIR CPEN ITEM f/ i l . __ _ . - . _.- _ __-_ _ . ~ , , . _ _ _ . . . _ _ _ _ . - _ _ - .
i I TABLE A-13-4 - - O l
~.
!! 54'-0* (' REACTOR SUILDING SEEAR STRESSES AT EL. I l-l Design Impedance ' I Basis Approach Allowable Wall Psi Psi Psi Location i
. 207 -430 North wall 323 224 630 South wall 333 261 630
' 29 6*
. East vall 268 430 West Wall 303 251 830 Cylindrical Shell 257 91 126 Pedestal 27 i
54'-0" SOUTE RADVASTE SHEAR STRESSES AT EL.
)
Design Impedance Basis Appr o ach Allowable Wall Psi Psi Psi Location L 183 207 630 North Wall 216 224 630 South Wall 208 276 630 East Wall 458 257 630 West Wall Notes: 1. Concrete f' c = 40 0 0 Psi
- 2. See F S AR Figur e s 1.2=2 for wall location.
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TABLE A-13-Sa (~
~
~
REACTOR /RADWASTE BUILDING - CBE SEISMIC MOMENTS AT EL. 54'0* 1spedance Design Basis Approach Me tho d Me tho d
~
Wall Location (Eip-Ft) (Kip-Ft) N o rth-Reacto r 359,200 414,500
^
N orth-Radwa s te South-Reacto r 847,700 South-Radwa s te 517,400 461,000 421,900 Ea st-Radwa ste 329,000 290,700 We s t-Radwa s t e 434,500 276,900 L' Ea s t-Reacto r 588,600 432,900 We s t- Reacto r cylindrical 2,772,000 (N-S) 1,847,000 (N-s )- Shell 1,723,000 (E-w) 1,639,000 (E-w) Note See FSAR Figure 1. 2-2' fo r wall location. l l t t DSER OPEN ITZM f/
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-- TABLE A-13-5b U
REACTOR /RADUASTE BUILDING - SSE SEISMIC MCMENTS AT EL. 54'0" I i Impedance Design Basis Appr oach wall Location Me tho d Metho d (Kip =Ft) (Kip-Ft) . . . North-Reacto r 499,100 N orth-Radwa ste 912,100
' i South-Reacto r l South-Radwe ste 1,344,000 1,429,000 475,000 732,300 E a st-Radwa s te 554,000 504,500 We st-Radwa s t e 909,000 480,200 Ea st-Re acto r vest-Reactor 1,320,000 _. s 3.7 , 4 0 0 _ .
i cylindrical 4,4 71,000 (N-s ) 3,092,000 (H-5) i shell 3,054,000 (E-V) 2,668,000 (1-"I Note: See FSAR Figure 1.2=2 for wall location. l 4 l I DSER OPEN ITEM [/
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- TnELE A-1>4 ,
k., l
.. POWER BLOCK SEISMIC CATEGORY Z EQUIPMEttT Egn13usent Method of Equipment Pregimencies Seimnic Applicab:
i er Incation Camponent Tag Bo. Elgd./31. (Es) 9aalification Rote Beactor 31dg. Borisontal- 10,12 1 Dertical - 23 Testing EPCE Turbine 341-C002 31. 54 i 1 1 Besidual Beat . Esactor 31dg. Borizontal- S.7, 9.7 Analysis 3
,[ Eamoval Pump / E11-C002 Estor 31. 54
- 9ertical - 233 E11-M 17 ~ '
E11-M 18 Aux. Eidg. Control mean Borizontal- 11.5, 14 Testing 1 Panels E11-M40 E1. 102 - 511-M41 vertical - >33 E11-M2C (, ' throngt: E11-M23 Aux. Eldg., Borizontal- 21, 29 Control Roan Testing Panels 111-M28 E1. 102 9ertical - >33 _ 1 E22-M31
~ ~ ~ -
4 111-P635 Aux. Ridg. Borizontal- 19, 37 control Roam Testing i Panels E11-M36 El. 137 Vertical - >33 1
~~^
' ' ~ " ~ ~
Aux. Eldg. Borizontal- 7, 12 Control mean Testing 1 Panels 111-608 II. 137 vertical - >33 E11-409 Aux. Eldg. Borizontal- 22, 37 i control Roan Panels E11-411 E1. 137 vertical - >33 Testizg 1 Esecr:or 3149 . Borizontal- 16 Analysis a 1, 2 RCIC Turbine ES1-C002 II. 54 vertical - 18 Testing I.. i l I,7CS Pump / E21-C001 Reactor 31dg. Borizontal- 11.5, 12.7 Dertical - >33 Analysis 2 Ester 31. 54 DSER OPEN ITEM f/ 1 l O
- - - - - ~-- -- - . . . . . . , __. _. _ _ _
Th3LE A-13-6 (Cont'd) . (.
- powsa ELocK SEISMIC CATEORY Z E2UIPMENT i
I~ s T l
' Equipment Method of Equipment Seismic or Imcation Freguoncies Applii l
Omeponent Tag Bo. 31gd./E1. (Es) Qualification No-
~ ~ ~
~ ~ ~ ' ~
Chiller Water ' IhT, D. G .
- Eorisontal - >33 '
' Tank ET 410, 413 31. 178 Dertical - >33 Analysis ECCS Jodney IAP, EP, Reactor 31dg. Borisontal - 233 Analysis Pump CP, D7 22B E1. 54 Dertical - >33 l
~~
SACS Expansion IAT, Reactor Bldg. Borisontal - 12.5 Tank yr 205 31. 201 Dertical - >33 Analysis Reactor 31dg. Morisontal - 8, 14 5.0 Ky Stitch- IAN , El , CN, DM 205 E1. 102 Vertical - 30 Testing year __ s
'a DC Stitchgear ~~
& Control 100 251, Reactor 31dg. Borizontal - 8, 35 EL. 54 vertical - 20 Testing Canter 261 i
Anx. Bldg. Horizontal - 14, 16 Batteries IQD 421, Testing Racks 431 El. 54 Vertical - 28 Anus. 31dg. Horizontal - 17, 21 Inst. AC Power ' ITF 401-4 07 Testing Panel ITF 209 E1 102 vertical - 6
' IAC, 3C 201 Reactor 31dg. Horizontal - 8, 17 Control Panel E1. 102 Horizontal - >33 Analysis l Nota:
- D.G. - Diesel generator area of the auxiliary buildinq i
i l DSER OPEN ITEM g/ i I I
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- TABt.E A-13-6 *(Cont'd!
BOWER BLDCK SEISMIC CATEGOltY Z E2UIPMENT m Equipment Bet 2nd of Equipment Seimic Applica i or Incation Freguancies ' (Es) Qualification ' Noti Campenent Tag Bo. 31gd./El. Standby Diesel ' 1(A-D)G 400 D. G . Morisontal - >15 vertical - >15 Analysis 2 Generator Set 31. 102 Analysis 2
SACS Esat 1A15, 1&2E201 Reactor 314g. Morisontal - 8, 10.4 Enchanger 1ME, 1523201 E1. 54 Dertical - 21 1
SACS Fumps 1(A-D)F210 Reactor 31dg. Borizontal - >33 Analysis 2
- 31. 201 Dertical - 333 4
Control Panal ICC, DC201 Reactor 31dg. Morizontal- 12.7, 17.4 Analysis 2
- 31. 102 Dertical - 29 1AT, E 412 D. G. Morizontal - 31, 33 I
Acaamulator Analysis 2 Tank EL. 54 vertical - 35 l .- Air Bandling 1AW407 D. G . Borisontal - 16.6, 18 V Analysis 2 Units 1BW407 E1. 178 9ertical - 19 , A/C Units i Unit Cooler 1AM208 Reactor 31dg. Borizontal - 9.4, 21 11 102 Vertical - 26.4 - Analysis 2 i 1AM209
- 15V5208 tBv5209
[ l 1AC, CC2s5 D. G . Borizontal - 12.7, 16.4 l EVAC Control 1AC, CC251 11 178 Vertical - 16.9 Analysis 2 !' Panels 1AC, DC48 3 Borizontal - >3 0 - - D. G . Centrifugal 1AK, EK403 El. 178 vertical - >30 Analysis 2 Water chiller l Motes: 1. TRS envelopes impedance appoach spectra.
- 2. Impedance appoach spectral acceleration is lower than that of the design-basis respnee spectra in the mapr equipment featusacies.
- 3. Although impedance appoach spectral acceleration accesis that
- of desip basis respose spectra in the equipment frequency range, a more estalled caloalation showed that the aguipent stresses are utthin the code allowebles.
DSER OPEN ITEM g/
- SBLE A-13-7
+ . . M EIA S PIPE STRESS
SUMMARY
l milding cale. Max. Seismie Stress Ratios Asts Code Equation Max. Impedance Stress Evaluation Deador No. Equip. Nozzls Max. Design Basis Stress Eq. 93* , 3e. 90' Code Allowable Code Allotable Allowables M. WE SSE Usset ' Faulted C1549 0.51 0.76 0.29 0.46 us j l hamiliary ns 0.64 0.04 0.40 0.28 C1541 0.75 0.83 0 44 0.34 ns C110 ; 0.65 0.83 0.43 0.85 ns Drywell C1842 C120 0.30 0.52 0.49 0.39 us C988 0.88 0.75 0.54 0.35 33 C911 0.88 0.94 0.84 0.63 us C943 1.10 1 . 10 0.71 0.47 35 Reactor 0.29 0.39 0.33 0.21 33 C CSIS C937 0.90 1.15 0.70 0.38 33
'AsME Section III NC, ND-3652 4
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( [ E nat.: A-is-a , POWER ELD"5C PIPE SUPPORF 14AD
SUMMARY
( = Average Percentage Supprt hilding cale. Total No. No. of Design I No. of supprts wLth increase in Load i Load Zacrease Upset Faulted Adequate i Supports 0 WA WA ES C1549 5 auxiliary ES 15 6 11% NottE C1541 . 8 1 2% ~ 1% MS C118 C1842 34 0 WA N/A us Drywell . 18 2 7% NotEE ES C120 3 NOIE 14% TES CSSS 11 34 6 2 06 17 % 33 CS11 7 4 274 28 % MS Reactor CS63 10 0 N/A N/A ES b CSIS CS37 17 5 17 4 21% ns _w- .-- -- 4 a- ~ ---. ew+ - 1 O M
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<r DEVELOPMENT W PSAR CRITERIA l
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- DEVELOPMENT W RESPONSE l ' COORDINATION AND REVIEW ~
.l SPECTMA AND DESIGN OF IMPELL~ ANALYSIS LOADS USING FINITE ELEMENT APPROACH
- PERFORM INDEPENDENT VERIFICATION ANALYSIS USING f
- 1. FINITE ELEMENT
(~') (FLUSH) APPROACH ii. IMPEDANCE APPRCACH l Figure A-13-1 Division of Responsibility i
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Revision 1 July 10, 1984 Meeting Date January 10, 1984 Question No.: A.16 , Provide calculations of ductility ratios due to 3311193: . pipe break for key elements. ARE2515: r:AR section 3.e.4.s.2 discusses the allowable ductility ratios. used for the desip of pipe whip restraints. For flemure in beans, an allowable ductility ratio of 20 is used. As discussed with the NRC staff, originally the majority of the pipe whip rastraints had ductility ratios less than or equal to 10. However, the ductility ration for approximately 258 of the pipe whip restraints exceeded 10 under the original design basis. These restraints have been reevalu-ated based on ae-built conditions, final pipe break loads and actual het gap requirements. This reevaluation revealed that all flemural members for pipe whip restraints have an actual ductility ratio
- of less than or equal to 10.
k FSAR Section 3.8.4.8.2 will be revised to reflect compliance with SRP 3.5.3 for actual ductility ration of flexural seabers. s G l I e e A.16-1 essa ons ITEM f%
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DSER Open Item No. bb -(DSER Section I I. 8)
- Impedance analysis for the intake structure
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'* Meeting Date: January 11, 1984 .
h ~ Question No.: A-16 Perform an independent seismic verification analysis
- j. Question:
(impedance analysis) for the intake structure and ccampare the results with design basis results. Consider the ef fects of side boundaries, embedmont and the presence of water masses in the analysis. . Response .. In accordance with the requirements of the Standard Review Plan, Section 3.7.2 (NUREG 0800), impedance approach (half-space) seismic l soil-structure interaction verification analyses of the service The ana-water intake structure (SWIS) are performed by Bechtel. lytical method used for the impedance approach seismic soil-structu l, interaction analyses of the SWIS is described in FSAR Section 3.7.2.1. The ef fects of side boundaries and embedment are considere using the method described in References A-16-1 to A-16-3. The i ' effects of water masses are also accounted for by adding ef fective water mass to the related nodal points of the structural model in
' accordance with procedures described in Reference A-16-4.
Figures A-16-1 to A-16-18_ show the comparison of the 2 percent damping response spectra obtained from the design basis finite element and the impedance approach seismic soil-structure interac-tion analyses. The impedance approach response spectra generally are enveloped by those obtained from the design basis analyses at v elevation 114.0 feet of the SWIS. For other elevations, the impedance approach spectral accelerations exceed the design basis spectral accelerations in some frequency ranges. These ranges vary approximately between 1.5 and 10.0 Hz. l As discussed during the January 1984 NRC Structural Audit Meeting , sampling studies have been performed to confirm the adequacy of l the SWIS design. The criteria used in selection of the samples for this study is given in Table A-16-1. The results of the sampli studies are as follows :
- 1. Structure All major reinforced concrete shear walls at the base of the intake structure have been evaluated for seismic forces and '
moments obtained frcza the impedance approach analyses. The she stresses resulting from the impedance approach analyses wre compared with those of the design basis analyses. Table A-16-2 shows ccuparison of shear stresses. In all cases these revisec l l shear stresses were found to be within the allowables. The moments in the walls, obtained from the impedance analyses, were asaller than those of design basis analyses for both the Eas t-West OBE and SSE cases, therefore, no further evaluation c these walls is required. m osza crzs : zx (, /,
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. Response to Question A'-16 (cont'd)
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p C For North-South OBE .and SSE cases , the moments obtained from
- impedance approach analyses exceeded the design basis moments.
The increase in moments were mostly isolated to the eastern , portion of the intake structure. This portion of the intake ; structure was reevaluated and the resulting acments were found to be less than the allowables. Based on the above, it is concluded that the as-built SWIS can accommodate loads obtained from the impedance approach analyses
- 2. Equipment __
The ef fects of the impedance approach response spectra was evaluated on 8 types of seismic category I equipment located in the areas Wtere the impedance approach spectra were found to have higher spectral accelerations than those of the design basis response spectra. The equipment evaluated represents over 30% of all equipment located in the intake structure.
!' Table A-16-3 ' summarizes the results of the above evaluation for equipment in the Intake Structure. It is concluded that all category I equipment can accasmodate the response spectra
- obtained from the impedance analyses.
- 3. Cable Tray and dVAC Supports Q All cable tray and HVAC supports were evaluated using the impedance analysis results. All supports were found to meet the impedance approach spectral response requirements.
- 4. Pipinc and Pipine Supports Piping and pipe supports were evaluated using the screening techniques discussed in Table A-16-1. The results are summar-
! ized in Tables A-16-4 and A-16-5. The analysis results show that piping stresses and nozzle loads are within allowable limits. There was no load increase found on existing supports. 1 It is therefore concluded that the existing design margins associated with the present project design basis seismic loadin are not af fected by the consideration of the loads generated fran the impedance approach analyses as demonstrated by the SW: piping systems. l
References:
A-16-1, Ap sel , R. J. , l.1979) " Dynamic Green's Functions i for Layered Media and' Applications to Boundary Value j Problems", Ph.D Thesis, University of California, San Diego. i I DSER OPEN ITEM h[p I
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References:
(Cont' d) .
\. A-16-2, Wong , H. L. , and Luco , J.E. , (1978) " Tables of Impedance Functions and Input Motions for Rectangular Foundations", Report No. CE78-15: University of California, San Diego.
- A-16-3, Barneich, J.A. , Johns, D.H. , and McNeill, R. L. ,
(1974) " Soil-Structure Aseismic Design of Nuclear Ir.teraction Power Parameters for Stations", Preprint 2182, ASCE National Meeting on Water Resources Engineer: January 21-25. A-16-4, Newmark, N. and Rosenblueth, E. , " Fundamentals of Earthquake Engineering," Prentice-Hall, Englewood Cliffs, N.J. (1971) 1 f DSZR OPEN I*EM Q 5
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- TABLE A-16-1
- PROCEDURES FOR EVALUATION OF k;7 INTAKE STRUCTURES, EQUIPMENT & COMPONENTS USING IMPEDANCE ANALYSIS RESULTS INTRODUCTION The results of the impedance analysis are used to assess the existing design of the BCGS intake structure,The' equipment andfor this procedure ccamponents. A sampling approach is used.
evaluation is as follows:
- A. STRUCTURES:
Since the maximum shear and axial forces and the maximum overturning moments occur at the base of the structure , and l the design margins for the upper elevations are greater than I those of the base, the ef fects of these 1; ads at the base of ' the structure are evaluated. B. EQUIPMENT: The impedance analysis spectra in general are not completely enveloped by the design basis spectra in the 1.5 to 10.0 Hz and in the ZPA range throughout the intake structure. The following procedure is selected for reviews (! . . Review the significant frequencies of at least 30% of equipment located in the areas Wiere the impedance approac spectra were found to have higher spectral accelerations than those of the design basis response spectra. If the significant equipment frequencies fall in the range where the dif ference in the spectra exist, additional eva; uation is necessary. No further evaluation is necessary : the significant frequencies are outside the frequency ran; in question. The evaluation is performed either by ccuparing the test ' response spectra of the equipment with the impedance spec-(if the equipment is qualified by testing) or comparing t; actual-to-allowable stress ratios with the spectrum excee, ance ratios. If the above evaluation shows the equipment may not be ' qualified for the impedance spectra, detailed evaluation consisting of analysis and/or testing is performed. As a result of evaluation, if equipment requires sedifica i tions, the sample size for this evaluation is expanded as required. e e osza em :m g i I
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[- C. CABLE TRAY AND HVAC SUPPORTS . All cable tray and HVAC supports are evaluated for impedance ("' analysis results. . D. PIPING AND PIPE SUPPORTS In general, impedance curves resulted in significant reduction: in spectral accelerations as compared to those of the design basis curves. 'However, in some curves, the peak accele: i- ations showed small increases. To evaluate the of facts of the increase in psak accelerations a " biased" sample of af facted The sample is piping systems is reanalyzed and reevaluated. selected as follows: Individual impedance curves for various elevations and structu: are superimposed on their corresponding design basis curves to l identify those impedance curves which are not enveloped by des basis curves. Those impedance curves are then superimposed on j the design basis " enveloped" response spectra used for' various If the design basis envelo l piping systen design calcuhtions. t response spectra curves af fecting a calculation did not totall: envelop all the corresponding impedance curves, that particula calculation is then identified as "r.f fected" calculation and a candidate for sampling. A
- biased" sample of the "affected" calculations was selected which emphasized the following important piping parameters :
Ci 1. Stress levels in the existing pipe stress calculations. Samples included systems with high stress levels. 1
- 2. Dif ference in "g" level (ag) between impedance and S design ample sele l
' basis curves in the af fected frequency zones.
to include curves showing significant dif ferences.
- 3. High equipment nozzle loads in existing calculation.
The number of calculations included in the sample is: Total No. No . of Calcs No. of Calcs No . o f Calcs Reviewed _ affected in the sam;'. Building of 0-Cales Intake 5 1 S tructur e 11 11 1 I Results of the analysis including support loads are ccupared against the design basis values for acceptability. l
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- Table A-16-2 .
the Base Intake Structure Shear Stress at )
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l Design Impedance Allowable l Base Wall Approach (psi) Location Base Elevation' (psi) (psi) Column Line Col. A 124 630 80 (East Wall) 79'-8" 66 98 630 79'-8" Col. Ac 47 73 630 79'-8" col. Ak Col. C 77 126 (West Wall) 47 70'-0" Col. 5 ' 214 630 l 79'-8" (South Wall) 230 f-200 17 6 630 79'-8" Col. 7 Col. 9 630 230 214 79'-8" (North Wall) f Notes: 1. Concrete f'c = 4000 psi.
- 2. See FSAR Figures 1.2-40 and 1.2-41 for wall location. .
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Tablo A-16-3 2ntake Structure - Seismic category I squipment m Fundamental Method of Equipment Frequencies Beismic , Applicable er Tag No. Elev. Note Camponent (Es) 9aalification 70'-O '& Eorisontal - 7.4,14 Trstelling Analysis 2 m ter Screen 1(A-D)S501 114'-0* Vertical - >33 l (T.WJ. ) Borisontal - 21, 30 Control Panel Testing 1 107'-0* vertical - >33 l l (ser T.W.S.) 1(A-D)c515 Service Water Borisontal - 28.4 3 Vertical - >33 Analysis Pumps 1(A-D)P502 33 '-0
- Analysis 2 Supply Fans CAV554 123'-0* Eorizontal - >33 03v558 Dertical - >33 i
Analysis 3 inne Axial Fans 119-D7503 122'-0* Eerizontal - >33 1AT-Df504 vertical - >3 3 - J Borizontal - 15, 22 EVAC Control Analysis 2 Panel 1(A-D)C 541 93 '-0
- 9e rtical - >3 3 Travelling Screen Spray Borizontal - >33 Analysis 2 Mater Bocater T AP-DP5 07 79'-8
- Ve rtical - >3 3 Pumps 1
Borizontal - 29,31 Transformer Panel Board 10Y 501-504 93'-0* Vertical - 23 3 Testing 1 { I Notes: 1. TRS envelops impedance approach spectra. l
- 2. 2spedance appoach spectral acceleration is lower than that of the f design basis response spectra in the ma9e equipment frequencies.
- 3. Although impedance appoach spectral acceleration accents that of Assip basis response spectra in the equipment frequency range, a more detailed calculation showed that the equipent stresses are within the cods allowables.
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Intake Structure Pice Stress Sumary mx. Seimic Stress Ratics AStE Code Equation Calc. No. Evaluation Wrdor Equipier Max. Iiroodance Stress Eq. 9B* Eq. 9D* Neszle Allomble Max. Desian Basis Stress Code Alls. Code A11w. Het CBE i SSE Upset Faulted C2019 0.46 0.51 0.26 0.14 hs
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g DSER Open Item No. 1472.5.r.d}kDSERSection9.3.1)
COMPRESSED AIR SYSTEMS ;
$ The service air system consists of two 100 percent capacity trai*ns of compressors, aftercoolers, moistura separators, receivers, and associated piping and valves. Cooling is provided by the turbine auxiliary cooling system. One compressor runs automatically with the other compressor on standby. The standby compressor starts automatically on f ailure of the first system or f ailure of the This system first system to meet the demand for compressed air.
maintains a constand pressure in the instrument air system. [ Die applicapt has not provided an FSAR figure which identifies each air user, the location of each user, and all accumulators, check valves, and other appurtenances associated with safety related component *s, systems,.and equipment, such as the ADS. _The appli-cant has not provided readable figures in the FSAR, due to the drawing scale factor.] The service air compressor supplies air I
for the instrument air system by means of an intertie between the service air system and the instrument air system before the
' instrument air dryer package. The isolation between the two air systems is supplied air from the emergency air supply . system (consisting of one compressor, filter, af tercooler, moisture (3 separator, and receiver) for all accidents except a LOCA. Cooling is provided by the reacotr auxiliaries cooling system.
[The applicant has not identified the location of the equipment and the component classifications on the FSAR figures. Therefore, we cannot conclude that air systems satisfy the requirements of General Design Criterion 2, " Design Basis for Protection Against Natural Phenomena," and the guidelines of Regulatory Guide 1.29, Positions C.1 and C.2, " Seismic Design Classification. ] 7-
~~
A scheduled program of testing and inspection of the system will'
- be provided to ensure operability of the system components'and For compliance with the requirements-of-GDC 1, control systems.
see Section 3.2 of this SER. _ _ _
J The service air system has no functions necessary for achieving safe reactor shutdown condition nor for accident prevention or mitigation. [The applicant has not identified and demonstrated that all instruments, controls and services required for safe shutdown of the plant such as the MSIV and ADS valves are pro-vided with seismic Category I passive air accumulators to assure their proper function in a loss of the air system.] All other
- air-operated valves including the scram discharge inlet and outlet valves and other devices are designed to move to a safe position f on loss of instrument air and do not require a continuous air i
supply under emergency or abnormal conditions.
e 147-.t.:,d-1 K53/4 E
, . . . . . . . ~ . . . . _ . - -
. .I
[ Additionally, the applicant has not verified that all' station air system containment penetrations are provided with redundant Therefore, seismic Category I, Quality Group B isolation valves.
we cannot conclude the requirements of General Design Criterion L
2 and the guidelines of Regulatory Guide 1.29, Position C.2, are "'
satisfied.]
The service instrument air systems-will initially meet (ANSI) .the re-MCll.1-quirements of American National Standards Institute [The" applicant 1976, using non-oil-lubricated air compressors.
i has not committed to perform periodic air quality testing of the -
air system to assure compliance with the requirements of ANSI-MC11.1-197 6. ] .
I
! [ Based on the above, we cannot conclude that the safety-related the require-and non safety-related compressed cir systems meet ments of General Design Criterion 2 regarding the protection against natural phenomena andWe theguidelines will reportofresolution Regulatoryof Guide this
- 1. 29, Positions C.1 and C.2. The compressed air- system item in a supplement to this SER.
doesnotmeettheapplicableacceptancecriteria'ofl{RFSection r
I 9.3.1.1 ..
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i- The ADS valve actuators are supplied with nitrogen (air) from the primary containment instrument gas system (see Section 9.3. 6 for details of nitrogen (air) supply to ADS valves). ._
As described in FSAR Section 9.3.1.3 except for the containment ._
isolation valves and penetration, whose location and classifica- ~
tion is shown in Table 3.2-1 (Item XVII.a.3) ' ~~~
the . . . _
service air system is not safety related7 M e~Eefore, General Design Criteria 2 and Regulatory Guide 1.29, positions C.1 and __
C.2 jo not apply.
i As described in Section 6.2.4.3.2.4, " Containment Isolation System the Compressed Service Air Line" and Table 3.2-1 (Item . _ .
XVII.a.3) the contaiment penetration is provided with redundant seismic Category I, Quality Group B isolation valves. t K53/4 Af22## l
b.'
As described in revised Section 9.3.1. the quality of air supplied to the instrument air system will be periodically tested to see that it meets the requirements of ANSI MCll.1-1976
- Quality Standard for Instrument Air".
fn ddt'/-fon, -
. [rne is.st. w e.nt aie a ystem a S4er Cil4ee is design ed +a e emo ve o . 4 in ie., a m eh-e pa.ehcles with a 91 p er cent e-fic.u en cy . rh e s y.s fem ia of ss og n ed' -fo permit preven hi de o r-e.o re e e H v s m a.: n + ,. n o. n e e on a n e. c.; e c6 y e.e a.n d o. C4.se G: l+a.e +ead a w ;%ew1- e.G eck a g sys+,.m op er a b Il hy . % e.eefo r e., g u.a.Ae e Iy i n a p e c.Mo n. o F & a Aerf,'/6ee assa.re s tha.t o y-s e. aic
+se inoxim u a p o. , A e t e . ,1 2. e. ;o
.s +e e a. - a + v.4 a ;os fc- an d i s 3.o a sii s med,es. .
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l HCGS FSAR 3- (m.
l Category I and ASME B&PV Code, Section III, Class 2, requirements l as defined in Sections 3.7 and 6.2.
i .
9.3.1.4 Tests and Insoections 1-The containment penetration pcetions of the compressed air systems are preoperationally tested in accordance wth the requirements of Chapter 14. The instrument air system is tested in accordance with Regulatory Guide 1.68.3, Preoperational
> Testing of Instrument Air Systems. Compressors and dryers shall be tested in accordance with ASME and manufacturers' test procedures.
.1*/USfM A ~~**
9.3.1.5 Instrumentation Acolication .
Instrumentation is provided for each instrument air and service air compressor train to monitor and automatically control each compressor's operation.
j l'
1 The compressors are tripped on the following signals: low oil
- I O pressure, high oil temperature, high cooling water discharge j temperature, high air pressure in the receiver, high outlet air temperature, and high vibration. Most of these signals are annunciated in the main control room by common trouble alarms.
High air temperature in the aftercooler and moisture separators, low pressure in the air receivers, and high intake filter differential pressure are also alarmed on a local control panel and the main centrol room by a common trouble alarm.
i Instrumentation is also provided locally for each instrument air dryer package train to monitor the packages operation.
Service air compressor and emergency instrument air compressor trouble are individually annunciated and alarmed on the local common service air compressor control panel. These alarms also ,
indicate on the main control room computer, along with the air l dryer trouble alarms.
9.3.2 PROCESS AND POST-ACCIDENT SAMPLING SYSTEMS The process sampling system (PSS) is designed to monitor and provide grab samples of both radioactive and nonradioactive fluids used in the normal operation of Hope Creek Generating Station (HCGS).
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j' eg. yai.}er as speciReal ir, ivua air dryer
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in desiy41 t o remase . og mice.-etre p.<Mies with 4 19"/. encIm3 . T he g stem is Q destfned to permit preven +iae or carrec4in
. maintensu.A en one dry << and a.fte.rf' Iter
'tain wshok aJ+ee+T'ny system opeedethy, ThercOre quartt/lg thsfeeb c4 1!::be af+erfifter assores wt se me.wimm ,
l p a/ +\eles. s o i.<. (n the qie s+ ream aI-Os in5Y./Jment i$ L C m iceo metres. T64 SAYnhSSS reg olre ment %L us hMSC M C II l-I425 ,
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BCGS DSER Open Item No.150 (DSER Section 9.3)
PRIMA 1E CONTAINMENT INSTIE! MENT GAS SYSTEM
' The applicant has committed to have the PCIG and instrument air systems meet the requirements of ANSI MC11.1-1976, using non-oil-lubricated air compressors as part of the preoperaticnal startup tests. The applicant has not committed to perform air quality tesfing in accordance with ANSI MCll.1-1976. ' On f ailure to meet acceptable air quality, branch lines are to be tested to determine the extent of problems and corrective action needed.
The safety-related portions of the PCIG and instrument air sys-tems are tested in accordance with the guidelines in RG 1.68.3, "Preoperational Testing of Instrument Air Systems" (refer to Sect' ion 14 cf this SER). ,
We cannot conclude that the design conforms to the guidelines of ANSI MC 11.1-1976. We will The report re' solution of this item in a ccepressed air system does not meet supplement to this SER.
the applicable acceptance criteria of SRP Section 9.3.1.
nESPONSE [
(~') Testing of the PC a quality can be performed in accordance by taking samples through the various with ANSI MC 11.1-19 vents, drains or test connections downstream ofw!'e the PCJGS a
receivers. f.MR seeHan 9 3.(,. / Ji a s b e.ess To' provode >-/se. Peyue.sted is fer& fos i
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preclude damage from missiles generated by the other
, compressing train.
- d. Protection against dynamic effects associated with pipe l l ruptures - Section 3.5.
i
- e. Environmental design considerations are discussed in i Section 3.11. ,
i Failure of a single component will not intGerupt the operation of
. the PCIGS because of the redundant trains provided with separate sources of electric power fed from independent Class 1E sources.
l l 9.3.6.4 Tests'and Insoections The PCIGS components are tested and inspected before leaving the
{1 supplier's shop to ensure that the system will meet the design criteria. The system is preoperationally tested in accordance with the requirements of Chapter 14.
! Operability of the system is demonstrated by actual use during normal operation.
INSERT W 9.3.6.5 Instrumentation Aeolications i
Instrumentation is provided for each train of the PCIGS to monitor and automatically control the system's operation.
Further information on the system control and logic is discussed in Section 7.3.
I The compressor is instrumented to shut down under the following conditions: .
l a. Low lubricating oil pressure l
- b. High lubricating oil temperature
- c. High discharge gas temperature l DSER oPEN ITEM /IO 9.3-51 Amendment 4
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.] P0ZGS dew painf will be les/ed in accordanc<
wiNr ANSI MC11.1 - 197.r, Qualily Slanalard y
for Ins /ramen}' Air, a} a freyuency of once j ,oer g.uar/er as speciAeol in Hre air dryer
. }echnical manual. -
r The. PCIGS cutlet filter rem e ses .3 fa.rtic.\es wi%
m'ucro m e+re. s. 98 %
e E c.'i e n c $ sThe system is ele. desiped O to ee<~w Pr~<~
- c' wru m ~'***
cn on < ccmfessee, dryer aed $i'skte t.re.in wih& o. Gat ~cg sy stem oper.ki 12+3, -
There fare jvarterh sspection of tbe fiMer o.ssu.re 5 that- tae. m a.m.i m a m '
i parfi cle. s h.4 in 6e o.ie s+.rea m -
o,t h e th strument is 3, o mec co m etres .
j This sa6sfies req vdemen + 8+.2 e4 A8sr Mcn.1-1975.
j ..
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h DSER OPEN ITEM /d
T HCGS C
- DSER Open Item No. 186 ( DSER Section 7.2.2.3) l TESTABILITY OF PLANT PROTECTION SYSTEM POWER We will require that the applicant demonstrate the capability of the design for on-line testing of each instrumentation channel, logic, actuation device and actuated equipment in the ECCS and BOP ESF systems. All actuated contacts and devices should be considered and those which cannot be tested on-line should be identified and justification provided.
RESPONSE
The response to Question 421.22 will provide the requested information concerning on-line testability (at the contact level). This information will be provided by July 1984. 1 O w l 186-1
BCGS FSAR 4/34 l i OUESTION 421.22 (SECTIONS 7.-3, 7. 3, 7.4, 7.5, 7. 6, a 7.7 ) I
?
C - The design of the instrumentation channels, logic and actuation devices of nuclear plant safety systems should include provisions for surveillance testing. Guidance is included in Reg. Guide 1.113 and IEEE Standard 338 for implementing the requirements of IEEE Standard 279, which requires in part that
-systems be designed to permit periodic testing during reactor l
- operation.
Section 3.1.2.3.2 and 7.2.2.3.2 includes a brief description of the at-power testing capability of the reactor protection system. Bowever, sufficient information has not been provided to determine the acceptability of the at-power testing capabilities provided in the Bope Creek design. Provide a detailed discussion i with illustrations from applicable drawings on the at-power ' testing capability of the eeactor trip system, engineered safety i. features actuation system .9d auxiliary supporting features, the i actuation instrumentation for the reactor core isolation cooling ,
- system, and the instrumentation and controls that function to prevene. accidents (i.e., high pressure / low pressure interlocks) or terminate transients (i.e., level 3 - turbine trip). This discussion should include the sensors, signal conditioning circuitry, voting logic, actuation devices and actuated components. Include in the discussion those design features that will initiate protection systems automatically, if required during testing, upon receipt of a valid initiation signal.
C l RSpom i As required by IEEE Standard 279, capability for at-power testing has been provided in the design of the MCGS safety systems. Conformance to the guidance specified in Regulatory Guide 1.118 1 and correspondingly, IIII Standard 338, is as stated in Section i 1.8.1.118. I The analysis portions of the various system descriptions in Chapter 7 for the safety-related systems referenced in the question describe the methods by which the safety system designs satisfy the testability requirements of IEEE Standard 279. The specific sections covering the testability of these systems are listed below: RPS - 7.2.1.2 ECCS - PPCI 7.3.1.1.1.1(c)
- ADS 7.2.1.1.1.2(c)
- CCRE SPRAY 7.3.1.1.1.3(c)
- RHR-LPCI 7.3.1.1.1.4(c)
PCRVICS 7.3.1.1.2(d) RHR-CSCM 7.3.1.1.3(c) RHR-SPCM 7.3.1.1.4(c) , i l b 421.22-1 , Amendment 5 osER OPEN ITEM /8h
o m opsu I m fgle BCGS FSAR 4/84 (, PCIS 7.3.1.1.5(j)
. .. CACS - Supp. Chamber to 7.3.1.1.6.1(c)
Drywell Press. Relief L L - R3 to Supp. Chambec 7.3.1.1.6.2(c) Press. Relief Sys. *
- BOAS 7.3.1.1.6.)(c)
- CNRS 7.3.1.1.6.4(d). .
NCRHIS 7.3.1.1.7(j) MSIVSS 7.3.1.1.S(c) FRvS 7.3.1.1.9 RSVIS 7.3.1.1.10(h) EAS - SSWS 7.3.1.1.11.1(c)
- SACS 7.3.1.1.11.2(c)
PCIGS . 7.3.1.1.1.11.4(c) CACWS 7.3.1.1.1.11.5(c)
. EACS - RBEAC 7.3.1.1.11.6.1(c)
- ABDA 7.3.1.1.11.6.2(c)
- ABCA .
7.3.1.1.11.6.3(c) .i
- SW7.E 7.3.1.1.11.6.4(c)
RCIC 7.4.1.1.3 i SI.C 7.4.1.2.3 RRCS 7.6.2.7.2(b) 7.6.2.7.2(n) 7.6.2.7.4.1
. - Design drawings in the form.of elementary diagrams, P& ids, logic b diagrams, instrument location drawings, and electrical drawings that describe this capability are listed in Tables 1.7-1, 1.7-2, i and 1.7-3.
f I In response to the NRC's request for additional information
- during the meeting of January 11, 1984, review of the systems identified above, with the exception of the reactor protection system (RPS), reactor core isolation cooling (RCIC) system, standby liquid control (SLC) stem, and redundant reactivity
! control system (RRCS) rill performed. The review t.4 W " esse.edrwM
; 'rterin: the capability for the at-power testing of all circuits and sengeJs used in these syster.s. All actuated contacts and devices g. : 5: considered. FAny ster., suosyst T., or ompo 1 n.. cas e capa Atty e at-p er te *ing wi be i ntif da stifi tion w 1 be evided. The ults 11 be C men in a evisio to th resee _ e te sub-i ed by J u 1984 J v n , ,,,, , J.-J ,,e4 ,4 &. s,., .h >
4,.se s shs (a.ikd apsky h- e6ea e,w/-@k tesd %qw I?n,a,,,J..Js,d .neede 4., asF [ a4.ge,p kg.s..n.a w , w n., w , w , - J.eAs, .La r wiA c..inal44 f ph... As4 a a .r. % y 4 ,.w.a. w. % ( ,.u.g u .,na 9 6 s ,w. e , m .L A y
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- INSERT A - ,
n _ L D During the review, the at-power testability of an item was established if an affimative response could be verified for//bie following
-- three questions: .
- a. Is the item sufficilently accessible to conduct the test during normal operation? - .
- b. Is the item sufficiently isolatable to permit its safety-related function to be varified or is a safety-related system or subsystem encompassing the item isolatable and testable?
- c. Does any bypassing method that must be used to accomplish the test conform to position C6 of Regulatory Guide 1.1187*
F-g g5SS t SW y these criteria)two items were judged to be untestable at power, the ADS SRys, which would cause depressurization if tested, and the steam-tunnel temperature elements, which are inaccessible. The reliability and redundancy of the ADS instrumentation, logic, and actuation devices. and the multiplicity of the SRVs adequately justify the lack of ADS at-power testability. Adequate element multiplicity and comparison tests of at-power output signals and electrical characteristics preclude the need for change-of-state testability of the steam-tunnel temperature elements. O i I
+
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c4% -& logi c. cwuitry of +L 'TtIs
'rsd o a u+ in k 6i+ & -berkap ococs o? & ac.h t ae % -kom 6ga*(s Som
% Tcn 6 fb 'idivio6A acfnatrcf
/1 com weets. O l l 1 i DSER CPEN I m /g4 s
Jun :p as y g; 0 0 00 4
~~
3 RCGS DSER Open Ites No .' 187 (DSER he ction 7.2.2.4) LIFTING OF LEADS TO PERFORM SURVEILLANCE TESTIN To provide conformance with the recommendation , , plemented to provide the capability to perform surveillanceFurther, testing without lif ting leads. i ent that would the applicant identify and justify the. equ pmThe justification should addre not be tested at power. i capabilityThe of the design opening ofto circuitsatisfy breakers criterion 21 of 10monthly t9 perform CFR, Part surveillance 50. testing should also be discussed within the context of Regulatory Guide 1.22. RESPONSE . see the responses to For the information requested above, Questions 421.4 and 421.22. 4 C-4 . l O e 9 e e O O 1 . i - 187"1 F67(1)
. ~ . . .. . - .
.\
l RCGS FSAR 4/84 (' .
~
OUESTION 421.4 (SECTION 7.1) (' l I FSAR Section 7.1.2.4 provides a discussion'of design conformance , to Regulatory Guide 1.118, Periodic Testing of Electrical Power
, and Protection Systems, June 1976 as an endorsement of
- IEEE-338-1977 and provides clarifications to two positions of
, this RG. The version of R.G. 1.118 cited is incorrect, as is the two positions discussed. To comply the staff review and the ensuing evaluation, the discussion of the justification for
, deviation from R.G. 1.118 will have to be corrected by referencing the 1978 version and by providing a clarification of the design deviations from the RG positions. It should be noted . i that the use of jury-rigged bypasses such as temporary jumpers, the removal of fuses, or removal of connectors is not an i acceptable method for standard in-service testing. i Rzseonst
! Although the June 1978 revision of Regulatory Gu'ide 1.118 is not
, part of the design basis of the HCGS, er :::rre--t c' " l O cc-'r-----= h-- br:r n u r ;_i Jsection 7.i.a.4.g and Table 7.1-2] ! r J ' (naveseenrevtsedtoreflectthespecifieddateandtoclarify the conformance situation y % ! arn con 3 unction wun tne revi- ee eeibee in ene resoonse em 3 j O( - wouestien 421.22,fa review
'nypasses such as temporary jumper
.a. L""b 1,erformed the removal to determine of fuses, if or the l removal of_ connectors may be necessary for HCGS inservice testine J Instan wnere su oypasses ma nave to e used will roe cocu. nted in a vision to his response, and discu ens will be p vided to tify the e of these b asses. T e stificati will be sed on th exceptions a horized by po tion C14 o Regulato Guide 1. :
. Tempora umper wi may be ed with port le test
- eq ment where e safety stem equ ment to be t ted is i prov ed with fac ties spec ically igned for connec on of this t t equipme . Thes acilities s 11 consi red part of safety stem an hall meet a I t c'equir ents of this andard, ether t portable t t equ ment is isconnected a remains nected these facil tes.
f "b. Remo 1 of fu s or opening breaker i ermit d only { such act n cause (1) the trip f the asso ted I pro ction sy em cha el or (2) the ctuation ( etup d
- opera on) of t assoc ted Class 1E oad group."
l <The evised re nse will e subm ted by July, 1984 56 nveed anes L& and St resslh an. sneluded k & rufen.%
^ " " " '
csta o m z m /87
.u -. .- -.. - -
. EL -6, *84 0 2 6 7 3 61
~" ~
HCGS FSAR , *4/84 , (_.
- HCGS; however, equipment is qualified following the
'. guidelines of IEEE 323-1971 as discussed in Section 3.11.2. Also refer to Section 3.1'1 for discussion of the environmental qualification program.
; o. Assessment to Regulatory Guide 1.100, Seismic i Qualifications of Electrical Equipment for Nuclear ,
Power Plants, March 1976 - While not a design basis, the extent of conformance to Regulatory Guide 1.100 is discussed in Section 3.10.
- p. Assessment to Regulatory Guide 1.105, Instrument Setpoints, November 1975 - While not a design basis, the design supplied includes the trip setpoint (instrument setpoint), allowable value (Technical Specification limit), and the analytical or design basis limit, which are all contained in Chapter 16,
- Technical Specifications. These parameters are all appropriately separated from each other based on instrument accuracy, calibration capabilility, and design drift (estimated) allowance data. The setpoints
! are within the instrument accuracy range. I The established setpoints provide margin to satisfy both safety requirements and plant availability objectives. I. l Lusessment to Regulatory Guide 1.118, Periodic Testing
- q.
- i sf Electrical Power and Protection Systems, June 1978 -
This regulatory guide, which endorses modified IEEE 338-1977, is not part of the design basis for i HCGS. Discussion of IEEE 338 is presented on a system-
- by-system basis in the analysis portions of Sections 7.2, 7.3, 7.4, and 7.6.pttn e foil ing n L.o: the r . oval o fuses vaara cata n or osat an ' r be kers o pr ent t ope tion o equip nt d ing e pe orma e of sts e ld be ther ed
-Jd- nderA tric admi strati con ols an appro .d a-,
no t- o _ stoc(duresuf J12.+ : 0
- l' - .4w6W 7 ._? r y g y }f I'L1L 7.1.2.5 Indeoendence of Safety-Related Systems The safety-related 1&C required to provide protective actions are physically arranged and separated to retain the minimum required i -
Amendment 5 I em om I:n /S7 l
' -~ . . . 2. -. .. . . .
t HCGS FSAR 4/34 OUESTION 421.22 (SECTIONS 7.-2, 7. 3, 7. 4, 7. 5, 7. h, a 7.7 ) n, t" The design of the instrumentation channels, logic and actuation devices of nuclear plant safety systems should include provisions for surveillance testing. Guidance is included in Reg. Guide 1.113 and IEEE Standard 338 for implementing the requirements of IEEE Standard 279, whicts requires in part that systems be designed to permit periodic test.ing during reactor t operation. Section 3.1.2.3.2 and 7.2.2.3.2 includes a brief description of the at~ power testing capability of the reactor protection system. i However, sufficient information has not been provided to ,' determine the acceptability of the at-power testing capabilities provided in the Hope Creek design. Provide a detailed discussion with illustrations from applicable drawings on the at-power + . i testing capability of the reactor trip system, engineered safety features actuation system and auxiliary supporting features, the actuation instrumentation for the reactor core isolation cooling
- system, and the instrumentation and controls that function to i
prevent accidents (i.e., high pressure / low pressure interlocks) 4 or terminate transients (i.e., level 8 - turbine trip). This discussion should include the sensors, signal conditioning circuitry, voting logic, actuation devices and actuated components. Include in the discussion those design features that will initiate protection systems automatically, if required during testing, upon receipt of a valid initiation signal. i, (,.) i
RESPONSE
l As required by IIII Standard 279, capability for at-power testing has been provided in the design of the MCGS safety syster.s. l Conformance to the guidance specified in Regulatory Guide 1.118 and correspondingly, IEEE Standard 338, is as stated in Section 1.8.1.118. The analysis portions of the various system descriptions in Chapter 7 for the safety-related systems referenced in the question describe the methods by which the safety system designs satisfy the testability requirements of IEEE Standard 279. The specific sections covering the testability of these systems are listed oelows e RPS - 7.2.1.2 ECCS - EPCI 7.3.1.1.1.1(c)
- ADS 7.2.1.1.1.2(c)
- CORE SPRAY 7.3.1.1.1.3(c)
- RHR-LPCI 7.3.1.1.1.4(c)
PCRVICS , 7.3.1.1.2(d) RHR-CSCM 7.3.1.1.3(c) RHR-SPCM . 7.3.1.1.4(c) .
' 421.22-1 Amendment 5
. eszR OPEN M M M ,
l
.nssa ornu ITzu /87
~
PCIS 7.3.1.1.5(j)
..- CACS - Supp. Chamber to 7.3.1.1.6.1(c) -
Drywell Press. Relief
- R3 to Supp. Chamber 7.3.1.1.6.2(c)
Press. Relief Sys. .
- 50AS 7.3.1.1.6.)(c) 7.3.1.1.6.4(1), 8
- CNRS NCRHIS 7.3.1.1.7(j)
MSIVSS 7.3.1.1.5(c) FRVS 7.3.1.1.9 RSVIS 7.3.1.1.10(h) EAS - SSWS 7.3.1.1.11.1(c)
- SACS -
7.3.1.1.11.2(c) l PCICS . 7.3.1.1.1.11.4(c) CACWS 7.3.1.1.1.11.5(c) i
. EACS - RBEAC 7.3.1.1.11.6.1(c)
- ABDA 7.3.1.1.11.6.2(c)
- ABCA .
7.3.1.1.11.6.3(c)
- SWIS 7.3.1.1.11.6.4(c)
~
RCIC 7.4.1.1.3 SLC 7.4.1.2.3 RRCS 7.6.2.7.2(b) . 7.6.2.7.2(n)
. 7.6.2.7.4.1 Design drawings in the form of elementary diagrams, P& ids, logic V. diagrams, instrument location drawings, and electrical drawings that describe this capability are listed in Tables 1.7-1, 1.7-2, and 1.7-3.
> In response to the NRC's request for additional information during the meeting of January 1,1, 1984, review of the systems l identified above, with the exception of the reactor protection l system (RPS), reactor core isolation cooling (RCIC) system, standby liquid control (SLC) stem, and redundant reactivity control system (RRCS) rill E erformed. The review M c- LJ
; 'rter-in: the capability for T.he at-power testing of all circuits and sensegs used in these systems. All actuated contacts and devices %.11 5: considered. fAny s t er. , suosyst t, or ompo t1 l
anos cas e capa Atty e at-p er te *ing wi be i ntif j nd a stifi tion w 1 be vided. The ults 11 he C
$ umen in a evisie to th resee e to suEri ed by J l
- IJul 1984.J
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t
- INSERT A -
m , W During the review, the at-power testability of an item was established if an affirmative response could be verified foybw following -
-* three questions: .
- a. Is the item suffici6ntly accessible to conduct the test during normal operation? - .
,~.
. b. Is the item sufficiently isolatable to permit its safety-related function to be verified or is a safety-related system or subsystem encompassing the item isolatable and testable?
[ c. Does any bypassing method that must be used to accomplish the test conform to position C6 of Regulatory Guide 1.118? f g 4S55 i g6 ty$5M y theseADS SRVs. criteria)two itemscause which would weredepressurization judged to be untestable if tested, andat thepower, the I steam-tunnel temperature elements, which are inaccessible. The ~ reliability and redundancy of the ADS instrumentation, logic, and actuation devices and the multiplicity of the SRVs adequately justify the lack of ADS at-power testability. Adequate element multiplicity and comparison tests of at-power output signals and electrical characteristics preclude the need for change-of-state testability of the steam-tunnel temperature elements. O h , Qp 0
, om om 1m /87
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e m O . E 1)SER oPEN ITEM 20/ ( 3ec.}id n 7. Y A 3 ) Rcic/MPC/ /Nrenenoxs emw , The appli cant' i.s eeg uir ed +* conArm +Aa d Hi*c t opera-Has i .s ini4 is.ded' by a.4ocs .s y n a / - or m a n ws. ( a c.+, a n .a n s' sed by a R cd c. /ou.) ' flow sibna./ Re.s oc o s e. f'or the inhrmAbtbn y eIg.tesfe d c jesie, 3 e.g., b jy g respen.gc - L' +o guu+;en fat. 3 7 1 4 1
NCGS , ("% DSER Open Item No. 202 (DSER Section 7.5.2.1) f: LEVEL MEASUREMENT ERRORS AS A RESULT OF ENVIRONMENTAL TEMPERATURE EFFECTS ON LEVEL INSTRUMENTATION REFERENCE LEG The applicant is required to submit the results of this evaluation for staf f review and to implement any hardware and/or procedural changes that may evolve as a result of this evaluation.
RESPONSE
The response to MQuestion 421.21 vill be revised-by August 1984 to identify any necessary design changes to HCGS vater level monitor-ing instrumentation. th e Vor the. : n fermohon t-egyau/es/e) Al a Jg ue , .r e:.e pe.spanse k Qu.eshbo 1 4 l r j 202-1
- - . - , - - --.--.,,--,,e- , ---,>----------,,,---,-,--~e-,- , , - - ----m,--- - - , , . . -- - - - v----w---.,--xw-m--,,,-
,-s
\, 4/34
- NCGS FSAR-OUESTION 421.21 (SECTIONS 7.2,.7.3, 7.(, 7.5)
Provide an evaluation of the effects of high temperatures on reference legs of water level measuring instruments subsequent to high-energy line breaks, including the potential for reference leg flashing / boil off, the indication / annunciation available to alert the control room operator of erroneously high vessel l'evel indications resulting from high temperatures, and the effects on safety systems acuation (e.g., delays).
RESPONSE
evalua on of thi issue is progress. Based on th results oposed mod cations, f any, t;o t NCGS of this alysis, level nitoring strumenta on design ill be prov ed to t This i estimat to be about ugust, 1 4. NRC w n availab e. i I~as62r ( . t l
. l t'
421.21-1 Amendment 5 osaopa g;g
. .. ... .. .. .... . .. . e . ........ .
MCGS FSAR EST[oM421.2g(SECTIONS .2, 7.3, 7.4, 7.5) ) of high peratures on Pr ide an valuation the effe ments sub quent r erence ogs of we r level me uring in refer c high nergy lin reaks, in uding the otential i og fl hing/boi: f, the in e ion /an nciation a 11able the contr/. rocat oper of er accusly hi vessel evel aler indi ations resulting from gh tempo atures, an the off tsjon) fety systems actuation (e.g., delays). e -, o
- wasr A- .
An evaluation of the of facts of high temperatures on retirence legs of water level measuring instruments subsequent to High Energy Line Breaks (RELB) is divided into two parts: 1) the ef fects of temperature alone, and 2) the effects of flashing / boiloff. Mich Temoerature Ef fects (without flashino/bollof f)_ An increase in the temperature of the drywell will cause a heat-up of the fluid in the instrument sensing lines, contributing to sensor error. The ECGS instrument sensing line design reduces this error by routing the variable leg and the referenceThe legonly b lines with equivalent elevation drops in the drywell. exceptions to this design are the Upset Range transmitters reference leg sensing lines. Physical configuration prevents equivalent routing of these lines. However, these transmitters are used exclusively for indication and will not present any challenges to plant saf ety. A high drywell temperature alarm is coreputer generated frau isolated outputs of class 1E temperature transmitters. Class it temperature recorders located in the main control room provide a continuous display of drywell temperature. riashinc/soiloff Effects The ef fect of flashing /boilof f of the instrument line reference leg is to cause the level instruments to indicate erroneously high levels. The amount of error is directly related to the drop , in elevation of piping physically located within the dryvell and subject to flashing . MCGS has rerouted two channels of reactor pressure vessel (RPV) l 1evel instrumentation sensing lines to provide a maximum 3-it elevation drop in the drywell (maximum 1-f t drop for the reference i legs). A worst case analysis of the effects of boiloff of g that portion of the sensing line inside the dryvell, indicates I 421.21-1 Amendment [ f f osza oPrs Irzx g c1
*
- ECGS FSAR the instruments using the reVouted lines will indicate a level that is 1.3 ft higher than actual. Durig and after an HEL5 the operator is required to maintain RPV level within the normal operating range, is it above the top of active fuel. The 1.3 ft
- error is negligible with respect to the operating requirements.
Transmitters used for post accident monitoring use the rerouted lines. Threfore, the wide,. narrow, and fuel zone range recordere and indicators will provide an unambiguous display of level even after partial flashing of the reference legs. i As a result of an SELE in containment, the drywell temperature may reach a maximum of 340'F. Flashing /boiloff of the sensing lines may occur when the RFV pressure is less than 118 PSIA when the drywell temperature is 3A0'F. At the 118 PSIA RPV pressure the high pressure coolant injection system (MFCI) and the automatic depressurization system ( ADS) are not required. In response to a utta of a large or inteemediate sized line (see figure 15.9-43) low pressure coolant injection (LPCI) and core spray are initiated by low water Level 1 (L1) or high drywell j pressure signals. For these postulated events, MFCI and ADS are
' not required.
/' Two different response paths must be considered for a small break accident (SSA). ,
The first response path considers an SEA with RPCI available. The emergency core cooling system (ECCS) response to an SEA is outlined in TSAA chapter 15 in response to event 42 (rigure 15.9-43). Core spray and LPCI are initiated by high drywell pressure. HPCI is initiated on receipt of a low Level 2 or high drywell pressure signal. HPCI continues to operate until the enactor vessel pres-sure is below the pressure at whien LPCI or core spray operation l ' can maintain core cooling. LPCI and core spray are designed to begin injecting water into the RPV when the dif ferential pressure between the RPV and the suppression chamber is approximately 300 psid per design requirements (see TSAR Chapter 6.3). l The second response path considers a HFCI line $$A that incapacitates HPCI. Accident mitigation requires the actuation of the automatic depressurization system (ADS), LPCI, and core spray. LFCI and core spray are initiated on high drywell pressure or a L1 signal, i ADS in initiated by a L1 and high drywell pressure and a L3 per-missive signal when low pressure ICCS pumps are running. At the point flashing could occur, the RFV pressure will be low enough that ads will not be required before that point level signals /actuations will remain accurate. ' 421.21-2 Amendment EJ '" esan optw trzx ,20 4 k
.c 1
' a .' , ,
MCGS FSAR ( .. In the event of any credible NELB inside containment, the capa-bility of the BCCS to mitigate the accident is not compromised by
* - high drywell temperature or flashing of the RPV level instrumen-tation line reference legs.
O
% e, N '
I i SS 3 I 421.21-3 mendment[ ossa cern trax JoA
.. .? . . . . . . . _ .< .
BCGS DSER Open Item No. 203 ( DSER Se ction 7.5.2.2) REGULATORY GUIDE.1.97 The staff is presently reviewing the BWRod h'sition and the HCGS specific deviations which will be shown on revised FSAR Table 7.5-1. In addition, the applicant is required to resolve the inconsistency be tween FSAR Se ction 1.8.1.9.7 Ites C, and Footnote 11 to Table 7.5-1.
RESPONSE
The response to Question 421.41 provides the HCGS specific implementation (via FSAR Table 7.5-1) of Regulatory Guide 1.97. Jeedion f. g.I 17 haa 6 sen psahsed 4e removs 1the in een s it d ene-y hs.+sseen sier C of Se shan J .R. /. 9 7 d
- d fo dna't / / +a Ab /t 7..f- /,
p 1 e g h I b 203-1
,. .+.,-.,.cw, + ,- w -g ,--.- .,, , a, -p.es aw:mw---ee--r-- -9 e ,w-eemr u.-
em ---e. 3 -+s**vv==n= +--twt -*
[ . HCGS FSAR 8/84' n ks) c. 'i.8.1.97 Conformance to Reculatory Guide l'.97, Revis?.on 2,
.\.-
3ecember 1980: :nstrumentation for Licht-Water-Cooled fuelear Power P;, ants to Assess Plant and Environs [ Conditions Durino and Followino an Accident l 1.8.1.97.1 General Position Statement HCGS concurs with the intent of Regulatory Guide 1.97, Revision
- 2. The intent of the regulatory guide is to ensure that necessary and sufficient instrumentation exists at each nuclear power station for assessing plant and environmental conditions during and following an accident, as required by 10 CFR Part 50, Appendix A and General Design Criteria 13, 19, and 64.
Regulatory Guide 1.97 requirements are being implemented except in those instances in which differences from the letter of the guide are justified technically and when they can be implemented wit.hout disrupting the general intent of the regulatory guide, or other applicable design criteria. In assessing Regulatory Guide 1.97, HCGS has drawn upon information contained in several applicable documents, such as the BWROG Position (Reference 1.8-4) on Regulatory Guide 1.97, O(- ANS 4.5, NUREG/CR-2100, and the BWROG Emergency Procedures Guidelines, and on data derived from other analyses and studies. HCGS has attempted to meet the intent of, as opposed to the literal compliance with the provisions of the regulatory gu'ide, because of their specific nature. In general, HCGS intends to follow the criteria used by the NRC for establishing Category 1, 2, and 3 instruments. Where differences between the Regulatory Guide Categories exist, justification for the category chosen is provided. This approach is preferable as some Regulatory Guide 1.97 requirements call for excessive ranges or categories or both, others call for functions already available, and still others could adversely affect operater judgment under certain conditions. For example, research by S. Levy, Inc. , (SLI), show that core thermocouples vill provide conflicting information to BWR operators. HCGS intends to follow the criteria used by the NRC for establishing Category 1, 2, and 3 instruments. The following HCGS compliance statement is applicable to the regulatory positions defined in Regulatory Guide 1.97, Revision 2 (the paragraph numbers cited correspond to those in Regulatory , Guide 1.97). ( v DSER OPEN I m g 3 1.8-60 Amendment 7
.a. D
. a. ,
. ,-- =. .. .
HCGS FSAR 8/84 7 o i - Accident-Monitoring Instrumentation !i
!m (- a. l
! Par. 1.1: HCGS concurs with this definition. l
- . Par". 1.2: HCGS concurs with this definition. l Par. 1.3: Instruments used for accident monitoring to meet the provisions of Regulatory Guide ~1.97 will have the proper sensitivity, range, transient response, and accuracy to ensure that both during and.following a design basis accident the control room operator is able to perform his role in bringing the plant to, and maintaining it in, a safe shutdown condition and in assessing actual or possible releases of radioactive
- ' material.
Accident-monitoring instruments that are required to be environmentally qualified will be qualified as described in Section 3.11. The seismic qualification of instruments is described in Sectior4 3.10. Ok ' The HCGS quality assurance program ensures that accident-monitoring instruments comply with the applicable requirements of Title 10 CFR 50, Appendix B. Table 3.2-1 identifies where these requirements have
- been applied.
The HCGS program for periodic checking, testing, calibrating, and calibration verification of accident-monitoring instrument channels (Regulatory Guide 1.118) is identified in Chapter 16, " Technical Specifications." l i Par. 1.3.1: A third channel of instrumentation for ; l Category 1 instruments will be provided only if: I
- 1. a failure of one accident-monitoring channel results in information ambiguity that would lead i
' operators to . defeat or fail to accomplish a required safety function, and ( I.I"0I AEendmenD 7 DSER OPEN ITEM 476 3 1
,,. . y _,~,, , . . . _ _ _ . _ . . _m . -,.--,m .~.e._ _ . . _ , - . _ _ - __________4_ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _
.* -=
T HCGS FSAR 8/84 . le
. (. -
c if one of the following measures cannot provide E 2. {.- the'information: h (a) Cross-checking with an independent channel that monitors a different variable bearing a known relationship to the variable being monitored. (b) Providing the operator with the capability of perturbing the measured variable to determine which channel has failed by observing the response on each instrument. (c) Using portable instrumentation for validation. Category 1 instrument channels, which are designated as being part of a Class IE system, will meet the more stringent
- design requirements of either the system or the. regulatory guide.
The requirements for physical independence of electrical systems (Regulatory Guide 1.75) are
.~ identified in Section 1.8.1.75.
Par. 1.3.2: HCGS concurs with the regulatory position for Category 2 instrumentation, except as modified by Par. 1.3 above. i 4 l Par. 1.3.3: HCGS concurs with the regulatory position ' for Category 3 instrumentation. ! Par. 1.4 Instruments designated as Categories 1 and 2 for variable types A, B, and C should be identified in such a manner as to optimize the human factors engineering and presentation of information to the control room operator. This position is taken to j clarify the intent of Regulatory Guide 1.97, which l l specified that these instruments be easily discerned for use during accident conditions (see Issue 1 Section 1.8.1.97.4) l DSER OPEN ITEM Amendment 7 1.8-62 ,
,, -- . , . - . _ , - , , , . . - - , . . , - -,-e.---,--,-.. -- . . , . . . ~ . _ _ - , . . - ~ . - - , , - - . . . - - . . - - - , - - - . .. -. . . - . . .
l . s, HCGS.FSAR , 8/8'4 HCGS concurs with the' regulatory position ,. (" Par. 1.5: taken in this section, except as modified by Par. 1.3 above. I 1 Par. 1.6: It is the position of HCGS that in terms of , accident monitoring at HCGS, Table 1 of Regulatory Guide 1.97 is not representative of the optimum SPT of variables required and does not necessarily represent correct variable ranges or instrumentation categories. HCGS accident monitoring variables are identified in Table 7.5-1. The classification of instrumentation used to measure the variables as Category 1, 2, or 3 is in compliance with the intent and method used in Regulatory Guide 1.97. However, differences between the Regulatory Guide Categories and*HCGS categories for each variable described in Table 1 of Regulatory Guide 1.97 is described in Section 1.8.1.97.3. The HCGS position on the implementation of each variable described in Table 1 of Regulatory Guide 1.97
- - is presented in Section 1.8.1.97.3.
- b. Systems Operation Monitoring and Effluent Release Monitoring Instrumentation The HCGS position stated in Par. 1.3 above is applicable to the Type D and E variables described in Regulatory Guide 1.97, Par. 2.1: HCGS concurs with these definitions. l Par. 2.2: HCGS concurs with this regulatory position. l Par. 2.3: HCGS concurs with this regulatory position l Par. 2.4: HCGS concurs with this regulatory position. [
Par. 2.5: The HCGS position as stated in Par. 1.6 above is applicable to this regulatory position.
~
osza opzz Irzu 403 1.8-63 Amendment 7
.~.
. . . - . . . . . ~ , .
~
i l HCGS FSAR 8/84 s { - 1.8.1. 97.2 Proposed Type A Variables ~ l 4 Regulatory Guide 1.97, Revision 2, designates all Type A l . variables as Category I plant-specific, thereby defining none in l particular. The regulatory guide defines Type A variables as: o Those variables to be monitored that provide primary ! information required to permit the control room operator to take specific manually controlled actions i for which no automatic control is provided and that are l required for safety systems to accomplish their safety
; functions for design basis accident events.
I Regulatory Guide 1.97 defines primary information as "information that is essential for the direct accomplishment of the specified , safety funtions." Variables associated with contingency actions that may be identified in written procedures are excluded from this definition of primary information. As part of their review of Regulatory Guide 1.97, the BWROG 7 undertook the task of developing and analyzing a group of 3 variables that were determined to be potential candidates for
- > - inclusion in Regulatory Guide 1.97 as specific Type A variables.
HCGS has reviewed the generic BWROG identified variable and determined that the monitoring of the following noted safety functions for the listed operator actions are required to meet the intent of Regulatory Guide 1.97. The specific Type A variables are identified in Section 1.8.1.97.3.1: Variable A1. Oxycen or Hydrocen Concentration Safety Function: Maintain containment integrity by . controlling oxygen for inerted and hydroger. for r.on-inerted contaminants. , l ( Operator action: If containment atmosphere approaches the combustible limits, initiate combustible gas control systems. . Variable A2. RPV Pressure ( ( DSER OPEN ITEM 803 1.8-64 Amendment 7 n ----m. n,- , - _ - - ..n-.,..,n.,. ,__-,+g . - - - . , . . - ,,nne ,m_.,.. ,,e , ,,,,.n,.n,-g---r
.. g . _ -
w i HCGS FSAR 8/84 k- ~ {. Safety Function: -(1) Core cooling; (2) maintain reactor coolsnt system integrity. Operator action: (1) Depressurize RPV and maintain safe cooldown rate by any of several systems, such as
. main turbine bypass valves, HPCI, RCIC, and RWCU: (2) manually open one SRV to reduce press'ure to below SRV setpoint if an SRV is cycling.
Variable A3. RPV Water Level . t Safety Function: Core cooling. l Operator action: Restore and maintain RPV water level. l Variable A4. Sucoression Pool Water Temperature Safety Function: (1) Maintain containment integrity
, and (2) maintain reactor coolant system integrity.
Operator action: (1) Operate available suppression pool cooling system when pool temperature exceeds normal operating limits; (2) scram reactor if temperature reaches limit for scram; (3) if suppression pool temperature cannot be maintained below the heat capacity temperature limit, maintain RPV pressure below the corresponding limit; and (4) close any stuck-open relief valve. Variable A5. Suppression Poc1 Water Level Safety Function: Maintain containment integrity. l i I Operator action: Maintain suppression pool water level within normal operating limits: (1) transfer RCIC suction from the condensate storage tank (CST) to the suppression pool in the event of high suppression-pool level; and (2) if suppression pool water levc1 cannot be maintained below the suppression pool load limit, - maintain RPV pressure below corresponding limit. (
.C 1.8-65 Amendment 7 DSER OPEN ITEM g 6 3
--- , , - . . - - - - , , , .- --.--.-,.-,,,,-.--..m,-.,.n.e, mm,-,.,,- ,w, , , , , , , ,,,,__,-m, ympa , -,,-,,v.,-,-,---
- . . . ... . -, . . - . .-. - . . . .w . - . . . -
. HCGS FSAR 8)84 l ( ~
~(-
~ '
l Variable A6. Drywell Pressure Safety Function: (1) maintain containment integrity - and (2) maintain reactor coolant system integrity.
. Operator action: Control primary containment pressure by any of several systems, such as containment I atmosphere control systems, suppression pool sprays, drywell sprays, etc.
1.8.1.97.3 ~ Plant variables For Accident Monitoring l . In brief, the measurement of the following five variable types provides the noted required information to plant operators during and after an accident: (1) Type A--primary information, on the basis of which operators take planned specified manually controlled actions; ( 2') Type B--information about the accomplishment of plant safety functions; (3) Type C--information i about the breaching of barriers to fission product release; (4) Type D--information about the operation of individual safety
. systems; and (5) Type E--information about the magnitude of the release of radioactive materials.
({} The three categories (1,2,3) of required variables define the design and qualification criteria for the instrumentation that is to be used for their measurement. Category 1 imposes the most , stringent requirements; Categories 2 and 3 impose progressively j less stringent requirements. The categories are also related (per Regulatory Guide 1.97) to
" key variables." Key variables are defined differently for the different variable types. For Type B and Type C variables, the key variables are those variables that most directly indicate the accomplishment of a safety function; instrumentation for these i
key variables is designated Category 1. Key variables that are Type D variables are defined as those variables that most directly indicate the operation of a safety function; instrumentation for these key variables is usually Category 2. And key variables that are Type E variables are defined as those variables that most directly indicate the release of radioactive material; instrumentation for these key variables is also usually Category 2. Backup variables for Type B, C, D and E variables are generally Category 3. A complete discussion of the variable l types and instrumentation design criteria is presented in I Regulatory Guide 1.97. 1.8-66 Amendment 7 DSER oPmt ITD( MO3
**t S $
7 , . , . - , . _ ,,m
HCGS FSAR
- 8/84 ,
m
'<~ .
' - HCGS positions on the implementation of the variables listed in
- (. Table 1 of Regulatory Guide 1.97 and on the assignment of design .
l t- and qualification criteria for the instrumentation proposed for I their measurement is summarized in the tabulation that follows. k The variables are listed here in the same sequence used in Table 1, Regulatory Guide 1.97; however, for convenience in cross-referencing entries and supporting data, the variables are designated by letter and number. For example, the sixth B-type variable listed in Regulatory Guide 1.97 is denoted here as
- . variable B6. _
The HCGS variable category designated ("HC") and the Regulatory Guide 1.97 category designated ("RG") are shown for each variable and for its instrumentation design criteria and category. In general, there are three positions cited by HCGS: (A) the variable and required instrumentation was implemented in
- accordance with the regulatory position stated in Table 1, Regulatory Guide 1.97 (B) was implemented with qualifying l exceptions or revisions,; and (C) was not implemented.
As necessary, the HCGS positions are justified or substantiated
~
by the 11 " Issues" (identified in the tabulation of variables ({) {T where applicable) noted in Section 1.8.1.97.4. 1.8.1.97.3.1 Type A variables (Reference Section 1.8.1.97.2) ~ l \ l A1. H, or 0, concentration (HC Category 1, RG Category 1) Position: Monitor is plant-unique for H, or 0,. Needed for initiation of combustible gas controls. Implemented in accordance with NUREG-0737. See C11 and C12. A2. Reactor pressure (HC Category 1, RG Category 1) Position: Implemented. A3. Coolant level in reactor (HC Category 1, RG Category 1) Position: Implemented. See B4. A4. Suppression pool water temperature (HC Category 1, RG Category 1) Position: Implemented. See D6. 1 1.8-67 Amendment 7 DSER OPEN ITEM .20 3
..:.e -
HCGS FSAC 8/8'4 A5. Suppression pool water level (HC Category 1, RG (.. Category 1) Position: Implemented. See C7 and D5. A6. Drywell pressure (HC Category 1, RG Category 1) Position: Implemented. See B7, B9, C8, C.10, and D4. 1.8.1.97.3.2 Type B Variables l I
- a. Reactivity Control l Bl. Neutron Flux (HC Category 2; RG Category 1)
Position: Implemented, as Category 2 in accordance with data in Issue 2, Section 1.8.1.97.4.2. , B2. Control Rod Position (HC Category 3; RG Category 3) l Position: Implemented. , l B3. RCS Soluble Boron Concentration (sample) () (HC Category 3; RG Category 3) Position: Implemented.
- b. Core Cooling l B4. Coolant Level in Reactor (HC Category 1; RG Category 1)
Position: Implemented. See A3. B5. BWR Core Therr.occuples (RG Category 1) Position: Not implemented. See B4, C3, and SLI-8121 (December, 1981) (Appendix A to Reference 1.8-4). l
- c. Maintaining Reactor Coolant System Integrity l 1
B6. RCS 2ressure (HC Category 1; RG Category 1) l Position: Implemented. See A2, C4, C9, and Issue ! 3, Section 1.8.1.97.4.3. ( i I*0~60 Amendment 7 r DSER OPEN ITEM l l l l
~
.- . - . -- .c 1
HCGS FSAR 8/84 ,a k-L* B7. Drywell Pressure (HC' Category 1; RG Category 1) ('. l Position: Implemented. See A6, B9, C8, C10, and D4. B8. Drywell Sump Level (HC Category 3; RG Category 1) Position: Implemented as Category 3. See C6 and Issue 4, Section 1.8.1.97.4.4.
- d. Maintaining Containment Integrity [
B9. Primary Containment Pressure (HC Category 1; l RG Category 1) Position: Implemented. See A6, B7, CS, CIO, and D4. i B10. Primary Containment Isolation Valve Position (excluding check valves) (HC Category 1; RG Category.1) l_ Position: Implemented. Redundant indication is not required on each redundant isolation valve. ()' 1.8.1.97.3.3 type C Variables l
- a. Fuel Cladding l t
C1. Radioactivity Concentration or Radiation Level in Circulating Primary Coolant (RG Category 1) Position: Not implemented. See Issue 5, Section 1.8.1.97.4.5. C2. Analysis of Primary Coolant (gamma spectrum) (HC Category 3; RG Category 3) Position: Implemented. C3. BWR Core Thermocouples (RG Category 1) Position: Not implemented. See B4, B,5, and SLI-8121 (December, 1981) (Appendi: A to Reference a.8-4). I i DSER OPrN IrD( ;p g 3 1.8-69 Amendment 7 l
m IH3Mi FSAR- 8/84 . I
- b. Reactor Coolant Pressure Boundary l
,. 4
~
C4. RCS Pressure (HC Category 1; RG Category 1) Position: Implemented. See A2, B6, and C9. C5. Primary Containment Area Radiation (HC Category-1; RG Category 3) Position: Implemented as Category 1. See E1. C6. 'Drywell Drain Sumps Level (identified and unidentified leakage) (HC Category 3;,RG Category 1) Position: Implemented as Category 3. See B8 and Issue 4, Section 1.8.1.97.4.4. C7. Suppression Pool Water Level (HC Category 1; RG Category.1) Position: Implemented. See A5 and D5. ( ) ('\ . CB. Drywell Pressure (HC Category 1; RG Category 1) Position: Implemented. See A6, B7, and B9, CIO, and D4.
- c. Containment C9. RCS Pressure (HC Category 1; RG Category 1)
Position: Implemented. See A2, B6, and C4. C10. Primary Containment Pressure (HC Category 1; RG Category 1) - Position: Implemented. See A6, B7, B9, CS, and D4. C11. Containment and Drywell H, Concentration (HC Category 1; RG Category 1) Position: Implemented. See A1. C12. Containment and Drywell Oxygen Concentration (HC , Category 1; RG Category 1) , Position: Implemented. See A1. os a op u I m 1.8-70 Amendment 7 g3
HCGS FSAR 8/84 "" Cl (' " C13. Containment Effluent Radioactivity--Noble Gases (from identified release points including Filtration, Recirculation & Ventilation System Vent) (HC Category 3; RG Category 3) Position: Implemented. C14. Radiation Exposure Rate (inside buildings or areas, e.g., auxiliary building, reactor building, which are in direct contact with primary , containment where penetrations and hatches are located) (RG Category 2) Position: Not implemented. See E2, E3, and Issue 6, Section 1.8.1.97.4.6. C15. Effluent Radioactivity--Noble Gases (from
, buildings as indicated above) (HC Category 2; RG
, Category 2) Position: Implemented. 1.8.1.97.3.4 Type D Variables l
~
{ a. Condensate and Feedwater System l D1. Main Feedwater Flow (HC Category 3; RG Category 3) Position: Implemented. ll D2. Condensate Storage Tank Level (HC Category 3; RG Category 3) Position: Implemented.
- b. Primary Containment-Related System l D3. Suppression Chamber Spray Flow (HC Category 2; RG Category 2)
- Position
- Implemented.
D4. Drywell Pressure (HC Category 2; RG Category 2) Position: Implemented. See A6, B7, B9, C8 and C10. * ( yg 1.8-71 Amendment 7 e ,,-,,.y .---w,- -
,-c.
- (
I. [;i ._, HCGS FSAR 8/84 l 1 (L
~,
\
(, ~ ' D5. Suppression Pool Water Level (HC Category 2; RG L - Category 2) l: Pocition: Implemented. See A5 and C7. i'- D6. Suppression Pool Water Temperature (HC Category 1; l RG Category 2) Position: Implemented, but must be Category 1. Both local and bulk temperature. See A4. D7. Drywell Atmosphere Temperature (HC Category 2; RG l Category 2) Position: Implemented. l' D8. Drywell Spray Flow (HC Category 2; RG Category 2) Position: Implemented. t c. Main Steam System l D9. Main Steamline Isolation Valves' Leakage Control
. System Pressure (HC Category 2; RG Category 2)
(;) Position: Implemented. (System is identified as Main Steam Isolation Valve Sealing System at HCGS). ! D10. Primary System Safety Relief Valve Position, l Including ADS or Flow Through or Pressure in Valve Lines (HC Category 2; RG Category 2) Position: Implemented.
- d. Safety Systems l D11. Isolation Condenser System Shell-Side Water Level Position: Not applicable to HCGS.
D12. Isolation Condenser System Valve Position Position: Not applicable to HCGS. D13. RCIC Flow (HC Category 2; RG Category 2) Position: Implemented. See Issue 7, Section 1.8.1.97.4.7. ( ; v . DSER OPET Irmt y263
. . . . .~. .
- 1. 1. ., . . . . _ _ _ . ._
l'
~
HCGS FSAR
- 8/84 a e,
-d D14. HPCI Flow (HC Category 2 RG Category 2) l,
[. --
- Position: Implemented. See Issue 7, Section 1.8.1.97.4.7.
I D15. Core Spray System Flow (HC Category 2; RG Category 2)
; Position: Implemented. See Issue 7, I Section 1.8.1.97.4.7.
D16. LPCI System Flow (HC Category 2; RG Category 2) Position: Implemented. See Issue 7, Section 1.8.1.97.4.7. D17. SLC System Flow (HC Category 3; RG Category 2) Position: Implemented as Category 3. See Issue 7, Section 1.8.1.97.4.7. D18. SLC System Storage Tank Level (HC Category 2; RG Category 2)
' Position: Implemented.
(~ ; "
- e. Residual Heat Removal (RHR) Systems l D19. RER System Flow (HC Category 2; RG Category 2)
Positicn: Implemented. D20. RHR Heat Exchanger Outlet Temperature , (HC Category 2; RG Category 2)
- Position
- Implemented.
l
- f. Cooling Water System l D21.. Cooling Water Temperature to ESF System Components
.(HC Category 2; RG Category 2)
Position: Interpreted as Safety Auxiliaries - Cooling System (SACS) temperature and implemented. i D22. Cooling Water Flow to ESF System Components (HC [ Category 2; RG Category 2) Position: Interpreted as SACS flow and implemented. 1.8-73 Amendment 7 DSER OPEN ITzx .203 l er * *g __ _ , . . ..._._,_.-....____,,_.,m . _ _ . , _ , _ _ _ . . _ . _ _ , . _ _ , , . _ , _ . , - - _ . .,c .,,_ . -_, , . _ . . _ . _ .
. .: . . a- -.. . - . - - . . ..s . -. . _ . --- . . - . . . - .
~ HCGS FSAR 8/84
(
- g. Radwaste Systems
(. ' ' l D23. High Radioactivity Liquid Tank Level (HC Category 3; RG Category 3) Position: Implemented. 1 1
- h. Ventilation Systems l D24. Emergency Ventilation Damper Position (HC Category 2; RG Category 2)
Position: Interpreted as meaning dampers actuated under accident conditions and whose failure could result in radioactive discharge to the environment. Control room damper position is indicated. Implemented.
- i. Power Supplies l D25. Status of Standby Power and Other Energy Sources (s . Important to Safety (hydraulic, pneumatic)
(HC Category 2: RG Category 2) Position: Implemented; on-site sources only. (Note: HCGS has implemented the following D-type . variables as recoax. ended by the BWROG; see Issue 8, Section 1.8.1.37.4.8.) D26. Turbine Bypass Valve Position (HC Category 3) Position: Implemented. See Issue 8, . Section 1.8.1.97.4.8. I D27. Condenser Hotwell Level (HC Category 3) Position: Implemented. See Issue 8, Section 1.8.1.97.4.8. 3 D28. Condenser Vacuum (HC Category 3) i Position: Implemented. See Issue 8; t
, Section 1.8.1.97.4.8.
D29. Condenser Cooling Water Flow (HC Category 3) l m, ( osER OPEN ITEM .2d 3 I*0*74 Amendment 7 l
.-w - - - -,-,-,,e---, ---en # +--y,, e s e- we+-r-'--1>p --v'*vw-=em'w= --*--eee--w----*=-=e-~~~--w--w--e--"'-*--o---m--=---+w-=-e***+-New====*-ewe--
s HCGS FSAR 8/84
~x
(-- Position: Interpreted as cooling water AT across the~ condenser and implemented. See Issue 8, Section 1.8.1.97.4.8. i D30. Primary Loop Recirculation (HC Category 3) Position: Implemented. See Issue 8, Section 1.8.1.97.4.8. ! 1.8.1.97.3.5 Type E Variables i
- a. Containment Radiation l E1. Primary Containment Area Radiation--High Range (HC Category 1 RG Category 1)
Position: Implemented in accordance with NUREG-0737 commitment. See C5. 4 E2. Reactor Building or Secondary Containment Area Radiation (RG Category 2 for Mark I and II containments). ('; . Position: Not implemented for HCGS (Mark I) containment. See C14, E3, and Issue 9, Section 1.8.1.97.4.9..
- b. Area Radiation l
! E3. Radiation Exposure Rate (inside buildings or areas 1 where access is required to service equipment important to safety) (HC Category 3; RG Category 2) l Position: Implemented as Category 3, using , existing instrumentation. See C14, E2, and Issue 10, Section 1.8.1.97.4.10. j (
- c. Airborne Radioactive Materials Released From Plant l E4. Noble Gases and Vent Flow Rate (HC Category 2; RG Category 2)
Position: Implemented. [ us osn oPEN ITu 203 1.s-75 Amendment 7 l I l c , _ , - . . _ , , - , ,n,-,,.e,v.,,,,-,-.n.,,,..n,,.
~~ . HCGS FSAR . 8/84 C.: ~
E5. Particulates and Halogens (HC Category 3; ( ~ '- RG~ Category 3) l lj Position: Implemented.
- d. Environs Radiation and Radioactivity .
l E6. Radiation Exposure Meters (continuous indication at fixed locations) Position: Deleted. See NRC errata of July 1981. E7. Airborne Radiohalogens and Particulates (portable sampling with on-site analysis capability) (HC Category 3; RG Category 3)
, Position: Implemented.
E8. Plant Environs Radiation (portable instrumentation) (HC Category 3; RG Category 3) Position: Implemented (portable equipment). E9. Plant and Environs Radioactivity (portable (c ' ., instrumentation) (HC Category 3; RG Category 3) Position: Implemented (portable equipment).
- e. Meteorology l E10. Wind Direction (HC Category 3; RG Category 3)
Position: Implemented. l r Ell. Wind Speed (HC Category 3; RG Category 2) Position: Implemented. t E12. Estimation of Atmospheric Stability (HC Category 3; RG Category 3) L Position: Implemented,
- f. Accident-Sampling Capability (Analysis Capability On-Site)
E13. Primary Coolant and Sump (HC Category 3--Primary Coolant only; RG Category 3) 1.8-76 Amendment 7 DSER OPEN Im BC 3 ,
i: , HCGS FSAR 8284 t/ ~
~
Implement'd e Primary Coolant. (- Position: (Dissolved hydrogen or Total Gas not implemented). - Sump not implemented. See Issue 11, Section 1.8.1.97.4.11. p . E14. Containment Air (HC Category 3; RG Category 3) Position: Implemented. The instrumentation for monitoring and display of type A, B, C, D, and E variables at HCGS is identified on Table 7.5-1. 1.8.1.97.4 Supplementary Analyses l These supplementary analyses support positions cited in Section 1.8.1.97.1 (Issue 1) and Section 1.8.1.97.3 (Issues 2-12). 1.8.1.97.4.1 ISSUE 1.- INSTRUMENT IDENTIFIChTION l
,, Regulatory Guide 1.97 specifies, in paragraph 1.4.b, the following: "The instruments designated as Types A, B, and C and
('T n ' Categories 1 and 2 should be specifically identified on the control panels so that the operator can easily discern that they are intended for use under accident ecnditions." The objective of tnis regulatory position is the achievement of good human factors engineering in the presentation of information to tne centrol room operator. This objective is best achieved by evaluating current practices and procedures that provide for identifying instruments in a manner that aids the operator; redundant labels would tend to distract the operator and cause confusion. Instruments designated as Categories 1 and 2 for monitoring variable types A, B, and C should be identified in such a manner as to optimize applicable human factors engineering and presentation of information to the control room operator. This pc3ition is taken to clarify the intent of Regulatory Guide 1.97, which specifies that these instruments be easily discerned for use during accident conditions. The method of identification used at HCGS will be based on the results of a human factors analysis performed on the HCGS main control room (See Chapter 18). (
. 1.8-77 Amendment 7 DSER 0?mt Irnt ,2C ]
l~ l' HCGS FSAR 8/84 ( " 1.8.1.97.4.2 ISSUE 2 - VARIABLE BI l; The measurement of neutron flux is specified as the key variable in monitoring the status of reactivity. Neutron flux is classified as a Type B variable, Category 1. The specified range is 10-* percent to 100 percent full power (SRM, APRM). The stated purpose is " function detection; accomplishment of mitigation." The lower end of the specified range,10-* percent full power, is intended to allow detection of an approach to criticality *by some undefined and noncontrollable mechanism after shutdown. In attempting to analyze the performance of the neutron-flux monitoring systems, a scenario was postulated to obtain the required approach to criticality. Basically, it assumes an increase in reactivity from dilution of boron concentration in the reactor water. The accident scenario incorporates the following factors: l ' Df
- a. The control rods fail (completely or partially) to insert, and the operator. actuates the standby liquid control system (SLCS).
- b. .The SLCS shuts the reactor down. l 4
- c. A leak in the primary system results in a dilution of borated water and replacement by water that contains no boron.
1
- d. A range of leak rates up tc 20 gpm was considered (see ,
Table 1.8-2). Calculations were made to evaluate the rise in neuteen population i- as a function of different leak rates. The calculations were
- made for a shutdown neutron level of 5 x 10-e percent of full power. The choice of 5 x 10-8 is based on measurement at two nuclear plants. The shutdown level was assumed to have a negative reactivity of 10 dollars, an assumption that is representative of a shutdown with all rods inserted. The results of the calculations are presented in Table 1.8-2. The numbers in ,
- ss l 1.8-78 Amendment 7 DSER OPEN ITEM M3
. , , p -3,m.- - - -
e wc .g- .#.-. ,- ---.wy-- * --y,-.y,wwm w--.u.ewgwme,u.---- - -+--_
. ._ . . .c HCGS FSAR 8/84
( .. ( the table refer to the time in hours required to increase the
flux by 1 decade. For example, with a leak of.5 gpm, it takes 100 he to increase the power from 5 x 10-s percent to 5 x 10-7 percent, and 10 he to increase it from 5 x 10-7 percent to 5 x 10-*' percent.
The reactor is suberitical and the neutron level is given by l ~_ Neutron level = S x M, l where S is the source strength and M is the multiplication which is given by M = 1/(1-k). l For k = 0.9, M is 10; for k = 0.99, M is 100 and so forth. For p criticality, the denominator approaches 0, as k approaches 1.0.
- Thus, the calculation model used the above equation to calculate i
relative neutron flux levels for a suberitical reactor until the reactor was near critical; then the critical equation of power (_j' l.. with excess reactivity was used. Reactor power is directly proportional to neutron level. The increase in reactivity toward criticality can be terminated by actuating the SLCS. Operating procedures provide for refilling the SLCS tank with borated water soon after its actuation. A second actuation of the SLCS would cause a decrease
; in reactivity because of the high concentration of boron in the injected SLCS fluid relative to that in the leaking fluid (nominally 400 ppm). The sensitivity of the detector nust allow adequate time for the operator to.act. Ten minutes is considered sufficient time for operator action for accident prevention and mitigation.
t j Table 1.8-2 shows that the detector sensitivity (i.e., lower l range) requirement is a function of leak rate and therefore, of reactivity-addition rate. On the basis of a 20-gpm leak rate, 1 Table 1.8-2 shows that a detector that is on scale within 3 decades of the shutdown power would allow 0.18 hr (10.8 min) for operator action before reactor power increased another decade. A total of 0.36 he (21.6 min) would be available for operator
- action from the time the indicator comes on scale to the time p reactor power reaches 0.5 percent of full power.
, (
1.4-79 Amendment 7 osER OPEN ITEM 203 l { l i
~ - . - . . . - . - , .
HCGS ESAR 8/84 us
% ? ..
l (, - The 20-gpm leak rate, which was assumed to continue for 27.75 he, was used to define the sensitivity of the detector, It should be noted that the assumed leak rate, extended over the 27.75-hr
; period, would result in a loss of inventory so large that it i would be detected by the operator. Moreover, detection of the I
".. ': reactivity addition caused by this gradual boron dilution will be
- i noted via boron concentration sampling and measurement. Again, the conservative 20-gpm leak rate was used only to obtain a mechanistic and conservative approach for selection of instrument i
l sensitivity. An absolute criterion for the lower range includes consideration
+
of the neutron source level. The use of the neutron level 100 days after shutdown is conservative. Conditions would be stable and controllable 2 days after the emergency shutdown, as the core-decay heat is at a low level and the b'ron o monitoring system is functional. The actual neutron leved will vary with fuel design, fuel history, and shutdown control strength. il Measurements of shutdown neutron flux (with all rods inserted) at two BWR reactors show readings of 30 to 80 counts /sec (1000 counts /sec corresponds to 10-a of full power). Measurements on other BWR reactors and for different fuel histories would show some variation, but those variations would be small compared with a criterion that is concerned with units of decades. Q ('? Regulatory Guide 1.97 classifies the instrumentation for measuring a variable as Category 1 on the basis of (1) whether it is a key variable (defined in Section 1.8.1.75.3), and (2) its importance to safety. Neutron flux is the key variable for
- measuring reactivity control, thus meeting the requirement of i criterion (1). The degree to which this variable is important to
! safety is another censideration. The large numoet of detectors (i.e., source-range monitors and intermediate-range monitors)
- that are driven into the core soon after shutdown makes it highly probable that one or more of the eticting NMS detectors will be inserted. On the other hand, there is little protability thac l
there would be, simultaneously, a need for this measurement (in terms of operator action to be taken) and an accident environment in which the NHS would be rendered inoperable. Further, the operator can actuate the SLCS on loss of instrumentation, i A rigorous Category 1 requirement is not justified when the purpose and use of the measurement are analyzed as they relate to the criterion of "importance to safety." A Category 2 classification of this variable fully meets the intent of Regulatory 1.97. ( n/ 1.8-80 Amendment 7 l l osza cPrN Irrn ,20 3 . f I- ---
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u (, -- A the range from flux neutron 5 x 10-s levelpercent 100 days of after full power (within shutdown) to 3 decades 100 percent of of L ; L j full power is recommended. It is concluded that a Category 2
- classification is responsive to the intent of Regulatory 1.97.
1 As defined in this issue, instrumentation for long term. monitoring of the lower end of the Regulatory Guide 1.97 i specified range is only needed during an anticipated transient without scram (ATWS) event. ATWS events do not require the J consideration of loss-of-coolant accident (LOCA) conditions. It is estimated that the environmental conditions existing during an ATWS event would be similar to the environmental conditions existing during normal operation, at least in~the short term ' during operation of equipment such as the SRM and IRM drive mechanisms. Further, ATWS mitigation features have a lower importance to safety than safety systems, making a Category 2 classification for neutron flux instrumentation in lieu of Category 1 as specified in Regulatory Guide 11.97, more appropriate and more consistent with the requirements applicable to other ATWS mitigation features. It is the HCGS position that since the HCGS neutron flux instrumentation consists of a large number of neutron monitoring channels (4 SRM, 8 IRM, and 6 APRMs plus individual LPRM C'i channels) of proven high reliability, designed to operate in environmental conditions similar to those postulated to exist during an ATWS event, and since the ATWS mitigation features have a lower importance to safety than safety systems, a Category 2 classification for neutron flux monitoring instrumentation is justified and the existing instrumentation at HCGS is satisfactory for this monitoring function without modification.
- 1.8.1.97.4.3 ISSUE 3 - TREND RECORDING ]
I The purpcse of addressing Issue 3 is to determine which variables set fceth in Pclulatcry Guide 1.97 require trend recording. [ Regulatory Guide 1.97, paragraph 1.3.2f, states the general requirement for trend recording as follows: "Where direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously available for dedicated recorders." Using the BWROG Emergency Procedures Guidelines (EPG's) as a basis, the only trended variables required for operator action are reactor water level and reactor vessel pressure.
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Other variables at NCGS are recorded as identified on Table 7.5-1.
l l' ISSUE 4 - VARIABLES B8 AND C6 l t 1.8.1.97.4.4 Regulatory Guide 1.97 requires Category 1 ' instrumentation to monitor drywell sump level (variable B8) and drywell drain sumps level (variable C6). These designations refer to the drywell equipment and floor-drain tank levels. Category 1 instrumentation indicates that the variable being monitored is a key variable. In Regulatory Guide 1.97, a key variable is. defined as "... that single variable (or minimum number of variables) that most directly indicates the accomplishment of a saf ety funct ion. . . " The following discussion supports the HCGS , position that drywell sump level and drywell drain-sumps levels should be designated as Category 3 instrumentation requirements. The HCGS drywell has two drain sumps. One drain is the equipment drain sump, which collects identified leakage; the other is the floor drain sump, which collects unidentified leakage.
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(_/ Although the level of the drain sumps can be a direct indication of breach of the reactor coolant system pressure boundary, the indication is not unambiguous, be~cause there can be water in those sumps during normal operation. There is other instrumentation required by Regulatory Gaide 1.97 that would indicate leakage in the drywell: I
- a. Drywell pressure--variable B7, Categcry 1 l
- b. Drywell temperature--variable D7, Category 2 l ;
l
- c. Primary containment area radiation-~ variable C5, ;
Category 1 , The drywell-sump levels signal neither automatic protection control circuitry nor the operator to take safety-related actions. Both sumps have level detectors that provide only the following nonsafety indications: i
- a. Continuous level indication l l
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- b. Rate of. rise indication l
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l I l l c.- High-level alarm (starts first sump pump) l l i d. High-high-level alarm (starts second sump pump) - l Regulatory Guide 1.97 requires instrumentation to function during j and after an accident. The drywell sump systems are deliberately
; isolated at the primary containment penetration upon receipt of l
an accident signal to establish containment integrity. This fact
. renders the drywell-sump-level signal irrelevant. Therefore, by i design
- drywell-level instrumentation serves no useful accident-monitoring function.
l I The Emergency Procedure Guidelines use the RPV level and the i drywell pressure as entry conditions for the Level Control l . Guideline. A small line break will cause the drywell pressure to increase before a noticeable increase in the sump level, i Therefore, the drywell sumps will provide a " lagging" versus
* "early" indication of a leak.
b; ' Based on the above considerations, HCGS believes that the drywell-sump level and drywell-drain-sumps level instrumentation i should be designated as Category 3, "high-quality off-the-shelf I instrumentation." 1.9.1.97.4.5 ISSUE 5 - VARIABLE C1 l I Regulatory Guide 1.97 speci.#ies that the states of the fuel cladding be monitored during and after an accident. The specified variable to accomplish this monitoring is variable Cl-- radioactivity concentration or radiation level in circulating primary coolant. The tange is given as "1/2 Tech. Spec. Limit to 100 times Tech. Spec. Limit, R/hr." In Table 1 of Regulatory Guide 1.97, instrumentation for measuring variable C1 is designated as Category 1. The purpose for monitoring this variable is given as " detection of breach," referring, in this case, to breach of fuel cladding. , The usefulness of the information obtained by monitoring variable , C1, in terms of helping the operator in his efforts to prevent 1 and mitigate accidents, has not been substantiated. The g particular planned operator action to be taken based on
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monitoring this variable is not specified in'the current draft of the Emergency Procedure Guidelines (EPGs). The critical actions that must be taken to prevent and mitigate a gross breach of fuel cladding are (1) shut down the reactor and (2) maintain water i
level. Monitoring variable C1, as directed in Regulatory Guide 1.97, will have no influence on either of these actions. The
- purpose of this monitor falls in the category of "information that the barriers to release of radioactive material are being challenged" and " identification of degraded conditions and their i magnitude, so the operator can take actions that are available to mitigate the consequences." Additional operator actions to mitigate the consequences of fuel barriers being challenged, other than those based on Type A and B variables, have not been identified.
Regulatory Guide 1.97 specifies measurement of the radioacti'rity of the circulating primary coolant as the key variable in
- monitoring fuel cladding status during isolation of the NSSS.
, The words " circulating primary coolant" are interpreted to mean coolant, or a representative sample of such coolant, that flows past the core. A basic criterion for a valid measurement of the specified variable is that the coolant being monitored is coolant that is in active contact with the fuel, that is, flowing past
.4 the failed fuel. Monitoring the active coolant (or a sample h
thereof) is the dominant consideration. The post-accident sampling system (PASS) provides a representative sample which can be monitored.
; The subject of concern in the Regulatory Guide 1.97 requirement i is assumed to be an isolated NSSS that is shutdown. This asrumption is justified as current monitors in the condenser off-gas and main steam lines provide reliable and accurate information on the status of fuel cladding when the plant is not {
isolated. Further, the PASS will provide an accurate status of coolant radioactivity, and hence cladding status, once the PAS 3 is activated. In che interim between NSSS isolation and operation of the PASS, monitoring of the primary containment radiation and containment hydrogen will provide information on the status of the fuel cladding. The use of a portable gross gamma monitor on the PASS sample line could likely be used to , monitor primary coolant before the analytical statien can be put l in operation (a period.of more than 2 hours). Later in the sequence, the PASS sample can be augmented by area 7 radiation monitors when the RHR system is being used to remove core decay heat. l ( v 1.8-84 Amendment 7 osER OPEN ITEM g3
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I' .. The designation of instrumentation for measuring variable C1 [. should be Category 3, because no planned operator actions are identified and no operator actions are anticipated based on thic variable serving as the key variable. Existing Category 3 instrumentation is adequate for monitoring fuel cladding status. L
; 1.8.1.97.4.6 ISSUE 6 - VARIABLE C14 -l Varable C14 is defined in Table 1 of Regulatory 1.97 as follows:
" Radiation exposure rate (inside buildings or areas, e.g.,
3 auxiliary building, fuel handling building, secondary j' cont d nment), which are in direct contact with primary containment where penetrations and hatches are located." The reason for monitoring variable C14 is given as " Indication of breach." The use of local radiation exposure rate' monitors to detect breach or leakage through primary containment penetrations is impractical and unneces.sary. In general, radiation exposure rate
. in the reactor building will be largely a function of radioactivity in primary containment and in the fluids flowing in ECCS piping, which will cause direct radiation shine on the area ; monitors. Also, because of the amount of piping and the number j h,',7 of electrical penetrations and hatches and their widely scattered locations, local radiation exposure rate monitors could give
, ambiguous indications. The proper way to detect breach of j containment is by using the plant noble gas effluent monitors. Therefore, it is the position of HCGS that this parameter not be l i implemented. , 1.3.1.97.4.7 ISSUE 7 - VARIABLES D13-D17 l , i Regulatory Guide 1.97 specifies flow measurements of the l i following systems: reactor core isolation cooling (RCIC) ( (variable D13), high-pressure coolant injection (HPCI) (variable D14), core spray (variable Dis), Icw-pressure coolart injection i (LPCI) (variable D16), and standby liquid control (SLC) (variable D17). The purpose is for monitoring the operation of individual safety systems. Instrumentation for measuring these variables is designated as Category 2; the range is specified as 0 to 110 percent of design flow. These variables are related to flow into the reactor pressure vessel (RPV). ,
' i 1.:-ss Amendment 7 osaa orzu z:za 410 3 .
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HCGS FSAR 8/84 m p (. .. The RCIC, HPCI, and core spray systems each have one branch line-
-the test line--downstream of the flow-measuring element. The test line is provided with a motor-operated valve that is normally closed (HPCI and RCIC also share a motor-operated valve that is normally open). Further, the valve in the test line automatically closes when the emergency system is actuated, thereby ensuring that indicated flow is not being diverted by the f test line. Proper valve position can be verified by a direct indication of valve position on the main control board.
Although the LPCI has several branch lines located downstream of I each flow-measuring element, upon initiation of the LPCI, the valves in the system automatically line up for proper operation
! and prevent flow diversion by branch lines. Proper valve position can be verified by the operator using main control board indication of valve position.
For all of t'he above systems, there are valid primary indicators other than flow measurement to verify the performance of the emergency system; for example, reactor vessel water level. Flow-measuring devices are not provided for the SLC system. The pump-discharge header pressure, which is indicated in the control [s}(f. room, will indicate SLC pump operation. Besides the discharge header pressure observation, the operator can verify the proper functioning of the SLC system by monitoring the followings
- a. The decrease in the level of the SLC stcrage cank, l
- b. The boron injection induced reactivity change in the reactor as measured by neutron flux I c. The main control room motor status indicating lights (or motor current),
l l d. Squib valve continuity indicating lights. l The use of these indications is believed to be a valid
- alternative to SLC system flow indication.
The flow-measurement schemes for the RCIC, HPCI, core spray, and LPCI meet the Category 2 requirements of Regulatory Guide 1.97.
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- Monitoring the SLC system can be adequately done by measuring the
( above named Category 3 variables rather than the actual flow. 1.8.1.97.4.8 ISSUE 8 - VARIABLES D26-D30 l Regulatory Guide 1.97 states that "The plant designer should select variables and information display channels required by his design to enable the control room personnel to ascertain the operating status of each individual safety system and other systems important to safety to that extent necessary to determine if each system is operating or can be placed in operation..." The purpose of this analysis was to determine whether certain other D-type variables should be added to Table 1, Regulatory Guide 1.97. Regulatory Guide 1.97 addressed safety systems and systems important to safety to mitigate consequences of an accident. Another list of variables has been compiled for the BWR in NUREG/CR-2100 (Boiling . Water Reactor Status . Monitoring during Accident Conditions, April 1981). That report and a companion report, NUREG/CR-1440 (Light Water Reactor Status Monitoring during Accident Conditions, June 1980), address plant systems not (~N (- important to safety, as well as systems that are important to safety. In particular,- these reports consider the potential role of the turbine generator system in mitigating certain accidents. These two reports.were reviewed in determining whether the listed variables (D26-D30) should be added to the Regulatory Guide 1.97 list. The NUREG evaluations used a systematic approach to derive a variables list. The basic approach of the analysis was to focus on those accident conditions under which the operator is most likely to be confronted with "and/or" accident conditions which result in the most serious consequences should the operator f ail to accomplish his required tasks. This is a probabilistic event-tree-type of study, and the reports used the sequences of the Reactor Safety Study (WASH 1400), and similar studies. The l events in each sequence that involved operator action were identified; also, events were added to the event tree to include additional operator actions that could mitigate the accident. The event tree defines a series of key plant states that could L evolve as the accident progresses and as the operator attempts to respond. Thus the operator's informational needs are linked to these plant states. l 1.8-87 Amendment 7 DSER OPEN ITEM g3 ~ FEN
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( " NUREG/CR-2100 is a BWR evaluation undertaken to address appropriate operator actions, the information needed to take those actions, and the instrumentation necessary--and sufficient-- to provide the required information. { The sequences evaluated were l ? a. ' Anticipated transient followed by loss of decay-heat removal,
- b. Anticipatut transients without scram ( ATWS), l
- c. Anticipated transient together with failure of HPCI, RCIC, and low-pressure ECCS,
- d. Large loss of coolant accident (LOCA) with failure of emergency core-cooling systems, e . e. Small LOCA with failure of emergency core-cooling
.F systems.
The Regulatory Guide 1.97 list is based on accidents that result in an isolated NSSS. The NUREG documents considered accidents that could be prevented or mitigated by using water inventory and the heat sink in the turbine plant. Five of the 15 variables identified in the NUREG, but not in Regulatory Guide 1.97, are recommended as Type D, Category 3 additions to the Regulatory Guide 1.97 list. Four of these variables are in the turbine plants the turbine bypass valve position, condenser hotwell level, condenser vacuum, and condenser cooling water flow. These variables provide a primary measure of the status of a heat sink or water inventory in the l turbine plant. The turbine-plant systems are not to be classed as " safety systems" or as systems important to safety. The addition of reactor primary-loop recirculation as a variable is also recommended. HCGS has implemented these four variables plus reactor primary loop recirculation (Variable D26-D30) as plant specific l Category 3 items in accordance with Regulatory Guide 1.97 considerations. C osxx oPEN ITEM [03 l _ . , , 9 -.. _, , ,,-,--r.,y-w---- m .-,,,,.,.-mw,
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l HCGS FSAR 8/84 ,, (' Note that HCGS has implemented variable D29 (condenser cooling water flow) by monitoring the circulating water temperature rise across the condenser as a positive AT across the condenser coupled with no decrease in condenser vacuum is an adequate l- indication of condenser cooling water flow. 1.8.1.97.4.9 ISSUE 9 - VARIABLE E2 l Regulatory Guide 1.97 specifies that " Reactor building or secondary containment area radiation" (variable E2) should be monitored over the range of 10-5 to 10* R/h for Mark I and II containments, and over the range of 1 to 107 R/hr for Mark III containments. The classification for Hope Creek is Category 2; for Mark III, the classification is Category 1. As discussed in the variable C14 position statement (Issue 6), reactor building area radiation is an inappropriate parameter to (- use to detect or assess primary containment leakage. Therefore, it is the position of HCGS that the specified reactor (rs building area radiation monitors are not required for HCGS. 1.8.1.97.4.10 ISSUE 10 - VARIABLE E3 l Regulatory Guide 1.97 specifies in Table 1, variable E3, that radiation exposure rate (inside buildings or areas where access is required to service equipment important to safety) be
> monitored over the range of 10-1 to 10* R/hr for detection of significant releases, for release assessment, and for long-term surveillance.
In general, access is not required to any area of the reactor building in order to service safety-related equipment in a post-accident situation. When accessibility is reestablished in the long term, it will be done. tar a combination of portable radiation survey instruments and post-accident sampling of the reactor building atmosphere. The existing lower-range (typically 3 decades lower than the Regulatory Guide 1.97 range) area radiation monitors would be used only in those instances in which . anticipated radiation levels were within measurable instrument t , i ranges. _( Dsn opa rm g03 1.8-89 Amendment 7 l l
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i HCGS FSAR 8/84 m I, ('"creditforexistingarearadiationmonitors.It is HCGS's Thatposit. ion that this parame is, this - parameter should be reclassified as Category 3 with the ranges specified on Table 11.5-1. 1.8.1.97.4.11 ISSUE 11 - VARIABLE E13 l Regulatory Guide 1.97 requires installation of the capability for " obtaining grab samples (variable E13) of the containment sumps and the reactor building sumps for the purpose of release assessment,. verification, and analysis. i' The need for sampling a particular sump must take into account , its location and the design of the plant in which it is installed. For all accidents in which radioactive material would be in the HCGS drywell sumps, these sumps will be isolated and
- will overflow to the suppression pool. A suppression pool sample can therefore be used as a valid alternative to a drywell sump sample.' -
'~
The analysis of reactor building sumps liquid samples can be used for release assessment, as suggested in Regulatory Guide 1.97 (
. only for those designs in which potentially radioactive water can be pumped out of a controlled area to an area such as radwaste.
For designs in which sump pump-out is not allowed on a high-radiation or a LOCA signal, or in which the water is pumped to ! the suppression pool, a sump sample does not contribute to i release assessment. The use of the subject sump samples for verification and analysis is of little value; a sample of the suppression pool and reactor water, as required by other portions of Regulatory Guide 1.97 provides a much better measurement for ' these purposes. The guidelines recommended by the BWR Owners Group and GE shall be followed in lieu of Total Dissolved Gas ; Analysis. This was agreed to in a meeting between NRC managemen.t (R. Vollmer) and GE (F. Quick) dated December 12, 1983. See Section 1.8.2 for the NSSS assessment of this Regulatory Guide.
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