ML20095H940

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Forwards Cold Shutdown/Power Bus Failure Analysis Rept, High Energy Line Break/Control Sys Failure Analysis, Common Power/Control Sys Failures Evaluation Rept & Common Sensor Failure..., Per Draft SER Open Items
ML20095H940
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 08/24/1984
From: Mittl R
Public Service Enterprise Group
To: Schwencer A
Office of Nuclear Reactor Regulation
Shared Package
ML20095H943 List:
References
NUDOCS 8408290134
Download: ML20095H940 (57)
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Text

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s Pub 2 Service O PS G Cornpany Electric and Gas 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation August 24, 1984 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:

HOPE CREEK GENERATING STATION DOCKET NO. 50-354 DRAFT SAFETY EVALUATION REPORT OPEN ITEM STATUS Attachment 1 is a current list which provides a status of the open items identified in Section 1.7 of the Draft Safety Evaluation Report (SER). Items identified as " complete" are those for which PSE&G has provided responses and no confir-mation of status has been received from the staff. We will consider these items closed unless notified otherwise. In order to permit timely resolution of items identified as

" complete" which may not be resolved to the staff's satis-faction, please provide a specific description of the issue which remains to be resolved.

Attachment 2 is a current list which identifies Draft SER Sections not yet provided.

In addition, enclosed for your review and approval (see Attachment 4) are the resolutions to the Draft SER open items listed in Attachment 3, and per your request in the July 30, 1984 meeting with the Geosciences Branch, Attachment 5 contains a copy of our comments (telecopied to 8408290134 840824 PDR ADOCK 05000354

  • E PDR The Energy People g 954912 (4M) 7 63

Director'of Nuclear

- Reactor ' Regulhtion '2 8/24/84 D.' Wagner on August 10, 1984), on the Lawrence Livermore National Laboratory draft report entitled " Site Spectra for the Hope Creek Site." 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,

- by,1 Attachments / Enclosure C D. H. Wagner USNRC Licensing Project Manager W. H. Bateman USNRC Senior Resident Inspector FM05 1/2

__q UNITED STATES OF AMERICA -

NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-354 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Public Service Electric and Gas Company hereby submits the enclosed Hope Creek Generating Station Draf t Safety Evalua-tion Report open item responses and comments on the Lawrence Livermore National Laboratory draft report entitled " Site Spectra for the Hope Creek Site."

The matters set forth in this submittal are true to the best of my knowledge , information, and belief.

Respectfully submitted, Public Service Electric and Gas Company

/

By:

Thomas J. rtin Vice Pre dent -

Engineering and Construction Sworn to and subscribed before me, a Notary Pub ic of New Jersey, this & day of August 1984.

n/Uh Mf l DAVID K. BURD NOTARYPUBLIC 0F NEW JERSEY My Comm. Empires 10-23-85 I

l GJ02 l-l

1 IRTE: 8/24/84 ATTACHMENT 1 DSER ,

R. L. MITTL 'IO OPEN SECTION A. SWWENCER -

ITEM NUMBER SUBJECT STATUS IEITER DATED 1 2.3.1 Design-basis temperatures for safety- Cmplete 8/15/84 related anxiliary systes 2a 2.3.3 Accuracies of meteorological Cmplete 8/15/84 -

measurements (Rev. 1)

  • 2b 2.3.3 Accuracies of meteorological Cmplete 8/15/84 -

measurements (Rev. 1) 2c 2.3.3 Accuracies of meteorological Cmplete 8/15/84 measurements (Rev. 2) 2d 2.3.3 Accuracies of neteorological Cmplete 8/15/84 measurements (Rev. 2) 3a 2.3.3 Upgrading of onsite meteorological Co elete 8/15/84 '

I measurements progran (III.A.2) (Rev. 2)

J 3b 2.3.3 Upgrading cf onsite treteorological Canplete 8/15/84 rreasurements program (III.A.2) (Rev. 2) 3c 2.3.3 Upgradirg of onsite treteorological NRC Action measurements progran (III.A.2) 4 2.4.2.2 Ponding levels Carplete 8/03/84 Sa 2.4.5 Wave impact and runup on service Carplete 8/20/84 Water Intake Structure (Rev. 1)

Sb 2.4.5 Wave impact ard runup on service Canplete 8/20/84 water intake structure (Rev. 1)

Sc 2.4.5 Wave impact and rurup on service Cmplete 7/27/84 water intake structure 5d 2.4.5 Wave impact and runup on service Canplete 8/20/84 water intake structure (Rev. 1) 6a 2.4.10 Stability of erosion protection Carplete 8/20/84 structures 6b 2.4.10 Stability cf erosion protection Canplete 8/20/84 structures 6c 2.4.10 Stability of erosion protection Carplete 8/03/84 structures 14 P84 80/12 1-gs

, ,,.n .-p

ATTACHMENT 1 (Cont'd)

DSER R. L. MITTL 10 OPEN SECTION' A. SQiWENCER ITEM NUMBER SUBJECT STATUS IEITER DATED e 7a 2.4.11.2 Thermal aspects of ultimate heat sink Cmplete 8/3/84 7b 2.4.11.2 Itermal aspects of ultimate heat sink C mplete 8/3/84 ,

8 2.5.2.2 Choice of maximum earthquake for New Cmplete 8/15/84 England - Piedmont Tectonic Province 9 2.5.4 Soil danpirg values Cmplete 6/1/84 10 2.5.4 Foundation level response spectra Complete 6/1/84 ,

11 2.5.4 Soil shear moduli variation Complete 6/1/84 12 2.5.4 Combination of soil layer properties Complete 6/1/84 13 2.5.4 Lab test shear moduli values Complete 6/1/84 14 2.5.4 Liquefaction analysis of river bottom Complete 6/1/84 sands 15 2.5.4 Tabulations of shear moduli Complete 6/1/84 16 2.5.4 Dryirg and wettirq effect on Cmplete 6/1/84 Vincentown 17 2.5.4 Power block settlement monitoring Complete 6/1/84 18 2.5.4 Maximtsn earth at rest pressure Complete 6/1/84 coefficient 19 2.5.4 Liquefaction analysis for service Cmplete 6/1/84 water pipirg 20 2.5.4 Explanation of observed power block Complete 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 M P84 80/12 2 - gs 1

i ATTACHMENT 1 (Cont'd)

DSER R. L. MITTL 'IO [

OPEN SECTICN A. SOMENCER I ITEM NUMBER SUBJECT STATUS IErrER DATED .

23 2.5.4 Clarification of ESAR Tables 2.5.13 Complete 6/1/84 I and 2.5.14 24 2.5.4 Soil depth mdels for intake Ccanlete 6/1/84 structure {.

25 2.5.4 Intake structure soil modeling Cm plete 8/10/84 -

26 2.5.4.4 Intake structure sliding stability Cmplete 8/20/84 f 27 2.5.5 Slope stability Complete 6/1/84 28a 3.4.1 Flood protection Cmplete 7/27/84 28b 3.4.1 Flood protection Complete 7/27/84 28c 3.4.1 Flood protection Complete 7/27/84 28d 3.4.1 Flood protection Complete 7/27/84 .

28e 3.4.1 Flood protection Cmplete 7/27/84 28f 3.4.1 Flood protection Complete 7/27/84 28g 3.4.1 Flood protection Complete 7/27/84 29 3.5.1.1 Internally generated missiles (outside Complete 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 Ccaplete 7/18/84 32 3.5.1.4 Missiles generated by natural phenmena Ccuplete 7/27/84 33 3.5.2 Structures, systems, and cmponents to Complete 7/27/84 be protected fram externally generated missiles M P84 80/12 3 - gs

-. . - - . -__.~.- -_- . . - -

ATTACHMENT 1 (Cont'd)

DSER R. L. MITIL 'IO .-

OPEN SECTION A. SGWENCER ITEM NUMBER SUBJECT STA'IUS ETTER DATED r 34 3.6.2 Unrestrained whipping pipe inside Cmplete 7/18/84 containment 35 3.6.2 ISI program for pipe welch in Cmplete 6/29/84 }

break exclusion zone l 36 3.6.2 Postulated pipe ruptures Cmplete 6/29/d4 ,

37 3.6.2 Feedwater isolaticn check valve Cmplete 8/20/84 cperability ,

38 3.6.2 Design of pipe rupture mstraints Cmplete 8/20/84 .

39 3.7.2.3 SSI analysis results using finite Cmplete 8/3/84 element method and elastic half-space approach for contairment structure 40 3.7.2.3 SSI analysis results using finite Cmplete 8/3/84 element method and elastic half-space approach for intake structure 41 3.8.2 Steel contairment buckling analysis Cmplete 6/1/84 42 3.8.2 Steel contairment ultimate mpacity Cmplete 8/20/84 analysis (Rev. 1) 43 3.8.2 SRV/IDCA pool dynamic loads Cmplete 6/1/84 44 3.8.3 ACI 349 deviations for internal Cmplete 6/1/84 structures 45 3.8.4 ACI 349 deviations for Category I Cmplete 8/20/84 structures (Rev. 1) 46 3.8.5 ACI 349 deviations for foundations Cmplete 8/20/84 (Rev. 1) 47 3.8.6 Base mat response spectra Cmplete 8/10/84 (Rev. 1) 48 3.8.6 Rocking tirre histories Cmplete 8/20/84 (Rev. 1) l M P84 80/12 4 - gs

ATTACHMENT 1 (Cont'd)

DSER R. L. MITIL 'IO OPEN SECTICDI A. SCHWENCER i ITEM NUMBER SUBJECT STATIJS LETTER DMED 1 49 3.8.6 Gems concrete section Cmplete 8/20/84 i (Rev. 1) 50 3.8.6 Vertical floor flexibility response Cmplete 8/20/84 ,

spectra (Rev. 1) 51 3.8.6 Cmparison of Bechtel independet.t Cmplete 8/20/84 verification results with the design- (Rev. 2) basis results 52 3.8.6 Ductility ratics due to pipe break Cmplete 8/3/84 53 3.8.6 Design cf seismic Category I tanks Cmplete 8/20/84 (Rev. 1) 54 3.8.6 cmbination d vertical responses Cmplete 8/10/84 (Rev. 1) 55 3.8.6 Torsional stiffness calculation Cmplete 6/1/84 56 3.8.6 Drywell stick model develcpment Cmplete 8/20/84 (Rev. 1) 57 3.8.6 Rotational time history irputs Cmplete 6/1/84 58 3.8.6 "O" reference point for auxiliary Cmplete 6/1/84 building model 59 3.8.6 overturning mment cf reactor Cmplete 8/20/84 building foundation mat (Rev. 1) 60 3.8.6 IEAP element size limitations Cmplete 8/20/84 (Rev. 1) 61 3.8.6 Seismic nodeling cf drywell shield Caplete 6/1/84 wall 62 3.8.6 Drywell shield wall boundary Cmplete 6/1/84 i conditions 63 3.8.6 Reactor building dme bcundary Cmplete 6/1/84 conditions M P84 80/12 5 - gs

ATTACHMENT 1 (Cont'd) 1

. DSER R. L. MITIL TO OPEN SECTION A. SGWENCER  !

ITEM NUMBER SUR7ECT STATUS IEITER DATED .

64 3.8.6 SSI analysis 12 Hz cutoff frequency Cmplete 8/20/84

?

(Rev. 1) 65 3.8.6 Intake structure crane heavy load Ccmplete 6/1/84 drop 66 3.8.6 Impedance analysis for the intake Cmplete 8/10/84 structure -

(Rev. 1) ,

67 3.8.6 Critical loads calculation for Cmplete 6/1/84-reactor building cbme 68 3.8.6 Reactor building foundation mat Cmplete 6/1/84 contact pressures 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 distribution in Cmplete 6/1/84 cylinder wall 71 3.8.6 overturniry of cylinder wall Cmplete 6/1/84 72 3.8.6 Deep beam design of fuel pool walls Cmplete 6/1/84 73 3.8.6 ASHSD dNe nodel load inputs Cmplete 6/1/84 74 3.8.6 Tornado depressurization Cmplete 6/1/84 75 3.8.6 Auxiliary building abnormal pressure Cmplete 6/1/84 76 3.8.6 Targential shear stresses in drywell Cmplete 6/1/84 shield wall ard the cylinder wall 77 3.8.6 Factor of safety against overturning Cmplete 8/20/84 of intake structure (Rev. 1) 78 3.8.6 [nad load calculations Cmplete 6/1/84 .

79 3.8.6 Post-modification seismic loads for Cmplete 8/20/84 the torus (Rev. 1)

M P84 80/12 6 - gs

l ATTACHMENT 1 (Cont'd) i DSER OPD3 SECITON

_ ITEM NUMBER R. L. MITIL TO ,

SUBJECI' A. SOiWENCER 80 STATUS ,

3.8.6 IEITER IATED  :

Torus fluid-structure interactions Cmplete 81 3.8.6 6/1/84 ,

Seismic displacement of torus Canplete 8/20/84 82 3.8.6 (Rev. 1) i Review of seismic Category I tank Carplete  !

design 8/20/84 '

83 3.8.6 (Rev. 1)

Factors of safety for drywll Canplete bucklirg evaluation 6/1/84 84 3.8.6 Ultimate capacity of containment (materials) Ca plete 8/20/84 85 3.8.6 (Rev. 1)

Load conbination consistency Canplete 86 3.9.1 6/1/84 Canputer code validation 87 Canplete 8/20/84 3.9.1 Infonnation on transients Canplete 88 3.9.1 8/20/84 l Stress analysis and elastic plastic analysis Canplete 6/29/84 j 89

' 3.9.2.1 Vibration systems levels for NSSS piping Carplete 6/29/84 90 3.9.2.1 Vibration testing nonitorirg program during Carplete 7/18/84 91 3.9.2.2 Pipirg supports ard anchors 92 Carplete 6/29/84 3.9.2.2 Triple flued-head containment p netrations Carplete 6/15/84 93 3.9.3.1 Ioad canbinations ard allowable Canplete stress limits 6/29 /84 94 3.9.3.2 Design of SRVs and SRV discharge piping Catplete 6/29/84 M P84 80/12 7 gs

i ATTACHMENT 1 (Cont'd)

DSER R. L. MITIL TO  ;

OPEN SECTION A. SCHMNCER ITEM NUMBER SUBJECT STATUS LETTER DATED -

t 95 3.9.3.2 Fatigue evaluation cn SRV piping Cmplete 6/15/84 +

and IDCA downcomers ,

96 3.9.3.3 IE Information Notice 83-80 ccuplete 8/20/84 i (Rev. 1)  !

97 3.9.3.3 Buckling criteria used for cmponent Cmplete 6/29/84 supp3rts 98 3.9.3.3 Design of bolts Ccmplete 6/15/84 99a 3.9.5 Stress categories ard limits for Ccmplete 6/15/84 core support structures 99b 3.9.5 Stress categories and limits for Ccmplete 6/15/84 core support structures 100a 3.9.6 10CFR50.55a paragraph (g) Cmplete 6/29/84 100b 3.9.6 10CFR50.55a paragraph (g) Ccmplete 8/20/84 101 3.9.6 PSI and ISI programs for pumps and Ccmplete 8/20/84 valves 102 3.9.6 Isak testing of pressure isolation Cmplete 6/29/84 valves 103a1 3.10 Seismic ard dynamic qualification of Ccmplete 8/20/84 mechanical ard electrical equipment 103a2 3.10 Seismic and dynamic qualification of Cmplete 8/20/84 mechanical and electrical equipment 103a3 3.10 Seismic and dynamic qualification of Ccuplete 8/20/84 mechanical and electrical equipment 103a4 3.10 Seismic and dynamic qualification cf Ccmplete 8/20/84 mechanical and. electrical equipment l

l M P84 80/12 8 - gs l

l

i l

!4 ATTACHMENT 1 (Cont'd)

DSER R. L. MITTL 10 OPEN SECTICN A. SCHWENCER ITEM NUMBER SUBJECT STATUS IETIER DATED '

t 103a5 3.10 Seismic and dynamic qualification of Cmplete 8/20/84 mechanical and electrical equipnent 103a6 3.10 Seismic and dynamic qualification of Complete 8/20/84  :

mechanical and electrical equipment ,

103a7 3.10 Seismic and dynamic qualification of Complete 8/20/84 mechanical and electrical equipnent  :

103bl 3.10 Seismic and dynamic qualification of Complete 8/20/84 mechanical and electrical equipment '

103b2 3.10 Seismic and dynamic qualification of Caplete 8/20/84 i mechanical and electrical equipnent

+

103b3 3.10 Seismic and dynamic qualification of Complete 8/20/84  :

mechanical and electrical equipment 103b4 3.10 Seismic and dynamic qualification of complete 8/20/84 mechanical and electrical equignent 103b5 3.10 Seismic and dynamic qualification'of Complete 8/20/84 mechanical and electrical equipment 103b6 3.10 Seismic and dynamic qualification of Complete 8/20/84 mechanical and electrical equipment 103c1 3.10 Seismic and dynamic qualification of Complete 8/20/84 niechanical and electrical equipment 103c2 3.10 Seismic and dynamic qualification of Complete 8/20/84 mechanical and electrical equipment 103c3 3.10 Seismic and dynamic qualification of Complete 8/20/84 mechanical and electrical equipnent 103c4 3.10 Seismic and dynamic qualification of Complete 8/20/84 mechanical and electrical equipnent 104 3.11 Environmental qualification of NRC Action mechanical and electrical equipnent M P84 80/12 9 - gs

l ATTACHMENT 1 (Cont'd)

DCER R. L. MITIL 'I0 {

OPEN SECTION A. SCHWENCER I'I1!M NUMBER SUR1ECT STA'IUS IEITER DATED i

105 4.2 Plant-specific mechanical fracturing Conplete 8/20/84 analysis ( Rev. 1) i 106 4.2 Applicability of seismic andd IDCA Couplete 8/20/84 4 loading evaluation (Rev. 1) l .

107 4.2 Minimal post-irradiation fuel Ccuplete 6/29/84 surveillance progran 108 4.2 Gadolina thermal conductivity Ccuplete 6/29/84 equation

109a 4.4.7 'IMI-2 Item II.F.2 Ccmplete 8/20/84 109b 4.4.7 'IMI-2 Item II.F.2 Ccaplete 8/20/84 110a 4.6 Functional design of reactivity Canplete 7/27/84 control systems 110b 4.6 Functional design of reactivity Cceplete 7/27/84 control systems lila 5.2.4.3 Preservice inspection program Ccuplete 6/29/84 (components within reactor pressum boundary) 111b 5.2.4.3 Preservice inspection progran Complete 6/29/84 (ccuponents within reactor pressure boundary) llic 5.2.4.3 Preservice inspection progran Ccuplete 6/29/84 (ccaponents within reactor pressure boundary) 112a 5.2.5 Reactor coolant pressure boundary Complete 7/27/84 leakage detection 112b 5.2.5 Reactor coolant pressure boundary Ccuplete 7/27/84 leakage detection 1

(

M P84 80/12 10- gs u- - _ _ _ - - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ ___ _ _ ___ _ _ _ _________ _ _ _

e ATTACHMENT 1 (Cont'd) ,

DSER R. L. MITTL 'IO b OPEN SECTION A. SCHWENCER f ITEM NUMBER SUBJECT STA'IUS LETTER DATED l 112c 5.2.5 Reactor coolant pressure boundary Caplete 7/27/84 leakage detection 112d 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 {

leakage detection  ;

112e 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 l leakage detection ,

113 5.3.4 GE procedum applicability Cmplete 7/18/84 ,

114 5.3.4 Compliance with NB 2360 of the Sunmer Complete 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 i for closum flange materials 116 5.3.4 Charpy v-notch test data for base Cmplete 7/18/84 materials as used in shell course No. 1 117 5.3.4 Crpliance with NB 2332 of Winter 1972 Cmplete 8/20/84 addenda of the ASME Code 118 5.3.4 Lead factors and neutron fluence for Complete 8/20/84 surveillance capsules I 119 6.2 'IMI item II.E.4.1 Cmplete 6/29/84 120a 6.2 'IMI Item II.E.4.2 Cmplete 8/20/84 l

l 120b 6.2 'IMI Item II.E.4.2 Cmplete 8/20/84 121 6.2.1.3.3 Use of NUREG-0588 Cmplete 7/27/84 122 6.2.1.3.3 Temperature profile Cmplete 7/27/84 123 6.2.1.4 Butterfly valve operation (post Cmplete 6/29/84 accident) l M P84 80/12 11- gs l

l 1

ATTACHMENT 1 (Cont'd) i<

DSER R. L. MITTL '10 '

OPEN SECTICN A. SOlWENCER g ITEM NUMBER SURTECT STA'1US LETTER DATED ,

124a 6.2.1.5.1 RW shield annulus analysis Ccmplete 8/20/84 '

( Rev. 1) 124b 6.2.1.5.1 RPV shield annulus analysis Complete 8/20/84 f

( Rev. 1) i 124c 6.2.1.5.1 RW shield annulus analysis Complete 8/20/84

( Rev. 1) 125 6.2.1.5.2 Design drywell head differential Conplete 6/15/84 pressure 126a 6.2.1.6 Redundant position indicators for Cmplete 8/20/84 vacuum breakers (and control rom alarms) 126b 6.2.1.6 Red.indant position indicators for Cmplete 8/20/84 vacuun breakers (and control rom alarms) 127 6.2.1.6 Operability testing of vacuum breakers Cmplete 8/20/84 (Rev. 1) 128 6.2.2 Air ingestion Complete 7/27/84 129 6.2.2 Insulation ingestion Ccmplete 6/1/84 130 6.2.3 Potential bypass leakage paths Complete 6/29/84 131 6.2.3 Administration of secondary contain- Complete 7/18/84 ment openings 132 6.2.4 Containment isolation review Ccmplete 6/15/84 133a 6.2.4.1 Containment purge system Complete 8/20/84 133b 6.2.4.1 Containment purge system Cmplete 8/20/84 133c 6.2.4.1 Containrmnt purge system Cmplete 8/20/84 M P84 80/12 12- gs

I 1 I

l I

l ATTACIMENT 1 (Cont'd) f DSER R. L. MITTL 'IO OPEN SECTION A. SCHWENCER  ;

ITEM NUMBER SUR7ECT STA'IUS LEPIER DATED 134 6.2.6 Containment leakage testing Caplete 6/15/84 135 6.3.3 IPCS and LPCI injection valve Caplete 8/20/84  ;

interlocks i

136 6.3.5 Plant-specific IDCA (see Section Conplete 8/20/84 15.9.13) (Rev. 1) 137a 6.4 Control room habitability Cmplete 8/20/84 137b 6.4 Control rom habitability Cmplete 8/20/84 137c 6.4 Control rom habitability Cmplete 8/20/84 138 6.6 Preservice inspection program for Cmplete 6/29/84 Class 2 and 3 conponents 139 6.7 MSIV leakage control system Cmplete 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 Complete 8/15/84 (Rev. 1) 140d 9.1.2 Spent fuel pool storage Cmplete 8/15/84

( Rev. 1) 141a 9.1.3 Spent fuel cooling and cleanup Cmplete 8/1/84 system 141b 9.1.3 Spent fuel cooling and cleanup Caplete 8/1/84 system 141c 9.1.3 Spent fuel pool cooling and cleanup Complete 8/1/84 system M P84 80/12 13- gs

,1 ATTACHMENT 1 (Cont'd)

DSER R. L. MITIL 'IO j OPEN SECTIQ4 A. SCHWENCER S ITEM NUMBER SUR7ECT STKIUS IEITER DATED .

141d 9.1.3 Spent fuel pool cooling and cleanup Cmplete 8/1/84 syste -

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 Canplete 8/1/84 syste 141g 9.1.3 Spent fuel pool cooling and cleanup Couplete 8/1/84 system 142a 9.1.4 Light load handling system (related Cmplete 8/15/84 to refueling) (Bev. 1) 142b 9.1.4 Light load handling syste (related Complete 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 Station service water syste Cmplete 8/15/84 (Rev. 1) 144c 9.2.1 Station service water system Cmplete 8/15/84 (Fev. 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.)

M P84 80/12 14- gs

ATTACHMENT 1 (Cont'd) .

DSER R. L. MITTL 10 OPEN SECTION A. SCHWDiCER ITEM NUMBER SUBJECT STATUS L3'rER DATED 147a 9.3.1 Cmpressed air systems Cmplete 8/3/84 (Rev 1) 147b 9.3.1 Coupressed air systems Cmplete 8/3/84 i (Rev 1) <

147c 9.3.1 Conpressed air systems Complete 8/3/84 (Rev 1) 147d 9.3.1 Cmpressed air systems Cmplete 8/3/84 (Rev 1) 148 9.3.2 Post-accident sampling syste Cmplete 8/20/84 (II.B.3) 149a 9.3.3 Equipment and floor drainage system Complete 7/27/84 149b 9.3.3 Equipnent and floor drainage system Cmplete 7/27/84 150 9.3.6 Primary containment instrtsnent gas Cmplete 8/3/84 syste ( Rev. 1) 151a 9.4.1 Control structure ventilation systm Complete 7/27/84 151b 9.4.1 Control structure ventilation systm Cmpleto 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- Complete 8/1/84 tion system (Rev 1) 154 9.5.1.4.a Metal roof deck construction Complete 6/1/84 classificiation 155 9.5.1.4.b Ongoing review of safe shutdown NHC Action capability 156 9.5.1.4.c Ongoing review of alternate shutdown NRC Action capability M P84 80/12 15- gs c-_____--___

& I ATTACHMENT 1 (Cont'd)

D6ER R. L. MITIL 'IO {

OPEN SECTICN A. SCHWENCER r ITEM NUMBER SUBJECT STA7US IErrER DATED j 157 9.5.1.4.e Cable tray protection Cmplete 8/20/84 {

158 9.5.1.5.a Class B fire detection system Cmplete 6/15/84 159 9.5.1.5.a Primary and secondary power supplies Cmplete 6/1/84 for fire detection system 160 9.5.1.5.b Fire water pung capacity Complete 8/13/84 161 9.5.1.5.b Fire water valve supervision Cmplete 6/1/84 162 9.5.1.5.c Deluge valves Cmplete 6/1/84 163 9.5.1.5.c Manual hose station pipe sizing Ccmplete 6/1/84 164 9.5.1.6.e Remote shutdown panel ventilation Ccmplete 6/1/84 165 9.5.1.6.g Emergency diesel generator day tank Ccmplete 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 monitors Cmplete 7/18/84 168 12.5.2 Equipment, training, and procedures Cmplete 6/29/84 for inplant iodine instrtmentation 169 12.5.3 Guidance of Division B Regulatory Complete , 7/18/84 Guides 170 13.5.2 Procedures generation package Cmplete 6/29/84 submittal 171 13.5.2 TMI Item I.C.1 Complete 6/29/84 172 13.5.2 PGP Ccanitment Complete 6/29/84 173 13.5.2 Procedures covering abnormal releases Completo 6/29/84 of radioactivity M P84 80/12 16- gs

I

?.

l!

t ATTA0iMENT 1 (Cont'd)

DSER R. L. MITIL TO l OPEN SECTION A. SCHnENCER  ;

-ITEM NUMBER SUR7ECT STATUS IEITER IRTED

}

174 13.5.2 Resolution explanation in FSAR of Cmplete 6/15/84  ?

TMI Items I.C.7 and I.C.8 175 13.6 Physical security Open j t

176a 14.2 Initial plant test program Caplete 8/13/84  ;

176b 14.2 Initial plant test program Cmplete 8/13/84 i

176c 14.2 Initial plant test program Cmplete 7/27/84 176d 14.2 Initial plant test program Cmplete 8/4/84 (Rev. A) 176e 14.2 Initial plant test program Cmplete 7/27/84 176f 14.2 Initial plant test program Cmplete 8/13/84 176g 14.2 Initial plant test program Cmplete 8/20/84 176h 14.2 Initial plant test program Cmplcte 8/13/84 1761 14.2 Initial plant test program Cmplete 7/27 /84 177 15.1.1 Partial feedwater heating Cmplete 8/20/84 (Rev. 1) 178 15.6.5 EDCA resulting frcm spectrum of NBC Action postulated piping breaks within RCP 179 15.7.4 Radiological consequences (f fuel NRC Action handling accidents 180 15.7.5 Spent fuel cask drcp accidents NBC Action ,

181 15.9.5 TMI-2 Item II.K.3.3 Cmplete 6/29/84 182 15.9.10 TitI-2 Item II.K.3.18 Cmplete 6/1/84 183 18 Hope Creek DCRDR Cmplete 8/15/84 M P84 80/1217 - gs

l ATTAGMElff 1 (Cont'd) d DSER R. L. MITIL 'IO l OPDi SECTION A. SGENCER  ;

I'IEM NUMBER SUEkTECT STA'IUS LETTER DATED 184 7.2.2.1.e Failures in reactor vessel level Cmplete 8/1/84 i 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 systems at power (Rev. 1) 187 7.2.2.4 Lifting d leads to perform surveil- Caplete 8/3/84 lance testing 188 7.2.2.5 Setpoint nethodology Cmplete 8/1/84 189 7.2.2.6 Isolation devices Cmplete 8/1/84  ;

190 7.2.2.7 Regulatory Guide 1.75 Cmplete 6/1/84 191 7.2.2.8 Scram discharge volume Cmplete 6/29/84 192 7.2.2.9 Reactor node switch Cmplete 8/15/84 (Rev. 1) 193 7.3.2.1.10 Manual initiation cf safety systems Cmplete 8/1/84 194 7.3.2.2 Standard review plan deviations Ccmplete 8/1/84 (Rev 1) 195a 7.3.2.3 Freeze protection / water filled Cmplete 8/1/84 instrument and sanpliry lines and cabinet temperature control 195b 7.3.2.3 Freeze protection / water f11 led Cmplete 8/1/84 instrument and sanpling lines and cabinet tenperature control 196 7.3.2.4 Sharing cf comen instrunent t@s Cmplete 8/1/84 197 7.3.2.5 Micrcprocessor, multiplexer ard Cmplete 8/1/84 cmputer systems (Rev 1)

M P84 80/1218 - gs

ATIACNGENT 1 (Cont'd)

DSER R. L. MITIL 10 .

OPEN SECTICE A. SCHWENCER e ITEM NUMBER SURTECT STATUS IEITER DATED i

198 7.3.2.6 TMI Item II.K.3.18-ADS actuation Cmplete 8/20/84 4 199 7.4.2.1 IE Bulletin 79-27-Ioss of non-class Cmplete 3/A//fY IE instrunentation and control power CAs/4) '

system bus during operation 200 7.4.2.2 Remote shutdown system Complete 8/15/84 (Rev 1) 201 7.4.2.3 RCIC/HPCI interactions Complete 8/3/84 202 7.5.2.1 Level measurement errors as a result Complete 8/3/84 of environmental tenperature effects on level instrunentation reference leg  ;

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 camputer 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 Complete $/a//p/

failures tAar1) 208 7.7.2.2 Multiple control system failures Complete r/Av/ff cR.s v .L) 209 7.7.2.3 Credit 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 Ccmplete 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 M P84 80/12 19- gs 1

I ATTACHMENT 1 (Cont'd)

DSER R. L. MITTL 'IO i OPEN SECTICN A. SCHWENCER i' ITEM NUMBER SUR7ECT STATUS LETTER DATED ,

t 211d 4.5.1 Control rod drive structural materials ccmplete 7/27/84 i 1

211e 4.5.1 Control rod drive structural materials Ccaplete 7/27/84 212 4.5.2 Reactor internals materials Ccmplete 7/27/84 213 5.2.3 Reactor coolant pressure boundary Ccmplete 7/27/84 material l 214 6.1.1 Engineered safety features materials Couplete 7/27/84 215 10.3.6 Main steam and feedwater system Conplete 7/27/84 materials 216a 5.3.1 Reactor vessel materials Ccmplete 7/27/84 l

216b 5.3.1 Reactor vessel materials Ccmplete 7/27/84 217 9.5.1.1 Fire protection organization Ccmplete 8/15/84 218 9.5.1.1 Fire hazards analysis Ccmplete 6/1/84 219 9.5.1.2 Fire protection administrative Ccmplete 8/15/84 controls 220 9.5.1.3 Fire brigade and fire brigade Complete 8/15/84 training 221 8.2.2.1 Physical separation of offsite Complete 8/1/84 transmission lines 222 8.2.2.2 Design provisions for re-establish- Complete 8/1/84 ment of an offsite power source 223 8.2.2.3 Independence of offsite circuits Ccmplete 8/1/84 i between the switchyard and class IB -

[ Was 224 8.2.2.4 Ccmmon failure node between onsite Canplete 8/1/84 and offaite power circuits i

M P84 80/12 20- os

l l

ATTACHMENT 1 (Cont'd)

DSER R. L. MITIL 'IO OPEN SECTION A. SOMNCER -

ITEM NUMBER SUBJECT STA'IUS LETTER DATED 225 8.2.3.1 Testability of automatic transfer of Cmplete 8/1/84 power frm the normal to preferred power source 2 26 8.2.2.5 Grid stability Cm plete 8/13/84 .

(Rev. 1) 2 25 8.2.2.6 Capacity and capability of offsite Cmplete 8/1/84 circuits 2 28 8.3.1.l(1) Voltage drop during transient condi- Cmplete 8/1/84 tions 229 8.3.1.l(2) Basis for using bus voltage versus Caplete 8/1/84 actual connected load voltage in the voltage dr@ analysis 230 8.3.1.l(3) Clarification of Table 8.3-11 Caplete 8/1/84 231 8.3.1.l(4) Undervoltage trip setpoints Caplete 8/1/84 232 8.3.1.l(5) Load configuration used for the Cmplete 8/1/84 voltage drcp analysis 233 8.3.3.4.1 Ebriodic system testing C m plete 8/1/84 234 8.3.1.3 C@acity and capability of onsite Cmplete 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 Cmpletu 8/1/84 test 2 36 8.3.1.6 Cmpliance with position C.6 of Cmplete 8/1/84 10 1.9 237 8.3.1.7 Decription of the load sequencer Cmplete 8/1/84 238 8.2.2.7 Sequencing cf loads on the offsite Cmplete 8/1/84 power system M P84 80/12 21 - gs

s f

ATTACIMENr 1 (Cont'd)  !~

DSER R. L. MITTL 10 i OPEN SECTION A. SCHWENCER j' ITEM NUMBER SUR7ECT STATUS LETTER DATED i 239 8.3.1.8 Testing to verify 80% mininnan Caplete 8/15/84

' voltage 240 8.3.1.9 Compliance with BTP-PS&2 Caplete 8/1/84 f' 241 8.3.1.10 Ioad acceptance test after prolonged Complete 8/20/84 i no load operation of the diesel (Dev. 1) generator 242 8.3.2.1 Ca pliance with position 1 of Begula- Cmplete 8/1/84 tory Guide 1.128 243 8.3.3.1.3 Protection or qualification of Class Caplete 8/1/84 lE equipnent frun the effects of fire suppression systems 244 8.3.3.3.1 Analysis and test to demonstrate Cenplete 8/1/84 adequacy of less than specified separation 245 8.3.3.3.2 The use of 18 versus 36 inches of Complete 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- Complete 8/1/84 stand long duration short circuits at less than maxinun or worst case j short circuit l 248 8.3.3.5.2 Separation of penetration primary Ccunplete 8/1/84 and backup protections l

249 8.3.3.5.3 The use of bypassed thermal overload Complete 8/1/84

! protective devices for penetration protections 250 8.3.3.5.4 Testing of fuses in accordance with Complete 8/1/84 R.G. 1.63 M P84 80/12 22- gs ,

ATTACHMENT 1 (Cont'd)

DSER R. L. MITTL 10 '

OPEN SECTION A. SOMENCER ITEN NUMBER SUBJECT STAltJS [EPTER DATED 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 Caplete 8/1/84 penetration protection 253 8.3.3.1.4 Ccanitment to protect all Class 1E Cmplete 8/1/84 equipnent fra external hazards versus l

only class IE equipment in one division l 254 8.3.3.1.5 Protection of class lE power supplies Cmplete 8/1/84 fra failure of unqualified class 1E loads 255 8.3.2.2 Battery capacity Ccaplete 8/1/84 256 8.3.2.3 Automatic trip of loads to maintain Ccuplete 8/20/84 sufficient battery' capacity 257 8.3.2.5 Justification for a 0 to 13 second Complete 8/1/84 load cycle 8.3.2.6 Design and qualification of DC Cmplete 258 8/1/84 system loads to operate between minimum and maximum voltage levels 259 8.3.3.3.4 Use of an inverter as an isolation Cmplete 8/1/84 device 260 8.3.3.3.5 Use of a single breaker tripped by Conglete 8/1/84 a IDCA signal used as an isolation device 261 8.3.3.3.6 Automatic transfer of loads and Complete 8/1/84 interconnection between redundant

divisions 262 11.4.2.d Solid waste control progre Cmplete 8/20/84 t

M P84 80/12 23- gs i

i a

1 I ATTADMENT 1 (Oont'd)

DSER R. L. MITIL 'IO OPEN SECTION A. SCHMNCER ITEM NUMBER SURIECT STNIUS IEPIER DATED i

263 11.4.2.e Fire protection for solid radwaste Conglete 8/13/84  !

storage area 264 6.2.5 Sources of oxygen Conglete 8/20/84 >

265 6.8 .1.4 ESF Filter Testing Conglete 8/13/84 266 6.8.1.4 Field leak tests Caplete 8/13/84 267 6.4.1 Control rom toxic chemical Complete 8/13/84 detectors 268 Air filtration unit drains Conplete 8/20/84 269 5.2.2 Orde cases N-242 and N-242-1 Cmplete 8/20/84 270 5.2.2 Code case N-252 Caplete 8/20/84 TS-1 2.4.14 Closure of watortight chats to safety- Ogwn related structures TS-2 4.4.4 Single recirculation loop operation Open

'IS-3 4.4.5 Core flow monitoring for crud of fects Cmplete 6/1/84 TS-4 4.4.6 Loose parts monitoring syst m Open TS-5 4.4.9 Natural circulation in normal Open operation TS-6 6.2.3 Secondary containment wJative Open pressum l TS-7 6.2.3 Inleakage and drawkMn time in qwn secordary containment TS-8 6.2.4.1 Imakage integrity testing Olen TS-9 6.3.4.2 ECCS nutmystem periodic omponont Olen ,

testing M P64 80/12 24- gs

ATTA0043rr 1 (Cont'd)

DSER R. L. MITIL 10 h OPSI SECTIGi A. SCHWENCER $

ITB4 NUpWER FAJEUECr STAltJS IErrER DATED '

TS-10 6.7 MSIV leakap rate

.TS-J,1 15.2.2 Availability, mutpoints, and testing open of turbine bypass systaan  :

TS-12 15.6.4 Primary coolant activity 14-1 4.2 Fuel rod internal pressuru criteria Conglete 6/1/84 14-2 4.4.4 Stability analysis sutaitted before Open second-cycle operation M P84 80/12 25- gn

l ATTACHMENT 2 DATE: 8/24/84 DRAFT SER SECTIONS AND DATES PROVIDED ,

SECTION DATE SECTION DATE  !

3.1 l 3.2.'l 11.4.1 See Notes 1&S I 3.2.2 11.4.2 See Notes 1&S 5.1 11.5.1 See Notes 1&S 5.2.1 11.5.2 See Notes 1&S  ?

6.5.1 See -Notes 1&5 ~ 13.1.1 See Note 4 8.1 See Note 2 13.1.2 See Note 4 8.2.1 See Note 2 13.2.1 See Note 4 8.2.2 See Note 2 13.2.2 See Note 4 8.2.3 See Note 2 13.3.1 See Note 4 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 Notes 1&5 9.5.7 See Note 3 17.1 8/3/84 9.5.8 See Note 3 17.2 8/3/84 10.1 See Note 3 17.3 8/3/84 10.2 See Note 3 17.4 8/3/84 10.2.3 See Note 3 10.3.2 See Note 3 10.4.1 See Note 3 10.4.2 See Notes 3&5 10.4.3 See Notes 3&5 10.4.4 See Note 3 11.1.1 See Notes 1&S Notes:

11.1.2 See Notes 1&S 11.2.1 See Notes 1&5 1. Open items provided in 11.2.2 See Notes 1&5 s letter dated July 24, 1984 11.3.1 See Notes 1&5 (Schwencer to Mittl) 11.3.2 See Notes 1&5

2. Open items provided in June 6, 1984 meeting
3. Open items provided in
v. April 17-18, 1984 meeting CT:db J 4. Open items provided in May 2. 1984 meting
5. Draft SER Section provided in letter dated . August 7, 1984 (Schwencer to Mittl)

MP 84 95/03 01 '

)

?

DATE: 8/J N'[8I -

ATTACHRENT 3 OPEN ITEM DSER SECTION SUBJECT 176d 14.2 Initial plant test prograrn 199 7.4.2.1 IE Bulletin 79-27-Loss of non-class IE instrunentation and control power system bus during operation 207 7.7.2.1 HELBs and consequential control system failures 208 7.7.2.2 Multiple control systen failures i

i S

g.

l 1

1

\

4

' 1 I I 1

4  !

t I;

?

e h

i ATTACHMENT 4 l

1 I

(

(

HCGS ggy R DSER Open Item 176d (Section 14.2)

INITIAL PLANT TEST PROGRAM The response does not address the concerns of IEE Information Notice Number 83-17, March 31,1983. The concern is that if a time delay prevents fuel from being supplied to the diesel generator following a shutdown signal, the air supply may be exhausted before the fuel supply is reinstated. The response to this item should be modified to address these concerns.

^

RESPONSE

The response to 0640.10 has been revised to provide the information requested abov.e.o.nd +0 a.dd e-e ss i+em a pu discussions u2 % *

  • M E C- -

M P84 126/07 2-dh I

HCGS FSAR 6/84

~

~ l 00ESTION 640.10 (SECTION 14.2.12)

Modify your FSAR submittal to address the following concerns regarding emergency diesel generator testing:

1. FSAR Subsections 1.8.1.108 and 14.2.13.5 state that Regulatory Guide 1.108 (Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants) is not applicable to Hope Creek. It is the staff's position that this guide is applicable to your facility.

Therefore, either delete or provide justification for this .

statement.

I

2. FSAR Subsections 1.8.1.108 and 14.2.13.5 take exception to i Position C.2.a(5) of Regulatory Guide 1.108. These subsections state that testing of the sequencing controls after the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test run does not subject the controls to more severe conditions than testing accomplished under other circumstances. Provide technical justification for your position or perform this test in accordance with this guide.

Additionally, modify FSAR Subsection 14.2.12.1.30 (KJ-Emergency Diesel Generators) to perform a restart simulating loss of ac directly after the 24-hour run in accordance with your statement in the aforementioned FSAR subsections.

3. Modify FSAR Subsections 14.2.12.1.30 (KJ-Emergenc) Diesel Generators), 14.2.12.3.30 (Loss of Turbine-Generator and r

Offsite Power), or other test abstracts as appropriate, to:

a. Perform the simultaneous, redundant diesel starts specified in Position C.2.b of Regulatory Guide 1.108.
b. Include prerequisite testing to ansure the satisfactory operability of all check valves in the flow path of cooling water for the diesel generators from the intake to the discharge (see IEE Bulletin No. 83-03: Check Valve Failures in Raw Water Cooling Systems of Diesel Generators).
c. Provide assurance that any time delays in the diesel generator's restart circuitry will not cause the supply of compressed air used to initially rotate the engine

' to be consumed in the presence of a safety injection signal (see IEE Information Notice Number 83-17, March 31, 1983).

RESPONSE f*M R Secdian3 f. g. /. tor a & M. A./3 F W 'A be re.viseol a s repe sied a bove Ti.is i; ~R ~_

NPC  :; _ tery C;id: 1.1^* 1 :t :;;1i0:51: te HCG 3.

j;;tified :: et=ted ir. !;pler: tetien Sectian n nr n ;rieto,y t __

l 640.10-1 Amendment 6

a l

l HCGS FSAR 6/84 A

Guia i . ii,;; -i icn providw. th;t th: ; ide is te be r::d in the reh ti;; ef sceitt:1; f;; ;; :trerHan P r=itr. %

Section 14.2.12.1.30.c.6 has been revised to state that a restart simulating loss of ac power will be performed following the 24-hour run.

Upon restart, a sequencing check will not be performed since the 24-hour run test has no effect on the sequencing circuit. The sequencing circuits are located in the emergency load sequencer panels remote from the diesel generator room. The circuits will l not have left their standby state since the 24-hour run is accomplished without a loss-of-power or loss-of-coolant accident condition, and is synchronized to the grid. However, the sequencing will be checked during the ECCS integrated initiation during loss-of-offsite power test described in Section h*"

  • b 14.2.12.1.47. H o u> e v e r- i m m e ch+e l y Eollowin3 %c Z4

%e. s im wl a.ted 105 5 o f et c pow e r will be fo llowco/ hy a n e mmedia Y C Simultaneous redundant diesel starts are accomplished as ma n u.3 loadiof described in Section 14.2.12.1.47.c.2.  % des ig n loo d c-o ncL l+ e n .

Section 14.2.12.1.30 has been revised to include prerequisite component testing on all diesel generator cooling water check valves.

The diesel generator control design has a time delay relay which holds the fuel racks closed to allow the unit to come to a complete stop. However, in the event of an emergency start signal due to ECCS requirements during the count down of the time delay relay, this relay is functionally overridden and the fuel racks open to allow the diesel to continue to run or restart through the normal starting air sequence described in Section 9.5.6.

640.10-2 Amendment 6 7

~- -

,n---- -- - - - - , , ,, - -, -.-- ,

. ,,.,--nom-----,m --

,, -n-e---

l l

~

HCGS FSAR 8/83 .

1.8.1.107 Conformance to Reculahory Guide 1.107, Revision 1, l February t977: Os alif: ,cattolm for Cement Grouttaa for Pres :ress: .no Teni ons .n Con ;ainment Structures l

Regulatory Guide IJ107 is not applicable to HCGS. J l

1.8.1.108 Conformance to Rooulatory Guide 1.108, Revision 1, l Aucust 1977: Per . odic Testino of Diesel Generator Units l Used as Onsite E; .ectric Power Systems at Nuclear Power Plants Althens,h R=ulateg Guit S ? ^" i. uv6 applive'uiw te "^G", ym.  ;

it; i;+1 :::t: tion :::tienf_.HCGS complies with with the .  !

following exceptions g & ,y 6 anold I I#I i m m e.diohelf P sition C.2.a(5) requires that the accident loading sequence to j dnsign load requirements be performed directly af ter the 24-hour i run. This does not test the sequencing controls under a more '

severe condition than if sequent.ially loaded at an earlier or . Ig Icter periodp A restart simulating loss of ac power -eeaR #  !

parformed2fi_ _ : 'y af ter the 24-hour run.g Sequencing, however, will be performed when the loads can be lined up for operation

] '

l cnd all four diesels are available. .

{

kY'm~ o$e$rk'O'eWsYt Ye eh Y 1.8.1.109 Contormance to nequiacory uutde 1.109, Revision 1, October 1977: Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluatina Comp',iance with 10 CFR Part 50, e llowec) by con im m su;hW

%na1 icae:t ng -te cle.s(3r\ l HCGS complies with Regulatory Guide 1.%oo.d c.ond' Mons-For further discussion, see Chapter 15. i 1.8.1.110 Conformance to Reaulatory Guide 1.110, Revision 0, March 1976: Cost-Benefit Analysis for Radwaste Systems For Licht-Water-Cooled Nuclear Power Reactors HCGS complies with Regulatory Guide 1.110.

i i

1.8-67 Amendment 1 I

l 1

HCGS FSAR 6/84

4. Demonstrate that manual and automatic operation of ,

the diesel generators is satisfactory, and that they start automatically upon simulated loss of ac voltage and attain the required freqJency and voltage.

5. Verify that proper response and operation of the i design basis accident loading sequence to design *

< basis load requirements, and verify that voltage and frequency are maintained within specified '

limits. This test may be accomplished in the preoperational test described in '

Section 14.2.12.1.47, ECCS integrated initiation during loss of offsite power.

6. Demonstrate full load carrying capability of the diesel generators for a period of not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, of which 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> are not less than the equivalent DBA full load for the respective bus, and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> are at the 2-hour 110% load rating.

Following the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run, an automatic restart due to simulated loss of ac power n ,rn - > J Mll be anol enual I

h M 3 .desi3 n - l e a.d regui re m erds w )l i ,,, rri e cA de l y j be perh,me.d

< 7. Verify that the diesel generators can be .

synchronized to an offsite power source while maintaining the Class 1E loads.

8. Verify that the standby diesel generator (SDG) system is capable of transferring the Class 1E i

load from the generator to the offsite power source, and of isolating the generator from the bus and returning it to standby status.

9. Verify that the rate of fuel consumption at design basis load for each diesel generator is such that the requirements for 7-day storage inventory are met.
10. During surveillance testing, verify the capability of the diesel generators to respond to an emergency signal and supply power to the Class IE bus, while monitoring time, frequency, and voltage.

I i

~

14.2-82 Amendment 6

HCGS FSAR 1/84 i

6. In response to DBA simulation the loading sequence -

is as specified in Table 8.3-1 and voltage and frequency are maintained within the values specified in Section 8.3.1.1.3.

o.nd loaoNJ

7. The diesel gener shall operate for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> under load as specified in Section 8.3.1.1.3. The automatic resta after the 24 hour run shows that the diesel generator attains rated speed and voltage as specified in Section 8.3.1.1.3.
8. The diesel generator will synchronize to offsite power while maintaining the Class IE loads, transfer the load to offsite power, and resume standby status following an operational mode.
9. The diesel generator fuel oil storage tanks have a demonstrated capacity as specified in Section 9.5.4.2.1, based upon engine fuel consumption.
10. With the diesel generator operating in the surveillance mode, it will respond to an emergency signal to supply power to Class IE bus loads.
11. Load rejection does not result in exceeding speeds or voltages which cause diesel generator tripping or mechanical damage.
12. The standby diesels start the number of times
specified in Section 9.5.6.3 without the air
receiver recharging compressor available.

1402.12.1.31 KP-Main Steam Isolation Valve Sealing

a. Objective The test objective is to verify flow paths, controls operation, interlocks, and alarms associated with the main steam isolation valve (MSIV) leakage control j system. 3 14.2-84 Amendment 4

1 HCGS FSAR 1/84 1 calibration completed prior to performing the preoperational test. .

14.2.13.5 SRP II.e, Reaulatory Guide 1.108, Revision 1, Aucust 1977: Periodic Testina of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants

..lther?

Dagn1 =&ary Gu Me 1 na 1: :st ;glic:ble te "CCC, pu b its invl,.. Ontat!On ::: tic =,"-HCGS complies withytt, with the ,

follcwing clarifications: [g W&ry 6asd c l 10E

a. Position C.2.a (5) requires that the accident loading sequeqce to design load requirements be performed directly after the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run. This does not test the sequencing controls under a more severe condition than if sequentially loaded at an earlier or later perio A restart simulating loss of ac p be performed imme d atel y -al_ _:tP; af ter the 24-hour rung L w: n followed bod c_onct; hens. imm eam an an as cLiu+ e (

d:nj + o O'* E Sequencing, nowever, will be performed when the loads can be lined up for operation and all four diesels are available.

d a e + o % e. e,n e rgency loa.d seg uen c e r par'cls ni, "PC lo c.cdeol remot e from %e d'. cs c i 3 enera.1er roe i

I MJ1-206 Amendment 4

s' HCGS h e. V b ,

DSER Open Item No. 199 ( DSER Se ction 7.4.2.1)

IE BULLETIN 79 LOSS OF NON-CLASS lE INSTRUMENTATION AND CONTROL POWER SYSTEN BUS DURING OPERATION.

I We will require the applicant to document the results of the  !

analysis, providing recommendation of hardware or procedural changes as appropriate in response to IEB 79-27. This is presently scheduled for submittal during the fourth quarter of 1984. 3

RESPONSE

The response to Question 421.42 has heed t *- v/8e d f' To pr-o va cle wh e. inferm Men regaes+eci ab o ve. A copf o f %e co ad sh utdowit / Power Ba s fallu r t. '

& naIy 5; s e e po ,.1 as ata.+e.d ny sf, 19 2 y ;s a ti a e.h e d % y o m.c u s e.

199-1

HCGS FSAR 4/84 syst s use in attai 'ng the ld shut wn con tion.

Id tify sses that ould a ect the bility o achieve ,

. Id sh down. Us plant erating rocedu s and ,

proce res devel ed for ertain er bus ailures to -

ens e the ide ificat' n of all ritica power busses  ;

I dentify t instr entation nd cont 1 devices connected

}2. to each ' entifi power b . Evalu e the effec)g of a each lo d, includi g the limi fig ef fects  ;

loss o power e cold sh down.

f on t abili to achi and the /

/

3. eate b trees d oting the b s hierarc cascadifg bus cordiguration o all bussps that power /

instr entation'and controls he oper tor would manipulate i in ing to cold shutdown. ,/

/

6. termine he annunciat s and alarms that would alert the \i perator to a failure f any of'the identifi,e~d busses. , ,

5 Dete ine the eff ts of a singlepowerduslosson'the abi ity to cont) ue in eagit particular.4hutdown path being '

u dd at the t me the bu loss occurs, Include the cascading ffects of bus lo , and conside'r alternate' indications'-

and cont s powere byunaffectedbussesthat'mayaidtJd/

, operat in the ev nt of a bus 4 css. Identify alternative /

/' shutd,osn paths gallable andMxisting peo'cedures for/

restoration the af f ectetd' bus.

6. Document the results 9ftheanalysis,providin[ i recommenfations of ha'rdware or procedural J:h'anges as i appropriate. /
The progr'ams descr ' ed in the responses yd this quesy on and to; uesti, ens 421.51 nd 421.52 wfll be c cted as a rombined 1

,ffort that wil be complej by Dece er, 1984 /

Analysis Of th: rep rt inf m ier show n sYtuationwherea single bus power failure would prevent plant personnel from achieving a safe shutdown -

condition. The results establisf that no single bus supplies power to all existing shutdown paths. The assignment of the instrument loads identified in this analysis is such that the loss of one bus would not prevent the minimum safety function from being performed.

ead d The failure of the buses id;;tificd ir T ble 2 ;f A;;:ndh t are annun-ciated and are displayed by the computer in the control room, thereb a%sa she giving the operator the knowledge of which power bus is lost. The 3 r:ti c that control room personnel will have knowledge of individual bus and/or circuit failures, and that the operator has alternat(instruments and shutdown paths available to achieve a cold shutdown condition.

( #dAIMUM 421.42-2 Amendment 5

1. "Celd %h w ~Poww stan %ffe vesma % [tilm Anal sts mc e,,_p$;" y Re Hopec,a Creek m Qcnerab

d

?

HCGS FSAR 4/84 OUESTION 421.42 (SECTION 7.5)

If reactor controls and vital instruments derive power form ,

. common electrical distribution systems, the failure of such t electrical distribution systems may result in an event requiring j operator action concurrent with failure of important  ;

instrumentation upon which these operator actions should be l-based. IE .9ulletin 79-27 addresses several concerns,related to the cbove subject. You are requested to provide information and a discussion based on each IE Bulletin 79-27 concern. Also, you  ;

are to:

1) Confirm that all a.c. and d.c. instrument buses that could affect the ability to achieve a cold shutdown condition were reviewed. Identify these buses.

2 ). Confirm that all instrumentation and controls required by emergency shutdown procedures were considered in review.

Identify these instruments and controls at the system level of detail.

3) Confirm that clear, simple unambiguous annunciation of loss of power is provided in the control room for each bus addressed in item I above. Identify any exceptions.
4) Confirm that the effect of loss of power to each load on each bus identified in item 1 above including ability to reach cold shutdown, was considered in the revtew.
5) Confirm that the re-review of IE Circular No. 79-32 which is required by Action Item 3 of Bulletin 79-27 was extended to include both Class 1E and non-Class 1E inverter suppited instrument or control buses. :dentify these buses or confirm that they are included in the listing required by Item 1 above.

,.

  • O RESPONSE } ' >

L ww.!c 41 es st..n =ts n.n (LGi-d (see RefewC' L)uA5 An analysis 4will b: conducted bas d on the Cen :a1 Clectrid iPfcach methed:1:gy for answering the con ' erns raised in IE Bulletin 79-27. Ihts Iethodology has been reviewed and approved cy the i

NRC via a report written for the ##p=6 project. jine .T nnadvi gy' pcv ide for sysLxmacic nd ev..pjenensi e anayysis o ens' e t t, 'n t event/ of a ngle pMer bus f ailure, s sficier ont ol e om in4(cator instru ents, d con rois xist * '

.ac eve cold thutdo . 1 jL- n tline the thodol y follo s:

. Re ew the Class 1 and non lass E bu es incl ding

/ i verter supplyi power ins ument tion an cont is i e

4 421.42-1 Amendment 5

'O HCGS fe/

DSER Open Item No. 208 ( DSER Se ction 7.7.2.2)

MULTIPLE CONTROL SYSTEM FAILURES ,;

The applicant is required to submit the analysis and its con- }

clusions concerning multiple control system failures to the ,

NRC for staf f review. This is scheduled for submittal during  ;

the fourth quarter of 1984.

RESPONSE l

%* re SPonse +o PsAK Ques + ion # s ..s-l has been te.ve'a ed to 'proVi d e d % e. i n br mo tio n regue3+Ed 0I80VC' l A e o py o f %e. So llow.*nj re. pert s are 0 n- o C b e d to % ;s re.sponse kr gou r uS*

Fa.* lures i) common Pouser/ c.ontrol 53 s4 em.s s ml ua+t o n Refo rTj %+ed : Aas u.s +, l 9 U4 0 C-O m mo e S e n so r- F a l t v. es C va lar+io n R e- P o r t , bm+cd: A 9 st,39: V.

I i

l 208-1

t

HCGS FSAR 4/84 i iB QUESTION 421.51 (SECTION 7.7)
  • intended to demonstrate the adequacy of safety systems in  ;

mitigating anticipated operational occurrences and accidents. {

Based on the conservative assumptions made in defining these }

" design bases" events and the detailed review of the analyses by )

the staff, it is likely that they adequately bound the consequences of single control system failures. To provide

  • assurance that the design basis event analysis for Hope Creek adequately bounds other more fundamental credible failures, r provide the following:

. (1) Identify those control systems whose failu're or malfunction could seriously impact plant safety.

(2) Indicate which, if any, of the control systems identified in (1) receive power from common power sources. The power sources considered should include all power sources whose failure or malfunction could lead to failure or malfunction of more than one control system and should extend to the effects of cascading power losses due to the failure of higher level distribution panels and load centers. *

(3) Indicate which, if any, of the control system identified in

  • 1) receive input signals from common sensors. The sensors considered should include common taps, hydraulic headers and impulse lines feedtnc pressure, temperature, level or other signals to two or more control systems.

(4) Provide Justification that any alfunctions of the control systems identified in (2) and (3) resulting from failures or

- T.alfunctions of the appitcable common power source or sensor

! Including hydraulic components are bounded by the analyses j in Chapter 15 and would not require action or response beyond the capability of operators or safety systems.

RESPONSE f pd 2,) waft TwoAn analys. c (see b

  • tsN1.11 b: 3conducted based on the General Electric 4

methodology for answering NRC concerns for common power source i

failures and common sensor or sensing line failures. This i

methodology, which received NRC concurrangg via reports for the Grand Gulf, Shoreham, and WNP-2 projects,3will Sc used for the Hope Creek project. gn met.0a01 gy t sys matt and cump ene tv and xam nes onte I sy tems nter tion to es bli t li. ti -cas eve ts. The onse ence of ing p er- ur or en ng-l'ne f ilur s wi be valu ed ith

' esp tt con ol rade syst ms d wi 1 en ure e1 iti g-

' cas eve ts a e b unde by ee nts nal ed i Cha er 5 . __

421.51-1 Amendmend 5 .,

l l

l

, HCGS FSAR 4/84 .

[A. on er So ce F lure

  • An tline f th method gy for e common power source {

lure alys follo  :  ;

1. Iden y all onsafet grade co 1 systems t at have i th poten ' 1 of a cting t ritical r,ead, tor
  • ramet s of wa r level, ressure, op power.

I

/ vel;

\

. Re 'ew thes control stems at the component ,

entify the ef cts of th oss of pow o each {

system omponen andthespbs[equentint ctions with i

oth compon s and ms.

. 3. Generat us treeg denoting th us hierarchy u case tng config6 cation of power bus at supply c onents o control syst ms under stgd .

4 Perfor a combined, effects analypt's. Evaluate fap re of each po,wer bus (1 pad center, motop ontrol nter, etc. starting h the lowest- el source common to .ultiple co .ol systems a orking up each,,,

bus t to the h est common po level. At eacV lov examine Jhe,j.effects of single bus fapufe and e conseq control.sy,getrces of cascading stems' compopet1'es . bus failures

' on'all /-

5. P ulate the limi' ting transiertt' events as a-iesult of he combine eff ects analyps'and compapeihese event,s-to those alyzed in Ch pter 15.
6. orm any additional transi.en calculations or analyses es'sary to enspr(the po lated limiting events ac bounded by ttose analyza in Chapter 15 7
7. ument the ta c'fts of th nalyses of com werf source fail , providi recommendation as j apprope e.

B. Commo ensor or sing Line Fail _

n outline the methodolo for the com sensor or sensin tne failure an sis follow -

. Identify . nonsafety e control sys s that have[

the ntial of af ing the criti reactor meters of w level, press , or power.

Identif 1 instrument sa ing lines an ensors i L. _.

uti ' ed by two or mor of these con systems.

)

421.51-2 Amendment 5-

l HCGS FSAR 4/04

}

( 3. Ana ze the ects failure a commo sensor a '

c plete p g or a illotine eak in ch of ese- t

, t ommon i trument ines. Ex ine the ffects f '

errone s signa on each ~ strumen and on ach func on (sce s, trips, ermissi sign , etc. hat co d be ac ated or re dered i perati . f, e i

4. xamine e interac ve eff ts amo g all s tems  :

affect by the c on se ing li e or se or fail e -

and t consequ tial .bined fects the cr icalj rea or param ers. ['

5. mpare t conse ences these ostula d events with th e anal ed in to en ce the conse ences the pp ,C apter tulate events ce bound by the esults f the napter event and to e re the p ulate events ould n requir actions /:

espons beyo he capa ilitle of the o rators,6rj the s ety sy ems. P,erform a addition trans'ent cal lation or anal,y'ses nec sary to e ure the p tulate imiting events te bounde by thos i nalyze in Chapter 15. j Docu nt the.results the anal' es of egmmon sen g\

li . or senp6r fail es and pr ide recoynendatio s asl ,

propria 4. f The pr rams de ribed i, the respor es to titis questj.o'n and to ques 'ons 421.,4 and 42f.52 vill " conducted as a ccmbined/j

' eff et that f ll be coe.pleted b'

/ .;

ecember /198 4. /

? -

/ -

\

that the- limits of minimum critical The power conclusion ratio (MCPR), peakof thWeYakMion" vessel an (d main steamline pressures, and peak fuel cladding temperature for the expected operational occurrence category of events would not be exceeded as a result of common power source fri h x - Although transient category events h:1 been#pistulated as a g restr f th W stud d the net effects [z: brc # positively determined to be less severe than those of the original cons e rvative , Chapter 15 events. It should be noted that useal the event-consequence logic of the Chapter 15 analysis,thidI' studf,Y thelogic chain from a but start specific source (e.g., a single bus failure) rather than a system condi-tion (e.g., feedwater runout). By ap oaching the study in this manner, a great deal of confidence can be plac d in the study conclusions. Ihr The soundness of the total p1 nt design,iss demonstrated by its f

h eing tolerant of these interactions. e< scaser we ec % s w c failuu s k co. wl as % s.

R_eyeneocas d

1. "Ca= wen Pourec/Co druly S s/ ems faldnas Evalud% Hope Creele qem sw , Po.u,c. serv,he, sub3. Mas co Au gsf e/.
t. "co. m s .u . A imm. s u w s Hog Rep d " q ,cu.h q m ,J ,h & 6 %

\, Pu.leth, ferde EluMe ud GAsfyjgyp3Aacad Msd ,

s M s C @ .5

. :.. e :: a .

t:

1 t

ATTACHMENT 5 t

I l

t

40310 ?4026'9368 COMMENTS 04 LAWRENCE LIVERMORE NATIONAL LABORATORY REPORT

\,

Ref erence: "Draf t Report: Site Spectra for the Hope Creek Site",

}I 1 prepared by the Lawrence Livermore National Laboratory ,

(LLNL) for the NRC. July 16, 1984. .

e i

Comments j Earthquake magnitudes 6.2,

1. 6.3 and 6.5 were used by the LLNL in preparing site specific spectra. These magnitudes are j unnecessarily high.
2. The distance weighting scheme has a data set with a magnitude bias. Mean magnitudes for the distance subsets are as follows: ,

0 - 10 5.07 10 - 15 5.04 15 - 20 5.43 This difference is probably severe enough to invalidate the weighting process.

3. Bernreuter prepared a report for NRC on the Wolf Creek Site (March 5, 1982). In this report he performed a study

' similar to Hope Creek. Using 30 soil site records (mean Magnitude, ML = 5.210.2, Table 6 of that report) he pro-duced median and 84th percentile spectra. These are plotted together with the median and 84th percentile base case spectra from the 27 sites used for Hope Creek in the July 16, 1984 report to the NRC on the attached figure (mean Magnitude ML = 5.210.3). The spectra on the attached figure dif fer f rom each other significantly. The choice of an almost entirely new data set for the recent report and rejection of records used for the Wolf Creek site study should be explained. This is especially important because of the implications being made for the higher recent spectral

values.
4. Hope Creek design spectra is applied at elevation 40 f t which is approximately 60 feet below grade. The site specific spectra proposed by the LLNL correspond to a control point j at or near the ground surface. See Appendix A for reconciliation.

i

?

k i

t PE5/23 i

. . - - . . - . - . , - . - . . , , ... ._,~ ,_,, - - - . -

r-

.-  ! se; w e4 0 .1 6 9 3 C 6 l

l l

- i i 8 I 3I I I l 4 i I I_  !

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,4

- _" l

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@ '4 NBC (0.29) i k rpj iblf #- \j D Creek *o 8A -

/ ~s nope 9

t -

s Creek 2

h

,, /

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d i i i il i i eii e i i ier o e i

.'01 .1 1. 10.

UNDPMPED NMtJRAL PERIOD (seconds) l COMPARISN OF SITE SPIIIFIC SPECITB FOR WOIE & HOPE CREEK ' SITES l Acceleration = 0.2g, Dartping = 5.0%

us 10 74 0 P.69 3 6 6 APPENDIX A EVALUATION OF HOPE CREEK SITE SPECTRA Introduction A meeting was held between the NRC (Geosciences Branch) and  !

PSE&G on July 30,.1984 to discuss the "Draf t Report Site '

l Spectra for the Hope Creek Site" prepared by Lawrence Livermore National Laboratory (LLNL) under contract from the NRC. The

  • LLNL developed the preliminary site specific spoetra to help assess the conservatism of the Hope Creek seismic input spectra i which are based on Regulatory Guide 1.60 spectra. The purpose of this~ discussion is to evaluate the impact of the LLNL study on the Hope Creek design spectra.

Lawrence- Livermore National Laboratory Proposed Hope Creek Site Spectra

The following three criteria were used by the LLNL in selecting
earthquake records for possible inclusion in the development of Hope Creek Spectra
a. The magnitude of earthquake corresponds to the ranges of mbLg = 5.25 10.5 and mbLg = 5.75 10.5
b. The distance between the epicenter and the recording station is approximately less than 20 km, and i
c. The recording stations are located at a " deep soil" site (soil depth greater than 200 f t). '

) The recommended preliminary median and one sigma spectra in both the horizontal and vertical directions (54 damping) for

! magnitude 5.25 are shown in Figures 5 and 12C of Reference 1.

, A direct comparison of the LLNL spectra (mbLa = 5.25) with j the Hope Creek design spectra without considefing the design earthquake control point elevation reveals the following:

  • Horizontal Spectra: LLNL spectral velocities are higher than those of the Hope Creek design spectra at periods below 0.35 seconds.
  • Hope Creek design spectra generally Vertical Spectra:

envelop the LLNL spectra except at periods below 0.06 seconds.

, A , comparison of the spectra of mbLg = 5.75 at about 30km distance

! with the Hope Creek design spectra reveals no exceedance.

t A-1

Ati lo U D P. 6 9 3 6 6 Evaluation of Hope Creek Design Spectra The Hope Creek site specific spectra prepared by LLNL were j developed from an ensemble of earthquake records obtained at or '

near the ground surf ace of sites whose top soil layers have an !lL I average shear wave velocity cf approximately 1,500 ft/sec. and an average compressional wave velocity of 4,800 f t/seo.

I' As stated in the Hope Creek FSAR Section 3.7.1, the omitzol point of the design earthquake input at the Hope Creek site is t not at the ground surface but at the level of the foundation  :

i structure in the f ree field which is approximately 60.0 f t below '

ground surface. Therefore, a direct comparison of the LLNL spectra with the Hope Creek spectra is not appropriate. To pre-sent a more direct comparison, equivalent spectra are developed from the Hope Creek design spectra in accordance with the following procedures:

A. Horizontal Earthquake

i. A free-field soil column as shown in Figure 1 is used in deconvolution analysis. The top of the soil column is truncated at elevation 65.0 ft ,
since the soil below this elevation (Kirkwood and

! Vincentown formation) has shear wave velocity of 4

approximately 1,850 f t/sec which are comparable to the site condition of the LLNL Study. FLUSH computer code is used for the soil column analysis (Reference 2).

ii. The Hope Creek design spectra are applied at elevation 40.0 ft where the pcwer block founda-tion is located and the response spectra (54 damping) at the top of the soil column are generated.

Figure 2A shows the comparison plots between the regenerated spectra at the top of the soil column and the input spectra delineating amplification effect. Figure 2B shows comparison between the LLNL spectra (mbLg = 5.25) and the regenerated design spectra.

iii. In all cases, the equivalent regenerated Hope Creek design spectra at elevation 65.0 ft envelop the proposed LLNL spectra. Therefore, it is concluded j that the Hope Creek design basis spectra are adequate.

B. Vertical Earthquake

i. The soil column as shown in Figure 1 is used for the evaluation of the Hope Creek vertical site spec-tra. Due to the presence of water table near the ground surface, the compression wave velocity of soil is appropriately adjusted for the vertical analysis.

A- 2 l

l

m 10 'es 0 ? 69 3 66 ii. The Hope Creek design response spectra are applied i at elevation 40.0 ft of the soil column and the corresponding response spectra are regenerated at elevation 65.0 f t.

iii. Figure 3A-shows the comparison plots between the  !

velocity response spectra at the top of the soil column and the input spectra delineating amplifi- ,

cation effect. Figure 3B provides the comparison  ;

between LLNL spectra and the equivalent regenerated -

Hope Creek spectra. .The Hope Creek spectra envelop the LLNL spectra. Therefore, it is concluded that the Hope Creek input spectra are adequate. .

Conclusion Based on the above, it is concluded that the Hope Creek seismic input criteria meet or exceed the criteria proposed by LLNL.

References:

1. "Draf t Report: Site Spectra for the Hope Creek Site",

! prepared by the Lawrence Livermore National Laboratory for the Nuclear Regulatory Commission, July 16, 1984.

2. Lysmer, J. , Udaka, T. , Tsai, C.F., and Seed, H.B., " FLUSH -

A Computer Program for Approximate 3-D Analysis of Soil-Structure Interaction Problems", Report No. EERC 75-30, College of Engineering, University of California, Berkeley, California, November, 1975.

2 i

i i

?

l PE5/22 A- 3

n i o 8 4 0 ?. 6 9 3 0 8 .

I Elevation (ft.) Response Spectra 65.0 n  ;

d NRC Design Spectra R.G. 1.60 40.0 4

Reference:

FSAR Figure 3.7-5 and 3.7-6 i

1

-300.0 ,,,,,,,,,

Figure 1 Soil Column for Evaluation of Hope Creek Site Spectra A-4 l

PEr:Do sa ssconos <

GC' .

. ,. *M.

. .. h. , 8 .

  • r *. , , ,' t. o P ....

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doico FICURE 2A RESPONSE SPECTRA COhPARISON asse suas est e AT TOP OF FRE FIELD aweve. HORIZONTAL. SSC. DAMPING PERIDO IN $ ECON 05 O. O s , , , o.,0q , , g e , , , og s, , , ,e , o , , , p , , , ,1,0 ,,,Q,,,f%

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l - FIGWRE 28 RESPONSE SPECTRA C0hPARISON i ampe som use e AT TOP OF FREE FIELD I

em3_'an.se*

  • HORIZONTAL. SSE. $X OAhPING l

l A-5 i

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FIGURE 3A s'lo ' de'soo samma euen see e RESPONSE SPECTRA Con #ARISON AT TOP OF FREE FIELO

.".5.,.**** ,VERTICAL. SSE. $% DAMPING Ptanoo rw seconos

. * ,85

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    • .S.,T,** * *

,,, AT TOP OF FREE FIELO VERTICAL. SSE. Elt DaedPING A-6

  • l 1

_ __ _ . - - - . - - . _--. -- --- - - - - -- ~ - - - - - - . - -~ - -~- ---

i t

HCOS I e

gy 1 *

/ DSgR Open Item No. 207 (DSER section 7.7.2.1) i i

HELBs AND CONSEQUENTIAL CONTROL SYSTEN FAILURES . ,

The applicant is required to submit the analysis and its con-clusions concerning HEL8s and consequential control system .

failures to the NRC for staff review. This is scheduleu #,

for submittal during the fourth quarter of 1984. i

  • \

l

RESPONSE

%e. response to Fs AC ouestion VJ/.ra ha s b een t-evhed To p r o v o'd e He in foren a'fia n rej u e sf e o/ ca. h o VC.

e +c %b *'.s r*SPon'*

% e fo //ou>iny :r .por+ i.s u tfo-eh ed for y oa.e ase L.io t 0Vto.Nf 6On N l Y i

o} }ll3'h L'n e(3 y f a n )u.r e An aIy sis,' ba+ccl.' Aug u 27. I N Y 1

i 4

I e

207-1

HCGS FSAR 4/84 i P

OUESTION 421.52 (SECTION 7.7) .

)

, If control systems are exposed to the environmental resulting i

, from the rupture of reactor coolant lines, steam lines, or feedwater lines, the control systems may malfunction in a manner which would cause consequences to be.more severe than assumed in i safety analyses. I&E Information Notice 79-22 discusses certain 1

non-safety grade control equipment, which if subjected to the adverse environment of a high energy line break, could impact the l'

l safety analyses and the adequacy of.the protection functions  :

performed by the safety-related systems. i 1

The staff is concerned that a similar potential may exist at light water facilities now under construction. You are, therefore, requested to perform a review per the I&E Information Notice 79-22 concern to determine what, if any, design changes or operator actions would be necessary to assure that high energy line breaks will not cause control system failures to complicate the event beyond the FSAR analyses. Provide the results of your ,

review including all identified problems and the manner in which

you have resolved them, i

The specific " scenarios" discussed in the above referenced j Information Notice are to be considered as examples of the kinds  :

of interactions which might occur. Your review should consider (,

analogous interactions as relevant to the BWR design.

srsposst s 1)wa (ses NW' An analysis3 u;11 5: conducted based on the General Elertet:

methodolcqy for answering the concerns raised in IE Information 6 Notice 79-22. The NRC has concurred with this metnodology vta its review prepared for the Shoreham and Grand Gulf proje: s.

he m thodoiggy assu.,es a systemacic, c .pcenenst- . a.,alysjs or 1

, igh energy ine creaks and e conse ential - trol sy tems

! a' ures. An out (ne of is meth ology folfows:

/

. entify 4 1 non fety con .ol-gra systems d components .e j within,tnese s tems who failur could af'_et the c- ical ,

reac e para ters of ter lev , pressu , and po r.

t. ablis assumpti s and teria fo determin' g high energy ines an ipe be locatio and for valuat' g the cons uences pipe b aks. Pip whip, je imping .ent,
an environ. ntal pa meters su as high empera re, hic i essure, nd high umidity w 1 be cone deced the i analysi .

. Id ify fro appropria plant de ings t se p1

\s___ cations which hi -energy 1 es wit postu ed break y, ,; g7,411.52-1 . Amendment 5 i w C- .

H+L.,, ' -

l 1

HCGS FSAR 4/84 loc ions oexist with non fety com ents o control- rade s tems.

Cond a pl t walkd n to ver' y the cations control 4.

sys em com nents a to det ine t r proxim' y to hi e egy li e break ocation

5. Exami , one a a time, igh-e gy line eaks an est ish th worst-c e comb
  • ed effee of eac reak and th conse ntial cofitrol-s tem fail es.
6. Ensure hat the onsequ ces of.th e pipe- esk eve . are boun d by th e of t events lyzed i Chapter jk .

7'.' oose tw or mor of the w t-case enarios,a6d postujate <

for eac a wors ase add tonal fa ure in afsafety-rela 3ei, mitiqp ing sy . Ensur thatthyconseque,n/ ces of f

th ge.new events do n fall ou ide the , bounds o/of the f the  ;

ca'pabiliti,es of saf systJm or the c sequence (

aly::ed events hapter 1 .

e analyS's of th

/

Interact ns t

i 8. Doc ..ent the esults of l ween hig energy 1' e breaks,,a'nd conte , sys te.T and recommend ctions be taken s appro late.

i

$The progra descri)6d in the "_sponses o this estion a to j 421. 11 be c, ducted a combt d_ .s

'Questio effor hatwi[42an,d421.51be comppd by DeSemb,eL._.13 b-'"~~

The

.(nalysis . m. u , (Appm.dM described each of the postulated HELB event s and their limiting effects on the reactor parameters. In most cases, A g g ts of the postulated HELB/ control systems failures e

events p 4 1ess severe than the Unacceptable Results for Incidents of Moderate Frequency - Anticipated Operational Transients presented in Cgapter 15. In all cases, the ef fects of the postulated events

  1. g/$oudded a.ee by the Unacceptable Results for Limiting Faults - Design Basis (Postulated) Accidents presented in y Chapter 15. It se sos that sa fe reactor shutdown is assured for all :=.t:; postu-conclu lated'j gd ia and the consequences of these postulated events would w

not result in any significant risk to the health and safety of the public.

s 89fE2 E9C.E l, " N$ k Le$ LAe hak/ code ( Sgslew. Slags Ans/ 3,3," yo 5

Greek $Mt Sto.NeN hhhc hmee Elecke aed go.5, Au3usfl484 j

. m- - 421.52-i Amendment 5 (

71, .

d

-w - - - ,.,

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