ML20097B349
| ML20097B349 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek, Susquehanna, 05000000 |
| Issue date: | 08/30/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17054D526 | List: |
| References | |
| NUDOCS 8409140006 | |
| Download: ML20097B349 (241) | |
Text
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Pudic Service O PS G 3rnpany Dectnc and Gas 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitt; General Manager Nuclear Assurance and Regulation August 30, 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 1/
\_ #
Gentlemen: ?
HOPE CREEK GE ATING STATION DOCKET NO.57-354 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 id'entifies Draft SER \ ,
- k. Sections not yet provided.
R l The Energy People. p
- f os nmin m
4-
- r e
~
. Director-of Nuclear.
- Reactor: Regulation 2 8/30/84 l
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 revised FSAR Sectionsc1.8.1.26, 6.2.5.2.5, 7.6.1.4.3, and Figure 8.3-16.
A_. signed original of the required affidavit is provided to
' document'the submittal:of these items.
~
iShould'you1have any questions or require any additional information on.these open items, please contact us.
Very truly yours,
("
< hl I Attachments / Enclosure I
.C D. H, Wagner USNRC Licensing Project Manager W. H. Bateman USNRC Senior Resident Inspector FM05 1/2 r
O
k 3- e UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION t - DOCKET NO. 50-354 PUBLIC SERVICE. ELECTRIC AND' GAS COMPANY Public.. Service Electric and Gas Company ' hereby submits the
- enclosed Hope Creek Generating Station Draft Safety Evalua-tion: Report open ' item responses and revised FSAR Sections 1.8.1.~26, 6.2.5.2.5, 7.6.1.4.3, and Figure 8.3-16.
'The matters set f orth in this submittal are true to the best of my knowledge, information, and belief .
Respectfully submitted, Public Service Electric and Gas Company i -
By: M/A /
Yhdmas7 ~.M ih Vice Presi nt.-
Engineering and Construction Sworn to and subscribed
,bef ore me, a Notary Public of.New Jersey, this 8/8 day of August l1984.
k dlN 0 97 -
, DAVID K. BURD NOTARYPUBUC OF NEW JERSEY
. , My Comm. Empires 10-23 85 1
.it i GJ02/5 L_
IRTE: 8/30/84 ATTAOMENT 1 ,
DSER- R. L. MIT11 'IO OPEN SECTION A. SOINENCER '
s ITEM IGEER SUBJECT STA'IUS IETIER IRTED
.l'- 2.3.1 Design-basis taperatures for safety- Ccuplete 8/15/84 related auxiliary systes .
2a 2.3.3 Accuracies cf meteorological Ccmplete 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 cf meteorological Ccmplete 8/15/84 measurements (Rev. 2) 2d 2.3.3 Accuracies of meteorological Ccmplete 8/15/84 measurements (Rev. 2) 1
- 3a 2.3.3- Upgrading cf onsite neteorological Cmplete 8/15/84 measurements progra (III.A.2) (Rev. 2) 3b 2.3.3 Upgrading cf onsite meteorological Ccmplete 8/15/84 measurements program (III.A.2) (Rev. 2) ,
3c. 2.3.3 Upgrading cf onsite meteorological NRC Action measurements progra (III.A.2) 4 2.4.2.2 Ponding levels Ccmplete 8/03/84 Sa 2.4.5 Wave impact and rurup on service Ccmplete 8/20/84 Water Intake Structure (Rev. 1)
Sb 2.4.5 Wave impact and runup on service Ccmplete 8/20/84 water intake structure (Rev. 1)
Sc 2.4.5 Wave impact and rurup on servi Couplete 7/27/84 water intake structure 5d 2.4.5 Wave impact and runup on service Conplete 8/20/84 l water intake structure (Rev.'1) ]
6a 2.4.10 Stability cf erosion protection Ccmplete 8/20/84 i structures 6b 2.4.10 Stability cf erosion protection Canplete 8/20/84-structures 6c- 2.4.10 Stability cf erosion protectim tcmplete 8/03/84 structures M P84 L 80/12 1-gs
_*+v y _ _ , -r%-et--w=-* -,:w- cvq,-y-.a
ATTACHMENT 1 (Cont'd)
DSER R. L. MITTL TO OPEN SECTION A. SCHWENCER ITEM NUMBER SUBJECT 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 sittk Cmplete 8/3/84 8 2.5.2.2 Choice of maximtn earthquake for New Caq 1.ete 8/15/84 England - Piedmont Tectonic Province 9 2.5.4 Soil danping values Complete 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 Combinaticn 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 botts Ccmplete 6/1/84 sands 15 2.5.4 Tabulations of shear moduli Cmplete 6/1/84
-16 2.5.4 Drying and wetting effect on Cmplete 6/1/84 Vincentown 17 2.5.4 ~ Power block settlement monitoring Complete 6/1/84 18 2.5.4 Maxim m earth at rest pressure Complete 6/1/84 coefficient 19 2.5.4 Liquefaction analysis for service Conplete 6/1/84 water piping 20 ' 2.5.4 Explanation of observed power block Complete 6/1/84 settlement
-21 2.5.4 Service water pipe settlement records Complete 6/1/84 22 2.5.4 Cofferdam stability Cmplete 6/1/84 M P84 80/12 2 - gs
ATTACIMENT 1 (Cont'd)
DSER R. L. MITIL E)
OPEN -SECTICN - A. SOlWENCER ,
I'ITM NUMBER SUBJECT STATUS LEITER DATED 23 2.5.4. Clarification of ESAR Tables 2.5.13 Cmplete 6/1/84 and 2.5.14 24 2.5.4 Soil depth nodels for intake Cmplete 6/1/84 structure
-25 2.5.4 Intake structure soil modeling Carplete 8/10/84 26 2.5.4.4 Intake structure sliding stability Ccmplete 8/20/84 27 2.5.5 Slope stability Carplete 6/1/84
.28a 3.4.1 Flood protection Corplete 8/30/84 (Rev. 1) 28b 3.4.1 Flood protection Cmplete 8/30/84 (Rev. 1) 28c 3.4.1 Flood protection Cmplete 8/30/84 (Rev. 1) 28d> -- 3.4.1 Flood protection' Canplete 8/30/84 (Rev. 1) 28e 3.4.1 Flood protection Carplete 8/30/84 (Rev. 1) 28f 3.4.1 Flood protection Cmplete 7/27/84 28g- 3.4.1 Flood protection Cmplete 7/27/84 29 3.5.1.1 Internally generated missiles (cutside Cmplete 8/3/84 containment) (Rev. 1) 30 3.5.1.2 Internally generated missiles (inside Closed 6/1/84 containnent) (5/30/84-Aux.Sys.Mtg.)
31 3.5.1.3 'Ibrbine missiles Cartplete 7/18/84
- 32 3.5.1.4 ' Missiles generated by natural phenarena Carplete 7/27/84 33 3.5.2 Structures, systems, and cmponents to Carplete 7/27/84 be protected frcm externally generated missiles M P84 80/12 3 - gs
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.n ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL TO
' OPEN SECTION A. SOINENCER ITEM NUMBER SUBJECT STATiS IEITER DATED 34 3.6.2 Unrestrained whippirg pipe inside Cmplete 7/18/84 containment 35 3.6.2 ISI program for pipe welchs in Ccmplete 6/29/84 break exclusion zone 36 3.6.2. Postulated pipe ruptures Cmp]ete 6/29/84 37 3.6.2 Feedwater isolaticm check valve Cmplete 8/20/84 cperability 38 3.6.2 Design of pipe rupture restraints Cmplete 8/20/84 39 3.7.2.3 SSI analysis results usirg finite Ccmplete 8/3/84 element method and elastic half-space approach for contairment structure 40 3.7.2.3 SSI analysis results usirg finite Ccriplete 8/3/84 element method and elastic half-spa approach for intake structure 41 3.8.2 Steel contairment bucklirg 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/LOCA pool dynamic loads Ccmplete 6/1/84 44 3.8.3 ACI 349 deviations for internal Ccmplete 6/1/84 structures 45 3.8.4 ACI 349 deviations for Category I Cmplete 8/20/84 structures (Rev. 1) 46 3.8.5 ACI 349 deviations for foundations Ccmplete 8/20/84 (Rev. 1) 47 3.8.6 Base mat response spectra Cmplete 8/10/84 (Rev. 1) 48 3.8.6 Rocking time histories Ccmplete 8/20/84 (Rev. 1) ..l i
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M P84 80/12 4 - gs l
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A'lTACHMENT 1 (Cont'd) !
'DSER R. L. MITIL 'IO OPEN SECTION- A. SOlWENCER ITEM NUMBER SUBJECT STA'IUS LETIER DATED
.49 3.8.6 Gross concrete section Cmplete 8/20/84 (Rev. 1) 501 3.8.6' Vertical floor flexibility response Conglete 8/20/84 spectra (Rev. 1) 51 3.8.6 Cmparison cf Bechtel independent Cmplete 8/20/84 verification results with the design- (Rev. 2) basis results 52 -3.8.6 Ductility ratim due to pipe break Cmplete 8/3/84
-53 3.8.6 Design cf seismic Category I tanks Ccmplete 8/20/84 (Rev. 1)
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'54 3.8.6 Ccnbination of vertical responses Cmplete 8/10/84 (Rev. 1)
.55 3.8.6 Torsional stiffness' calculation Cmplete 6/1/84 f
3.8.6 Drywell stick model development Ccmplete 8/20/84 (Rev. 1) 57 3.8.6 actational 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 moment cf reactor Cmplete 8/20/84 building foundation' mat (Rev. 1) 60 3.8.6 BSAP element size limitations Cmplete 8/20/84 (Rev. 1) 61 3.8.6 Seismic modeling cf drywell shield Complete 6/1/84 well 62 '3.8.6 Drywell shield wall boundary Cmplete 6/1/84 conditions r- 63 3.8.6 Reactor building dme boundary Ccmplete 6/1/84 conditions M P84 80/12 5 - ga w __ .. _ .- . _ . - _ _ _ _ _
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ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL 'IU ,
OPEN SECTION A. SOINENCER ITEM NLNBER SUBJECT STATUS LETTER IRTED 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 Cmplete 6/1/84 drop 66 3.8.6 Impedan analysis for the intake Cm.plete 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 slidity and Cmplete 6/1/84 overturning of drywell shield wall 70 3.8.6 Seismic shear for distribution in Cmplete 6/1/84 cylinder wall 71 3.8.6 Overturning cf 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 dme trodel load inputs Cmplete 6/1/84 74 3.8.6 Tornado depressurization C m plete 6/1/ 84 75 3.8.6 Auxiliary building abnormal pressure Cmplete 6/1/84 76 3.8.6 Targentizi shear stresses in drywell Canplete 6/1/84 shield wall and the cylinder wall 77 3.8.6 Factor cf safety against overturning Cmplete 8/20/84 of intake structure (Rev. 1) 78 3.8.6 Dead 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
ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL TO
-OPEN SECTION A. SOMNCER ITEM NUMBER SUBJECT STATUS IEITER DATED 80- 3.8.6 Torus fluid-structure interactions C mplete 6/1/84 81 3.8.6 Seismic displacement cf torus Cmplete 8/20/84 (Rev. 1) 82 3.8.6 Review cf seismic Category I tank Catplete 8/20/84 design (Rev. 1) 83 3.8.6 Factors cf safety for drywell Ccmplete 6/1/84 buckling evaluation 84 3.8.6 Ultimate capacity d containment Ccmplete 8/20/84 (materials) (Rev. 1) 85 3.8.6 toad combination consistency Cmplete 6/1/84 86 3.9.1 Cmputer code validation Cmplete 8/20/84 87 3.9.1 Information on transients C mplete 8/20/84 88 3.9.1 Stress analysis and elastic plastic Cmplete 6/29/84 analysis 89 3.9.2.1 Vibration levels for NSSS piping Ccmplete 6/29/84 systems 90 3.9.2.1 Vibration monitoring program during Ccmplete 7/18/84 testing 91 3.9.2.2 Piping supports and anchors Ccmplete 6/29/84 92 3.9.2.2 Triple flued-head containment Carplete 6/15/84 penetrations 93 3.9.3.1 Inad cmbinations and allowable Cmplete 6/29/84 stress limits 94 3.9.3.2 Desi@ of SRVs and SRV discharge Ccmplete 6/29/84 Pi ping M P84 80/12 7 - gs
y
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ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL TO L OPEN SECTION A. SOIMNCER I'IEM NUMBER SUBJECT STATUS IEITER DATED 95 .3.9.3.2 Fatigue evaluation m SRV piping Cmplete 6/15/84 and IDCA downcomers
% 3.9.3.3 IE Infonnation Notice 83-80 Cmplete 8/20/84 (Rev. 1) 97 3.9.3.3 Buckling criteria used for cmponent Cmplete 6/29/84 supports 98 3.9.3.3 Design cf bolts Cmplete 6/15/84 99a 3.9.5 Stress categories and limits for Cmplete 6/15/84 core support structures 99b 3.9.5 Stress mtegories and limits for Cmplete 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) Cmplete 8/20/84 101 3.9.6 PSI and ISI programs for punps and Cmplete 8/20/84 valves 102 3.9.6 Isak testing af pressure isolation Cmplete 6/29/84 valves 103al 3.10 Seismic and dynamic qualification of Cmplete 8/20/84 mechanical ard electrical equipment 103a2 3.10 Seismic ard dynamic qualification cf Cmplete 8/20/84 i mechanical and electrical equipment j 103a3 3.10 Seismic and dynamic qualification of Cmplete 8/20/84 mechanical and electrical equipment 103a4 3.10 Seismic and dynamic qualification cf Cmplete 8/20/84 mechanical and electrical equipnent ;
M P84 80/12 8 - gs 1
1
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ATTACHMENT 1 (Cont'd)
DSlix R. L. MITTL *10 OPEN SECTION A. SOMENCER i ITEM NUMBER SURIECT STARIS IErrER DATED 103a5 3.10 Seismic and dynamic qualification of C aplete 8/20/84 mechanical and electrical equipnent 103a6 3.10 Seismic and dynamic qualification of Complete 8/20/84 mechanical and electrical equipnent 103a7 3.10 Seismic and dynamic qualification of Couplete 8/20/84 mechanical and electrical equipnent 103bl 3.10 Seismic and dynamic qualification of Cmplete 8/20/84 mechanical and electrical equipnent 103b2 3.10 Seismic and dynamic qualification of Caplete 8/20/84 mechanical and electrical equipnent 103b3 3.10 Seismic and dynamic qualification of Couplete 8/20/84 mechanical and electrical equipnent 103b4 3.10 Seismic and dynamic qualification of Caplete 8/20/84 mechanical and electrical equipnent 103b5 3.10 Seismic and dynamic qualification of Cmplete 8/20/84 mechanical and electrical equipnent 103b6 3.10 Seismic and dynamic qualification of Complete 8/20/84 mechanical and electrical equipnent 103c1 3.10 Seismic and dynamic qualification of Caplete 8/20/84 mechanical and electrical equipnent 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 Conplete 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
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l ATDOMENT 1 (Cont'd)
DSER R. L. MITIL TO OPEN SECTICE A. SODENCER ITEM NLMBER SUBJECT STATUS LETIER DATED .
105 4.2 Plant-specific mechanical fracturing Cmplete 8/20/84 analysis (Rev. 1) 106 4.2 Applicability cf seismic andd IDCA Ccrplete 8/20/84 loading evaluation (Rev. 1) 107 4.2 Minimal post-irradiation fuel Cmplete 6/29/84 surveillance program 108 4.2 Gadolina thermal conductivity Ccmplete 6/29/84 equation 109a 4.4.7 TMI-2 It s II.F.2 Cmplete 8/20/84 109b 4.4.7 7MI-2 It s II.F.2 Cmplete 8/20/84 110a 4.6 Functional design cf reactivity Ccmplete 8/30/ 84 control systems (Rev. 1) 110b 4.6 Functional desicp cf reactivity Cceplete 8/30/84 control systems (Rev. 1) lila 5.2.4.3 Preservice inspection program Cmplete 6/29 / 84 (cmponents within reactor pressure boundary) lllb 5.2.4.3 Preservice inspection program Ccmplete 6/29/84 (cmponents within reactor pressure boundary) lllc 5.2.4.3 Preservice inspection prcgram Ccmplete 6/29/84 (components within reactor pressure boundary) ll2a 5.2.5 Reactor coolant pressure boundary Ccmplete 8/30/84 leakage detection (Rev. 1) 112b 5.2.5 Reactor coolant pressure boundary Cmplete 8/30/84 leakage detection (Rev. 1)
M P84 80/1210 - gs
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I ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL 70 f OPEN SECTION A. SCHENCER ',
-ITEM NLMBER SUBJECT STATUS LETIER IRTED I
ll2c 5.2.5 Reactor coolant pressure boundary Cmplete 8/30/84 leakage detection (Rev. 1)
Reactor coolant pressure boundary 8/30/84 i ll2d 5.2.5 Cmplete leakage detection (Rev. 1) l 112e 5.2.5 Reactor coolant pressure boundary Cmplete 8/30/ 84 leakage detection (Rev. 1) 113 5.3.4 GE procedure applicability Cmplete 7/18/84 114 5.3.4 Capliance with NB 2360 cf the Sumer Cmplete 7/18/84 1972 Addenda to the 1971 ASME Code 1 15 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 test data for base Cmplete 7/18/84 materials as used in shell ocurse No. I 117 5.3.4 . Cmplian with NB 2332 of Winter 1972 Cmplete 8/20/84 Addenda cf the ASE Code 118 5.3.4 Lead factors and neutron fluen for Cmplete 8/20/84 surveillance capsules 119 6.2 1MI item II.E.4.1 Cmplete 6/29/84 120a 6.2 TMI Item II.E.4.2 Cmplete 8/20/84 120b 6.2 1MI Item II.E.4.2 Cmplete 8/20/84 121- 6.2.1.3.3 Use cf 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)
M P84 80/1211 - gs
ATTACHMENT 1 (Cont'd)
DSER R. L. MITTL 10 OPEN- SECTION A. SOD elCER ITEM NUMBER SUBJECT STATUS IErrER IRTED 124a 6.2.1.5.1 RW shield annulus analysis Cmplete 8/20/84 (Rev. 1) 124b 6.2.1.5.1 RW shield annulus analysis Cmplete 8/20/84 (Rev. 1) 124c- 6.2.1.5.1 RW shield annulus analysis Cmplete 8/20/84 (Rev. 1) 125 6.2.1.5.2 Design drywell head differential Couplete 6/J5/84 pressure 126a 6.2.1.6 Redundant position indicators for Cmplete 8/20/84 vacute breakers (and control rom alarms) 126b 6.2.1.6 Redundant position indicators for Cm plete 8/20/84 vacuta breakers (and control rom alarms) 127 6.2.1.6 Operability testing of vacuta breakers Cmplete 8/20/84 (Rev. 1) 128 6.2.2 Air ingestion Complete 7/27/84 129 6.2.2 Insulation ingestion Conplete 6/1/84 130 6.2.3 Potential bypass leakage paths Cmplete 6/29/84 131 6.2.3 Idministration of secondary contain- Complete 7/18/84 ment openings 132 6.2.4 Containment isolation review Cmplete 6/15/84 133a 6.2.4.1 Containment purge system Cmplete 8/20/84 133b 6.2.4.1 Containment purge system Cmplete 8/20/84 133c 6.2.4.1 Containment purge system Cmplete 8/20/84 M P84 80/12 12- gs L
ATTACEMENT 1 (Cmt'd)
DSER R. L. MITIL TO ,.)
OPEN SECTIOi A. SOMNCER '
ITEM NLMBER SUBJECT STATUS IETIER DATED i
134 6.2.6 cmtairnent leakage testing Cmplete 6/15/84 1351 6.3.3 IPCS and LPCI injection valve Cmplete 8/20/84 interlocks 136 6.3.5 Plant-specific LOCA (see Section Ccmplete 8/20/84 15.9.13) (Rev. 1) 137a 6.4 Cmtrol rom habitability Cmplete 8/20/84
- 137b 6.4 Control roon habitability Cmplete 8/20/84 137c 6.4 Control roon habitability Cm plete 8/20/ 84 138 6.6 Preservice inspection program for Canplete 6/29/84 Class 2 and 3 camponents 139 6.7 MSIV leakage control system Cceplete 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 Canplete 8/15 / 84 (Rev. 1) 140c 9.1.2 Spent fuel pool storage Ccmplete 8/15/84 (Rev. 1) 140d 9.1.2 Spent fuel pool storage Canplete 8/15/84 (Rev. 1) 141a 9.1.3 Spent fuel cooling and cleanup Ccmplete 8/30/84 system (Rev. 1) 141b 9.1.3 Spent fuel cooling and clearup Canplete 8/30/84 systern (Rev. 1)
'141c 9.1.3 Spent fuel pool cooling and cleanup Ccmplete 8/30/84 system (Rev. 1)
M P84 80/1213 - gs
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ATTACHMENT 1 (Cont'd)
DSER- R. L. MITIL 10 ,
OPEN SECTICN A. SODENCER
-ITEM NtNBER SUBJECT STATUS LETIER DATED Spent fuel rool cooling ard cleamp C mplete 141d 9.1.3 8/30/84 system (Rev. 1) 9.1.3 Spent fuel pool cooling ard cleamp Cm plete 8/30/84 141e
- system (Rev. 1) 141f 9.1.3 Spent fuel pool cooling ard cleamp Cmplete 8/30/ 84 system (Rev. 1) 141g 9.1.3 Spent fuel pool cooling and cleanup Cmplete 8/30/84 system (Rev. 1) 142a 9.1.4 Light load handling syste (related Cmplete 8/15/84 to refueling) (Rev. 1) 142b 9.1.4 Light load handling system (related Cm plete 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 systm Cmplete 8/15/84 (Rev. 1) 144b 9.2.1 Station service water system Cmplete 8/15/84 (Rev. 1) 144c' 9.2.1 Station service water syst m Cmplete 8/15/84 (Rev. 1) 145 9.2.2 ISI program ard functional testing Closed 6/15/84 of safety and turbine auxiliaries (5/30/84- l cooling systems Aux.Sys.Mtg.) I l
146 9.2.6 Switches and wiring associated with closed 6/15/84 ;
HPCI/RCIC torus suction (5/30/84- 1 Aux.Sys.Mtg.) l M P84 80/1214 - gs
ATTACthENT 1 (Cont'd)
DSER R. L. MITIL 'IO OPEN SECTION A. SODENCER ITEM NUMBER SUR7ECT STA'IUS IETIER DATED 147a 9.3.1 Cmpressed air systems Cmplete 8/3/84 (Rev 1) 147b .9.3.1 Cmpressed air systems Ccmplete 8/3/34 (Rev 1) 147c 9.3.1 Ccmpressed air systems Ccmplete 8/3/84 (Rev 1) 147d 9.3.1 Cmpressed air systems Cmplete 8/3/84 (Rev 1) 148 9.3.2 Post-accident sanpling system Ccmplete 8/20/84 (II.B.3) 149a 9.3.3 Equipment and floor drainage syste Ccmplete 7/27/84 149b 9.3.3 Equipment and floor drainage system Cmplete 7/27/84 150 9.3.6 Primary contaiment istrument gas Cmplete 8/3/84 system (Rev. 1) 151a 9.4.1 Control structure ventilation syste Ccmplete 8/30/84 (Rev. 1) 151b 9.4.1 Control structure ventilation systm Cmplete 8/30/84 (Rev. 1) 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- Ccmplete 8/30/84 tion syste (Rev 2) 154 9.5.1.4.a Metal rocf deck construction Cmplete 6/1/ 84 classificiation i 155 9.5.1.4.b ongoing review cf safe shutdown NRC Action capability
-156 9.5.1.4.c Ongoing review cf alternate stutdown NRC Action capability M P84 80/12 15 - gs L
ATTACHMDrr 1 (Cont'd)
DSER R. L. MITIL 10 OPEN SECTICH A. SOfWENCER
-ITEM NUMBER SUBJECT STATUS LETTER DPGED 157 9.5.1.4.e Cable tray protection Cmplete 8/20/84 158 9.5.1.5.a Class B fire detection system Caplete 6/15/84 159 9.5.1.5.a Primary and secondary power supplies Cmplete 6/1/84
-for fire detection systen 160 9.5.1.5.b Fire water ptmp capacity Cmplete 8/13/84 161 9.5.1.5.b Fire water valve supervision Ccmplete 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 Ca plete 6/1/84 164 9.5.1.6.e Remote shutdown panel ventilation Cmplete 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 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 Cmmitment Cmplete 6/29/84 173 13.5.2 Procedures covering abnormal releases Complete 6/29/84 of radioactivity M P84 80/12 16- gs
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ATTACINENT 1 (Cont'd)
DSER R. L. MITIL 10 OPEN SECTION A. SODelCER ITEM NLMBER SUBJECT STATUS IEITER IRTED 174 13.5.2 Resolution explanation in ESAR 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 progra Catplete 8/13/84
-176b 14.2 Initial plant test progra Ccuplete 8/13/84 176c 14.2 Initial plant test progra Cmplete 7/27/84 176d 14.2 Initial plant test program Cmplete 8/24/84 (Rev. 2) 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 trogram Cmplete 8/20/84 176h 14.2 Initial plant test program Conplete 8/13/84 1761 14.2 Initial plant test program Cmplete 7/27/84 17 7 15.1.1 Partial feedwater heating Cmplete 8/20/84 (Rev. 1) 178 15.6.5 IDCA resulting fran spectnan of NRC Action postulated piping breaks within RCP 179 15.7.4 Radiological consequences cf fuel NRC Action handling accidents 180 15.7.5 Spent fuel cask drcp accidents NRC Action 181 15.9.5 1MI-2 Item II.K.3.3 Ccmplete 6/29/84 182 15.9.10 INI-2 Item II.K.3.18 Ccmplete 6/1/84 183 18 Hope Creek DCRDR Cmplets 8/15/84 M P84 80/12 17 - gs
ATTAGMENT 1 (Cont'd)
DSER R. L. MITIL 10 OPEN SECTION A. SGENCER !
Ill!M NtMBER SUBJECT STATUS ETIER DATED ;
i Failures in reactor vessel level Cmplete 184 7.2.2.1.e 8/1/84 sensing lines (Rev 1)
-185 7.2.2.2 Trip system sensors and cabling in Ccmplete 6/1/84 turbine building 186 7.2.2.3 Testability cf plant protection Cmplete 8/13/84 systems at power (Rev. 1) 187 7.2.2.4 Lifting cf leads to perfom surveil- Ccmplete 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 Ccmplete 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 sanpling lines and cabinet temperature control 195b 7.3.2.3 Freeze-protection / water filled Ccrplete 8/1/84 instrument and sanpling lines and cabinet teroperature control 196 7.3.2.4 Sharing cf ccnnon instrument taps Cmplete 8/1/84 197 7.3.2.5 Micrcprocessor, multiplexer and Cmplete 8/1/84 cmputer systems (Rev 1)
M P84 80/1218 - gs L_
A'ITAOiMENT 1 (Cont'd)
DSER R. L. MITIL 'IO CPDi SECTICN A. SQiENCER ITEM NUMBER SUBJECT STATUS LETIER DA'IED 198 7.3.2.6 'IMI Its II.K.3.18-ADS actuation Cmplete 8/20/84 199 7.4.2.1 IE Bulletin 79-27-Ioss cf non-class Cmplete 8/24/84 l
IE instruentation and control power (Rev. 1) systs bus during cperation 200 7.4.2.2 - Rsote shutdown syste C m plete 8/15/84 (Rev 1) 201 7.4.2.3 RCIC/HPCI interactions Ccmplete 8/3/84 202 7.5.2.1 tevel measuremnt errors as a result C m plete 8/3/84 of enviromental 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 'IMI Its II.F.1 - Accident nonitoring Cmplete 8/1/84 205 7.5.2.4 Plant process cmputer systm 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 syste Cmplete 8/24/84 failures (Rev. 1) 208 7.7.2.2 Multiple control syste failures Cmplete 8/24/84 (Rev. 1) 209 7.7.2.3 Credit for non-safety related systes Cmplete 8/1/84 in Chapter 15 cf 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 Ccmplete 7/27/84 211c 4.5.1 Control rod drive structural materials Ccmplete 7/27/84 M P84 80/1219 - gs L
= ,
tt l
ATTACHPENT 1 (Cont'd) ,
i DSER R. L. MITTL 10 t,
- CEW SBCTICN A. SOMNCER ITEM NUMBER SUBJECT STATUS LETTER IRTED l' t
211d 4.5.1 Control rod drive structural materials Canplete 7/27/84 211e. '4.5.1 Control rod drive structural materials Cm plete 7/27/84 ,
1' 212 4.5.2 - Reactor internals materials '
Caplete 7/27/84 213 5.2.3 Reactor coolant pressure boundary Cmplete 7/27/84 material 214 6.1.1 Engineered safety features materials Conglete 7/27/84 215 10.3.6 Main steau and feedwater syste Cmplete 7/27/84 materials 216a 5.3.1 Reactor vessel materials Cm plete 7/27/84 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 Conglete 8/15/84 training 221 8.2.2.1 Physical separation of offsite Cmplete 8/1/84 trananission lines 222 8.2.2.2 Design provisions for re-establish- Cmplete 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 Coninon failure mode between onsite Cmplete 8/1/84 and offaite power circuits M P84 80/12 20- gs
i ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL 'IO OPEN SECTIOi A. SOMENCER ITEM NLMBER SUBJECT STA'IUS IEFIER DATED 225 8.2.3.1 Testability cf autmatic transfer cf Cmplete 8/1/84 power frm the normal to preferred power source 2 26 ' 8.2.2.5 Grid stability Ccaplete 8/13/84 (Rev. 1) 227 8.2.2.6 Capacity and capability <f offeite Ccmplete 8/1/84 circuits 2 28 8.3.1.l(1) Voltage deep during transient condi- Cmplete 8/1/84 tions 229 8.3.1.l(2) Basis for using bus voltage versus Ccmplete 8/1/84 actual connected load mitage in the voltage drop analysis 230 8.3.1.l(3) Clarification of Table 8.3-11 Ccmplete 8/1/84 231 8.3.1.l(4) Undervoltage trip setpoints Ccmplete 8/1/84 232 8.3.1.l(5) Load configuration used for the Ccmplete 8/1/84 voltage decp analysis 233 8.3.3.4.1 Periodic systs testing Cm plete 8/1/84 234 8.3.1.3 Capacity and capability cf onsite Cmplete 8/1/84 AC power supplies and use cf ad-ministrative controls to prevent overloading cf the diesel generators 235 8.3.1.5 Diesel generators load acceptance Cmplete 8/1/84 test 2 36 8.3.1.6 C mpliance with position C.6 of Cmplete 8/1/84 IG 1.9 237 8.3.1.7 Decription cf the load sequencer Ccmplete 8/1/84 238 8.2.2.7 Sequencing cf loads on the offsite Ccmplete 8/1/84 power syste M P84 80/12 21 - gs
ATTAOIMENT 1 (Cont'd)
DSER R. L. MITIL 'IO OPEN SECTION A. SODENCER ITEM NUpeER SUBJECT STATtJS IEr1ER DA'IED 2 39 8.3.1.8 Testing to verify 80% minimum Cmplete 8/15/84 voltage 240 8.3.1.9 Ccmpliance with BIP-PSB-2 Cmplete 8/1/84 241 8.3.1.10 Ioad acceptance test after prolonged Ccmplete 8/20/84 no load operation cf the diesel (Rev. 1) generator -
242 8.3.2.1 Ccupliance with position 1 of Regula- Cmplete 8/1/84 tory Guide 1.128 243 8.3.3.1.3 Protection or qualification of Class Cmplete 8/1/84 lE equipnent frcm the effects of fire suppression systems 244 8.3.3.3.1 Analysis and test to demonstrate Ccmplete 8/30/84 adequacy of less than specified (Rev. 1) separation 245 8.3.3.3.2 The use cf 18 versus 36 inches of Ccaplete 8/15/84 separation between raceways (Rev. 1) 246 8.3.3.3.3 Specified separation cf raceways by Ccuplete 8/1/84 analysis and test 247 8.3.3.5.1 Capability cf penetrations to with- Ccmplete 8/1/84 stand long duration short circuits at less than maximum or worst case short circuit 248 8.3.3.5.2 Separation cf penetration primary Ccmplete 8/1/84 and backup protections l 249 8.3.3.5.3 The use cf bypassed thermal cuerload Ccmplete 8/1/84 protective devices for penetration protections l l
250 8.3.3.5.4 Testing cf fuses in accordance with Ccaplete 8/1/84 R.G. 1.63 M P84 80/12 22 gs
i ATTAOSEST 1 (Cont'd)
DSER R. L. MITFL 10 CPIBi SECTICBI A. SONENCER I11M NLBSER SUBJBCr STATUS LET1ER DA2?D 251 8.3.3.5.5 Fault current analysis for all Caplete 8/1/84 representative penetration circuits 252 8.3.3.5.6 The use of a single breakor to provide Caplete 8/1/84 penetration protection 253 8.3.3.1.4 Canaltment to protect all Class lE Caplete 8/1/84 equipment from external hazards versus only class lE equipment in one division
'254 8.3.3.1.5 Protection of class lE power supplies Caplete 8/1/84 from failure of unqualified class lE loads 255 8.3.2.2 Battery capacity Cmplete 8/1/84 256 8.3.2.3 Autanatic trip of loads to maintain Cmplete 8/20/84 sufficient battery capacity 257 8.3.2.5 Justification for a 0 to 13 second Caplete 8/1/84 load cycle 258 8.3.2.6 Design and qualification of DC Cmplete 8/1/84 system loads to operate between minimum and maximin voltage levels 259 8.3.3.3.4 Use of an inverter as an isolation Omplete 8/1/84 device 260 8.3.3.3.5 Use of a single breaker tripped by Conglete 8/1/84 a IACA signal used as an isolation device 261 8.3.3.3.6 Autanatic transfer of loads and Conglete 8/1/84 interconnection between redundant divisions 262 11.4.2.d Solid waste control progran Caplete 8/20/84 M P84 80/12 23- gs
ATDO9ENr 1 (Cont'd)
DSER R. L. MITIL 10 '
OPIM SECTICH A. SCHWENCER :
ITEM NUBBER SUBJBCr STATUS IErrER DATED l 263 11.4.2.e Fire protection for solid radwaste Ctmplete 8/13/84 storage area 264 6.2.5 Sources of oxygen Couplete 8/20/84 265 6.8 .1.4 ESF Filter Testing Conglete 0/13/84 266 6.8 .1.4 Field leak tests Conglete 8/13/84 267 6.4.1 Control roon toxic chemical Couplete 8/13/84 detectors 268 Air filtration unit drains Caplete 8/20/84 269 5.2.2 Code cases N-242 and N-242-1 Ca glete 8/20/84 270 5.2.2 code case N-252 ccuplete 8/20/84 TS-1 2.4.14 Closure of watertight doors to safety- Open related structures TS-2 4.4.4 Single recirculation loop cperation Open TS-3 4.4.5 Core flow monitoring for crud effects Caplete 6/1/84 TS-4 4.4.6 Imse 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 15-7 6.2.3 Inleakage and drawdown time in Open secondary containment TS-8 6.2.4.1 Imakage integrity testing Open TS-9 6.3.4.2 EECS subsystem periodic cxmponent Open testing M P64 80/12 24- ga i
ATUODU!NT 1 (Cbnt'd)
D8ER R. L. MITFL 10 l' OMN SBCTIGI A. SCHWENCER ITIBt IDWER SUBJECT STATUS IEr!1!lR DM1llD t
TS-10 6.7 MSIV leakage rate T9-11 15.2.2 Availability, setpoints, and testing Open i of turbine bypass systemi 15-12' 15.6.4 Primary coolant activity 14-1 4.2 Fuel rod internal pressure criteria Complete 6/1/84 tr-2 4.4.4 Stability analysis substitted before open second-cycle operation i l
M P84 80/12 25- ge f
l l
l ATTACHMENT 2 DATE: 8/30/84 DRAFT SEP. SECTIONS AND DATES PROVIDED SECTION DATE SECTION DATE 3.1 3.2.l' 11.4.1 See Notes 1&5 3.2.2 11.4.2 See Notes 1&5 5.1 11.5.1 See Notes 1&5 5.2.1 11.5.2 See Notes 1&S 6 . 5 . l' 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 S.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 Fute 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&5 Notes:
11.1.2 See Notes 1&5 11.2.1 See Notes 1&S 1. Open items provided in 11.2.2 See Notes 1&S letter dated July 24, 1984 11.3.1 See Notes 1&S (Schwencer to Mittl) 11.3.2 See Notes 1&S
- 2. Open items provided in June 6, 1984 meeting
- 3. Open items provided in April 17-18, 1984 meeting CT:db
- 4. Open items provided in May 2. 1984 meting
- 5. Draft SER Section provided in letter dated August 7, 1984 (schwencer to Mittl)
MP 84 95/03 01,
- m - ._ -__ _ _ ._ ._ - _ . _ _ - - , _ - _ , -.
o 2 Datos Augu=t 30, 1984 ATTACHMENT 3 Open DSER Item Section Subject 28A-e~ 3.4.1 Flood Protection 110 4.6 Functional design of reactivity control systems 112 5.2.5 Reactor coolant pressure boundary leakage detection 141 9.1. 3 - Spent fuel pool cooling and cleanup system
- 151 9.4.1 Control structure ventilation system 153 9.4.5 Engineered safety features ventilation system 244- 8.3.3.3.1 Analysis and test to demonstrate adequacy of less than specified separation s
1
e I
l i
l ATTACHMENT 4 k
9
4
. HCGS FSAR 6/84 1.8.1.23 Conformance~to Reculatory Guide 1.23 (Safety Guide 8),
Revision 0, February 17, 1972: Orisite Meteorological Procrams HCGS complies with Regulatory Guide 1.23.
i 1.8.1.24 Conformance to Reaulatory Guide 1.24 (Safety Guide 24),
Rev:.sion 0, March 23, 1972: Assumptions Used for Eva;.uatina the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storace Tank Failure Regulatory-Guide 1.24 is not applicable to HCGS.
1.8.1.25 'Conformance to Reculatory Guide 1.'25 (Safety Guide 25),
Rev:,sion 0, March 23, 1972: Assumptions Used for
- va;.uating the Potential Radiolocical Consecuences of a
'ue Hand .ino Accident in the Fuel Handlina and Storace lac:.lity dor Boilina and Pressurized Water Reactors HCGS complies with Regulatory Guide 1.25.
1.8.1.26 Conformance to Reculatory Guide 1.26, Revision 3, February 1976: Quality Group Classifications and Standards for Water , Steam , and Radioactive-Waste-Containino Components of Nuclear Power Plants HCGS complies with Regulatory Guide 1.26, with the clarifications outlined below.
7 Ze~'e ;::iti:= is thsb-egtt r--at that it i ;::ter.t te ::fet-j i: ' -
__ . 1 .... anu uu-6 fer d::: ::t distir.guish b;ture th :: t~ ,
" PSE&G does recognize the need for the assurance of the specified operation of certain non-safety-related structures,.
systems and components, such as fire protection systems, radioactive waste treatment, handling and storage systems, and
. Seismic Category II/I items. Such assurance is documented through-the specification of limited quality assurance programs (described in Table 3.2-1, footnotes (22), (50) and (52). In addition, items designated "D+" in Table 3.2-1 will be included in the QA program during operations.-
The exception to Position C.2.b is that since the reactor recirculation pumps do not perform any safety function and since failure of the reactor coolant pumps due to seal or cooling water failure does not have serious safety implications, the control rod drive (CRD) seal purge-supply and reactor auxiliaries cooling 1.8-11 Amendment 6
_ _ _ _ . ~ . _ _ _ _ _ _ . . _ . _ _ _ ..._. _ .__ ._ _ _ . _ . _ . _ .._
i .
ius 24 860 2 7 0 0 4 6 HCGS FSAR isolation signal. The isolation signal to the valves can be
, overridden manually from the main control room. Containment isolation is discussed further in Section 6.2.4. r-Jseg_r A
-~
Each analyzer package can only sample one sample point at one time. The selection of a specific sample point is determined by the operator.. Gases from the selected sample point are routed in parallel through a hydrogen analyzer cell and oxygen analyzer cell located in the analyzer panel inside the reactor building.
The operation of the hydrogen and oxygen analyzer cells is based on the measurement of thermal conductivity of the gas sample.
The thermal conductivity of the gas mixture changes proportionally to the changes in the concentration of the individual ~ gas constituents of the mixture. The thermal conductivity of hydrogen is far greater (approximately seven times the thermal conductivity of air) than any other gas expected to be present in the primary containment. The hydrogen analyzer cell incorporates a catalytic combustion feature in which hydrogen in the sample is removed by catalytic recombination with a reagent gas (oxygen). The thermal conductivity of the sample is measured before and after '
1 recombination, and the two measurements are compared. The difference in thermal conductivity is proportional to the concentration of hydrogen originally in the sample. The oxygen analyzer operates simultaneously in a similar manner, except that the reagent gas is hydrogen. .
The hydrogen analyzer has dual range capability of 0 to 10% by volume and 0 to 30% by volume.- The oxygen analyzer has dual range capability of 0 to 10% by volume and 0 to 25% by volume.
The hydrogen and oxygen concentrations in the sample gas are indicated at the analyzer panel in the reactor building and at the remote control panel in the main control room. The concentrations are also recorded in the main control room. An additional oxygen indication is provided at the entrance to the drywell service hatch.
l Sample gases are drawn throUgh the analyzer cells by the diaphragm pump located in the analyzer panel. Sample gases and any excess moisture, either from the sample or created by the catalytic recombination, are routed back to the suppression chamber.
i l
l HOAS design and performance data is included in Table 6.2-17.
l The HOAS environmental qualification program is found in 6.2-75
Aus 24 84 0 27 0 0 4 6
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HCGS FSAR four LPRM strings (16 detectors) surrounding the selected rod are t used in the RBM to provide protection against local fuel overpower conditions.
7.6.1.4.3 Average Power Range Monitor Subsystem The APRM subsystem monitors neutron flux from approximately 1% to above 100% power. There are six APRM channels, each receiving core flus level signals from 21 or 22 LPRM detectors. Each APRM
- channel averages the 21 or 22 separate neutron flux signals from the LPRMs assigned to it, and generates a signal representing core average power.
9 This' signal is used to drive a local meter and a remote recorder located.on the main control room vertical board. It is also applied to a trip unit to provide APRM downscale, inoperative and upscale alarms, and upscale reactor trip signals for use in the RPS or RMCS.
Refer to Section 7.2.1.1 for a description of the APRM inputs to the RPS, and Figure 7.6-5 for the RPS trip circuit input arrangement. APRM trips are summarized in Table 7.6-2.
The APRM scram units are set for a reactor scram at 15% core power in " refuel" and "startup" modes. When the mode switch is in "run," the APRM trip reference signal is provided by a signal ,
that varies with recirculation flow. This provides a power following reactor scram setpoint. As power increases, the reactor scram setpoint also increases up to a fixed setpoint L above 100%. Reactor power is always bounded with a reactor scram, yet the change in power required to generate the reactor scram does not vary greatly with the operating power level.
Provision is made for manually bypassing one APRM channel at a time. Calibration or maintenance can be performed without tripping the RPS. Removal of an APRM channel from service without bypassing it, by unplugging a card, by taking the APRM function switch out of " operate," or by having too few assigned LPRM signals to the APRM, will result in an APRM " inoperative" condition which causes a half scram, a rod block, and annunciation i The APRM channels receive power from non-Class 1E uninterruptible power source . Power for each APRM trip unit is supplied from F 5-il 3 skt 3D (See Fgun7.6-11
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. HCGS FSAR- .
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the same power supply as its associated APRM. The ac bus used for a given APRM channel also supplies power to its associated LPRMs.
hlMSE W h ,
APRM signals are sent to redundant reactivity control system (RRCS) to enable the logic if additional reactivity control is necessary following an ATWS event. The use of this signal is discussed in Section 7.6.1.7.
The APRMs are designed to remain accurately functional for at least 20 minutes after an ATWS feedwater run-back is initiated.
7.6.1.5 Bacirculation Pump Trio System - Instrumentation and Controls 7.6.-1.5.1 RPT Purpose The reason for tripping the recirculation pumps is to reduce the impact on the fuel of thermal transients caused by turbine trip, generator trip, or load rejection. The rapid core flow reduction increases void content and thereby introduces negative reactivity in conjunction with control rod insertion.
7.6.1.5.2 RPT Logic and Operation The RPS detects turbine contro.1 valve fast closure and main stop valve closure, using four channels of sensor logic. This is :
. . combined into two channelized two-out-of-two trip logic for RPT.
Trip signal initiation requires confirmation from at least two sensor channels. No single failure will prevent RPT trip.
Each trip logic channel will trip both recirculation pumps. ,
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(See Fig. 8.3-11, sht. 3)
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, PCWER RANCE POWER RANCE NHS NHS BUS A BUS 3 10C608 10C608 I
l 640PE CREEK
, GENERATING STATION FINAL SAFETY ANALYSIS REPORT l
ELECTRICAL PROTECTION ASSEMBLIES (EPAs) IN THE POWER RANCE NEUTRON MONITORING SYSTEM l FIGURE BJ-16 Amead***'
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DSER Open Item No. 28 (DSER Section 3.4.1) 9 FLOOD PROTECTION . ,
The design of the facility for flood protection was reviewed in ,
.accordance with Section 3.4'.1 of the Standard Review Plan (SRP) !
NUREG-0800. Jus audit . review of each of the areas listed in the i
'" Areas of Review" portion of the SRP section was performed according to the guidelines provided in t!e " Review Procedures" portion of the SRP section. Conformance with the acceptance criteria formed the basis for our evaluation.of the design of the facility for flood protection with respect to the applicable regulations of 10 CFR Part 50.
- .In order to assure conformance with the requirements of General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena," our review of the overall flood protection design included all systems and components whose failure due to flooding
-could prevent safe shutdown of_the plant or result in uncontrolled j release of significant radioactivity.
The applicant has sited the plant (at elevation 22.5 feet Mean Sea Level (MSL)) along the Delaware River near the point where the river flows into the_ Atlantic Ocean. The design basis flood is
! 'the result of the probable maximum hurricane (PMH) surge with wave runup coincident with the 10% exceedance high tide. The design basis flood level for all structures is 34.8 feet MSL, including wave activity.(refer to Section 2.4.2 of this SER).
The design basis flood level of. 34.8 feet MSL represents plant L
submergence at the plant site by 12 feet 3.6 inches. Vertical '
l and horizontal construction joints are provided with waterstop to l: elevation 32 feet MSL.[fThe applicant must water-proof all safety-l related structures and all penetrations to those structures to a higher elevation than the flood elevation of the design basis flood (PMH) .}. g g4 The probable maximum flood which results in over 12.3 feet of water onsite is due to the PMH and is greater than the flooding due to'the probable maximum precipitation.
The personnel access doors to areas where flood protection must be provided doors 318 and are 158. all[ submarine In order to doors which comply open with theoutward, guidelines except of Regulatory Guide 1.102, " Flood Protection for Nuclear Power Plants", Position Cl, the applicant must modify doors 31B and il5B to be submarine doors or equivalent for these doors to open outward or assume the doors are open during the design basis flood and verify that no safety-related equipment will be flooded}LPRh (The applicant has not provided information requested concerning Regulatory Guide 1.102, Position C.2, and therefore no conclusions 28-1
. 1 4
Item No. 28 (Cont'd) g canbemadeconcerningcompliancesatthistime[7/[Theapplicant i has not committed to providing sensors on all doors and hatches in exterior. walls which are below the desgin basis flood elevation, plus wind-generated wave ef fects to alarm in the control room when they are opened. As an alternative, the applicant may provide the_rgsults of a flooding analysis with the administra-tively controlled doors open and which shows that no safety-related
, equipment .will be flooded.]- age, (The site contains non-seismic Category I tanks. The applicant has stated that the site drainage system will prevent the contents of the failed tanks (as the result of a safe shutdown earthquake) from flooding the safety-related structures. The applicant has not identified the site drainage system as safety-related, seismic Category I. The site drainage system must be safety-related and seismic Category I in order to take credit for the system af ter a design basis event. Similarly, the site drainage system should be tornado and tornado missile protected if the drainege system is needed to prevent any flooding resulting from tank (s) failure due
= to a tornadic even or due to tornado generated missiles.] -JFd The applicant has stated that the electrical cables will continue
. to function properly even if the manholes and duct banks are flooded. The ability of the cables to perform the function if they are flooded with sea water and the long-term effects of continued submergence in sea water is discussed in Section 8.3 of this SER.
['In response to our concern regarding internal flood protection, the applicant indicated that their discussion of plant features to prevent internal flooding of redundant safety-related equipment was in Section 6.1. 3.e of the FSAR. There is no Section 6.1. 3.e in the FSAR.].JIge,.
[The applicant has not addressed our concern associated with the structural integrity of the safety-related structures during the design basis flood and the effects of " floating" missiles. Since the Delaware River is a navigable waterway with the refineries and naval shipyard in Philadephia, the applicant must address the ef fects of ships and boats with a draf t of less than 12 feet hitting the walls and penetrations of safety-related structures.
Some ships which do travel up and down the Delaware River and can have a draft of less than 12 feet are the " Newport" class LSTs 4
(LST-ll79 series), the "DeSoto County" class LSTs (LST-ll73 series),- the " Anchorage" class LSDs (LSD-36 series), submaries (especially the non-nuclear power submaries), tug boats, visiting "American" ships from foreign countries, oil tankers (when they are empty), and a large host of pleasure craf t.]. g3 $
28-2
4 Itan No. 28 (Cont'd)
Because the applicant has not adequately addressed the staff's concerns identified above, we cannot conclude compliance with
- General Design Criterion 2 and the guidelines of Regulatory Guides 1.102, " Flood Protection for Nuclear Power Plants," Positions C.1 and 1.59, " Design Basis Floods for Nuclear Power Plants", Positions C.1 and C.2 and Branch Technical Position ASB 3-1, " Protection Against Piping Failures in Fluid systems Outside Containment".
We will report resolution of these items in a supplement to this
, SER.. -The design of the facility for providing protection.from flooding does not meet the acceptance criteria of SRP Section 3.4.1.
RESPONSE
- a. The requested information with respect to waterproofing all
-safety-related structures to a higher elevation than the flood elevation of the design basis flood (PMH) has been provided in response to Question 240.8.
- b. Doors 3331B and 3315B are watertight .(submarine) doors and although they are installed in an unseated position (they swing inward), both doors have been designed for specified unseating pressure of 19 feet of water. To assure that these doors will l not be inadvertently opened or left open, both doors are locked
. closed and administratively controlled during a flood event.
- c. HCGS procedure " Acts of Nature", will commit to ensure that exterior doors and hatches are closed and locked by administrative procedure under impending flood conditions. Add crnseed _f.
4
- d. .The response to FSAR Question 410.7 has been revised to state that the site drainage system is not required to prevent the contents of failed tanks (as the result of a safe shutdown earth-
- quake) from flooding the safety-related structures.
- a. The response to NRC Question 410.9 has been revised to refer to Section 3.6.1.e instead of 6.1.3.e.
t 1
l I
I l
6
XEROX TELECOPIER 495320- S-843 9:17AM i 20165241614 a tt 5 lh I ldsmAT i HCGS 18 ~3 DSER ITEM 28c In summary, the " Acts of Nature" procedure specifies an immediate check of all external doors to insure they are in the locked closed position upon receipt of a hurricane warning from the National Weather Service which may impact Artificial Island. The doors.will be checked once per shift to verify they remain locked closed during the hurricane period unless the river level reaches site grade, at which time the doors will be checked every 30 minutes.
l-l l
b b
HCGS FSAR 10/83 QUESTION 410.7 (SECTION 3.4.1)
~
For these nonseismic Category I vessels, pipes and tanks located outside of buildings, discuss the effect of failure of these items and any potential flooding of safety-related structures, systems and components. Provide a similar discussion for nontornado. protected vessels, tanks and piping.
RESPONSE .
The failure of non-Seismic Category I and non-tornado protected tanks, vessels, and major pipes located outside of buildings (Table 410.7-1) will not_ adversely affect safety-related structures, systems and components by flooding, as discussed below:
Failure of Tanks The locations of tanks in the yard area are shown on Fidure 1.2-1. Failure of the condensate storage tank, located on the south side of the power block (Table 410.7-1, Item 1), will not cause flooding. Any spillage due to failure of this tank will be contained within a reinforced concrete dike designed to be t
' Seismic Category I, aus discussed in Section 3.8.4.1.6. ,
.The tanks located on the north and west sides of the power clock (Table 410.7-1, Items 2.through 7) do not have Seismic Category I dikes around them. Failure of these tanks could cause local
-flooding. However, this flooding would not adversely affect safety-related facilities for the following reasons:
- a. The storm drainage. system in this area will drain the spillage to the Delaware River before it reaches the vfA"y wer plant complex.
- b. Seismic Category I electrical cables and duct banks located in the vicinity of these tanks are protected aga' inst flooding,.as discussed in the response to Question 410.8.
Failure of Coolino Tower Basin Wall (Table 410.7-1, Item 8)
The failure of the cooling tower basin wall would not adversely affect safety-related structures, systems and components, as discussed below:
The operating water level within the cooling tower basin is elevation 102.5 feet. The slabs and walls are conservatively designed for 3 feet of freeboard, allowing the water level to rise to' elevation 105.5 feet. The grade around the basin well is DSER O N I N 410.7-1 Amendment 2
4 .
" Insert A" -
- a. Any spillage will be conveyed to the Delaware River ny means of overland surface runoff without adversely affecting any safety-related structures, systems or components by flooding. There is a clear path to tne river from.the building which will assure that any surface water will not enter the building. In addition, storm drainage is provided to facilitate conveyance of runoff to the river which will furtner minimize tne potential for any local ponding. -
9
+
- i. OSER OPEN ITEM M E A
BCGS FSAR 10/83 at elevation 104.5 which is 2 feet above the operating water level in the basin. 1 i
The worst case flooding could result from the unlikely " wash-off" '
of the soil on the south side of the tower.. For this case,-the run%ff would be dispersed and intercepted by the storm drainage system before it could. reach the power block are The Seismic Category I duct banks located between the intake tructure and the power block will not be affected as they are not located in
~
the flow path of the water.
Failure of Circulatina Water Pipes (Table 410.7, Item 9)
Failure'of these pipes within the yard area between the cooling tower basin and the turbine building will cause flooding of this area. Water from the damaged pipes will erode the soil cover and flood the yard. No Seismic Category I equipment or compohents are located in this area of possible erosion. The storm drainage systemwouldeventuallydrainthewatertotheDelawareRivep the most severe case, all the water from the cooling tower
, basin could drain through the damaged pipe into the yard area
, between the circulating water pumphouse and the turbine building.
This could cause flooding of the lower level of the turbine building. However, safety-related systems and components would not be damaged, as discussed in the response to Question 410.115.
Failure of Maior Yard Pipino Failure of any of the pipes identified in Table 410.7-1, Items 10
-to 14, may cause local flooding. However, the intensity and volume of water discharge from any of these pipes is less than-
[
that of the circulating water pipes discussed above and would not cause damage to ,any safety-related facilities. Soil erosion caused by failure of_these pipes is discussed in the response to Question 410.64.
l l
or the u.Joder uJow/d f /c w aye.r/anol fa +h e
~l)ejawm E **Ver 66 di3 c uS sed for' SanN'
( Z hen s a ths-u 1 )
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- om om N Eb 410.7-2 Amendment 2
. . -_-.. . . . . . - . . __ . . ~
e 4
NCGS FSAR TABLE 410.7-1 YAkD TANKS At3D NAJOR PIPING (NON-SEISNIC), 18/83
- i. Iten Capacity' -Type of l Tornado -
4 tao. Tank ' of Pipe Description or Flow Location Containment Protection 1 CondensateLStorage Tank 500,000 gal South of power plant- Seismic Cat. None s
comples I Aelaforced Conc. walle 2 Fire Mater Tanks (2) 300,000 gal em North of power plant comples None Mone
! 3 Asphalt Storage Tank 9,000 gal North of power plant complex Concrete unit Mone j
Nasonry walla 4 Fuel Oil Day Tank 18,000 gal North of power plant comples Reinforced None 1
Conc. walle 5 Chemical Treatment Tanks 2 Sodium Nypochlorite 30,000 gal ea North of power plant comples Reinforced None
' 1 Sulfuric Acid 20,000 gal North us power plant compleu Concrete Mene 2 Sodium Hypochlorite 15,000 gal ea West of power plant comples Malls None 1
4
! 6 Sewage Treatment Plant [
t 1 Equalization Tank 20,000 oct l North of power plant comptes Buried None 2 Treatment Tanks 8,000 gal ra North of power plant s'oep) w Perig4 None 1 Treatment Tank 35,000 951 North of power plant comptes Earth bare None j 7 Puel Oil Storage Tank 1,000,000 gal North of power plant comple,a Earth dike None l -
8 Cooling Tower Basin 6,500,000 gal North of power plant comples Reinforced None j
. Conc. wall 9 144*f Circulating Mater Pressure 552,000 gpa Pipe s (2)
Between cooling tower and undergroued Soil cover turbine building y 10 48"# Nakeup Nater Pressure Pipe 30,000 gpm Reacror building to coollag underground Soil cover i
tower d
- Qq 11 36*p Nakeup Nater Pressure Pipe 21,000 gpa Reactor building to cooling underground Soil cover t
g towe r .
- 12 48'S Blowdown Mater Gravity Pipe 15,400 gym Cooling tower to Delaware underground soil cover River i H 13 36*# Deicing unter Pressure Pipe 12,000 gpa Circulating water pipe to Underground Soil cover
! g g intake structure i B 1. 126FireM.t., coop 2,500 gp. Around plant co.pi.= und.rground Soil cover i
ti=
8 A
j C6/3 i ,
]e.V b RCGS ,
I I
DSER'Open Item No.-110 A & B (Section 4.6)
FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS The control rod drive system was reviewed'in accordance with .
Section 4.6 of the Standard Review Plan (SRP), NUREG-0800.'
An audit review of each of the areas listed in the " Areas of
'- Review" portion of the SRP section was performed according ;
to the guidelines provided in the " Review Procedures" por-tion of_the SRP section. Conformance with'the acceptance criteria formed the basis for our evaluation of the control
. rod drive system.with respect to the applicable regulations
! of 10 CFR 50.
The applicant has not. addressed the recommendations of
'NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System' Piping."
The design does not utilize a CRDS return line to the reac-tor pressure vessel. In~accordance with NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drives Return Line Nozzle Cracking," dated November 1980, equalizing valves are installed between the cooling water header and exhaust water header, the-flow stabilizer loop is routed to the cooling water header, and both the exhaust header and flow stablizer loop _are stainless steel piping.
Ws have reviewed.the' extent of conformance of the Scram Discharge Volume (SDV) design with the NRC generic study, "BWR Scram Discharge System Safety Evaluation," dated December 1,1980. The design provides two seoarate SDV headers, with an integral instrumented volume (IV) at the end of each header, thus providing close hydraulic coupling.
Each IV has redundant and diverse level instrumentation ,
l (float sensing and pressure sensing) for the scram function .
!_ Vent and drain lines are com-attached directly to the IV. '
! pletely separated and contain redundant vent and drain val-ves with position indication provided in the main control room. With respect to Design Criterion 8, the applicant stated that the "SDV Piping is continuously sloped frca its high point to its low point." In order to provice a re-
, sponse to Design Criterion 8, the applicant must provide a l description of the SDV from the beginning of the SDV to the IV drain. The description should include piping geometry (i'.e., pitch, line size, orientation).
M~P84 126/05 1-mw l
d' 4
2
. DSER Open Item No. 110 A & B (Section 4.6) (Continued)
- Except for Design Criterion 8, we conclude that the design of the SDV fully meets the requirements of .the above referenced NRC generic SER and is therefore acceptable.
Additionally, the above-described design of the SDV satisfies LRG-II, Item 1-ASB, "BWR Scram Discharge Volume Modifications."
Based on our review, we conclude that th'e functional design of the reactivity c.ontrol system meets the requirements of General Design Criteria 23, 25, 26, 27, 28, and 29 with respect' to demonstrating.the ability to reliably control reactivity changes under normal operation, anticipated operational occurrences and accident conditions including single failures, and the guidelines of NUREG-0619 and is, therefore,-acceptable. 'We cannot conclude compliance with the guidelines of NUREG-0803 and the generic document dated
. December 1, 1980. The functional design of the reactivity control sytem does not meet the applicable acceptance Criteria of SRP 4.6. We will report resolution of these items in a supplement to this SER.
RESPONSE ,
- The concerns of NUREG-0803 are addressed in response to Q410.26.
, ' FSAR Section 4.6.1.2.4.2(f) has been revised to include a
' description of the SDV piping.
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M P84 129/05 2-mw l
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- room. Differential pressure between the reactor vessel
'and the cooling water header is indicated'in the main control room. -Although the drives can function without cooling water, seal life is shortened by long-term exposure to reactor temperatures. The temperature of each drive is indicated and rec ~orded, and excessive temperatures are annunciated in the main control room,
- e. Exhaust water header - The exhaust water header connects to each HCU and provides a low pressure plenum and discharge path for the fluid expelled from the drives during control rod insert and withdraw operations. The fluid injected into the exhaust water header during rod movements is discharged back up to the RPV via reverse flow through the insert exhaust directional solenoid valves of adjoining HCUs. The ,
i pressure in the exhaust water header is, therefore, maintained at essentially reactor pressure'. To ensure that the pressure in the exhaust water header is maintained near reactor pressure during the period of vessel pressurization, redundant pressure equalizing valves connect the exhaust water header to the cooling water hea' der.
i -
la i nc,h d e e**T
- f. Scram discharge volume - The scram discharge volume (SDV) consists of two sets o header piping, each of which connects to one-half o the HCUs and drains into IS Sad d' 7 scram discharge instrument volume (SDIV). Each set of header piping is sized to receive and contain all L the water discharged by one-half of the drives during a i scram, independent of the SDIV. .
The m inimhanda!
u.m p P'M*f i+c h a* f/*Pe '/s ' p2 e.r to f.a tlaasu).shown Poin6on " %As "-
ure +4 ~/-
The SDIV for each header set is directly connected.to ;
the low point of the header piping. The large-diameter pipe of each SDIV thus serves as a vertical extension of the SDV. A 2 p,' ping eo,,n se.+dn a eh e sance *f l He .s dHAin b d PraL' vedes dea,'a g:e -o'ifsrhs s loped ass w;w e.rw n.m*
" p ge.rskov GoatOSdl *2 Ad ' ***
During normal plant operation, the SDV is empty and is j vented to the atmosphere through its open vent and drain valves. When a scram occurs, upon a signal from j the safety circuit, these vent and drain valves are closed to conserve reactor water. Redundant vent and drain valves are provided to ensure against loss of reactor coolant from the SDV following a scram. Lights
! in the main control room indicate the position of these
- valves.
l
)
! ossa o m r = no l 4.6-13 l
l - --- - . - - - -._ - --___--
T HCGS FSAR 12/83 QUESTION 410.26 (SECTION 4.6)
Provide the information requested in our generic letter 81-34,-~-~
dated August 31, 1981, regarding NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping."
RESPONSE
HCGS is participating in the BWROG activit'ies related to the cram discharge pipe integrity. The BWROG's final response to
' "~
_,e NRC is being prepared for NRC review and approval. , . . - - ....
r- -. m avww. w u ,cr.2 m4ti s. n,.mu4a. 4 "
A HLSS Pl ant .5 pee if,'c responsc No' / e p7a vlde.d w;4 hin 40 day 1 o S MRC resolu.fion af dhe AWAo&
Pt .n by pas;Han . Mcss w.'il im plemed any regutyd w h i c.h i s o.+
Th e. e.n el o G 'th e n e.xt t-e.f u.e. ling o u.-hnge l e.a.s t i 2; rwen% s o.f+er N R c. y e so l u.4-. on . Pe. ndi og m o.-+ er sa.I
- o. v a.lla.b i llk y , % i s s c.heol u.I e. vno.y c ko n3 c. w 1% NRC. a.ppre ala.l.
4 1
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RCGS N f. V -
DSER Open Item No.112 (DSER Section 5.2.5) 7
'. REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTIdN
~
Provisions have not been made to monitor all of the systems con-nected, as identified in Table 1 of Section 5.2.5 of the Standard Review Plan, to the RCPB for monitoring and alarming intersystem l
' -leakage by using radioactivity and differential flow monitors.
Specifically, the applicant has not provided monitoring capabflity for intersystem leakage for the safety injection system (high and
- low pressure systems), residual heat removal system (inlet and
' discharge ), reactor core isolation cooling system, and the steam side of the high pressure coolant injection system. Thus, the guidelines of Regulatory Guide 1.45, Position C.4 are not met.
! Each leakage detection system has indicators and alarms either in the control room or at the local panels. The monitor signals pro- !
vided to the control room are generated through the plant computer system with no unprocessed signals available to the operators and ru) proce'dures to direct the operators where or how to obtain the information if the control room indications are lost. The appli-
< cant should provide a discussion of the capability to maintain suf ficient onsite manpower at all times to man all locaf panels
'100% of the time (this is in addition to the manpower requirements discussed in Section 9.5 of this SER) when the information is not available in the control room, to provide a seismic Category I i
communication system between the control room .and all local panels,
!! to provide procedures to guide the personnel at the local panels,
. and to propose a Technical Specification requiring the manning of the local panels when the control indications are not available.
Thus, the guidelines of Regulatory Guide 1.45, Position C.7 is not met. ,
f The applicant does not have a sump flow monitoring system, an airborne particulate radioactivity monitoring system, and a l
seismic Category I monitoring system and therefore does not meet the guidelines of Positions C.3 and C.6 of Regulatory Guide 1.45.
As recommended by Regulatory Guide 1.45, at least three separate detection methods should be employed and two of these methods are '
to be (1) sump level and flow monitoring, and ( 2) airbone parti-culate radioactivity monitoring. We will require the applicant. ;
level menitoring stated in the FSAR, in order to meet the first ,
part of Position C.3. The applicant has not provided an air-borne particulate radioactivity monitoring system. Not having an airborne particulate radioactivity monitoring system is accept-able provided that the applicant provides an alternate monitoring system which meets the qualifications of the airborne particulate !
system. The applicant has not proposed any alternate at this j time. In conformance with Regulatory Guide 1.45, Position C.3, i the third method of detecting leakage is the monitoring of drywell
! cooler condensate flows. Regulatory Guide 1.45, Position C.6, l 1 , requires the airborne particulate monitoring system to be seismic l- Category I. The applicant must provide a seismic Category I
+ airborne radioactivity monitoring system or a seismic Category I acceptable alternate leakage monitoring system.
112-1
BCGS ,
DSER Open Item No.112 (Cont'd)
The applicant has not 'provided . information concerning the systems testing and calibration frequency and capability during power operation of the plant in accordance with Regulatory Guide 1.45, Position C.8. The applicant has committed to specifying the
' maximum allowable identified and unidentified leakage rates as 25 gym and 5 gym, respectively, in .the technical specifications.
Thus, the guidelines of Regulatory Guide 1.45, Position C.9, are met. Until the applicant provides the information stated above on the leakage detection systems, we cannot make any con-clusions as to the acceptability of the systems. We will report l
resolution of this item in a supplement to this SER.
RESPONSE
For the HCGS definition of intersystem leakage, refer to Sec-1 tion 1.14.1.7.
For a discussion on leak detection for the four systems noted, refer to the following sections:
'l. Safety Injection System (high and low pressure systems) -
Se ction 5.2.5.2.1 (o ) .
- 2. Residual Heat Removal System (inlet and discharge ) -
} Section 5.2.5.2.1 (o).
- 3. Reactor Core Isolation Cooling Sfstem - Section 5.2.5.2.1 (m) ,
- 4. High Pressure Coolant Injection System (steam side ) -
Section 5. 2.5-2.1 ( 1) .
4 Section 5.2.5.2 has been revised to indicate that the drywell floor and equipment drain sump leakage rate indications are class lE and are located on main control room panel 10C604.
- Sections 1.8.1.45 and 5.2.5.2 have been revised to address the l
concerns of positions C.3 and C.6 of Regulatory Guide 1.45.
Section 5.1.5.2 has been revised to identify that the dzywell l
equipment and floor drain sump level monitoring instrumentation is seismic Category I.
l r Sections 5.2.5.9 and 11.5.2.2.15 have been revised to provide information concerning testability. .
4
( F65(4) 112-2
. 4
' -- , -- . . . . - . . , . _ . , - , . - - , , . . - , . . , - . . . . , - - - . - . . . - _ - - . - _ . _ _ . _ - ~ ~ -. - - - . -
_ __ . _ . . . _ ~ . _ _ . _ . _ _ .__ .- . - _ _ _ . _ _ _
. . . .- . -. . ..~ ... .-.
~
b}f gdy}6
' lpo.QcO, yt i 6
- f \
BCGS FSAR (j'.9 ,
10/83 y'
See Section 5.2.3 and 6.1 for further discussion and Section 1.8.2 for the NSSS assessment of this Regulatory Guide.
1.8.1.45 Conformance to Reculatory Guide 1.45, Revision 0 May
. 1973: Reactor Coolant Pressure Boundary Leakace i
Detection Systems l BCGS is designed to comply with Regulatory Guide 1.45, with the
< exceptions, clarifications, and amplifications discussed below.
Paragraph C.3 of Regulatory Guide 1.45 requires that three
- methods of leak detection be provided. BCGS does not employ an
! airborne particulate radioactivity monitor due to uncertainties in detecting 1 gym of RCPB leakage in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The uncertainties that affect the reliability, sensitivity, and response times of radiation monitors, especially iodine and particulate monitors,
! are discussed below.
The amount of activity becoming airborne following a 1-gpa leakage from the RCPB varies, depending upon the leak location and the coolant temperature and pressure, which affect the ,.
flashing fraction and partition factor for iodin'es and particulates. Thus, an airborne concentration cannot be correlated to a quantity of leakage without knowing the source of i the leakage.
- Coolant concentrations during operation can vary by as much as several orders of magnitude within several hours. These effects are mainly due to spiking during power transients or changes in
[
' the use of the reactor. water cleanup (RWCU) system. An increase in the coolant concentrations can give increased containment concentrations when no increase in unidentified leakage occurs.
Not all activity is from unidentified leakage. Changes in other sources result in changes in the containment airborne i
.oncentrations. For example, identified leakage is piped to the I
drywell equipment drain sump, but all sump and collection drains i are vented to the drywell atmosphere, thereby allowing
( particulates to escape, causing further measurement l uncertainties.
l The aswani, of activity that is detected depends upon the amount of plateout on drywell surfaces prior to reaching the detector )
intake. The amount of plateout is dependent on uncertain t ossa orsu IrzM //4 1.8-26 Amendment 2 j l
. l
^
.. , l l
ECGS FSAR 10/83 .,'
i
' antities, such.as location of the leak, distance from the
tectors, and the pathway to the detector.
Furthermore, background does under not normal exist. operating conditions a radiation-free 8 There is a buildup of activity concentration due to both identified and unidentified leakage.
- At high equilibrium activity levels, a small change in activity ;
level due to a small leak is hard to detect in the desired time
-interval. '
Although particulate monitors are available with sensitivities i covering concentrations expected in the drywell, previously l discussed uncertainties under operating conditions coupled with i any calibration and setpoint uncertainties make particulate i monitors a less reliable method of leak detection. . I five. i BCGSdoesemploy[theseseparateanddiverseleakdetection methods. The RCPB leak detection system consists of: '
u ns w e. ca 1 equy-
GE60R1 I AIR PARTICutAm perEc710N .sy. STEM).
- b. A drywell cooler condensate flow monitor
. I
- c. A noble gas monitorf IR UE.W 0F AN Mt P4KTicULATE CEE7EC10N SYSEO
~ /MSSA Y b '
i Paragraphs C.2 and 5 requirst that the leakage monitors be able to i
detect an increase in leakage of 1 gym in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The noble gas !
monitor can detect concentrations as low as 10-* eCi/ce, the l i
miniana activity concentration expected in the drywell based on the primary system coolant. Bowever, an' increase in 1 gym !
j leakage within an hour may be difficult to detect due to high i
equilibrium and buildup of background radiation.
activity levels for noble gases (10-* to 10-* .Ci/cc) !
The noble gas monitor is 4
capable of detecting leaks of approximately 10 gym and does so i very quickly due to the high diffusion rates of the noble gases. !
l The drywell floor drain sump level monitor and the drywell cooler condensate monitor can detect fluid flows of 1 gpa in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Bowever, fluid flow is not always a direct indication of RCPB !
leakage because of free communication between the suppression chamber and the drywell. The drywell atmosphere is not !
necessarily drywell coolers. saturated due to the water vapor removal by the l Hot water can evaporate from the torus and '
o DSER OPEN I M 1.8-27 Amendment 2 I
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Eul NCGS INSERT D
- d. Se.ismic category I drywell pressure monitors e, seismic category I drywell temperature monitors.,
Leakage flows into the drywell floor and equipment drain sumps are not measured directly due to physical configuration which makes it impractical'to do so. As stated in Section 5.~2.5.2, leakage flow into the sumps is calculated based on the rate of change of level in the sumps.
sump pump starts and stops and duration of pumpout are monitored by the Class IE radiation processor. An alarm is annunciated in the main control room whenever pumpout duration exceeds a predetermined time limit. Total sump pumpout can be calculated based on the
, duration of pumpout and the constant known flowrate of the sump pump provided that only one pump is required to lower the sump level. The starting of the second pump is a positive indication of excessive leakage inte the sump or is an indication that the first pump has failed with either event requiring operator action.
The high-high level condition which initiated the operation of the second pump is annunciated in the main control room.
4 I
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- T
, '? HCGS FSAR 10/83 enter the drywell. The water will condense and register on the drywell cooler cendensate monitor. The condensate drains into s the drywell-floor drain sump and will register on the sump level monitor. Therefece, during times of suppression pool transients,
' such as froetheat gp from main steam safety / relief valve (SRV) or HPCI system testing, evaporation from the suppression chamber will obscure values of RCPB leakage.
Pos requires that the leakage detection sys g
'S9 /,
capable of per heir functions afte c event that does not require plant s detection system is capable of operatin opera is earthquake (OBE) and a DBA. 'N. level monitor is used o _ eeculatory M 45 and 1.97 purposes. N __
DR., C.6 also suggests that at least one atra leax cet -
method s o in functional after an SSE. Thi ity t 'does not exist in design. The the RCPB leak
, detection system is to monito rity of the RCPB so that if there are any changes ant can ly shut down.
Since the plant aut down after an SSE, the etection
- system have to remain functional after an SSE, t
~
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l Position C.7 requires that indicators and alarms for each leakage
! detection system should be provided in the main control room. ,
!* Procedures for converting various indications to a common leakage l
_ equivalent :should be available to the operators. The calibration
of.the indicators should account for needed independent M
1 variablesc . l m
Position C.7 is' further clarified by Standard Review Plan .
Section 5.1.5, III.5 which requires that if monitoring is computerized, backup procsdures should be available to the operator. !
L' -
/A/ SE/27~ A- -
drywell sumps and drywell air coolers leakage moni i L systems, and level change is electronically ttted h from level sen a local esdiation proc LRP) which processes these signa in turn t s processed data for L- indication and alarms, leve alculated flow rates to the a
centrnE radiation proc ) in omputer room. Data in
!. the CKP-is avai o the operator on the a keyboard :
printer erminal CRTc and/or annunciated in control l C99 V l c
.-l l % , osER OPEN ITEM // M 1.8-28 Amendment 2 1____.._._.__. -
~.s .. . . - - ~ . . . -, ....__..... .
. ' Wey j HCGS INSERT A i The drywell air coolers' leakage monitoring and noble gas monitoring systems' signals are processed by local radiation processors which then transmit the processed data to the main control room via the central radiation processor (CRP). The CRP in turn makes this indicating.and alarming information available to the control room operator via CRT. displays.
These signals are processed locally by local radiation processors
-(LRPs) which are provided with digital readout indicators. These indicators provide information to the operator in the same format (using the same engineering units) as the information provided by the CRP through the CRTs in the main control room. S'ince these indications are of the same format, procedures for converting
- the LRP indication to a common leakage equivalent (to that nor-mally provided in the main control room) are unnecessary.
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incocthe leakage signal's are processed locally with capability 8 i for Idcal readout, procedures for converting various indications a r d fc to a ec'amon leakage equiv,alent are not provided to the operatorsj n l
Jackup procedures ;;: nr. provided to the operator 1..;; ;;; "-
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Position . es that leakage detection s equipped to readily permit t or operabil calibration during plant operation. This capa not provided on RCPB leak containment, because detection instrumen' nside the calibratio sting cannot be performed in e e ent during reactor operation.
For further discussion of the RCPB leak detection system, see Section 5.2.5.
! 1.8.1.46 Conformance to Reculatory Guide 1.46, Revision 0, May 1973: Protection Acainst Pipe Whip Inside Containment The criteria set forth in Regulatory Guide 1.46 are design bases for BCGS. See Section 3.6.2 for further discussion of pipe break design and Section 1.8.2 for the NSSS assessment of this Regulatory Guide.
I oszR OPEN MEM ///
1.8-28a Amendment 2
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HCGS INSERT B As described in section 5.2.5.2, displays of drywell equipment and floor drain sump levels (which are not dependent of the non-lE plant computer systems) are provided on panel 10c604 in the main control room.
Position C.B requires that the leakage detection systens should be equipped with provisions to readily permit testing for opera-bility and calibration during plant operation. This is interpreted to mean channel functional testing as defined in the technical specifications (Chapter 16) . Calibration of the leakage detection systems is performed during plant outages per the techn4 cal speci-fications. Calibration of the drywell floor and equipment drain sump level monitoring systems can not be performed at power due to the fact that the sensors are located inside the drywell and are therefore inaccessible during power operation. Rosemount 1153 transmitters are used throughout the plant and are typically
. calibrated on an 18 month cycle (reference MUREG'-0123) . This ~
model transmitter is used for the sump level' transmitter. In addition, the calibration accuracy of these transmitters can be observed on an ongoing basis by comparing the level readings with known independently measured sump levels at which the sump pumps start or stop. The pumps are started and stopped using electro-mechanical float switches. It should also be noted that the rate of change readings (sump inflow) obtained from these trans-mitters will be substantiallyThe free from the effects of drift due sensors for the drywell cooler to the sampling frequency.
condensate flow monitoring systems and the drywell temperature j
monitoring system are also located inside the drywell (and L
therefore inaccessible during power operation). However, these sensors-are RTDs and access to them for normal instrument channel calibration is not required. The remaining leak detection monitoring h-E systeme discussed above have the capability of being calibrated during operation.
l r
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D5EE 12. 9/i1
NCGS FSAR 3/33 the RWCU pump heat exchangers and the reactor recirculation pump .
seal and jacket cooling heat exchangers. The RACS sensor monitors radiation emanating from a continuously flowing RACS water sample which is taken at a point downstream of the RACS pumps.
Eigh radiation in the SACS water or the RACS water indicates intersystem leakage. The affected sensor and its associated monitoring channel will activate an alarm in the main control I room when the radiation exceeds a predetermined limit.~ No isolation trip functions are performed by these channels.
These radiation channels are part of the process radiation I monitoring system described in Section 11.5. . <
l' High levels in the SACS or RACS head tanks may also indicate intersystem leakages from the sources given above. High level in ,
either head tank will activate an alarm in the main control room. l 5.2.5.2 Leak Detection Instrumentation and Monitorina ,
5.2.5.2.1 Leak Detection Instrumentation and .%nitoring Inside Primary Containment
- a. Floor drain sump level and flow - The normal design leakage collected in the floor drain sump includes unidentified leakage from the control rod drives (CRDs), valve flange leakage, component cooling water, service water, air cooler drains, and any leakage not connected to the' equipment drain sump.
- /A/SE.A~ T~ C. -
1 1 transmitter is used in the drywell floo n sumps fed into a local microproces . level change'in the will be convert ow rate by the processor. Abn les ates are alarmed in the main control room 'on in excess of background leak uld indicat nerease in reactor - " n leakage from an uniden source in j
gym within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- b. Equipment drain sump level and flow - The equipment drain sump collects only identifled leakage and valve stem packing leakoff collectively. This at:sp receives .
oszaorzu m //A 5.2-46 Amendment 2 i
i
. Ib 1 BCGS I
INSERT C A Class 1E level transmitter is used to monitor the drywell floor drain sump with the level signal being supplied to a Class IE radiation processor of the Class lE radiation monitoring system (RMS) (panel 10C604) located in the main control room. A level change in the sump is converted to a flow rate by the processor and leakage rates can be displayed continuously at panel 10C604 and are available, via data link, at the operator's console CRT.
An increase in unidentified leakage in excess of technical speci-fication limits is alarmed in the main control room.
The floor drain sump level monitoring instrumentation is qualified to remain functional following a safe shutdown earthquake (SSE) o e
4 j
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i i
f l
~ b 5 *i t- Il' L -
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b/
g 3 ygAR 8/83 l piped desinage from pump seal leakoff and reactor j vessel head flange vent drainage. The equipeent daain sump instrumentation is identical to the floor = drain ;
sump instrumentation. )
f
- c. Drywell air cooler condensate drain flow - Condensate
- from the drywell air cooler is routed to the floor l drain sump.
ght e in each of two drain lines f and is trapped l drywell a e drains int ntrolled by a local '
by a closing solen a the drain line is steroprocess risin ignal to een evel transmit er that Flow in each of the two drain headers from the eight drywell coolers (four coolers per header) is monitored by a flow sensor. The flow signal from each flow sensor is processed by a local radiation processor which transmits the flow data to the main control .
room, via the central radiation processor, for indi-cating and alarm functions. Any flowrate increase exceeding technical specification limits will be alarmed in the main control room.
This flow monitoring instrumentation is capable of operation following seismic events wh{ch do not require ,
plant shutdown.
l l
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- 1
i BCGS FSAR 8/83 C to differentiate between, identified and unidentified leakage is discussed in Sections 5.2.5.4, 5.2.5.5, and 7.6.
5.2.5.7 Sensitivity and coerability Tests a i .
Sensitivity, including sensitivity testing and response time of the leak detection system, and the criteria for shutdown if i leakage limits are exceeded, is covered in Section 7.6.
Testability of the leakage detection system is contained in Section 7.6.
5.2.5.8 Safety Interfaces .
The Balan'ce of plant-GE Nuclear Steam Supply System (NSSS) safety l interfaces for the leak detection system are the signals from the monitored balance of the plant equipment and systems that are part of the nuclear system process barrier, and associated wiring '
and cable lying outside the NSSS equipment.'
C.
5.2.5.9 Testina and Calibration
- - /A/$68.T~ E 5.2.5.10 Conformance to Reculatory Guide 1.45 For a discussion of compliance with Regulatory Guide 1.45, see Section 1.8.1.45. .
5.2.5.11 SRP Rule Review
- SRP 5.2.5 acceptance criterion 11.1 requires that leak detection system integrity must be maintained following an earthquake, as per GDC2. This is met through Regulatory Guide 1.29 positions C-1 and C-2.
j f*
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i INSERT E l
Testing and calibration will be in conformance with the Technical Specifications and will consist of channel checks and channel functional tests during power operation. Channel calibration will be done during refueling outages.
Testing and calibration of the noble gas monitor is discussed in Section 11.5.2.15.
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. c information about the HEPA and charcoal filter efficiency and
' 1 condition. )
11.5.2.2.12 Radwaste Area Exhaust Radiation Monitoring System The RAE RMS is located in the exhaust duct for radwaste area compartments in which there is equipment that has a possibility of releasing airborne radioactive materials (Refer to <
Figure l'1.5-1). The RAE RMS is upstream of the filters and will be exposed to higher concentrations than the RES RMS, thus 4
- allowing earlier detection of any problems in the radwaste areas of the auxiliary building. The RAE RMS has the same components l
and functions as the RBVSE RMS described in Section 11.5.2.2.8.
11.5.2.2.13 Gaseous Radwaste Area Exhaust Radiation Monitoring System
- The gaseous radwaste area exhaust (GRAE) RMS is located in the exhaust duct for the recombiner compartments (Refer to Figure 11.5-1). This allows earlier detection of airborne
' radioactive materials than is possible by downstream monitors where the concentrations are more diluted. -The GRAE RMS has the same components and functions as the RBVSE RMS described in Section 11.5.2.2.8. There are no filters upstress of the I location.
I 11.5.2.2.14 Technical Support Center Ventilation Radiation Monitoring System The technical support ce'nter ventilation (TSCV) RMS is located in i the inlet plenum for the technical support center (Refer to Figure 11.5-1) The purp6se of the TSCV RMS is to detect radioactive materials ih the inlet air. The TSCV RMS has the same components as the RBVSE RMS described in Section 11.5.2.2.8.
If the concentration exceeds the trip setpoint, an alarm at the CRP alerts the operator to manually transfer from the normal air supply to an emergency recirculation and filtration mode.
11.3.2.2.15 Drywell Leak Detection Radiation Monitoring System The drywell leak detection (DLD) RMS sonitors the gaseous radioactive materials in the drywell (Refer to Figure 11.5-3).
The design objective of this system is to monitor reactor coolant .
ossa orzu rrzu //p 11.5-18 Amendsent 1 -
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pressure boundary (RCPB) leakuge in accordance with Regulatory Guide 1.45. Conformance to Regulatory Guide 1.45 is discussed in Section 1.8. The capability to do so declines as the normal in-containment background of gaseous radioactive materials increases because of the accumulation from identified leaks. An air sample is extracted and returned through penetrations that are isolated
- by the PCIS described in Section 7.3.1.1.5. The DLD RMS components are one inlet and one outlet stub on the east side of the drywell, penetrations, and isolation valves. There is also a shield sample chamber, a beta scintillation detector, and an LRP.
The high-high alarm indicates excessive leakage from the RCP3.
The DLD RMS is seismically qualified to operate under conditions during which the reactor is operated. The functional ,
requirements and descriptions of other leak detection equipment are discussed in Sections 5.2.5 and 7.6.1.3. Provision for a grab sample is included.
- msge r f -
11.5.2.2.16 Reactor Auxiliaries Cooling System Radiation Monitoring System The reactor auxiliaries cooling system (RACS) RMS monitors a sample extracted from the RACS (Refer to Figure 11.5-1). The .
RACS RMS has the same components as the liquid radwaste RMS. The high-high alarm indicates leakage into the RACS from the heat exchangers that are serviced by the RACS.
11.5.2.2.17 Safety Auxiliaries Cooling System Radiation Monitoring System The safety auxiliaries cooling system (SACS) RMS has two monitors, A and B, one for each of the two SACS loops (Refer to Figure 11.5-1). The SACS RMS monitor samples extracted from the SACS. The SACS RMS has the liquid radwaste RMS.
l l
sample chambers are part of the SACS pressure dary boun,The and.are SACS seismically qualified. The high-high alarm indicates leakage into the SACS heat exchangers from the safety auxiliaries served by the safety auxiliaries cooling system.
l 11.5.2.2.18 Heating Steam Condensate, Waste Radiation Monitoring System The heating steam condensate, waste (HSCW) RMS monitors a sample of the condensate flow from the liquid waste management system (Refer to Figure 11.2-4). The high-high alarm / trip indicates both leakage of radioactive materials from one or both of the osER OPEN ITEM //g 11.5-19
%[ \
HCGS INSERT F Testing and calibration of the DLD RMS will be in conformance with the Technical Specifications and will consist of channel checks and channel functional tests during power operation. Channel cali-bration will be done during refueling outages.
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O
[ e. V 1-BCGS DSER Open Ites No.141 (D6ER Section 9.1.3)
SPENT PUBL POOk. COOLING AND CLEANUP SYSTEN The applicant has not provided a discussion of the means to provide cooling to the spent fuel pool after a safe shutdown earthquake which fails the non-seismic Category I skimmer tanks in such a manner as to plug the tank drains. Thereforr , we cannot conclude that this design satisfies the requirements of General Design Cri-terion 2, " Design Bases for Protection Against Natural Phenomena,"
and the guidelines of Regulatory Guides 1.13, " Spent Fuel Storage Facility Design Basis," Positions C.1, C.7 and C.8, and 1.29,
" Seismic Design Classification," Positions C.1 and C.2.
The applicant has not adequately addressed the concern of high-and moderate-energy piping system failures and the means to pro-tect these systems (refer to Section 3.6.1 of this SER.) Thus, we cannot conclude that the requirements of General Design Cri-terion 4, " Environmental and Missile Design Bases," and the guide-lines of Regulatory Guide 1.13, Positions C.2, are satisfied.
The system is accessible for routine visual inspection of the system components. Both fuel pool cooling pumps are required to operate at all times to remove the maximum normal heat load.
Thus, the cooling system does not meet the single failure cri-terion. The applicant has not committed to include the portions of the cooling and cleanup systems which are not normally operat-ing in the inservice inspection and periodic functional testing The programs as described in Sections 6.6 and 3.6.6 of the SRP. Thus, applicant has not specified the frequency of the testing.
j the requirements of General Design Criterion 45, "Inspe ction of Cooling Water Systems," and 46, "Tec ting of Cooling Water Systems,"
are not satisfied.
i The spent fuel pool cooling system will maintain the fuel pool water t emperature at .135'F, with a heat load of 16.0 MBtu/ hour based on decay heat generation from 3,668 fuel bundles (maximum storage) and both cooling trains in operation. This is the normal discharge from 15 fuel cycles. The spent fuel pool cooling system I
l consists of two pumps with a common suction line and a common dis-charge line, which feeds two heat exchangers with a common inlet i
line and a conson outlet line. Each pump and each heat ex cha nger have a manual isolation valve on the inlet and manual isolation valve on the outlet; thus, each component can be independently
! isola ted . If one pump or one pump and heat exchanger were not available under these conditions, the pool temperature would ex-l coed the 140*F specified in the Standard Review Plan. The pool t
cooling must maintain a pool temperature of less than 140*F with ,
l any single active failure.
l 1
i F67(1) 141-1 l
6 l
BCGS i
DSER Open Item No.141 (Cont'd)
Bowever, the full flow by-pass line around the non-safety-related
~
_ lc ea nup water system has not been clearly indicated in the FSAR fig ures. Therefore, we cannot conclude that the requirements of General Design Criterion 44,' ' Cooling Nater," are met.
Until the applicant provides acceptable responses, we cannot con-clude that the system conforms to the requirements of General Design Criteria 2, 4, 44, 45, and 46 as they relate to protection from natural phenomena, missile and environmental ef fects, cooling water capability, inservice inspection, and functional testing and the guidelines of Regulatory Guides 1.13, Positions C.1, C.2, !
C.7, and C.8,1.29, Positions C.1 and C.2 relating to the system's f unctional design and seismic classification. The spent fuel pool cooling and cleanup system does not meet the applicable acceptance criteria of SRP-9.1.3. 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 its evaluation of the spent fuel pool sampling and monitoring. To complete the review, the following information is needed:
(1) Describe the sampling procedure, analytical instrumentation, and sampling frequency for monitoring spent fuel pool purity.
( 2) . State the radiochemic.11 limits for initiating corrective action.
The applicant's response should consider permissible gross gamma and iodine activities and the demineralizer decontamination factor.
~
RESPONSE
See the revised response to FSAR Question 410.55 and revised Se ction 9.1.3.3 for a discussion of the seismic response of the skimmer surge tanks.
Section 3.6 describes the method of protection against dynamic ef fects associated with postulated ruptures in high and moderate energy piping located both inside and outside the primary con-tai nme nt. The FPCC and Tcrus Water Cleanup Systems are classified as moderate energy systems. The failure of high and other moderate energy piping on FPCC and torus water cleanup systems has been evaluated in Section 3.6. Because of the physical separation of
' the FPCC and torus water cleanup systems from high and other moderate energy piping, it has' been concluded that s ' postulated piping failure in high and/or other moderate energy piping will not adversely affect the operation of these systems. Therefore, insert /b F67(1) 141-2
eEPDx TELECCPIER 495230- 8-84; 9:33CN 2 3016524161-* in 4
%I Ihmert A Page 141-2 1here are no high or moderate energy lines above or near the spent fuel pool whose failure would adversely affect the spent fuel pool or the storage racks. Piping within the spent fuel pool is set =nie=11y designed noderata energy piping. A crack in this piping would not have an adverse effect on either the spent fuel pool or the storage racks. -
a
HCGS DSER open Item No.141 (Cont'd) it can be concluded that the systema design meets the requirements of GDC No. 4, " Environmental and Missile Design Bases", and the
.3 guidelines of Regulatory Guide 1.13, Position C.2. For discussion on moderate energy leakage in the common spent fuel pool cooling 4
pump discharge line, see response to guestion 410.48.
The spent fuel ~ cooling system does not perform a specific function in shutting down the reactor or in mitigating the consequences of an accident; therefore, does not meet the criteria for being included in~ASME B&PV Code Section XI testing requirements.- A n h suscAr 8 '
As discussed below, there is no single active failure within the FPCC system which will result in the loss of a FPCC heat exchanger.
Bowever, two system configurations (one FPCC Pump and two FPCC heat exchangers and one FPCC pump and one FPCC heat exchanger) have been evaluated as requested. The results are provided in Table 141-1.
The evaluation indicates that in the event of a single active failure of one FPCC pump, the spent fuel pool temperature 'could reach 152*F, which exceeds the SRP 9.1.3 limit of 140*F. It is conservatively estimated that the fuel pool temperature could exceed 140*F for 26 days under these conditions. This is based ;
on worst-case assumptions. A maximum SACS water temperature of 95*F is assumed. In , addition, a maximum accumulation of spent f uel is assumed stored in the fuel pool, i.e. , 16 consecutive '
ref uelings at 18 month inters 's, . to fill the high density racks
' to their maximum capacity of 184 spent f uel assemblies (which exceeds the SRP 9.1.3 requirements). It is also assumed that the last 1/3 core is placed in the spent fuel pool as quickly as i practical after-shutdown, i.e. 8 days. This is slightly longer than the 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> recommended by SRP 9.1.3 and is based on the BWR servicing and refueling improvement program - Phase 1 Summary Report prepared by GE (NEDG-21860).
Review of Figure 9.1-5, Sheet 1 of 2, confirms that there is no single active failure mechanism within the FPCC system which will render one heat exchanger unavailable (e.g., inadvertent valve actuation. ). In e$dition, preventive maintenance on the FPCC heat exchangers can be scheduled prior to the ref ueling outage to ensure the availability of both heat exchangers The plate type to :
remove the calculated maximum normal heat load. ,
l 767(1) 141-3 )
. - _ -.. _ .-~- - - - . - .. _ -
l
-XEROX TELECOPIER 49533D- C-843 9:35AM 3016524161, 3 IN 6 IdSEET'd , EJ (
, f.141-3 HCGS I
. 2**nsert B l Spent fuel pool cooling and cleanup system piping will be visually inspected once every 18 months. System pumps will be start /stop tested once every 30 days if they have not been used within the previous 30 days.
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HCGS DSER Open Item No.141 (Cont'd) h'est exchanger is a low maintenance component with long life
, gaskets that are expected to be replaced about every 5 years.
4 In addition, .the manufacturer has performed a reliability and maintainability analysis on the plate-type heat exchangers which l indicates that failures are extremely unlikely. Therefore, i failure of one FPCC heat exchanger is not considered to be a credible event. Table 141-1 also provides the maximum pool heatup rate for these postulated events. The time to reach !
the maximum temperature is conservatively based on a constant heatup rate.
A single active failure of one of the SACS cooling loop inlet valves to the FPCC system heat exchangers has also been evaluated. '
This could render the FPCC system heat exchangen unavailable for a s ho rt period of time. However, the fuel pool cooling is re-established in a short period of time by either manually re-open- ;
ing the af fected valve or providing cooling from the standby SACS l loop. It is anticipated that the fuel pool heat-up rate during this short period will not cause the fuel pool temperature to -
exceed 1400F.
During normal operation, the of fsite doses from the fuel pool are negligible. - Elevating the fuel pool ~ temperature to 152*F l or 174*F would result in a slight increase in the evaporation i
rate. This slight increase would result in a slight increase in the offsite doses, however, the doses would still be negli-gible and well below the 10CFR20 limits.
A bo lt- II* S.f' j uf to thf*6 Elevated pool temperature [will not cignificen:l/# affect the per-formance of the fuel pool filter domineralizer. grhe only adverse factor is a slightly reduced capacity for ion exchange. Up to 175'F, approximately a 10% reduction in run length of the dominer-alization cycle is expected with no change in the filter capacity.
The response to Question 410.46 has been revised to address the failure of one FPCC pump. As stated in Section 9.1.3.1.j, normal makeup capability is provided to makeup evaporation losses and to ensure that fuel pool cooling is maintained.
FSAR Figure 9.1-5, Sheet 1 of 2, identifies the f ull flow bydpass lines around the non safety-related filter-demineralizer system (10'-HBC-06 2, 6"-HBC-062, 6*-HCC-015 ) . This mode of operation is discussed in Section 9.1.3.2.3, and meets the requirements of General Design Criterion 44, " Cooling Wate r" .
FSAR 5ection 9.1.3.2.2.4 has been revised to provide'the requested information on spent fuel pool sampling.
[ F67(1) 141-4
1 I
I TABLE 141-1
- Single Active Failure Analysis for FPCC System l l l System configuration l I I I 1 FPCC Pump 1 FPCC Pump Description and 1 BX of Parameter l and 2 HX l l
_ l l
I l
l
- 1. Normal Max. heat load l 16.1 x 10 6 BTU /hr ! 16.1 5 106 BTU /hr lI 1 l l 95'F -
l Cooling Water (SACS) 95*F l 2. Tempe rature l
i .
l l I
- I 152*F l 174*F
- 3. Maximum Fuel Pool Tempe rature Heat-up Rate 1.02*F/hr 2.26*F/hr 4.
l- l 2.13 gpm 5.99 gpm l l 5. Evaporation Rate l l I
~
l i l 17.3 hrs l Time to reach the 16.7 hrs l 6. Maximum Tempe rature l
l l l l assuming the Fuel Pool l l l Tempe rature at 135'F. l l l
l 1
l l
i l
l i
I
- . _ _ _ _ .. _ __., _ ._ . _ . . _ _ _ . . , _ , . _ _ . . . . . _ . _ _ . _ _ . . . . _ _ _ _ _ _ _ - . . _ _ _ _ _ , . . _ . _ _ _ _ . _ _ . ~ . _ _ _
i
- b. The FPCC system cooling loop (consisting of skimmers,
, surge tanks, fuel pool. cooling pumps, fuel pool heat enchangers, and interconnecting loop piping) and the emergency fuel pool water makeup piping are designed to i meet Seismic Category I requirements, except for the surge tanks. The surge tanks are of non-Seismic
- Category I design, but are embedded in a Seismic
- Category I concrete structure that provides the pressure boundary for this part of the FPCC system
- cooling loop. The FPCC system purification loop, consisting of the filter-demineralizers, their interconnecting piping, and associated equipment, is non-Seismic Category I.
- c. The FPCC system is designed to handle the decay heat released by all anticipated combinations of spent fuel that could be stored in the fuel pool. The pool water temperature is maintain at a maximum of 13581' under M*n t.no design load oz z 108 Stu/h. This heat load is l6 hamed onJf consecutive refuelings with one-third of the core removed during each refueling, and on a 4' refueling frequency of 18 months.
- d. The FPCC system is designed to permit the residual heat removal (RNR) system to be operated in parallel with the FPCC system through a crosstie, to remove the maximum heat load and to maintain the bulk water temperature in the spent fuel pool at or below 1500F, with a maximum anticipated heat load of 34'1Ms 108 Btu /h. This is based on one full core load of fuel at the end of a fuel cycle, plus the decay heat of the spent fuel discharged at the d previous refuelings. An b :c.nse.r c 1. tglytm
- e. The FPCC system is designed with additional capability to provide a source of makeup water to ensure against l
loss of fuel pool cooling, in compliance witit Regulatory Guide 1.13.
i
- f. The FPCC system is designed to monitor fuel pool water i
level and potential leakage paths and maintain a sufficient level above the spent fuel elements to i provide radiation shielding for normal building occupancy. . ,
l :
tssa opsu ITun /
/9/
l 9.1-17 l
.fROx TELECCPIER 495120- S-84; 9:39Gt1 2 30165241314 2010 YY*
DSER 141 ,
Insart 1: Page 9.1-17 parey..pi d.
If required, cne Nm pung and one RHR heat exchanger can be aligned to auynant the FPCC systasa through the system crosstie. Ibr this system configuration, a heat load greater than 45 millicm Btu /hr can be nmoved from the spent fuel pool with a SACS inlet tenparature of 95'F and a spent fuel pool tyrature of 150'F.
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. = . - _ _ - . - - - -. - . - . - _ _ - - - - - _ - - _ _ _ - -
l l
l acos rsAn l T
l 9.1.3.2.2.2- Fuel Fool Cooling Pumps Two single-stage, horizontal, motor-driven, centrifugal, half-capacity recirculation pumps circulate water through the FPCC system. The pumps are piped in parallel and take suction from .
the skimmer surge tanks through a common header. The pump motors, pump control circuits, and power supplies are Class 1E.
Each pump is provided with controls for starting and stopping the motor as follows: For normal and accident operation, the primary control in the NCR is used. If it is necessary to start or stog either pump when the NCR is inaccessible, the control in the '
' remote shutdown panel (RSP) is used. Each pump is automatica113 stopped by skimmer surge tank low-low level, low suction ;
pressure, or low discharge' flow. :
9.1.3.2.2.3 Fuel Fool Heat Exchangers Two half-capacity, plate-type heat exchangers are provided for the FPCC system. They are designed to transfer the system design M7 eat load of M108 Stu/h from 1358F pool water, flowing at the system design flow rate of 1400 gym, to the safety ausiliaries cooling system (SACS) at its maximum temperature of 95'F.
The heat exchangers are arranged in parallel. Fuel pool heat exchanger inlet and outlet temperatures are monitored and recorded'by the control room integrated display system (CRIDS).
1 9.1.3.2.2.4 Fuel Pool Filter-Demineralizer System L
l The cleanup loop of the FPCC system includes a filter-l demineralizer system located in the auxiliary building. The ,
' filter-deaineralizer system consists of two vessels, located separately in shielded cells, and two holding pueps. One of the vessels, including its holding pump, normally serves as a spare.
The holding pumps and the equipment common to the two vessels, including the resin tank with agitator, dust evacuator, and resin eductor, and the associated piping, valves, and instrumentation, I
are located in a separate room adjacent to the vessel cells.
The filter-demineralizer system also services the torus water ;
cleanup system for the purification of suppression pool water. I l
,/
l ossa orms Itsu /9 /
/
9.1-20 j
~
HCGS FSAR The stainless steel filter-demineraliser vessels are of the pressure precoat type. A tube nest assembly consisting of the tube sheet, clamping plate, filter elements, and su p et grid is inserted as a unit between the flanges of the vessel. The filter elements are stainless steel and are mounted vertically in the <
vessel. Air scour connections are provided below the tube sheet, and vents are provided in the upper head of each vessel. The '
filter elements are installed and removed through the top of each vessel. The holding elements are designed to be coated with powdered ion exchange resin as the filtering medium.
The fuel pool filter-demineralizers maintain the following effluent water quality specifications:
Specific conductivity at 25*C, micronho/cm 50.1 ,
, pH at 25*C 6.0 to 7.5 ,
4 Heavy elements (Fe, Ng, Cu, Ni), ppe 0.05 1
Silica (as Sio n), ppe <0.05 Chloride (as Cl-), ppe <0.02 Total insolubles, ppm 90% removal to a
. minimum of 0.01 ppe LMst&
The filter-demineralizers are designed to be backwashed periodically with water to remove resin and accumulated sludge
, from the holding elements. Service air pressure loosens the material from the holding elements and the backwash slurry drains 1
through the gravity drainline to the waste sludge phase separator l in the solid waste management system.
l The resin tank provides adequate volume for one precoating of one j filter dominera11:er vessel.
The resin eductor transfers the precoat mixture of resin to the holding pump suction line at a flow rate of 4 gpe.
The holding pumps are designed to recirculate a uniform mixture of resin through the filter-demineralizer vessel L.ing precoated '
! at a flow rate of 1.5 gym /fts of filter element surface area, and .
to automatically start and maintain the precoat material on the ,
filter elements when the system flow rate f alls below the value necessary to keep the precoat on the elements.
i ossa opsu ItsM /W
.,/
HCGS Insert 4
' The influent and ef fluent water of the Spent Fuel Pool Damineralizer is continuously monitored by on-line pH and conductivity instrumentation. In addition grab samples of the influent water will be analyzed 1/ week for C1 and for gamma isotopic and 1/ month for heavy metals, and the ef flu-ent water will be analyzed weekly for C1, SiO ,2suspended solids, . H-3 and for gamma isotopic, when the cleanup system is in operation.
Decontamination factors (df) of ) 10 are expected for any C1 present and 75 for isotopes of I and Co. Resin beds will be regenerated and/or replaced when these df's are not gchieved.
The pressure drop across the Domineralizer is continously monitored and when the DP increases to a predetermined level the ion exchange media will be replaced. Typically this level is 30 PSID.
.xnsert B rhe. Spe.<,4 +~ael Pool bes,,1<, e radr a. u sa lIl b e.
oy e uded a s eeg u o reaf
- ma an da.in eodia.b en Isas/s an th e. r e.fu. sjo;,y p /a.4/or<n /e.ss tha n '.
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- __-~___-__________-- ____ _ ___ _____________
l . .
! NCGS FSAR I Y s .
b*
Figure 9.2-13 shows the refueling water transfer pumps. Manual valves are aligned.to establish the fill flow path, and the pumps are manually stas:ted. provision is made to permit filling the *
- cask pool or the reactor well independently.
i Aer o After refueling or spent fue hipping activities are completed,
! either the reactor well and the ryer and separator storage pool, 1 or the cask pool are drained via gravity drain lines to the
! refueling water transfer pumps' uction header from which the 4 water is pumped through t to ondensate dee.ineraliser and back to tto CST. Alternately,.the reactor well, dryer and separator storage pool, and cask pool can be drained via gravity drain lines to the fuel pool pumps' suction header from which the
- water is pus through the fuel pool filter-domineralisers and back to the . During refueling operations, a portion of the cooling system flow is diverted from the fuel pool return line to the reactor well via the reactor well diffusers. The
recirculation pattern established by the diverted flow allows the l i
heated water that rises above the reactor core to be cooled in
, the fuel pool heat enchangers. This supplements the parallel RHR system (operating in the shutdown cooling mode) decay heat removal from the core region. When the shipping cask contains I
) spent fuel and is in the cask pool, a portion of the FPCC system flow is diverted from the fuel pool diffusers to the cask pool ,
. via the cask pool diffuser. When the RHR system is operated in i parallel with the FPCC system to provide fuel pool cooling during the full core unload case, one RHR pump takes the suction from the skimmer surge tanks, circulates the water through one RHR ,
heat exchanger, and returns it to the spent fuel pool via the two i j
RHR intertie return diffusers.
1 The cask pool is filled via the refueling fill line and drained through a condensate domineraliser or the fuel pool filter- ,
domineralisers in the same manner as the refueling volume is filled and drained. Filling of the cask pool is normally done i
prior to spent fuel loading into the cask, and draining is normally accomplished after cask loading.
108 Stu/h. This is the 4
S The FPCC system design heat load is 16.secay heat espected gros' # conse af 2 t.t ... ;t -?!: This Str M The FPCC e stem's maximum
! 341 Asat load is htdd;10* Stu/h. is the decay est espected if it becomes necessary to unload the entire core from the reactor andjstore it in the pool, which already tontains spent fuel from
- thifidARWyrevious refuelings. For this core unload design condition, an RNR heat eschanger is operated in parallel with the FPCC system. The RER system is only interconnected when the reactor ),
is shut down, and larger-than-normal batches of spent fuel, such 1
i oszA OPEN ITEM / #/[ I'l"IO i
- . . - - - - _ ~ _ - _ _ - - _ _ _ - - - . _ - _ - - _ . . . --
4 1
BCGS FSAR 1
i draining the suppression pool if it is ever necessary. In this mode of operation, the torus water cleanup pump takes suction from the torus and circulates the water through a fuel pool filter-domineraliser and to the CST. Operator action is l necessary to terminate torus water cleanup operation, escept on low pump suction flow.
9.1.3.3 Safety Evaluation j
The FPCC system cooling loop (skimmers, skisser surge tanks, fuel pool cooling pumps, fuel pool heat enchangers, interconnecting loop piping), and the emergency fuel' pool water makeup system are
- designed to the requirements of Seismic Category I, except for
! the surge tanks. The surge tanks are of non-Seismic Category I .
l design, but are embedded in a Seismic Category I concrete i structure that provides the pressure boundary for this part of the FPCC system cooling loop. AThe interconnecting piping between I RHR and the FPCC system is designed]to Seismic Category I ,
l requirements. M sutn The cooling water return lines to the spent fuel pool, associated with both the FPCC and the RNR systems, penetrate the walls of the spent fuel pool horizontally above the normal pool water
); l level. Each of these cooling water return lines is provided with
}
two vacuum breakers to prevent the water from being siphoned out i of the pool. No piping connections are made to the pool below j the normal water level to prevent any accidental lowering of the l
l water level. Therefore, there is no operator error or FPCC i system malfunction that could result in draining the spent fuel i pool and uncovering the stored spent fuel. The fuel pool structures are also designed to Seismic Category I requirements.
If a line break occurs in the non-Seismic Category I purification i loop, the remotely operated purification loop isolation valves close automatically on surge tank low-low level or by operator action. ;
or Any leakage be, tween the fuel pool gates, cask pool gates,'well
- through the vessel to drywell seal or drywell to reactor
l seal is alarmed in the NCR. A segmented leak channel system i' behind the liner veld seams is provided to detect fuel pool, cask pool, reactor well, and dryer and separator pool leakage. .
I I The torus water cleanup system suction and return piping from the
' torus, out through and including the prisary contaiheent isolation valves on each lir.e, is designed to Seismic Category I l requirements. The torus water cleanup system piping to and from i
i
! 9.1-26 '
i osER OPEN ITEM ///[
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l 4
-~ - . i l 1"!1" - 1 4 ;
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)
The swge tape were designed to withstand an external loadi '
of 699 lb/ft during congtruction.
(approaimately 300 lb/f t ) was This lower enternal than the loadingdesign induced valuex due >
to the'eso of a Rover pour rate.
a stress level less than one half of the design stress level in "
' the tank shell.
As analysis has been performed to detensine the ef fect og This analysis seismic loads on the skimmer surge tanks.
, indicates that the induced stresses resulting from the seismic loads are insignificant (approximately it of the stresses due ll to concrete placement) : - - w and
.e tothat plustheth skimmer surge tanks wi
- M Mei-3.*~
not fail 1.. :: hc. ,
(o ll,w;n a so.fc s hutd ow n c_arth t
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i G
e
l' l BCGS FSAR '
10/83 l I
! The combined use of the techniques mentioned allows an accurate )l' i assessment of the SFPFD and permits the determination of when a specific unit should be changed.
i 3-Reactor well water level is monitored in the CRIDS, and an !
annunciator alare is provided in the NCR to indicate a low l reactor well water level during refueling. An interlock trips i the refueling water transfer pumps on low reactor well level when
! the well is draining back to the CST after fuel transfer.
I i The torus water cleanup pump fa started and stopped from the FPCC filter-domineraliser panel and the pump is stopped automatically hp low suction flow. Low suction flow is alarmed on the FPCC i falter-domineraliser panel. A pressure indicator is located in !
the pump discharge line. !
ll l 9.1.3.6 SRP Rule Review t
.i '
i
! Acceptance Criterion II.1.d.(4) of SRP 9.1.3 limits the water -
i temperature in the fuel pool to 140*F at the maximum heat load
- with the normal cooling system operating in a single active failure condition. )
The bulk water temperature in the fuel pool grmyt B =l*k-
- r.l ;r.
- 5::t ::cher.;;; ir. :er" ira after the f :"b:::e:r:f """
-t '"M ing i
i e,;y . However, the RHR system can be manually aligned to i
provide supplemental cocling in order to avoid bulk water temperatures in excess of 140er. !
l l 9.1.4 FUEL HANDLING SYSTEM i i 9.1.4.1 Desian Bases The fuel handling system is designed to provide a safe and
! effective means for transporting and handling fuel from the time it reaches the plant until the time it leaves the plant after
! post-irradiation cooling. Safe handling of fuel includes design l considerations for maintaining occupational radiation esposures as low as reasonably achievable (ALARA) during transportation and l
) handling. ,
j
. i l i 4
! 9.1-30 Amendment 2 ossa orsa nun /V/
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>EROX TELECOPIER 495330- S-84; 9:48AM 3 20165241614 3220 VeV 1.
HCGS FSAR TABLE 9.1-1, Page 1 of 3 FUEL POOL COOLING AND CLEANUP SYSTEM AND TORUS WATER-CLEANUP SYSTEM DESIGN PARAMETERS Skimm'er Surae Tanks Type Vertical, cylindrical Quantity 2 Design pressure, psig 0 Design temperature,~*F 212 Capacity, gallons 3750 Fuel Pool Coolina and Cleanuo System Pumos Type Horizontal, centrifugal, single-stage Quantity 2 Design pressure, psig 150 Design temperature, OF 212 Rated flow per pump, gpm 700 l Developed head (TDH) at~ rated 257 l flow, feet l- Motor horsepower, each 75 L
! Fuel Pool Heat Exchancers l
Type Plate Quantity 2 Design pressure,~psig
, Cold side 150 L Hot side 175 l- '
Design temperature, *F Cold side 150 Hot-side 212 Btu /h dehxIO*
Rating,
. (,ctt 9s 93+cs ecl 13FY5d Q Flow, each, gpm 700 V
DSIR CPEN ITEM /Y/ ,
MGS MER .
TABLE 9.1-2 ,
' F1tEL POOL COOLtMG AND CLEAT 4UP SYSTEM MEAy RR800 VAL CAFA:'ITy idD st4EEUF 9808tDastaWTS a
Parameter value_at _ esormal Weat ImaJ Wales at lessames seat Lsad 1/3 of core: 4-1/2 yr irradiation time 1/3 et cores 4-1/2 yr irradiatten time j geantity of feet it days decay time t
8 days decay time 1/3 of cores 4-1/2 yr irradiation time 1/3 of core 3 yr irradiatten time l 5% days decay time 10 days decay time 4
1/3 of core: 4-1/2 yr arradiation time 1/3 of cores 1-1/2 yr irradiatise time i
1104 days decay time it days doce, time 1/3 of core 4-1/2 yr arradiation time 1/3 of cores 4-1/2 yr arradiattaa ties 1652 days decay time 558 deve decay time 4 1/3 of core: 4-1/2 yr irradiation time 1/3 of cores 4-1/2 yr 1rradiation time 3200 days decay tisse 1105 days decay time l
1/3 of core 4-1/2 yr irradiation time 9/3 ef cove + 4. y,Egy (WaJimb*d# *E***"
2748 days decay time j gg b'a**a=
1/3 of core: 4-1/2 yr 1rradiation time g f b ,,,, gg a.
g,,,, g . + YA Y 3296 days decay time 1/3 of core 4-1/2 yr irradiation time 1.T.*
-Yo* j1 MM N#
JyCYTAds*M b
! 1844 days decay time 73 af C8T8' t 4g.g g 4 JA %4 **'*-
1/3 of core: 4-1/2 yr arradiation time l 4392 days decay time 'g . gats M 1
1/3 of core 4-1/2 yr arradiation tie. Ys *
- w^ : 4 '/ y g.
j 4940 days decay time 3T % . g.
1/3 of core: 4-1/2 yr irradiation time of Carts q. .K y TT g Stet days decay time p y g JA.
- 1/3 of core
- 4-1/2 yr irradiation time 6034 days decay time IS 4 U T4-
- 4 - Va. '$V U f gNg . . ,k , g ,
1 4-1/2 yr 1rradiation time 1 S9 1. M *dl8 g ,, e,y, . 4y3 p mahal*'** p
- 1/3 of cores 6ste days decay time i Ja Nar***-
1 1
1/3 of core 4-1/2 yr irradiation time 7132 days decay time NU 4 (198 g, kW 4-1/2 yr arcadiation time
- 4ys95 gg, ,
1/3 of cores Ys ,f g,,.
) p g g,g,,. @ ;;.ite F
- 8i Can: noe days35decay ,t(meLivaa.m ma. Ha*
4
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A J
norsat aesten heat load Sl x 10* Btu /h S i.'L1 g g g, b y*3,ofd*T4
- g j
j
- - - r -r-m.mber of heat enchan,ere required 3
2 r, n eMa erstes I g4 a 19e ste/h
! news.am design heat load sf 2 ,
y ,ett v=w , m 7,,
.%e w,gf jgy t.c m.ie.p -treme.es d to evaporation losses ( g ogg,,a 4 a 4}*"**I*
2 com . g gy (t Summ.
' Makeup during normal operation Makeup rate for retuating 5 gem ]Sn *f p gg g,*%
I i
l
~
. l BCGS FSAR <tu/s3-TABLE 9.1-18 DECAY HEAT AND EVAPORATION RATES FOR LOSS OF SPENT FUEL POOL COOLING Description of Normal heat load in Maximum heat load @g/
the event the spent fuel pool in the spent fuel (1M z 10* BTU /hr) cool L&O z 10* BTU /hr)
- 9. o4 3 . 31 e ~2.
A Time to 17.2 hrs (*) .-9 hrs reach 212*F 73 5 3 Evaporation 34.4 gym 4&:t gym
~
rate .
Time required 2 hrsO) 1,'2 5 ra W
- C g 9,(O ,
to initiate 1/2 hr(s) l makeup water 20 hrs (8) g g (g) ..
. Notes:
O) An estimated time of 2 hrs would be required to couple the fire hose fill connections to the Seismic Category I SSWS loops to provide fresh water makeup to the fuel pool.
(a) It has been conservatively estimated that the SSWS can be initiated within 1/2 hr by operator action in the MCR to provide makeup to the fuel pool.
O) It has been conservatively estimated that after 20 hrs one RHR pump loop.and the associated heat exchanger can be used for fuel pool cooling. = cerlin;; :;; i: ~ initMad fr W
- 2. G% .
(*) Sincetheentirecoreisinthefuelpool,theRyforfuelsystem can be made availabe " i 1;;;; i..;.ly ^ '2 L ,
pool cooling.%> ;:::f:: ::tir: in t M "'"' - lie ts) This assumes a normal maximum heat load after J rconsecutive refuelings.
D8zR OPEN ITEM /h Amendment 2
i
~
l BCGS FSAR _ in/st j OUESTION 410.46 (SECTION 9.1.3)
Verify that the normal heat load after refueling can be removed by using one spent fuel pool cooling systes pump and both heat exchangers. With this system configuration, verify that the pool '
water temperature will remain less than 140*F and specify the length of time that that (SIC) second heat exchanger is required.
If this cannot be verified, justify this deviation from the Standard Review Plan.
RESPONSE ,e g fr2* F The fuel pool temperature could :::::d M00 Ywith normal maximum heat load in the fuel pool, one spent fuel pool cooling pump and both heat exchangers operating for fuel pool cooling. Twsw t c With the above system configuration it has been conservatively estimated that after 90 days the fuel pool heat load will be such that only.one fuel pool heat exchanger is required for fuel pool cooling.
4
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410.46-1 Amendment 2
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BCGS FSAR 10/83 OUESTION 410.47 (SECTION 9.1.3)
Verify that the decay heat loads are based on NUREG-0800, Standard Review Plan, Section 9.1.3 and Branch Technical Position ASB 9-2.
RESP M .
The fuel pool heat loads are calculated based on NUREG-0800, Standard Review Plan, Section 9.1.3 and Branch Technical Position ASB 9-2.O.xxup g g,,p,g,wimq.
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DSER OPEM ITEM /p/ 410.47-1 Amendment 2 l
l I .-, . . . - - - - - . . - - . . . . . - - - - . - - . - _ _ - . - _ _-. - - - - - - - _ _ . _ . - .
. HCGS FSAR 20.':3-OUESTION 410.55 (SECTICE 9.1.3)
Provide a discussion of the means to provide cooling to the spent fuel pool after a safe shutdown earthquake which fails the skimmer surge tanks and plugs the tank drains. The results will be the loss of the spent fuel pool cooling system pumps due to cavitation from an iscilated suction line, loss of offsite power i from the earthquake, and,the unavailability of the RRR system from the loss of the common suction with the spent fuel pool cooling pumps. The worst single active failure should be considered as part of the discussion. If the pool is allowed to boll, then consideration must be given to the time required to clear the skimmer tank drains as compared to the minimum time required to achieve boiling; the continued reduction in worker efficiency as the ambient air temperature, humidity, and radioactivity increases; and the time required to bring the reactor to cold shutdown and thereby have an RHR cooling loop available to cool the pool. ,
1
RESPONSE
I nsideration of multiple failures of non-Seismic Category I
- components following a safe shutdown earthquake is beyond the design basis for HCGS. In particular, the postulated failure of both skimmer surge tanks is not considered credible because these
- " tanks" are, in fact, steel-lined voids in the Seismic Category I spent fuel pool wall.
Section 9.1.3.2 discusses the backup sources of makeup water available to supply the pool in the event normal cboling is lost and RHR cooling is not available.
3 M b* WM q,g,3 y Ms W S h' 0 '
l l
DSER OPEN ITEM /
l 410.55-1 Amendment 2
~
t
- - - - , , , , - - - - = , , , , . , ~ , - , ~ , - - , - , - - - . - - - , . - - . _ . - - - , ~ . ~ - - . _ . - . - _ . - - - , . - - . - - - . _ - - . - - - - . - -
- HCGS h( E Y J ~ ,
DSER Open Item No. 151 ( DSER Se ction 9.4.1)
CONTROL STRUCTURE ~ VENTILATION SYSTEM.
The CRS and CREF systems take outside air from a common tornado-missile-protected air intake. The air intake for the CERS system is also tornado missile protected; however, there is no protection for the nonsafety-related WAS system intake. The exhaust for the CABE,- WAE, CASE, and CAE systems are tornado missile prote cted.
Thus, the staf f concludes that the requirements of GDC 4, " Environ-mental and Missile Design Bases," are satistied. The air intakes have no chlorine" monitoring capability but do have radiation monitor-ing capability. ' Signals from the radiation detectors alarm in the
-control room, automatically isolate the fresh air intake from the control room HVAC system, and automatically start the CREF system to purify the fresh alr. There is no automatic operation associated with the redundant CREF system train upon loss of the operating system. The CRS and CREF systems are designed to maintain the operability of the equipment in the control room. The control room systems are designed to maintain the control room u6 der a positive pressure to minimize infiltration of gases intoThus, the con- the trol room except during 100% recirculation operation.
staf f concludes that the requirements of GDC 19, " Co n trol Ro om , "
and the guidelines of Regulatory Guide 1.78, " Assumptions for
- Evaluating the Habitability of a Nuclear Power PlantPositions . Control C.3, Room i During a' Postula ted Hazardous Chemical Release ,"
C.7, and C.14, are satisfied. We cannot conclude that the guide-lines of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release,"
Positions' C.4a and C.4d are satisfied.
The CR S, CREF, and CERS systems consist of two 100% ca pa ci ty tra in s of-filters. The CREF system consists of a prefilter, a HEPA filter,
- a charcoalThe filter, and a fan in series for the removal of radio-CRS and CERS systems consist of a prefilter, high activity.
efficiency filter, and a fan. There is no filtration of the ex-haust; however, it is isolated upon a high radiation signal.
Chilled water is supplied to the two 50% capacity cooling coils in each of the air handler units. The maximum ambient tempe is ra ture 94*F.
for which one train will maintain the one proper environment train of ventilation sys-
.The applicant must demonstrate that tems can maintain the compartment environmental . conditions within l
the qualification limits with an outside ambient tempera ture of l
lO2*F for all design basis accidents with the loss of the redundant ventila tion systems. Based on the above, we cannot conclude that the requirements of General Design Criterion 60, " Control of Releases.cf Radioactive Materials to the Environment," and the
- 1 151-1
un to v*uAUvvvU
, ; HCGS l
' DSER Open Item No.151 (Cont'd) l guidelines of Regula tory Guides 1.52, " Design, Testing , and Mainten- l
' ance criteria for Atmospheric Cleanup System Air Filtration and Adsorption Units of Light Watee-Cooled Nuclear Power Plants,"
<~ Posi t io n . C. 2, _ a nd 1.14 0, " Design, Testing and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light Water-cooled Nuclear Power Plants," Positions C.1 and C.2, are satisfied with respect to ensuring environmental limits for proper operation of plant controls under all normal and accidgent conditions, including LOCA conditions.
Ba' sed on _the above , the staf f concludes that the CSV systems are in conformance with the requirements of the GDC 2, 4, and 19 with respe ct to protection against natural phenomena, tornado missile protection, and control room environmental conditions and the guidelines of RGs 1.29, Positions C.1 and C.2, and ' 1.78, Positions C.3, C.7, and C.14, relating to the seismic classifi-cation and protection against hazardous chemical relea se ..and is, therefore, acceptable. We cannot conclude that the CSV systems are in conformance with the requirements of General Design Cri-terion 60 with_ respect to control of radioactive releases and 1.95, the guidelines of Regulatory Guide 1.52, Position C* 2, Positions C.4.a and C.4.d, and 1.140, Positions C.1 and C.2, re-lating to the design for emergency operation, protection of pe rsonnel against a chlorine gas release,'and normal operation.
We will report resolution of this item in a supplement to this SER. The HVAC systems which make up the.CSV systems do not meet the acceptance criteria of SRP Se ction 9.4.1.
RESPONSE
Evaluation o't accidents relating to the release of toxic chemicals including chlorine is addressed in FSAR Section 2.2.3.1.3.
Also, per DSER Section 6.4, Page 6-3:
"Wi th r e's pe ct to toxic gas protection, the staf f's evaluation in accordance with SRP Section 6.4, P.Gs 1.78 and 1.95 indicated that there is no danger to control room personnel from toxic chemicals, including chlorine, stored onsite or offsite, or tra nsported nearby -(See Se ctio n 2. 2. 3 ) . "
Section 9.4.1.3 has been revised to include reference to Sec-tion 2.2.3.L3.
The CRS* system provides cooling (with chilled water cooling coils) during normal operating conditions. The system also provides 4 -cooling, in conjunction with the CREF unit, in the event of an acc ide nt condition. i 151-2
APR 26 '84 0 2 6 3 3 3 9 HCGS DSER Open Item No . 151 ( Con t' d )
The function is either: ,
ea.
- 1. 1000 cfm outside p makeup mixed with 3000 cfm of room return air diverted through the CREF unit. The balance of air is recirculated from the air conditioned space or, 4
- 2. A 1004 recirculation mode, i.e., without outside air and with the use of the CREF unit. ,
See FS AR Se ction 9.4.1.2.3.
.L Function Mode Jf is . selected in the event of an accident condition.
When the outside ambient temperature condition is 102* F, 1000 cfm air is a minimal quantity (Xpproximat31 yJ.4% of the total air supply) which will increase the suppliestgtemperature by less than l'F. Therefore, this increase in temperature will not affect the operation of the plant ' controls due to the u of cooling coils as stated above. Since neither outside air is brought into the system nor is the control room exposed to sola load, outside ambient temperature of 102*F has no effect on unction Mode 2.
c.on+o f -com j re suNiny in a t emp e ro.+u.re a f 17 *fo +0fLI * $
p ers isin'ny for
~ /20 hou ru l i e , 4, hau rs P er d o-y b r so d=3s )
l l-F64/5 151-3
.,ArN 20 cu tJ g g a a a y
'i HCGS FSAR 3af er to the fcliewi..g sec:i: .s f::
- . ;r:.sc :::1-/ :: i23:sn ,, .
included in the design of the safety-related c:n:rol area HVAc systems:'
- a. Protection from wind and tornado effects - Section 3.3
- b. Flood design - Section 3.4
- c. Missile protection - Section 3.5 l
- d. Protection against dynamic effects associated with the l
postulated rupture of piping - Sec. tion 3.6
- e. Environmental design - Section 3.11
- f. Fire protection - Section 9.5.1. ..
Topc,C&nttb\s-$tNICA 4
- 2.S I 5 f.
9.4.1.4 Tests and Inspections The CRS, CERS, CREF, and CABE systems and their components ar;-
tested in a program consisting of the following: '
- a. Factory and in-situ qualification tests (see ,'
4 4 Table 9.4-6)
- b. Onsite preoperational testing (see Chapter 14)
- c. Onsite operational periodic testing (see Chapter 16).
Written test procedures establish minimum acceptable values for all tests. Test results are recorded as a matter of perforsance record, thus enabling early detection of faulty operating performance.
All equipment is f actory inspected and tested in accordance with the applicable equ'ipment specifications, codes, and quality assurande requirements. Refer to Table 9.4-6 for details of inspection and testing.
9.4-15 G
-g,-m,w,,,,.,.,,._w-r,_ .-w w ,,-,-w w ,r.-,n
- HCGS t -
DSER'Open Item No.153 ( DSER Section 9.4.5) bEV '
BNGINEERED SAFETY FEATURES VENTIIATION SYSTEM The safety related systems are designed to Seismic Ca'tegory I, Quality Group C requirements and are housed in the seismic Category.I, flood and tornado protected auxiliary building, thereby satisfying the requirements of GDC 2 and the guidelines 4
of RG 1.29, Positions C.1 and C.2. The applicant has provided tornado missile protection for the inlet and outlet louvers.
The system is separated from high-energy piping systems and internally generated missiles. [The applicant has not specified- l the maximum temperature inside the structure with all equipment I running during a 102*F summer day. The 102*F day is the maximum summer temperature recorded between 1948 and 1981 (refer to FSAR
- Table 2. 3-13 ) . Therefore, we cannot c'onclude that the . require- !
monts of General Design Criterion 4 are satisfied.] The inlet '
louvers have tornado-missle-protected barriers and are more than 30 ft above plant grader thus, the staf f concludes that the guidance of Item 2, Subsection A, of NUREG-0660, " Enhancement of onsite Emergency Diesel Generator Reliability," and therefore, the pertinent requirements of GDC 17, " Electric Power System,"
relating to the protection of essential electrical components from failure due to the accumulation of dust and particulate material, are satisfied.
RESPONSE
A. Service Water Intake Structure (SWIS)
With an extreme outdoor air temperature of 102*F, the SWIS room ambient temperature will rise ( from 104*F with design outdoor air temperature of 94 *F) to approximately 113 *F.
The manufacturer's design information and/or the equipment environmental qualification reports for all active, safety-related equipment and instrumentation in the service water intake structure which could be af fected by temperature has been reviewed. A temperature-of 113*F will not cause the failure of any of this equipment or instrumentation. This temperature persisting for a total of 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> (i.e. , 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per day for 30 days) will not have a signifcant impac t on the life of this equipment or instrumentation.
B. Standby Diesel Generator (SDG) Area l
Section 9.4.6.1 has been revised to indicate maximum space design temperatures. An extreme outdoor air temperature of l
153-1
_ 102*F would hcVe littlo cr_ no of fact en SDG arco HVAC Cyateen or safety-related equipment. Individual HVAC systems within the SDG area are discussed below:
- 1. IE Panel Rom Supply ,
The IE panel rom supply unit mixes 7000 cfm outside air with 34000 cfm return air and further cools this mixture using cooling coils before it is distributed. A rise in outdoor temperature from 94 *F to 102 *F would result
-in less than a 1.5'F rise in the mixed air temperature entering .the cooling coil. Because of reserve capacity in the cooling . coil, space temperatures will rise less than 1.5'F. / !
- 2. SDG Air Recirculation .
The SDG nir recirculation system recirculates 100 percent room air and is designed to maintain a space maximum of 120*F; thus, the system would be unaffected by a rise in outside air temperature to 102*F.
- 3. Switchgear Room Cooling The switchgear vom cooling units each six 1840 cfm out-side air with 9360 cfm return air and further cool this mixture using cooling coils before it is distributed. A rise in outdoor temperature from 94
- F to 102 *F would result in less than a 1.5*F rise in the mixed air tempera-cure entering the cooling coil. Because of reserve capacity in the cooling coil y space temperatures will rise less than 1.5*F.
- 4. Safety-Related Battery Room Exhaust i
e Air l's supplied to safety-related battery rooms by either the IE panel rom supply system or the switchgear room cooling system and is then exhausted by this system.
Based on discussions above, the temperature in the safety-related battery rooms will rise no more than 1.5'F above the design maximum temperature.
Temperature increases of less than 1.5'F
_ per4eds would have no ef fect on safety-related equipment 1
_ operation or environmental qualification.
Refer to DSER Open Item No.1. ,
p er s isti o[9o r cm. -toto I o E I20 hou PS C 1. C , le hou.e 3 p e.e cl.a.3 for- so clay 3 h d
h osum OPEN rfEN /5~.5. 153-2
HCGS FSAR indication _for these locations is provided in the main control
( room.
9.4.5.6 SRP Rule Review Justifications for deviations with SRP Section 9.4.5, Engineered Safety Feature (EST) Ventilation Systems, are presented in SRP Sections 9.4.1 and 9.4.2, which address the specific ESF ventilation systems of HCGS.
9.4.6 STANDEY DIESEL GENERATOR AREA VENTILATION SYSTEMS 9.4.6.1 Desion Bases The standby diesel generator (SDG) area ventilation systems maintain a suitable _ operating environment for the SDG rooms, the safety- and nonsafety-related battery rooms, switchgear rooms, .
SDG fuel oil storage rooms, electrical chases, corridors in the diesel area, and the SDG Class IE panel room during all modes of i plant operation. The heating, cooling, and ventilating systems for the SDG area consist of both safety-related and nonsafety-t,' related systems. The seismic classification and corresponding codes and standards that apply to the design of the system are discussed in Section 3.2. ,.
System Description
' '~~~ --
A t" vJ' 9.4.6.2 The SDG area is provided with the separate ventilation systems .
listed below and shown on Figures 9.4-15 and 9.4-16. Equipment design parameters are listed in Table 9.4-16. The systems are:
f
- a. Diesel area supply system - This system is nonsafety-related. It is composed of two 50%-capacity heating '
and ventilating units. It supplies air to the SDG area corridor, stairwells, and the electrical chases.
Outside air is taken from a Seismic Category I plenum and passed through an automatic outside air intake damper, low officiency and high officiency f11ters, an electric heating coil, a centrifugal supply fan provided with automatic inlet vanes, and an automatic 4 supply air sh'utoff damper.
osaa opsu Irax / f 3 9.4-74 Y
e 4
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QS G 8.T A pbEa-7 To FM MC- G d S. /
TZe. kea/m aoods; end vesHadon sfa},,#a
</and-Ap ywafor area is A ay >ad b iroausfa<i, ge f//oaiy .ajuce /e ys<aAnt .clueug noema/p/as/
, oukide d'st15s femesswe. o( 9@
has.afions,6aaJondu/4/ 7s *r av de/4: .
- a. /04 'E mximum m We Hk V yu,pme / ecoms, eweidor=,
efec/eies/ cAase.x, dese/fel deox rooms a n x 1 4s e die.re/yeseads- re w cu 4 do., f n:a n.
J. /so *F ew>aximum m ke. diese/y aade m du ks. oGesaf y a< ders an. esepiza.L.
.e. 8S*s msximum sn $e.. denfo/ egayme.,f rosas, inwAime-
^
rooms, haRay aAay ,coms a.a 4. atae/ysM sondo/ roans.
- d. 9a *E maximum m dia. <<snc'hjear rooms.
2
- 2. 77*ic .t ff /n We Nb4-y swams.
i e
DSER OPEN ITEM / 33
-e-- ,--,,--y- , - ~.--,me,.. _ , . _ - - _ , , ,,-.__,,,,,,,_m._w.,...-m...r,,,%-- --~.e, --v.---_.-
HCGS FSAR The discharge shutoff damper automatically opens when
\. the fan starts. The fans are started by handswitches located on the local panel.
The two safety-related battery rooms at elevation
- 163 feet 6 inches are provided with two 100%-capacity l exhaust fans. Makeup air to these battery rooms is provided by the diesel area Class 1E panel room supply system. Each fan is provided with a manual inlet shutoff damper and an automatic discharge shutoff damper, and a tornado protection, damper. During LOP, the fans are automatically connected to emergency Class 1E power from the SDG. The automatic disenarge shutoff damper opens when the fan starts. A low flow computer input actuates an alarm in the main control room upon loss of airflow, and starts the redundant fan i automatically. The fans are started by handswitches ,
located on the local panel. l
- d. Diesel area nonsafety-related battery room exhaust r system - The two nonsafety-related battery rooms at elevation 163 feet 6 inches are provided with two 100%~
capacity exhaust fans. Each fan has a manual inlet g shutoff damper, and an automatic discharge shutoff
- k. damper, and a tornado protection damper. Makeup air to 1 these battery rooms is provided by the diesel arem Class 1E panel room supply system. During LOP, the fans can be manually connected to SDG-backed non-Class IE power from the SDG. A low flow computer input actuates an alarm in the main control roos upon loss of -
l airflow, and automatically starts the redundant fan.
The automatic discharge damper opens when the fan starts. The fans are started by handswitches located on the local panel,
- e. Switchgear room cooling systems - These are safety-related systems. Each of the four switchgear rooms is )
provided with one Seismic Category I, full-capacity air cooling unit that has a centrifugal supply fan, a l
tor protection check damper at its outside ai in h __
l Juct , low efficiency filter, and two 100%-cap u ty ;
I chi ___ ater cooling coils. The air cooling unit can be isolated by the automatic outside air shutoff damper and by manual dampers located in the discharge and return ducts. A mixture of outside air and return air enters the switchgear room unit cooler for processing.
l The conditioned air is supplied to the switchgear room, battery charger room, battery room, and SDG control osaa orzu IrzM /f3 9.4-76 l
- * ~
HCGS TSAR room.of each respective SDG. Cooling coils are k, supplied with chilled water from the safety-related control area chilled water system. Chilled water piping is arranged so that one coil in each unit receives chilled water from loop A, and the other coil receives chilled water from loop B. During LOP, the cooling units are autosatically connected to Class 1E power from the respective SDG that they serve. Each unit cooler can be started by a handswitch located at the local. panel. .
f The low-flow switch for each fan actuates an alarm at the local panel, and in the main control room upon IcLe
.of airflow, and stops the operating fan. Alarms are also provided for high-pressure differential across the filter and for high or low return air temperature.
. y stina
- f. Diesel area Class 1E panel non suppl system - This d' 1upplie conditioned air sysces is safety-relatedto the four 7==battery orter rooms rooms, dC and two heating, ventilating conditioning (NVAC) rooms at elevation 163 feet, and the elevator machine roce at elevation 178 feet. One It is composed of two 100%-capacity HVAC units. unit runs while the
( - othee is on standby. The standby unit will automatically start upon failure of the operating unit.
-Outside air for each unit is taken from a separate Seismic Category I p'lenus. Each unit has a low and a high efficiency filter, an electric heating coil, a chilled water coil, and a centrifugal supply fan provided with automatic inlet vanes. The outside air raturn duct and discharge air ducts are provided with automatic shutoff dampers. The outside air duct is also provided with a tornado protection check damper.
A flow controller is provided that ensures a constant air volume. The cooling coil is supplied with chilled water from the auxiliary building control area chilled water system. Water piping is arranged so that the coil of one unit receives chilled water from loop A and the coil of the other unit receives chilled water from loop B. During LOP, the units are automatically connected to emergency Class 1E power from the SDG.
2ach unit cooler can be started by a handswitch located at the local panel.
The low
- flow switch actuates a local alarm upon loss of airflow and starts the standby units. Local alarms are also provided for high-pressure differential across the osaa orsu Irnx ff,y g.4-77
,n.,., - - , , . v.,-, . - , . , - - , .----.,,,.,,--,,__---.,,--._..._.____n, , _ _ _ . - - , , - _ _ _ , _ , - . - , , - . . , , , . . _ _ - . . . - - - , .
I 1
DSER Open Item No. 244 (DSER Section 8.3.3.3.1)
. ANALYSIS AND TEST TO DEMONSTR ATE ADEQUACY OF LESS TH AN SPECIFIED SEPARATION The applicant, by Amendment 4 to the FSAR, provided a description .
of physical separation between redundant enclosed raceways (covered I trays and open top raceways, and between non-Class lE trays and Class 18 conduit, as follows: i
- 1. In the cable spreading rooms, the main control room, relay room, and control equipment room, the separation is twelve inches (12") horizontal, and eighteen inches (18") vertical.
- 2. In all other plant areas, the separation is three feet
, horizontal and five feet vertical.
The applicant further stated that where the separation distances specified above can not be maintained, cable trays shall either be covered with metal tray covers or an analysis, based on test results, will be performed.
The staff concludes that the above separation meets the guide-lines of Regulatory Guide 1.75 and is acceptable except for the following:
(1) The use of 18 versus 36 inches of separation between race-ways is evaluated in Section 8.3.3.3.2 of this report, and (2) The use of an analysis to justify less than specified separation will be pursued with the applicant.
RESPONSE
The response to Question 430.52 has been revised to provide the requested analysis. dire cc,cy af co.cs o f' yfe je' )/ow,,, re,ooo t.s a.r e a. Ha.c.h ec/ fa r you.r- usc:
) u) y le t.Jo r aSe s , 7~s s+ Repoet No. s c, 7 i 9, bo hol Novemb e r 30, /980 i pecpo. red ke- S u s g u e. ha.nn s o,. nd S 4 co-m E l e.c.1rs' c. 3+o.+4 o n Foe e. \ e.c.:fr,* c a. I w .' r c.
- c. odole. i s o f Afi o n bo.cei e e M mode r .'oJ s te sf-r n s & k>.+ c. R es eo.vc k L. o.ho r o.+e <s' t. S , W
.0 F r- a.n M : a h+t.ch . m o,e c_k s o, l97 ?, freP aeccl %e Tol ed 0 ho e C.o n el w.M S e.po e o.:Noa NSi E cl; So o c.o m po.q
? e o } ro-m .
F70(8) e
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