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1 May 2, 1984 meting | 1 May 2, 1984 meting | ||
: 5. Draft SER Section"provided | : 5. Draft SER Section"provided | ||
.c in letter dated August 7, 1984 (Schwencer to Mitti) | .c in {{letter dated|date=August 7, 1984|text=letter dated August 7, 1984}} (Schwencer to Mitti) | ||
MP 84 95/03 01 , | MP 84 95/03 01 , | ||
- ~ ~ ' | - ~ ~ ' | ||
Latest revision as of 13:44, 24 September 2022
| ML20096E860 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/05/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8409070203 | |
| Download: ML20096E860 (88) | |
Text
7_.. 3
. Public Service
. Electnc and Gas Cornpany 80 Park Picza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570. Newark, NJ 07101 Robert L. Mitti General Manager Nuciear As urance and Regubtion September 5, 1984 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
'7920 Norfolk Avenue Bethesda, MD 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of' Licensing Gentlemen:
HOPE CREEK GENERATING STATION DOCKET NO. 50-354 DRAFT. SAFETY EVALUATION REPORT OPEN ITEM STATUS Attachment 1 is a current list which provides a status of the open items identified in Section 1.7 of the Draft Safety Evaluation Report (SER). Items identified as " complete" are those for which PSE&G has provided responses and no confir-mation of status has been received from the staff. We will consider these items closed unless notified otherwise. In order to permit timely resolution of items identified as
" complete" which may not be resolved to the staf f 's satis-taction, please provide a specific description of the issue which remains to be resolved.
~ Attachment 2 is a current list which identifies Draft SER Sections not yet provided. -
8409070203 840905 PDR ADOCK 05000354 a E PDR f e, l The Energy People um sw m
Directorfof Nuclear Reactor Regulation 2 9/5/84 s
In addition, enclosed for your review and approval-(see JAttachment'4) are the resolutions- to the Draf t SER open
' items listed in Attachment 3. A signed original of the required affidavit is provided to document the submittal of these items.-
.Should!you have'any questions or' require any additional information'on these open items, please contact us.
Very truly yours,
?
5
-Attachments / Enclosure I
C. D. H. Wagner USNRC Licensing Project Manager W. H. Bateman USNRC Senior Resident Inspector FM05 1/2' Wh, '
r:-
x UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-354 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Public Service-Electric and Gas Company hereby submits the
-enclosed Hope Creek Generating Station Draf t Safety Evalua-
~ tion Report open' item responses.
The matters set'forth in this submittal are true to the best
-of?my knowledge, intormation, and belief.
Respectfully submitted, Public Service Electric
-and Gas Company s
By: ~
//
T.'J. Martin Vice Presi t - Engineering-and Constr ction
- Sworn to and subscribed before me, -a Notary- Pub ic
- of-New Jersey, this f !- day of September 1984.
k /71H-
//f
' DAVID X. BURD NOTARYPUBUC OF NEW JERSEY Ny Comm. Expires 10-23 LGJ02/2 f
c;: -
DATE: 9/5/84:
ATTAONENT 1 DSER.' R. L. MITIL TO OPEN1 SECTION A. SOMENCER ITEM NLMBER' SUBJECT STATUS IEITER DATED 2.3.1 - Design-basis tenperatures for safety- Cmplete 8/15/84 related auxiliary systems
~ 2a' 2.3.3 - Accuracies of meteorological Cmplete .8/15/84
-measurments !(Rev. 1) 2b 2.3.3 ' Accuracies of meteorological Couplete 8/15/84
- measurements (Rev. 1) 2c 2.3.3' Accuracies of meteorological Ccmplete 8/15/84 measurements (Rev. 2)
-2d~ 2.3.3 Accuracies of meteorological Cm plete 8/15/84 measurements (Rev. 2)
-3a' 2.3.3 Upgrading _of onsite meteorological Cmplete 8/15/84 measurements progrm (III.A.2) (Rev.-2) 3b 2.3.3- Upgrading of onsite meteorological Cmplete 8/15/84 measurements program (III.A.2) (Rev. 2)-
- 3c- 2.3.3 Upgrading of ensite meteorological NRC Action measurements progra (III.A.2)
4 ~2.4.2.2 Ponding levels Cmplete 8/03/84 Sa 2.4.5 Wave inpact and runup cn service - Cm plete 8/20/84 Water Intake Structure (Rev. 1)
Sb 2.4.5 Wave inpact and runup on service Cmplete 8/20/84
. water intake structure (Rev. 1)
Sc 2.4.5 Wave impact and runup on service Cmplete 7/27/84 water intake structure 5d - 2.4.5 Wave inpact' and runup cn service Ccmplete 8/20/84 water intake strt 'ture (Rev. 1) l.
- 6a - 2.4.10 Stability of erosicn protection Cmplete 8/20/84 structures 6b -2.4.10 Stability of erosion protection Cmplete 8/20/84 structures 6c^ 2.4.10 Stability of erosicn protecticn Complete 8/03/84 structures l
M P84 80/12 1-gs e
- t. _
ATTACHMENT 1 (Cbnt'd)
DSER R. L. MITIL.TO OPEN SBCTIOi A. SOMENCER ,
ITEM NLMBER SUBJBCT STATUS IErrER DMED
! 7a 2.4.11.2 Thennal aspects of ultimate heat sink Conglete 8/3/84 7b 2.4.11.2 Thennal aspects of ultimate heat sink Omplete 8/3/84 i 8 2.5.2.2 Choice of maxinun earthquake for New Cmplete 8/15/84 l England - Piedinont Tectonic Province 9 2.5.4 Soil danping values Couplete 6/1/84 l 10 2.5.4 Foundatim level response spectra Couplete 6/1/84 11 2.5.4 Soil shear moduli variation Couplete 6/1/84 l
- 12. 2.5.4 Cmbination of soil layer properties Couplete 6/1/84 I 13 2.5.4 Lab test shear moduli values Conglete 6/1/84 14 2.5.4 Liquefaction analysis of river bottan Conglete 6/1/84 sands 15 2.5.4 Tabulations of shear rnoduli 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 Cmplete 6/1/84 i
7-18 2.5.4 Maxinun earth at rest pressure Ccuplete 6/1/84 coefficient i
19 2.5.4 Liquefaction analysis for service Cmplete 6/1/84 water piping 20 2.5.4 Explanation of observed power block Conglete 6/1/84 i 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
ATDONENT 1 (Cont'd)
DSER R. L. MITE 10 OPEN SECTIQ4 A. SOSENCER IT1!M IMEER SUBJECT STATUS IEFIER DPGED 23 2.5.4 - Clarification of ESAR Tablee 2.5.13 Cmplete 6/1/84 and 2.5.14 24 2.5.4 Soil depth nodels for intake Ccaplete 6/1/84 structure 25' 2.5.4 Intake structure soil modeling Ccmplete 8/10/84 26 ' 2.5.4.4 Intake structure sliding stability Cceplete 8/20/84 27 2.5.5 Slope stability Ccmplete 6/1/84 28a 3.4.1 Flood protection Ccmplete 8/30/84 (Rev. 1) 28b 3.4.1 Flood protection Cmplete _8/30/84 (Rev. 1) 28c 3.4.1 Flood protection Ccmplete 8/30/84 (Rev. 1)-
28d 3.4.1 Flood protection Coplete 8/30/84 (Rev. 1) 28e 3.4.1 Flood protection Ccmplete 8/30/84 (Rev. 1) 28f 3.4.1 Flood protection Cmplete 7/27/ 84
[
l 28g 3.4.1 Flood protection Cmplete 7/27/84 29 3.5.1.1 Internally generated missiles (cutside Ccmplete 8/3/84 containment) (Rev. 1) 30 3.5.1.2 Internally generated missiles (inside Closed 6/1/84 contairment) (5/30/84-Aux.Sys.Mtg.)
i 31 3.5.1.3 Turbine missiles Ccmplete 7/18/84 32 3.5.1.4 Missil'es generated by natural phenmena Ccmplete 7/27/84 33 3.5.2 Structurer , systems, and cuupxients to Ccuplete 7/27/84 be protected fern externally generated missiles s
l M P84 80/12 3 - gs
ATTACIMENT 1 (Cmt'd)
DSER R. L. MITIL 'IO CPEN SECTION A. SODENCER I'I1M NIMBER Sulk 7ECT STA'IUS IEITER DATED 34 3.6.2 Unrestrained whipping pipe inside Cmplete 7/18/84 containment 35' 3.6.2 ISI program for pipe welds in Ccmplete 6/29/84 break exclusion zone 36 3.6.2 Postulated pipe ruptures Ccmplete 6/29/84 37 3.6.2 Feedwater isolation check valve Ccmplete 8/20/84 cperability 38 3.6.2 Design cf pipe ruptree restraints Cmplete 8/20/84 39 3.7.2.3 SSI analysis results using finite Cmplete 8/3/84 element method and elastic half-space approach for contairment structure 40 3.7.2.3 SSI analysis results using finite Couplete 8/3/84 element method and elastic half-space approach for intake structure 41 3.8.2 Steel contaire.ent tuckling analysis Cmplete 6/1/84 l 42 3.8.2 Steel contairment ultimate capacity Cmplete 8/20/84 analysis (Rev. 1) 43 3.8.2 SRV/LIXbA pool dynamic loads Ccmplete 6/1/84 44 3.8.3 ACI 349 deviations for internal Ccmplete 6/1/84 i structures 45 3.8.4 ACI 349 deviations for Category I Cmplete 8/20/84 structures (Rev. 1) 46 3.8.5 ACI 349 deviations for fcundations Ccaplete 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)
I l
l M P84 80/12 4 - gs 1
ATTACSENr 1 (Cont'd)
DSER R. L. MITIL 10 TEN SECTIQi A. SOMNCER ITEM NUNIER SUBJBCE STATUS ETIER DMED 49 3.8.6 Gross concrete section Cmplete 8/20/84 (Rev. 1) 50 3.8.6 Vertical floor flexibility response Cmplete 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 Cmplete 8/20/84 (Rev. 1) 54 3.8.6 Ccubination of vertical responses Cmplete 8/10/84 (Rev.1) 55 3.8.6 Torsional stiffness calculation Cmplete- 6/1/84 56 3.8.6 Drywell stick model development Cmplete - 8/20/84 (Rev. 1) 57 3.8.6 Rotational tire history irputs Cmplete 6/1/84 58 3.8.6 "O" reference point for a2xiliary Cmplete 6/1/84 building model 59 3.8.6 Overturning mment cf reactor Caplete 8/20/84 building foundation mat (Rev. 1) 60 3.8.6 MAP element size limitations C mplete 8/20/84 (Rev. 1) 61 3.8.6 Seismic modeling cf drywell shield Cmplete 6/1/84 wall 62 3.8.6 Drywell shield wall boundary Cmplete 6/1/84 conditions 63 3.8.6 Rector buildina dme boundary Cmplete 6/1/84 cont.'tions M P84 80/12 5 - gs
ATEA09ENT 1 (Cont'd)
DSER R. L. MITIL 'IO CPEN SBCTION ,
A. SOSENCER TITM NLMBER SUEL7ECT STATUS LET11!lR DMED 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 Ccaplete 6/1/84 drop 66 3.8.6 Impedance analysis for the intake Cmplete 8/10/84 structure (Rev. 1) 67 3.8.6 Critical loads calculation for Ccmplete 6/1/84 reactor building ckme 68 3.8.6 Reactor building foundation mat Cmplete 6/1/84 contact pressures 69 3.8.6 Factors cf safety against sliding and Cmplete 6/1/84 l overturning cf drywell shield wall l
-70 3.8.6 Seismic shear form distribution in Ccmplete 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 Itodel load inputs Cmplete 6/1/84 j 74 3.8.6 Tornado depressurization Cmplete 6/1/ 84 j 75- 3.8.6 Auxiliary building abnormal pressure cmplete 6/1/84
- 76 3.8.6 Tangential stwar stresses in drywell C mplete 6/1/84 shield wall and the cylinder wall I 77 3.8.6 Factor cf safety.against overturning Ccmplete 8/20/84 cf intake structure (Rev. 1) 78 3.8.6 Dead load alculations Cm plete 6/1/84 79 3.8.6 Post-modification seismic loads for Cmplete 8/20/84 the torus (Rev. 1)
M P84 80/12 6 - gs l
ATDODENT 1 (Cont'd)
,DSER R. L. MITIL 'IO OPEN SECTION A. SOMNCER ITEN NL4eER SUBJECT STA'IUS [EITER DATED 80 3.8.6 Torus fluid-structure interactions Ccaplete 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 Ccmplete 8/20/84 design (Rev. 1) 83 3.8.6 Factors of safety for drywell Cmplete 6/1/84 buckling evaluation 84 3.8.6 Ultimate capacity cf containment Ccmplete 8/20/84 (materials) (Rev. 1) 85 3.8.6 Ioad ombination consistency Cmplete 6/1/84 86 3.9.1 Ccaputer code validation Cmplete 8/20/84 87 3.9.1 Information on transients Cmplete 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 Cmplete 6/29/84 systems-90 3.9.2.1 Vibration nonitoring program during Cmplete 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 Ccmplete 6/15/84 i penetrations l 93 3.9.3.1 Imd ombinations and allowable Ccaplete 6/29/84 stress limits l 94 3.9.3.2 Desic.p of SRVs and SRV discharge Ccmplete 6/29/84 Pi ping M P84.80/12 7 - gs i
ATIACHMENT 1 (Cont'd)
DSER R. L. MITIL TO OPEN SECTION A. SODEN3R ,
TIEM IOEER SUR7ECT STATtJS LETTER DA3ED
.95 3.9.3.2 Fatigue evaluation on SRV piping Cmplete 6/15/84 and IDCA downcmers 96 3.9.3.3 IE Information Notice 83-80 Ccaplete 8/20/84 (Rev. 1) 97 3.9.3.3- Buckling criteria used for cmponent Ccmplete 6/29/84 supports 98 3.9.3.3 Design cf bolts Ccmplete 6/15/84 99a 3.9.5 Stress categories and limits for. Ccmplete 6/15/84 core support structures 99b 3.9.5 Stress categories and limits for Conplete 6/15/84 core support structures 100a 3.9.6 10CFR50.55a paragraph (g) Ccmplete 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 Ccmplete 8/20/84 valves 102 3.9.6 Isak testing cf pressure isolation Cmplete 6/29 /84 valves
- 103al 3.10 Seismic and dynamic qualification of Ccmplete 8/20/84 l mechanical and electrical equipment l
103a2 3.10 Seismic and dynamic qualification cf Cmplete 8/20/84 mechanical and electrical equipment 103a3 3.10 Seismic and dynamic qualification of Ccmplete 8/20/84 f mechanical and electrical equipment l
t 103a4 3.10 Seismic and dynamic qualification cf Cmplete 8/20/84 mechanical ard electrical equipment M P84 80/12 8 - gs
ATrJO90NT 1 (Cont'd)
ESER~ R. L. MITTL 10 CPEN SIITICBI A. SOlWENCER IT1!N NLDEBER SURTECT STATUS IEFIER DMED 103a5 3.10 Seismic and dynmic qualification of Cmplete 8/20/84 mechanical and electrical equipment 103a6 3.10 Seismic and dynamic qualificaticm of Complete 8/20/84 mechanical and electrical equipnent 103a7 3.10 Seisnic and dynamic qualification of Caplete 8/20/84 mechanical and electrical equipnent 103bl 3.10 Seisnic and dynamic qualification of Couplete 8/20/84 mechanical and electrical equipnent 103b2 3.10 Seisnic and dynamic qualification of Cmplete 8/20/84 mechanical and electrical equipnent 103b3 3.10 Seisnic and dynamic qualification of Couplete 8/20/84 mechanical and electrical equipnent 103b4 3.10 Seisnic and dynamic qualification of Caplete 8/20/84 mechanical and electrical equipnent 103b5 3.10 Seisnic and dynamic qualification of Conglete 8/20/84 mechanical and electrical equipnent 103b6 3.10 Seisnic and dynamic qualification of Complete 8/20/84 mechariical and electrical equipnent 103cl 3.10 Seisnic and dynamic qualification of Cmplete 8/20/84 mechanical and electrical equipnent 103c2 3.10 Seisnic and dynamic qualification of Complete 8/20/84 mechanical and electrical equipnent 103c3 3.10 Seismic and dynamic qualification of Ca plete 8/20/84 mechanical and electrical equipnent 103c4 3.10 Seisnic and dynamic qualification of Ca plete 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
ATDOMENT 1 (Cont'd)
DSER R. L. MITIL TO OPS 6 SBCTIGE A. SOSENCER ITEM IN3eER SUBJEC.T STATUS IETIER DAITD
.105 4.2 Plant-specific nedianical fracturing Canplete 8/20/84 analysis (Rev. 1) 106 4.2 Applicability d seismic andd IDCA Caglete 8/20/84 loading evaluation (Rev. 1) 107 4.2 Minimal post-irradiation fuel Canplete 6/29/84 surveillance program
'108 4.2 Gadolina thennal conductivity Canplete 6/29/84 equation 109a 4.4.7 'IMI-2 Iten II.F.2 Conplete 8/20/84 109b 4.4.7 'IMI-2 Itan II.F.2 Canplete 8/20/84 110a 4.6 Nnctional design d reactivity Canplete 8/30/84 control systems (Rev. 1) 110b 4.6 Nnctional design d reactivity Cmplete 8/30/84 control systens (Rev. 1) lila 5.2.4.3 Preservice inspection program Canplete 6/29/84 (cmponents within reactor pressure boundary) 111b 5.2.4.3 Preservice inspection program Cmplete 6/29/84 1
(conponents within reactor pressure boundary) lile 5.2.4.3 Preservice inspection program Caplete 6/29/84 (canponents within reactor pressure l
boundary) ll2a 5.2.5 Reactor coolant pressure boundary Cmplete 8/30/84 leakage detection (Rtrv. 1) ll2b 5.2.5 Reactor coolant pressure boundary Conplete 8/30/84 leakage detection (Rev. 1) l l
M P84 80/1210 - gs i
+
i ATTAOMENT 1 (Cont'd)
DSER R. L. MITTL TD OPIN SECTION A. SCHWENCER I11M - NIMBER SUBJECT- STATUS LEITER [ATED 112c 5.2.5 Reactor coolant pressure boundary Cmplete 8/30/84 leakage detection (Rev. 1) 112d' 5.2.5 Reactor coolant pressure boundary Couplete 8/30/84
' leakage detection (Rev. 1) 112e 5.2.5 Reactor coolant pressure boundary Cmplete 8/30/84 leakage detection (Rev. 1) 113 5.3.4 GE procedure applicability Conplete 7/18/84 114 5.3.4 Coupliance with NB 2360 of the Sunener Conplete 7/18/84 1972 Addenda to the 1971 ASME Code 115- 5.3.4 Drop weight and Charpy .v-notch tests Complete 9/5/84
.for closure flange materials (Rev. 1) 116 5.3.4 01arpy v-notch test data for base Conglete 7/18/84 materials as used in shell course No. I 117 5.3.4 Ocmpliance with NB 2332 of Winter 1972 Canplete 8/20/84 Addenda of the ASME Code 118 ".3.4'
> Isad factors and neutron fluence for Cmplete 8/20/84 surveillance capsules 119 6.2 1MI it s II.E.4.1 Otmplete 6/29/84 120a 6.2 1MI Item II.E.4.2 Complete 8/20/84 120b 6.2 1MI Item II.E.4.2 Ocmplete 8/20/84 121 6.2.1.3.3 Use of NUREG-0588 Otmplete 7/27/84 122 6.2.1.3.3 Tenporature profile Ccmplete 7/27/84 123 6.2.1.4 Butterfly valve cperation (post Ccmplete 6/29/84 accident)
. M P84 80/12 11 - gs
. 1 ATDOBENT 1 (Cont'd)
D6ER R. L. MITIL 10 CPSI SECTIGI A. SOlWENCER ITBI NINGER SI&7ECT STATUS IJrffER DA2TD 124a 6.2.1.5.1 RPV shield annulus analysis Complete 8/20/84
( Rev. 1) 124b 6.2.1.5.1 RPV shield annulus analysis Co g lete 8/20/84 (Rev. 1) 124c 6.2.1.5.1 RPV shield annulus analysis Caplete 8/20/84
( Rev. 1) 125 6.2.1.5.2 Design drywell head differential Complete 6/15/84 pressure 126a 6.2.1.6 Redundant position indicators for Caplete 8/20/84 vacutsa breakers (and control roam alarms) 126b 6.2.1.6 Redundant position indicators for Caplete 8/20/84 vacutan breakers (and control rocan alarms) 127 6.2.1.6 Operability testing of vacuta breakers Caplete 8/20/84
( Rev. 1) 128 6.2.2 Air ingestion Conplete 7/27/84 129 6.2.2 Insulation ingestion Conglete 6/1/84 130 6.2.3 Potential bypass leakage paths Conglete 6/29/84 131 6.2.3 Administratics of secondary contain- Complete 7/18/84 ment openings 132 6.2.4 Containment isolation review Caplete 6/15/84 133a 6.2.4.1 Containment purge system Couplete 8/20/84 133b 6.2.4.1 Containment purge system Caplete 8/20/84 133c 6.2.4.1 Containment purge system Ccanplete 8/20/84 M P84 80/12 12- gs
.e ATIAO NENT"1 (Cont'd)
DSER R. L. MITIL '1D CPEN SECTION A. SOMlNCER ITEM NLDEER SUBJECT STATUS IErI1R DmD 134 6.2.6 Contairment leakage testing Complete 6/15/84 135 6.3.3 I.PCs and LPCI injection valve Cmplete 8/20/84 interlocks 136 6.3.5 ' Plant-specific IDCA (see Section Ccmplete 8/20/84 15.9.13) (Rev. 1) 137a 6.4 Control rom habitability Cmplete 8/20/84 137b 6.4 Control rocm habitability Cmplete 8/20/84 137c 6.4, Control rom habitability Cm plete 8/20/84 1 38 6.6 Preservice inspection program for Cmplete 6/29/84 Class 2 and 3 cmponents 139 6.7 MSIV leakage control systess Cmplete 6/29/84 140a 9.1.2 Spent fuel pool storage Cmplete 8/15/84 (Rev. 1) 140b 9.1.2 Spent fuel pool storage Coplete 8/15/84 (Rev. 1) 140c 9.1.2 Spent fuel pool storage Ccuplete 8/15/84 (Rev. 1) n 140d 9.1.2 Spent fuel pool storage Cmplete 8/15/84 (Rev. 1) 141a 9.1.3 Spent fuel cooling ard cleanup Ccmplete 8/30/84 ,
system (Rev. 1) l 141b 9.1.3 Spent fuel cooling and clearup Cmplete 8/30/84 system (Rev. 1) 141c 9.1.3 Spent fuel pool cooling and cleanup Cmplete 8/30/84 system (Rev. 1) w M P64 80/1213 - gs
_ - . . _ . _ _ . . . . . . . . . , _ . _ - . . . _ , _ . - - . - , ._ -..._,_.~...-_.,.m.- , _ . - . , _ - - _ . . - _ _ . _ _
ATTACHMENT 1 (Cont'd)
I DSER R. L. MITIL 10 j OEW SECTIQ4 A. SONENCER !
ITEM Nt3EER SUELTECT STATUS IErNR IATTlD l 141d 9.1.3 Spent fuel pool cooling and cleamp Ccuplete 8/30/84 system (Rev. 1)
A.
141e 9.1.3 Spent fuel pool cooling and cleamp Ccaplete 8/30/84 systen (Rev. 1) 141f 9.1.3 Spent fuel pool cooling and cleamp Cmplete 8/30/84 system (Rev. 1) 141g 9.1.3 Spent fuel pool cooling and cleamp Ccmplete 8/30/84 syste (Rev. 1) 142a 9.1.4 Light load handling systen (related Carplete 8/15/ 84 to refueling) (Rev. 1) 142b 9.1.4 Light load handling systen (related Conglete 8/15/84 to refueling) (Rev. 1) 143a 9.1.5 Overhead heavy load handling own 143b 9.1.5 overhead heavy load handling Open 144a 9.2.1 Station service water systen Ccmplete 8/15/84 (Rev. 1) 144b 9.2.1 Station service water systen Ccnplete 8/15/84 (Rev. 1) 144c 9.2.1 Station service water systen Canplete 8/15/84 (Rev. 1) t I
145 9.2.2 ISI progra and functional testing Closed 6/15/84 of safety and turbine auxiliaries (5/30/84-l cooling systems Aux.Sys.Mtg.)
i 146 9.2.6 Switches and wiring associated with closed 6/15/84 HPCI/RCIC torus suction (5/30/84-Aux.Sys.Mtg.)
i l
M P84 80/1214 - gs l
1 l
_ _ . _ _ . _ _ _ . . _ . _ _ _ L.
ATTA08 err 1 (Cont'd)
DSER R. L. MITIL '10 GWI SECTAGI A. SGIENCER TISE leeER SUBJECT S'IXtts IEFIEt DMED 147a 9.3.1 Ccapressed air systens Coplete 8/3/84 (Rev 1) 147b 9.3.1 . Ccapressed air systess Complete 8/3/84 (Rev 1) 147c 9.3.1 Ccapressed air systems Ccaplete 8/3/84 (Rev 1) 147d 9.3.1 Ccapressed air systems Cmplete 8/3/84 (Rev 1) 148 9.3.2 Post-accident sanpling system Ccuplete 8/20/84 (II.B.3)
IQa 9.3.3 Equipment and floor drainage systen Ccaplete 7/27/84 149b 9.3.3 Equipent and floor drainage systen Cmplete 7/27/84
-150 9.3.6 Primary contalment instrunent gas Ccaplete 8/3/84 system (Rev. 1) 151a 9.4.1 Control structure ventilation systen Ccmplete 8/30/84 (Rev. 1) 151b 9.4.1 Control' structure ventilation systen Complete 8/30/84 (Rev. 1) 152 9.4.4 Radioactivity monitoring elenents Closed 6/1/84 (5/30/84-Aux.Sys.Mtg.)
153 9.4.5 Engineered safety features ventile- Ccmplete 8/30/84 tion system (Rev 2) -
154 9.5.1.4.a Metal roof deck construction Cmplete 6/1/ 84 classificiation 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 PG4 80/1215 - go
l ATDOMENT 1 (Cont'd)
DSER R. L. MITTL 10 OPIBl SECTIGI A. SOfNENCER I11M NLMBER SUETECT STATUS LETIER DMED 157 9.5.1.4.e Cable tray protection Caglete 8/20/84 158 9.5.1.5.a Class B fire detection system Omelete 6/15/84 159 9.5.1.5.a Primary and secondary power supplies Complete 6/1/84 for fire detection systen 160 9.5.1.5.b Fire water pwp capacity Caplete 8/13/84 161 9.5.1.5.b Fire water valve supervisim Caplete 6/1/84 162 9.5.1.5.c Deluge valves Cmplete 6/1/84 163 9.5.1.5.c Manual hose station pipe sizing Cmplete 6/1/84 164 9.5.1.6.e Remote shutdown panel ventilation Omplete 6/1/84 165 9.5.1.6.g Emergency diesel generator day tank Cm plete 6/1/84 protection 166 12.3.4.2 Airborne radioactivity monitor Cm plete 7/18/84 positioning 167 12.3.4.2 Portable continuous air monitors Caplete 7/18/84 168 12.5.2 Equiprient, training, and procedures Cmplete 6/29/84 for inplant iodine instrumentation 169 12.5.3 Guidance of Division B Regulatory Cmplete 7/18/84 Guides 170 13.5.2 Procedures generation package Cm plete 6/29/84 submittal 171 13.5.2 TMI Item I.C.1 cmplete 6/29/84 172 13.5.2 PGP Commitment Caplete 6/29/84 173 13.5.2 Procedures covering abnormal releases Cmplete 6/29/84 of radioactivity M P64 80/12 16- gs
h
- i. ,
e ,
c . ATrJOMENT 1 (Cont'd)
' ^
, DSER '
R. L. MI'ITL 'ID GPEN SECTICN . A. SCHWENCER ITEM NLMBER SUEk7ECT : STATUS IETIER DMED c174 13.5.2: Resolution explanation in FSAR of Caplete . 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 Omplete 8/13/84-
'17.6b 14.2 Initial plant test progra Couplete 9/5/84 (Rev. 1)
'176c 14 .2 Initial plant test program Caplete .7/27/84 176d 14.2 Initial plant . test program 0cuplete 8/24/84 (Rev. 2) 176e- 14.2 Initial plant test program Caplete . 7/27/84 176t 14.2 Initial plant test progra couplete 8/13/84
=176g 14.2 ' . Initial plant test program Omplete 8/20/84 176h 14.2 , Initial plant test progran Omplete 8/13/84 1761 14.2 Initial plant test program Ca plete 7/27/84-177 15.1.1 Partial feedwater heating Caplete 8/20/84 (Rev. 1) 178- 15.6.5 IOCA resulting fran spectrum of NRC Action postulated piping breaks within RCP 1179 15.7.4 Radiological consequences of fuel NRC Action handling accidents 180 15.7.5 Spent fuel cask drop accidents NRC Action
. 181 15.9.5 ' IMI-2 Item II.K.3.3 Caplete 6/29/84
'182 15.9 .10 'IMI-2 Item II.K.3.18 Cmplete - 6/1/84
-183 18 Hope Creek DCRDR Cmplete 8/15/84 M PU4 80/12 17.- gs s
_..._.m_ __ ________.m. __..________.__.l____._._________.___.. . ___,.___m_.__ _ _ ._.s
ATTAOM!NT 1 (Cont'd) ,
1 DSER R. L. MITIL 10 l TEN SECTICH A. SOD elCER ITEM NIDEER SUBJECT STA1tB IETIER DMED 184 7.2.2.1.e Failures in reactor vessel level Ccuplete 8/1/84 sensing lines (Rev 1) 185 7.2.2.2 Trip systen sensors and cabling in Ccsplete 6/1/84 turbine building 186 7.2.2.3 Testability d plant protection Cmplete 8/13/84 systems at power (Rev. 1) 187 7.2.2.4 Lif ting cf leads to perfonn surveil- Caplete 8/3/84 lance testirg 188 7.2.2.5 Setpoint nethodology C mplete 8/1/84 189 7.2.2.6 Isolation devices Cmplete 8/1/84 190 7.2.2.7 Regulatory Guide 1.75 Ccmplete 6/1/84 191 7.2.2.8 Scram discharge volune Cmplete 6/29 /84 192 7.2.2.9 Reactor node switch Cm plete 8/15/84 (Rev. 1) 193 7.3.2.1.10 Manual initiation d safety systems Cmplete 8/1/84 194 7.3.2.2 Standard review plan deviations Cmplete 8/1/84 (Rev 1) 195a 7.3.2.3 Freeze protection / water filled Canplete 8/1/84 instrument and sangling lines and cabinet tenperature control 195b 7.3.2.3 Freeze protection / water filled Canplete 8/1/84 instrument and sangling lines and cabinet temperature control 196 7.3.2.4 Sharing cf comon instrument t@s Canplete 8/1/84 197 7.3.2.5 Micrcprocessor, multiplexer ard Canplete 8/1/84 ccuputer systems (Rev 1)
M P84 80/12 18 - gs
.-_ .=- . . - - . . . - - - _ - . - . - . - - - - - - . . . _ = _ _ - - _ _ , _ - _ _ _
i A2TAOBENT 1 (Cont'd)
DSER R. L. MITIL TO GWI SECTIGi A. SODENCER ITBE NLSEER SLEMBCT STATtB IEFITR DMED 198 7.3.2.6 1MI Item II.K.3.18-ADS actuation Ccaplete 8/20/84 199 7.4.2.1 IE Bulletin 79-27-Imss of non-class Ccaplete 8/24/84 IE instrumentation and control power (Rev. 1) systen bus charing cperation 200 7.4.2.2 Roote slutdown systen Camplete 8/15/84 (Rev 1) 201 7.4.2.3 RCICAIPCI interactions Cmplete 8/3/84 202 7.5.2.1 Imvel measurenant errors as a result Ccmplete 8/3/84 of envirornental tangerature offects cm level instrunentation reference leg 203 7.5.2.2 Regulatory Guide 1.97 Cmplete 8/3/84 204 7.5.2.3 1MI Itan II.F.1 - Accident nonitoring Ccnplete 8/1/84 205 7.5.2.4 Plant process emputer systen Cmplete 6/1/84 206 7.6.2.1 High pressure / low pressure interlocks Ccmplete 7/27/84 207 7.7.2.1 HELBs and consequential control systen Cmplete 8/24/84 failures (Rev. 1) 208 7.7.2.2 Multiple control systen failures Cmplete 8/24/84 (Rev. 1) 209 7.7.2.3 Credit for non-safety related systens Camplete 8/1/84 in Chapter 15 (f the FSAR (Rev 1) 210 7.7.2.4 Transient analysis recording system Ccaplete 7/27/84
'211a 4.5.1 Control rod driw structural materials Ccmplete 7/27/84
- 211b 4.5.1 Control rod driw structural materials Ccmplete 7/27/84 211c 4.5.1 Control rod driw structural materials Ccmplete 7/27/84 M P84 80/1219 - gs
ATDOMENT 1 (Cont'd)
DSER R. L. MITIL 10 GWI S8CTIGI A. S0tWENCER 111M NLDGER SUBJECT STATUS IErfER DA31D 211d 4.5.1 Control rod drive structural materials Ccuplete 7/27/84 211e 4.5.1 Oontrol rod drive structural materials Cmplete 7/27/84 212 4.5.2 Reactor internals materials Ca plete 7/27/84 213 5.2.3 anactor coolant pressure boundary Cmplete 7/27/84 material 214 6.1.1 Engineered safety features materials Complete 7/27/84 215 10.3.6 Main steam and feedwater systen Caplete 7/27/84 materials 216a 5.3.1 Reactor vessel materials Caplete 7/27/84 216b 5.3.1 Reactor vessel materials Caplete 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 Caplete 8/15/84 controls 220 9.5.1.3 Fire brigade and fire brigade Caplete 8/15/84 training 221 8.2.2.1 Physical separation of offsite Caplete 8/1/84 transmissicn lines 222 8.2.2.2 Design provisions for re-establish- Caplete 8/1/84 ment of an offsite power source 223 8.2.2.3 Independence of offsite circuits Caplete 8/1/84 between the switchyard and class IE buses 224 8.2.2.4 Camon failure node between onsite Caplete 8/1/84 and offsite power circuits M P84 80/12 20- gs
ATTAOBENf 1 (Cont'd)
DSER R. L. MITIL TO OPIBl. SECf1Gl. A. SODEN3R 11131 NLBGER SLEUBCf STARE IEPIER DPRD 225 8.2.3.1 Testability cf astanatic transfer cf Canplete 8/1/84 power from the normal to preferred power source ,
226 8.2.7.5 Grid stability Canplete 8/13/84 (Rev. 1) 227 8.2.2.6 Capacity and capability of offsite Ccmplete 8/1/84 circuits 2 28 8.3.1.l(1) Voltage &cp during transient condi- Canplete 8/1/84 tions 229 8.3.1.l(2) Basis for using bus voltage versus Ccmplete 8/1/84 actual connected load voltage in the voltage drcp 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 8.3.1.1(5) Load configuration used for the Ccmplete 8/1/84 voltage top analysis 233 8.3.3.4.1 Periodic systen testing Canplete 8/1/84 234 8.3.1.3 Capacity and capability cf onsite Canplete 8/1/84 E power supplies and use of ad-ministrative controls to prevent overloading d the diesel generators 235 8.3.1.5 Diesel generators load acceptance Ccmplete 8/1/84 test l 2 36 8.3.1.6 Conpliance with position C.6 of Canplete 8/1/84 l IG 1.9 237 8.3.1.7 Decription cf the load sequencer Ccmplete 8/1/84 238 8.2.2.7 Secp.encing cf loads en the offsite Ccmplete 8/1/84 power systen l
l M P84 80/12 21 - gs
--_e ,
. ., ,,~ _ , _ _ _ _ _ . _ , _ _ _ _ _ , . _ . . _ . _, , _ _ , , , , . _ _ , ,
ATTA09ENT 1 (Cont'd)
DSER R. L. MITIL TO CPIN SBCIIG6 A. SODENGR I1150 NUMER SUBJECT STKItJS IEPI1!lR DmD 2 39 8.3.1.8 Testing to verify 804 miniman cmplete 8/15/84 voltage 240 8.3.1.9 Capliance with BIP-PS&-2 C a plete 8/1/84 241 8.3.1.10 Ioad acceptance test af ter prolonged Cmplete 8/20/84 no load cperation of the diesel (Rev. 1) generator 242 8.3.2.1 Capliance with position 1 of Regula- Cmplete 8/1/84 tory Guide 1.128 243 8.3.3.1.3 Protection or qualification d Class Cmplete 8/1/84 lE equipment from the effects of fire suppression systems 244 8.3.3.3.1 Analysis and test to demonstrate Cmplete 8/30/84 adequacy d less than specifi l (Rev. 1) separation 245 8.3.3.3.2 The use d 18 versus 36 indles d Cmplete 8/15/84 separaticn between raceways (Rev. 1) 246 8.3.3.3.3 Specifi,ed separation d raceways by Cmplete 8/1/84 analysis and test 247 8.3.3.5.1 Capability cf penetrations to with- Cmplete 8/1/84 stand long duration short circuits at less than maximum or worst case short circuit .
20 8.3.3.5.2 Separation cf penetration primary Cmplete 8/1/84 and brickup protections 20 8.3.3.5.3 The uso d bypassed thermal overload Cmplete 8/1/84 protective devices for penetration protections 250 8.3.3.5.4 Testing cf fuses in accordance with Cmplete 8/1/84 R.G. 1.63 I
M P84 80/12 22 - gs 1
ATrJO9ert 1 (Cont'd)
D6ER R. L. MITfL 10 OMN SECTIG4 A. SOMNCER T198 IOSER SUBJECT STATUS IJrrtER DA2ED 251 8.3.3.5.5 Fault current analysis for all Ccaplete 8/1/84 representative penetration circuits 252 8.3.3.5.6 The use of a single breaker to provide cauglets 8/1/84 penetration protection 253 8.3.3.1.4 Ocmaitment to protect all Class 1E Ccuplete 8/1/84 equipment frta external hazards versus only class lE equipment in one division 254 8.3.3.1.5 Protection of class lE power supplies Ccuplete 8/1/84 from failure of unqualified class 1E loads 255 8.3.2.2 Battery capacity Otmplete 8/1/84 256 8.3.2.3 Automatic trip of loads to maintain Otmplete 8/20/84 sufficient battery capacity 257 8.3.2.5 Justification for a 0 to 13 second Cceplete 8/1/84 load cycle 258 8.3.2.6 Design and qualification of DC Ccmplete 8/1/84 system loads to operate between minimum and maximum voltage levels 259 8.3.3.3.4 Use of an inverter as an isolation Ccmplete 8/1/84 device 260 8.3.3.3.5 Use of a single breaker tripped by Ccmplete 8/1/84 a IDCA si pal used as an isolation device 261 8.3.3.3.6 Autcmatic transfer of loads and Ocuplete 8/1/84 interconnection between redundant divisions 262 ll.4.2.d Solid wasts control progran Ccmplete 8/20/84 M P84 40/12 23- gs
1 ATDO9ENT 1 (Q)nt'd)
DSER R. L. AITIL *10 OPBI. S8CIIGE , .
A. SO9 ENGR 1794 ! USER !KIIL7BCE SUmJS IEr11lR DATED 263 11.4.2.e Fim protection for solid radwaste Complete 8/13/84 storage area 264 6.2.5 Sources of oxygen O g lete 8/20/84 265 6.8.1.4 ESF Filter Testing Ccuplete 8/13/84 266 6.8.1.4 Field leak tests Ocuplete 8/13/84 267 6.4.1 Control room toxic cheatical Ccaplete 8/13/84 detectors 268 Air filtration unit drains Ccuplete 8/20/84 269 5.2.2 Code cases N-242 and M-242-1 Ca plete 8/20/84 270 5.2.2 Oxle case N-252 Ocuplete 8/20/84 TS-1 2.4.14 Closure of watertight (bors to safety- Open related structures TS-2 4.4.4 Single recirculation loop operation Open TS-3 4.4.5 Core flow monitoring for crixi effects Couplete 6/1/84 TU-4 4.4.6 Imse parts monitoring system Open TS-5 4.4.9 Natural circulation in nonnal Open operation TS-6 6.2.3 Secondary containment negative Open pressum TS-7 6.2.3 Inleakage and drawdown time in Open secondary containment TS-8 6.2.4.1 Imakage integrity testing Open TS-9 6.3.4.2 BCG subsystem periodic cagonent Open testing M P64 80/12 24- gs
ATDO9Wtr 1 ((bnt'd)
DSER R. L. MITIL 10 OISI SBCTICal A. SOSENGR ITBt ISBSER SEBCr STATUS IErtER Dh15 TS-10 6.7 M IV leakage rate TIP11 15.2.2 Availability, setpoints, and testing Open of turbine bypass system TS-12 15.6.4 Primary coolant activity IE-1 4.2 Fuel rod internal pressure criteria Conglete 6/1/84 14-2 4.4.4 Stability analysis submitted before open second-cycle operation 4
M M4 80/12 25- ga
ATTACHMENT 2 DATE: 9/5/84 DRAFT SER SECTIONS AND DATES PROVIDED SECTION DATE SECTION DATE 3.1
'3.2.1 11.4.1 See Notes 1&5 3.2.2 11.4.2 See Notes'l&5 5.1- 11.5.1 See Notes 1&S t
5.2.1. 11.5.2 See Notes 1&S 6.5.1 See Notes 1&S 13.1.1 See Note 4 8; .1 See Note 2 13.1.2 See Note 4 8.2.1 See Note 2 13.2.1 See Note 4 8.2.2 See Note 2 13.2.2 See Note 4
'8.2.3 See Note 2 13.3.1 See Note 4
.8.2.4 See Note 2 13.3.2 See Note 4 8.3.1 See Note 2 13.3.3 See Note 4 8.3.2 See Note 2 13.3.4 See Note 4 8.4.1- See Note 2 13.4 See Note 4 8.4.2 See Note 2 13.5.1 See Note 4 8.4.3 See Note'2 15.2.3 8.4.5 See Note 2 15.2.4 b.4.6 S9e 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-l&5 9.5.7 See Note 3 17.1 8/3/84 9.5.8 See Note 3 17.2 8/3/84 10.1 See Note 3 17.3 8/3/84 10.2 See Note 3 17.4 8/3/84 10.2.3 See Note 3 ^
10.3.2 See Note 3 10.4.1- See Note 3 10.4.2 See. Notes 3&5 10.4.3- See Notes 3&S 10.4.4 See Note 3 11.1.1 See. Notes 1&S . ', Notes:
~
11.1.2 See Notes 1&S 1.' Open items provided in
~
11.2.1 See Notes 1&5 11.2'.2 See Notes 1&S ,
letter-dated July 24, 1984 11.3.1 'See Notes 1&5 JUn(Schwencer to Mittl)' ~
11.3.2 See Notes 1&S .-
2..Open items provided in
' June 6, 1984, meeting , s '
- 7. .
'3. Open items provided in '
April 17-18, 1984 meeting s s CT:db ~~ t s 4.-_Open items provided in "v
1 May 2, 1984 meting
- 5. Draft SER Section"provided
.c in letter dated August 7, 1984 (Schwencer to Mitti)
MP 84 95/03 01 ,
- ~ ~ '
w in- m.
<, ;s'- DATE: 9/5/84 1 ATTACHMENT 3
.Open DSER-
? Item ~ Section Subject
^
115 5.3.4. Drop' weight and Charpy V-notch tests for. closure flange materials-
~
176b 14.2 Initial plant test program-
.4..
4 L 2
, ' '$8 k -.
s 4 m u
- ' _]
s
)
1 l- 1 N g
1 m
L f
s
, ; M P84 95/03 02 e
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~
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ATTACHMENT 4 G
a HCGS DSER Open Item No. 115 (Section 5.3.4) J DROP WEIGHT AND CHARPY V-NOTCH TESTS FOR CLOSURE FLANGE MATERIALS Provide drop weight test and Charpy V-notch test results from the closure flange region materials to demonstrats' compliance with the closure flange requirements of Appendix G, 10 CFR 50.
RESPONSE
For the information requested above, see the response to Question 251.4.
i F
4 l
l t
MP 84 112 15 06-bp 1
OUESTI0ff 251.4:
provide drop weight test and Charpy V-notch test results from the l closure flange region materials to demonstrate comp 11erice with the closure flange requirements of Appendix G, 10 CFR 50.
[
, RESP 0085E Available drop-weight and Charpy V-notch test results for the Hope Creek Unit 1 closure flange gnatorials are provided below:
- IST Test Lateral Temp. Temp. Absorbed Energy Expansion 01sterial Orientation
- (*F)
(*F) (Ft-lbs) (Mils) l l 5A508. C1.2 Longitudinal -20/ -40 64.1,70.6.20.8,77.1 48,51.11.58, l (Head Flange) -10 -10 93.1,114.7,106.6 64,78,62,55, 9180* 87.8,97.1,71.9 64,49 i AWAY 10 31.1,108.133.6, 49,68,78.95, 137.6.165.1 68,74 l
40 157.4.121.5.137.6 89,73,77,86, 134.9.144.3.137.6 79.85
! - 40 199.9.154.8.159.9 77,69.88,87, 195.4.144.3.170.1 82,73
- l. ,
SA508,C1.2 Longitudinal -10 10 -120.1.122.8,130.9, 77,81,83,81, (Shell Flange) '130.9.132.3.116.1 77,64 l -10 120.1,95.8.128.2, 72,58,80,75, 309.3.101.2,87.8 59.57 i- +40 141.6.134.9.141.6, 81,77,84,82, 145.6.167.6.182.4 85,89 l.
jj ,
-40 '3.1.4,69.3.59.0,55.2, 7,48,41,38, 74.5.101.2 54,68
! 44esdr.sC51 - ! .
C'er'Ris'l Pis1t sooneetcJ 6, Ifr.J Plsay) ---- --
(risc.s TzA) s ov3ited isl io 4.5, M.1, M.s %, H, e sos.s, 41. , ts.s vs, st, s4
{wisca Tas) ei seE %.),vo.6,,,.s ss, sa, 64
.Ma.s, TI.s, 616 11, ss,'so a
(naca TEC) so 67, 83, 62 es es41(es.10.6, 87.8, as.: ve, 6s, to (n aca Tap) H
- to as
...e,vs.z, 51, et va ss.,,S.t.s, s. ,,, suz sa s33,G r.8,Ct.1 ;
(UPPER.SHsu. CeMsCTep TD Shell Flseige) j so vs.e, 468, 68 5 A9, 89, r3 (macs Sic.) Len3rtedmsl _. u .t . v s a, 6s.4 sa, ps, 49 (m sc$ s tA) 10.
. M. s, t v.s, ss.O s?, v4, 45
.., ,+. s, m ss, ,5 a h" "N
- io N..o,' 9s.s, ts.s, ee.: p58 w. W, 19 go, 4 4, 11 1F Ivi 3ccor .dsace with the ASME C4c vid GE. Spec 4 cat,s,r rqui.cencoh, the well meisk 3* in in3 the fla.noe rejiori eniter:315 hsve C.v4 sbsorbed e.Acry vs.locs of st jes.St t o Pt-W s d +}o*F.
u-_. .. . , . - _ . - ~ . _ . .-
- ~
4 g 4
'O **
+ 4 m
M 6 \/ b 1 T.'
- c.;
.DSER OPEN ITEN 176bi(Section 14.2) x e
INITIAL PLANT' TEST PROGRAM _
+ 1
\
' 4 ~ 'The following FSAR Subsectiors 14.2.12 test abstracts should be modified as
.sthted-to provide 'dequate a acceptance criteria i ..
l
[. v.
'.TestiAbstract. Modification,
~
-1.5.d.1 A' reference 'should be provided for acceptable-M '
- closing times.
II.7.d.1-
~
!- _ A reference.should be provided for the design
~
1.'15.d.2 specifications.
1~
l ;.. c.l.23.d.4 <
' Reference should be 6.2.5.2.5 td.6' ,
Reference should.be E<2.5.2.3-
. l .3 f. d . 6 . 'A reference should belprovided regarding. sage levels
[ '
of- hydrogen' buildup ~
l
~\
11.41.d.l? -A reference should be provided regarding the, appropriate l
~ accuracy of; response
. l. 47.d.'4
' ~
. A' reference should be provided for the prescribed time.
~
il'.52.d.2-M
~
, .l.60.d.3. .The parameters in these tests should meet or exceed t ne q' 'll612d.1- -design values described ~in their respective references; M ,i < .1.65.d.2- .they should not simply "be comparable" or " compare 11.71.d.2 favorably.
3.24.~d.5 p 4 L ~ 'l .68.e .1 : .A reference should be_providAd regarding the' negative- l
-pressdre specification. ~ '
ix ,
DAdditionallyF all'startup tests'should be modified to specify the appropriate
- J .
-level of acceptance criteria (Level,1, 2, or 3) as defined in FSAR Subsection i s, <
=14.2.12.2..
Q
~
- . , RESPONSE- l
, : n: ,
U FSAR Section 14.2.12.1 was revised in~ Amendment 6 to provide the information-requested abo;ve.-
- g , 2,.13. . l* *7. d . I a n d
- In addition,-Secti 4.2.12.3.24.d.5 has been revisau to reflect the new SE Test-Specifications 4nd all the startup tests'in Section 14.2.12.3 have been' '!i t
modified to specify the' appropriate level of acceptance criteria.
- ,. , . . .. \ ' .anc$ ca db $+ ion &l N AC C omme n%S
.E _. ~-__.._.._._.._.__._....___._._,2__.._.._.. __ _ . .. J.. .!
. ) - ;
i
~d. Acceptance Criteria ;
o :
- 1. All valves, alarms, controls, interlocks, and logic shall function in accordance with the eye 4em ,
h for core spray. ,
GC Preopera b onal -tes-f sp ecifc a-
- 2. For the core spray test mode and core spray injection mode, the pump head / flow requirements, i l the NPSH requirements, and the system design flow i
requirements meet the GE preoperational test L specification acceptance criteria.
l
- 3. All modes of operation and flow paths shall be as i specified in the GE preoperational test specifications.
- 4. The jockey pump can fill and pressurize the core spray system . ,
' l 14.2.12.1.8 BF-Control Rod' Drive - Hydraulic l l l
- a. ~ Objective ,
i-r' The test objective is to demonstrate that the control i
rod. drive (CRD) system is fully operational, and that i al1~ components, including the hydraulic drive mechanism, manual control system, rod position indicator system, and all safety and control devices, ,
function per design.
- b. Prerequisites ,
l 1. All component tests have been completed and i -
approved.
L
- 2. AC and de power are available.
- 3. All instrumentation has been calibrated and
' instrument loop checks completed. i 14.2-40 Amendment 6--
[- -. _ . _ . . _ _ _ _ _ _ . _ _ _ ___ __.. _ _ _ _ i
. . ~ . . .- -
- their recommendations. This report must discuss alternatives of action, as well as the concluding recommendation, so that it can be evaluated by all related parties.
Level 3 If level 3 performance is not satisfied, plant operating or startup test plans would not necessarily be. altered. The numerical limits stated in this category are associated with expectations of individual component or inner control loop i transient performance.' Because overall system performance is a mathematical function of its individual components, one component whose' performance is slightly worse than specified can be accerted if a system adjustment elsewhere will positively overcome this small deficiency. Large-deviations from Level 3 performance are not allowable. Level 3 performance is also not specified in fuel'or vessel protective systems.
When a Level 3 performance is not satisfied, the subject component or inner loop must be analyzed closely. If all Level 1 and Level 2 criteria
- are. satisfied, then it is not required to repeat the transient test to satisfy Level 3 performance. The occurrence must be documented in the test report. Level 3 performance is to be viewed as highly desirable rather than required to be satisfied.
Good engineering judgement is necessary in the application of these-rules.
During performance of startup tests, technical specifications override any test in progress if plant conditions dictate.
~
14.2.12.3 Startuo Test Procedures 14.2.12.3.1 Chemical and Radiochemical Monitors and Sample Systems
- a. Objectives The tests provide verification of the sample systems' ability to:
- 1. Maintain-quality control of the plant systems' chemistry and ensure that sampling equipment,
. procedures, and analytical techniques supply the L
14.2-153 Amendment 4
1/84 data required *to demonstrate that fluids meet quality specifications and process requirements
- 2. Monitor fuel integrity,. operation of filters and demineralizers, condenser tube integrity, operation of the offgas system and steam separator-dryer, and tuning of system monitors.
- b. Prerequisites Intrument calibration and preoperational testing of chemical, radiation, and radiochemical monitors have been c.ompleted.
- c. Test Method Prior to fuel loading, a complete set of chemical and i
radiochemical samples are taken to ensure that all sample stations are functioning properly and to' l determine the initial concentrations. During reactor heatup, subsequent to fuel loading, samples are taken and measurements made at each major power level plateau to determine the chemical and radiochemical quality of reactor water and reactor feedwater, amount of radiolytic gas in the steam, gaseous activities after the air ejectors, decay time in the gaseous radwaste lines, and performance of filters and demineralizers.
Baseline data for the main steam process radiation monitoring subsystems and the offgas monitoring subsystems is also taken at each major power level plateau. Adjustments are made, as required, to monitors in the liquid waste management system (LWMS),
-liquid process lines, and offgas treatment system.
L
- d. Acceptance Criteria Level 1 :
The chemical and radiochemical, and water of quality
. factors are maintained within the technical specifications and fuel warranty requirements. Gaseous '
parbculale'
-and liquid effluents' activities shall conform with l Technical Specifications.
l i
14.2-154 Amendment 4
~ _ _ . . _ . _ . _ . . . _ _ _ _ _ _ _ . . _ . _ . _ _ _ _ _ _ . _ . _ _ _ _ _ _ . _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ , _ , . _,
No hh HCGS FSAR 4/84 14.2.12.3.2 Radiation Measurements
- a. Objective The test objective-i:s to monitor radiation at selected power levels during plant operation to ensure the adequacy of shielding for personnel protection, and to verify compliance with 10 CFR 20.
l
- b. Prerequisites Prior to fuel loading, a survey-of natural background radiation is made at selected locations throughout the plant site.
- c. Test Method During reactor heatup and at select.ed power levels subsequent to fuel loading, gamma dcse rates, and where appropriate, neutron dose rate measurements are made at specific locations around the plant including all
'potentially high radiation areas.
- d. Acceptance Criteria Level 11 Plant radiation doses and personnel occupancy times shall be within allowable limits, as defined in 10 CFR 20.
- 14. 2.12. 3.~ 3 Fuel Loading
- a. Objective The test objective is to load fuel safely and efficiently to the full core size.
14.2-155 Amendment 5 .
. _ . . . . ..-J,,..--.--......--..-.--.-.-..--_,.---.-_-. ..
4 o ff( ,
- b. Prerequisites
- Section 14.2.10 (initial fuel loading) describes the prerequisites for cor.mencing fuel. loading.
- c. Test Procedure The fuel loading procedure includes any tests performed.
during the fuel loading evolution, including subcriticality checks, shutdown niargin verifications, and control rod functional checks.
- d. Acceptance Criteria Level 1:
The core shall be fully loaded in accordance with established procedures and the core shall be subcritical by at.least 0.38% AK/K with the analytically determined strongest rod withdrawn.
14.2.12.3.4 Full Core Shutdown Margin
- a. Objective The test objective is to demonstrate that the reactor will remain subcritical throughout the first fuel cycle with the most reactive control rod withdrawn.
b._ Prerequisites The core is fully loaded at ambient temperature in the zenon-free condition.
c .- Test Method The shutdown margin is meas'ured by withdrawing selected control rods until criticality is reached. The empirical data is reviewed and compared with design data to determine the test results.
14.2-156 Amendment 4
fob l l
- d. Acceptance Crkteria j .Levei- 1 : .
The shutdown margin measurements shall verify that the j~
core remains suberitical with the most reactive control rod withdrawn and all other control rods, fully inserted
'O - g' , g sby at least 0.38% AK/K. J.Jd ; t.i... LL, ,'"dri ticali ty 4-7'p 3d 'should occur within 21.0% AK/K of the predicted critical.
I 14.2.12.3.5 Control Rod Drive System
- a. Objective 1
! The test objective is to obtain the baseline data for I the CRD system, and to demonstrate that the system
, operates over the full range of primary coolant conditions, from ambient to operating.
- b. Prerequisites Preoperational tiesting of the CRD system has been completed and the system is ready-for operation.
- c. Test tiethod l
l The startup tests performed on the CRD system are an L extension of the preoperational tests. Initial post fuel load tests with zero reactor pressure include position indication, normal insert / withdraw stroking, friction testing, and scram testing. Coupling checks are verified using station operating procedures.
Following initial heatup to rated reactor pressure, the friction and scram test is accomplished. Following initial heatup, the four slowest CRDs are measured for scram times following planned reactor scram as detailed on Figure 14.2-5. fa, a ff;) ion f rop r resfon se of 44.
CRD Plow co,dml vaIve win & y,,,-j;-g ,
- d. Acceptance Criteria The insert fiI withdrawal' times, scram,t'imes, and fricti test results,'shall meet the requirements of the GE startup test' specification l'imits. ffe CRD .e/ T / d sy' stem flow requfrements and flos control alve
/
14.2-157 Amendment 4
coMd i ti li
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1 l iLevel l2 l The windvau;J spets a ser -k>,.co sk.12 m u t % r e g u i< w e d .1 JA SE sL<kp +es+
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, h3kl HCGS FSAR 4/84 res e meet ,
e reg s eG pg 14.2.12.3.6 -
' Source Range Monitor Performance and Control Rod Sequence
- a. Objective The test objective is to demonstrate that the neutron sources, SRM instrumentation, and rod withdrawal sequences provide adequate information to achieve criticality and increase power in a safe and efficient manner. TE= cffect: cf 5 , i . _ ' . .J mv . . .. .. La en DM
- re rter ;;;;. __ - = J ' - :0 - - _1_ _1_ f .
- b. Prerequisites Fuel loading is complete, neutron sources have been installed, and all control rods have been inserted.
The CRD system is operational.
- c. Test Method l
With the neutron sources installed, source range monitor count rate data is taken and compared to the A j, L required signal count and signal count-to-noise count j d6C '
I ratio. Source range data is taken during rod ~ 2e I
withdrawals to the point of criticality.d"_During heatup to rated temperature, critical rod patterns are l
recorded. Rods will be withdrawn in accordance with a pre-established withdrawal sequence. Movement of rods in a prescribed sequence is monitored by the RWM and RSCS which prevents out of sequence movement. mar =Dbs ei*rre:1 f rr:5 red grer; 10 reepl:t:0 J.. ng rever I)cbdbI accer-4aa the electri;;l r.-.., ;f si = Ilu , 2n d .'.P"M raarer. : will Le ;;;r'-4
- d. Acceptance Criteria
)Qh% el A 3 The neut signal c - o-noise co ratio a fi"E
. mini counts o e SRMs shall t the re rements he GE s up test speciftfation. 9 h- GI+
14.2-158 Amendment 5
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$9E9h HCGS FSAR 4/84 14.2.12.3.7- Rod, Sequence Exchange l This Test Has Been Deleted l l
14.2.12.3.8 Intermediate Range Monitor Performance L a. Objective . .
l The test objective is to determine IRM system response to neutron flux and to optimize the IRM overlap with the SRMs and APRMs.
i
- b. . Prerequisites The reactor is critical and the IRM gains have been set L at maximum for conservatism.
- c. Test Method
'After criticality, and when flux level is sufficient, the IRM response to neutron flux and the IRM/SRM overlap is verified. Following the calibration of the APRM, the IRM gains are adjusted if necessary. If any adjustments are made, the overlap of the SRM and IRM is verified when flux levels are in the appropriate range.
1
- d. Acceptance Criteria l.
L:.n' s :L' L Each IRM channel must be on scale before the SRMs
- l. exceed their rod block setpoint. Each APRM must be on scale bpfore the IRMs exceed their rod block setpoint.
yf 3 p Abem, M ch IRM should be adjusted for half decade l overlap with SRMs and one decade overlap with APRMs.
l 14.2-159 Amendment 5
-t_--- - --.,--.-.,-<.,-.-*,.-,,,e-+.--, ....w. ,-.. _---,,-www --,--,-.-.-.--+.-w,,.w ,,.- w,,.-w- -.. , v,i..-.--,e -,,.w-e-v- w.-,-1.-
Ihhh HCGS FSAR 4/84 14.2.12.3.9 Local Power, Range Monitor Calibration
- a. Objective The test objective is to calibrate the LPRM.
- b. Prerequisites Reactor power and LPRM gains are sufficient to observe detector response. The process computer or other means are available for determining calibration factors.
- c. Test Method Core power is maintained at the specified level for a sufficient time to allow equilibrium conditions to be established. The process computer computes the average heat flux and calibration factor.for each LPRM. Each LPRM is calibrated in accordance with the calibration l procedure.
i
- d. Acceptance Criteria j Lcve l l1 *.
Each LPRM reading should be within 10% of its calculated value.
i 14.2-160 Amendment 5
. _ . _ . _ _ _ _ _ _ , , .I
Itof14 HCGS FSAR 4/84 14.2.12.3.10 Average Power Range Monitor Calibration e
- a. Objective The test objective is to calibrate the APRM.
1
'b. Prerequisite The core is.in a steady-state condition at the desired power level and core flow rate. Instrumentation used to determine core thermal power has been calibrated.
- c. Test Method A heat balance is taken at selected power levels. Each '
, APR'i channel reading is adjusted to agree with the core thermal power as determined from the heat balance. In addition, the APRM channels are calibrated at the
. frequency required by the Technical Specifications.
- d. Acceptance Criteria Le ve l d.;
- 1. The APRM channels must be calibrated to read equal to or greater than the actual-core thermal power.
- rod block must not be exceeded.
- 3. In the startup mode, all APRM channels must produce a scram at less than or equal to the thermal power setpoint required by technical specification.
Level a*.
% With the above criteria met, the APRMs are considered accurate if they agree with the heat balance r quir;d b Um 64 scarcup wat c=-- ~64 - or-N. m , nimum v n de a re t,uired i boS*8 on TPP , M t-H6ft , = B _fruc3 e c4 r f,&
pwer fo w;%)o 4 lt y;4 3 specTC ecQ ,, Q QC shefap ^ 4esb specillesbs n .
14.2-161 Amendment 5
-- _. ~ ._. . . _ . _ . _ . . _ . _ _ . _ _ _ . . _ _ _ _ _ _ _ _ . _ _ _ _ . . _ . . _ . _ _ _ _ - . . . _ . _
. HCGS FSAR 4/84 14.2.12.3.11 NSSS Process, Computer
- a. Objective t
l
( The test objective is to verify the performance of the
- process computer under plant operating conditions.
I
- b. Prerequisites Computer calculational programs have been verified using simulated input conditions. The computer room HVAC is' operational and plant data is available for computer processing.
!' c. Test Method During plant heatup and ascension to rated power following fuel loading, the NSSS and the balance-of-
. plant system process variables sensed by the computer j l become available. The validity of these variables is .
l verified and the results of performance calculations of l the NSSS and the balance-of-plant (BOP) are checked for accuracy,
- d. Acceptance Criteria Le v.e i 3 '-
[ 1. The process computer performance codes calculating i
the minimum critical power ratio (MCPR), linear heat generation rate (LHGR), and maximum average planar heat generation rate (MAPLHGR), and an independent method of calculation shall not differ in their results by more than the value specified in the GE startup test specification.
l t
I 14.2-162 Amendment 5
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- 2. The LPRM calibration factors calculated by the independent method and the process computer shall not differ by more than the value specified in the GE startup test specification.
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14.2.12.3.12- Reactor Core Isolation Cooling System
- a. Objective f
The test objective is to verify the proper operation'of the RCIC over its required operating pressure range.
- b. Prerequisite Fuel loading has been completed and sufficient nuclear heat is available to operate the RCIC pump.
Instrumentation has been installed and calibrated.
- c. Test Method The.RCIC system is designed to be tested in two ways:
- 1. By flow injection into a test line leading to-the condensate storage tank (CST), and i
- 2. By flow injection directly into the reactor vessel. 1 The earlier set of CST injection tests consist of manual and automatic mode starts at 150 psig and near i rated reactor pressure conditions. The pump discharge pressure during these tests is throttled to be 100 psi l
4 l
14.2-163 Amendment 5
N'fhh
4/84 abova che reactor. pressure to simulate the largest expected pipeline' pressure drop. This CST testing is done to demonstrate general system operability and for maki,ng'most controller adjustments.
Reactor vessel injection tests follow to complete the controller adjustments and to demonstrate automatic starting from a cold standby condition. " Cold" is
, defined as a minimum 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any kind of RCIC operation. Data will be taken to determine the RCIC i high steam flow isolation trip setpoint while injecting i at rated flow to the reactor vessel.
After all final controller and system adjustments have been determined, a defined set of demonstration tests must be performed with that one set of adjustments.
l Two consecutive reactor vessel injections starting from
! cold conditions in the automatic mode must satisfactorily be performed to demonstrate system reliability. Following these tests, a set of CST injections are'done to provide a benchmark for comparison with future surveillance tests.
After the auto start portion of certain of .the above tests is completed, and while the system is still l
. operating, small step disturbances in speed and flov command are input (in manual and automatic mode respectively) in order to demonstrate satisfactory stability. This is to be done at both low (above minimum turbine speed) and near rated flow initial conditions to span the RCIC operating range.-
l A demonstration of extended operation of up to two r hours (or until pump and turbine oil temperature is j stabilized) of continuous running at rated flow L conditions is to be scheduled at a convenient time.
during-the startup test program.
I Depressing the manual initiation pushbutton is defined L as automatic starting or automatic initiation of the RCIC system.
- d. Acceptance Criteria Level 1.2
- i. Following automatic initiation, the pump discharge flow must be equal to or greater than rated flow as specified in Section 5.4.6 within the time specified by the GE startup test specification.
l 14.2-164 Amendment 5 E
lffh .
- 2. The RCIC turbine shall not trip or isolate duririg 4
automatic or manual start tests.
k ve l 2 ',
{'. l . The turbine gland seal system is capable of preventing steam leakage to the environment.
i l p'.:. The delta-pressure setpoints for RCIC steam supply line high flow isolation trip shall be calibrated to the requirements of technical specifications using actual flow conditions.
F J. To provide overspeed and isolation trip avoidance j margin, the transient start speed peaks must not exceed the requirements of the GE startup test specification.
p.M. The speed and flow control loops are adjusted to
^
i
[ meet the decay ratio specified in the GE startup i test specification.
t-14.2.12.3.13 ,
High Pressure Coolant Injection System
- a. Objective The test objective is to verify the proper operation of the HPCI over its required operating pressure range.
l-l
- b. Prerequisite Fuel loading has been completed and sufficient nuclear heat is available to operate the HPCI pump.
Instrumentation has been installed and calibrated.
l
- c. Test Method l
The HPCI system is designed to be tested in two ways:
- 1. By flow injection into a test line leading to the condensate storage tank (CST), and i
14.2-165 Amendment 5
-a+--o-----,--_ -_w- - - - - - - -
s.
l
- 2. By flow injection directly into the reactor vessel.
The earlier set of CST injection-tests consist of manual and automatic mode starts at 150 psig and near rated reactor pressure conditions. The pump discharge pressure during:these tests is throttled to be 100 psi above the reactor pressure to simulate the largest expected pipeline pressure drop. -This CST testing is
.done to demonstrate general system operability and for mal.ng most controller adjustments.
Reactor vessel injection tests follow to complete.the controller adjustments and.to demonstrate automatic starting from a cold standby. condition. " Cold" is defined as a minimum 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any kind of HPCI operation. Data will be taken to determine the HPCI high steam flow isolation trip setpoint while injecting at rated flow to the reactor vessel. Dpressing the manual initiation pushbutton is defined as automatic starting or automatic initiation of the HPCI system.
After all final controller and system adjustments have been determined, a defined set of demonstration tests must be' performed with that one set of adjustments. l Two consecutive reactor vessel injections starting from cold conditions in the automatic mode must satisfactorily be performed to demonstrate system reliability. Following these tests, a set of CST injections are done to provide a benchmark for comparison with future. surveillance tests.
After the auto start portion'of certain of the above tests is completed, and while the system is still operating, small step disturbances in speed and flow command are input (in manual and automatic modes respectively) in order to demonstrate satisfactory stability. This is to be done at both low (above minimum turbine speed) and near rated flow initial conditions to span the HPCI operating range.
A continuous running test is to be scheduled at a convenient time during the startup test program. This demonstration of extended operation should be for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or until steady turbine and pump conditions are reached or until limits on plant operation are encountered.
14.2-166 Amendment 5
- d. Acceptance Criteria Level 1' -
- 1. Following automatic initiation, the pump discharge flow must be equal to or greater than the rated
_ flow, and within the time specified in Section 6.3.2.2.1.
, 2. The HPCI turbine shall not isolate or trip during automatic or manual start tests.
i ki vel 2 *,
3.,, Ine. speed and flow control loops are adjusted to meet the decay ratio specified in the GE startup test specification.
(.7, The turbine gland seal system is capable of preventing steam leakage to the atmosphere.
4 pff, The delta-pressure setpoints for HPCI steam supply line high flow shall be calibrated to technical specification requirements using actual flow conditions.
A'.I, In order to provide overspeed and isolation trip avoidance margin, the transient start speed peaks
' must not exceed the requirements of the GE startup test specification.
14.2.12.3.14 Selected Process and Water Level Reference Leg Temperatures
- a. Objectives i 1. To establish low speed limits for the recirculation pumps to avoid coolant temperature l
stratification in the reactor pressure vessel
- (RPV) bottom head region l' 2. To ensure that the measured bottom head drain -
temperature corresponds to bottom head coolant temperature during normal operation.
- 3. To measure the reactor water level instrument
- reference leg temperature and recalibrate the affected indicators if the measured temperature is different than expected.
14.2-167 Amendment 5
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-._.,%--m_,,.,,_,,,,,,,,_,m-_% .._-_%.. ..7, ,__. _ _ - -.,v,,--, , , - , , p..ew%c, - - _ . _,.,w-,., ,- -_e--.
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L
. HCGS FSAR 4/84 l b. Prerequisites .
The plant is in a hot standby condition. System and test. instrumentation have been installed. '
l .
- c. Test Method
! During initial heatup at hot standby conditions, the i bottom drain line temperature and applicable reactor l
parameters are monitored as the recirculation pump
!- speed is slowly lowered to determine the proper setting l of the low speed limiter. The parameters above are also monitored during planned recirculation pump trips l to. determine if temperature stratification occurs in the idle loop (s) and to assure that idle. loop-to-bulk l coolant temperature differentials are.within Technical Specification limits prior to restarting the pump (s).
The bottom drain line temperature and applicable parameters are monitored when core flow is 100% of rated flow.
4 A test.is also performed at rated temperature and
, pressure under steady state conditions to verify that ,
the reference leg temperature of.the level instrumentation is the value assumed during initial l calibration. Recalibration will be performed if -
necessary.
- d. Acceptance Criteria Le ve l l '
- 1. - The reactor recirculation pumps shall not be started unless the loop to loop delta-temperatures and steam dome to bottom drain delta-temperatures i are within the technical specification _ limits.
- Le ve l Z '.
I Iq, During two pump operation at-100% core flow, the difference between the bottom drain line
, thermocouple and recirculation loop thermocouple I - is within the delta-temperature required in the GE startup test specification.
7.2. The difference between actual reference leg temperature and the value used for calibration is less than the amount specified in the GE startup test specification.
4
}
. +
14.2-168 Amendment 5
. , - ~ - , , . --. , - - - - , , --,e-- --a .-- _ . , _ , , , , - , _n ,,,,,,,,,e-,, - - , . - - , ,m,,.eg,..rn.----e,._ _-y..--g----,------
11f HCGS FSAR 4/84 14.2.12 4sj.5 System Expansion
- a. Objective The test objective is to demonstrate that major
, components and. piping systems throughout the plant are free and unrestrained with regard to thermal expansion.
- b. Prerequisites Fuel loading has been completed and cold plant data has been recorded. Instrumentation required has been installed and calibrated.. The system' piping to be tested is supported and restrained properly.
- c. Test Method i
s i During heatup, observations and recordings of the-horizontal and vertical movements of major equipment and piping in the NSSS and auxiliary systems.are made in order to ensure that components are free to move as designed. Adjustments are made if necessary to allow
- freedom of movement. Snubbers, whose testing '
requirements are governed by technical specifications,
, will be monitored for thermal movement. The systems to
- be monitored are listed in Section 3.9.2.
l d. - Acceptance Criteria
\.3 Jci 2
~1. There shall be no evidence of blocking of.the-displacement of any system component caused by thermal expansion of the system.
l _.
- 2. Inspected hangers shall not be~' bottomed out or have the spring fully stretched.. .
- 3. The position of the shock suppressors shall be
- such as to allow adequate movement at operating
! temperature.
- 4. The piping' displacements at the established transducer locations shall not exceed the limits 14.2-169 Amebdment5 yn-,-,,,,m-,wam. ._-m--ww,- -
- HCGS FSAR 4/84 specified by the piping designer, which are based on not exceeding ASME Section III Code stress values. These specified displacements will be used as acceptance criteria in the appropriate startup test procedures.
14.2.12.3.16 C;. A- c utscrinution TIP L(ncerb ay l~'
- a. Objective l
The test objective is to demonstrate the reproducibility of the TIP system readings. ,
i l 4
- b. Prerequisites '
l The core is at steady-state power level with equilibrium menon, so as to require no rod motion er change in core flow to maintain power level during data
, acquisition by the TIP system.
- c. Test Method R
.. 1. Core power distribution data are obtained during the power ascension test program. Axial power
~ distribution data are obtained at each TIP i location. At intermediate and higher power l , levels, several sets of TIP data are obtained to j ,
determine the overall TIP uncertainty. i
': 2. TIP data are obtained with the reactor operating
! with a symmetric rod pattern and at steady-state
- cenditions. The total TIP uncertainty for the -
test is calculated by averaging the total TIP uncertainty determined from each set of TIP data. ,
The1TIP uncertainty is made up of random noise and l geometric components.
t if
- it 14.2-170 Amendment 5 k
1
- 3. Core power symmetry is also calculated using the TIP datar Any asymmetry, as determined from the analysis, will be accounted for in the calculations for MCPR.
o .
- d. Acceptance Criteria level Z I The total TIP uncertainty shall be within the specified limits required in the GE startup test specification.
14.2.12.3.17 Core Performance a.- Objective
'The test objective is to evaluate the principal thermal and hydraulic parameters associated with core behavior, b., Prerequisites
.The plant is operating at a steady-state power level.
c.- Test Method With.the core operating in a steady-state condition,
.the core performance evaluation is used to determine the following principal thermal and hydraulic parameters associated with core behavior:
- 1. Core flow rate
- 2. Core thermal power level
- 3. MLHGR
- 4. MCPR
- 5. MAPLNGR.
14.2-171 Amendment 5
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~d.- Acceptance Criteria Level 1*.
Core f, low rate, core thermal ~ power level,.MLHGR, MCPR, and MAPLHGR not exceed the; limits specified by the plant techt.ical specifications.
14.2.12.3.18 . Warranty Test
- a. . Objective s The test objective is to demonstrate the reliability of the NSSS and to measure,the steam production rate and plant heat rate.
- b. . Prerequisite 4
The plant has been stabilized at rated conditions. All required instrumentation has been installed and calibrated. _ l
- c. ~ Test Method The plant is-operated for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> at raced conditions. During the 100-hour run, the steam production rate and plant heat rate is measured.
- d. Acceptance Criteria Leve i 1 : -
The reliability of the NSSS and the ability of the NSSS to develop rated output shall be demonstrated to be within warranty specifications.
14.2.12.3.19 Core Power - Void Mcde
- a. . Objective l l
The objective of this test is to measure the stability l of the core power void dynamic response, and to ,
i 14.2-172 Amendment 5 l 1
l l
- - . . . . _ . - . . . _ . . . . . _ _ _ . . . _ . . _ ~ . . . _ _ . - . _ _ . . _ . . . - . . . _ _ _ . _ _ _ ... . .._......!
l 7Jhl l
\
demonstrate that its behavior is within specified l design limits. -
l
- b. Prerequisites The core is maintained in a steady-state condition prior to the starting of this test.
- c. Test Method The core power void loop mode, that results from a combination of the neutron kinetics and core thermal hydraulics dynamics, is least stable near the natural circulation end of the rated 100% power rod line. A fast change in the reactivity balance is obtained by two-methods: (1) pressure regulator step change, and (2) by moving a very high worth control rod one or,two notches. Both local flux and total core response will be evaluated by monitoring selected LPRMs during the transient.
- d. Acceptance Criteria Level 11 The transient response of any system-related variables to any test: input *must not diverge. System related variables are. heat flux and reactor pressure.
14.2.12.3.20 Pressure Regulator
- a. Objectives
- 1. To determine optimum pressure regulator setting to control transients induced in the reactor pressure control system.
- 2. To demonstrate the-takeover capability of the backup pressure regulator via simulated failure of the controlling pressure regulator and to' set the regulating pressure difference between the two regulators and an appropriate value.
14.2-173 Amendment 5
, ,. - ~, ~ ,, -.--- --,. -- .~.-,--n ..,----n-.,------- . - - ,-c ,-- .n., , , , - , n - - = - --
bY .
1 4
=3. To demonstrate smooth pressure control transition
. between the t'urbine control valves and bypass valves.
- b. Prer quisites.
Instrumentation has been checked and calibrated. The plant is at a steady-state power. level.
- c. ' Test Method The pressure setpoint is decreased rapidly and then !
increased rapidly by about 10 psi. The response of the
~
system is measured in each case. The backup pressure regulator is tested by simulating failure of the operating pressure regulator. The bypass valve is :
- tested by reducing the lead limit, which requires the
-bypass valves to open and control the bypass steam )
flow. At certain test conditions, the results of the '
backup regulator test will be included with the core l
power - void mode test report.
- d. Acceptance Criteria Level 1'
- 1. The transient response of any pressure control system related variable to any test input must not diverge.
Le ve l c2 -
- f. t , In the recirculation manual mode the response time j from initiation of pressure setpoint change to the j
( turbine inlet pressure peak should be less than that specified in the GE startup test specification.
l l --
- l. J .7 Pressure control system deadband should be small enough that steady state limit cycles shall l
produce steam flow variations no greater than
, specified in the GE startup test specification.
I i
l f. 3*For all pressure regulator transients the peak )
neutron flux / peak vessel pressure should remain j below the scram settings by~the margins specified in the GE startup test specification. )
l l
l c 14.2-174 Amendment 5 l I l 1
e -
HCGS FSAR 4/84 5'.0,Theratioofthemaximumtotheminimumvalueof the incremental change in pressure control signal divided by the incremental change in steam _ flow shall meet'the requirements of the GE startup test specification.
Leijel 3I
(.l. . Control or bypass valve motion responds to pressure input with deadband no greater than that-required in the GE'startup test specification.
- 7. Oj---i-- M M a prc::x ; . k.t rr -- -i-ile- D62c
'14.2.12.3.21 Feedwater Control System
- a. Objectives
- 1. To evaluate and adjust feedwater controls
- 2. To demonstrate capability of the automatic core flow runback feature to prevent low water' level scram following the trip of one feedwater pump at 100% power.
- 3. To calibrate the feedwater speed controller 1and to verify that the maximum feedwater flow during pump runout does not exceed the flows assumed in Section 15.1.2.
A. To demonstrate response to feedwater temperature loss
- 5. To demonstrate acceptable reactor water level control.
- b. Prerequisite Instrumentation has been checked and calibratcd as appropriate. The plant is operating at steady-state conditions.
14.2-175 Amendment 5
' - - - - * ' ~ ^' ' '-
-4 -
- c. Test Method
- 1. Reactor water level setpoint changes of several lnches are used to evaluate and adjust the feedwater control system (FCS) settings for all
_ power and feedwater pump modes. The level I. setpoint change also demonstrates core stability to subcooling changes.
- 2. From near 100% power, one of the operating-
[. feedwater pumps is tripped. The automatic recirculation runback circuit will reduce i recirculation pump speed to drop power to within the capacity of the remaining turbine driven feedwater pumps. It is not expected that the.
. reactor will scram on low water level.
- 3. The condensate /feedwater system will be subjected.
l to a loss of feedwater heating. .The initial powe'r i level will.be approximately 80% prior to the start l of the test. It is expected that the feedwater )
! temperature decrease will be less than 1000F.
l
- 4. Feedwater pumps and turbine parameters are monitored during the power ascension to demonstrate operability within specifications.
This test includes initial calibration of the l speed controllers, and verification'that maximum feedwater flows do not exceed the flows assumed in 1 the FSAR. i L
L d. Acceptance' Criteria Level 1
- 1. . TheE transient response of any level control system related variable must not diverge.
! Level 2 l 7.g Level control system oscillatory modes of response, open loop dynamic response, response to H step disturbances, and steady state operation i l' shall meet the requirements specified in the GE startup test specification.
14.2-176 Amendment 5 1
l
.. . .. = . .
App'\
J .7, For feedwater-heater loss the maximum feedwater temperature decrease due to single failure is less
,& than that specified in the GE startup test
.V specification, and the resultant MCPR must be greater than the fuel thermal safety limit b
specified in the FSAR.
yg f.2- On the. trip of one feedwater pump, the reactor
'l shall avoid low water level scram by-the margin specified by the GE startup test specification.
t , ')
. ) .3 Maximum speed attained shall deliver flows consistent with the requirements specified by the l
qy GE startup test specification limits. l 14.2.12.3.22' Turbine Valve Surveillance
- a. Objective The test objective is to demonstrate the methods to be used and the maximum power level for routine surveillance testing of the main stop, control, and bypass valves.
- b. Prerequisite I The plant has been stabilized at the required power level.
I
- c. Test Method Individual main stop, control, and bypass valves are manually closed and reset at selected power levels.
The response of the reactor is monitored and the maximum power level conditions for the performance of this test are determined. The rate of valve stroking i and timing of the closed-open sequence are chosen to minimize the disturbance introduced.
l l
l 14.2-177 Amendment 5
<4 . ,,,,._.._-,__,,,__m_...____m.,._,-___,,_-..._-,r, _...-,,,..m._,.,__,-._.,,.-,.m ,,.,,,_,__~.,--,m..,__.._mm.__ .
~d. Acceptance Criteria, Levei 2 :
Peak heat flux, vessel pressure, and steam flow shall remain below scram or isolation trip settings by a margin consistent with the GE startup test specification.
14.2.12.3.23 Main Steam Isolation Valves I
- a. Objectives
- 1. To functionally check the MSIVs at selected power levels and determine the maximum power level they can be tested at individually
- 2. To determine isolation valves' closure times. l l
- 3. To determine reactor transient behavior during and following simultaneous closure of all MSIVs. .
- b. Prerequisites The plant has-been stabilized at the required power level.
. c. Test Method
- 1. Individual closure of each MSIV is performed at selected. power levels to verify functional performance and to determine closure times. The
, maximum power level is determined for individual closure with ample margin to scram.
- 2. A test of the simultaneous full closure of all MSIVs-is performed at about 100% power. Operation of the RCIC system and the relief valves is demonstrated. Reactor parameters are monitored to determine transient behavior of the system during the' simultaneous full closure test. The reactor will immediately scram due to the actuation of the 14.2-178 Amendment 5
.wg... ., , +-.--,--v,, , ,,,,,_,.,,,.-m. -
. . , , r,.. .--e.._.., -~ - - , ,_.._. , . , _ _ _ _ _ _ . _ - - _ _ _ _ _ _ _ - - - - --
HCGS FSAR 4/84 MSIV position switches. Recirculation pumps will trip if Level 2 in the RPV.is reached. The feedwater control system will prevent.the RPV water level from reaching the steam lines.
d.- Acceptance Criteria Level 1:
Leve l' 2 :
- 7. /, Peak neutron flux, vessel pressure, and steam flow shall remain below scram or isolation trip ;
settings by a margin consistent with design I requirements when individually testing the MSIVs. j l
J.1. Following the full closure of all MSIVs, vessel pressure and heat flux level shall'tue as specified I in the GE startup test specification.
]
3 bv5D' 1
/.7,.The RCIC system and relief valves shall function J \ i
~ { yJ* " , .l . in accordance with the GE startup test 1' specification following the MSIV closure from-high
+ E.d 1' 4 ' 'A power. ,
( ;e. hd -;) s.3 The reactor must immediately scram and the C feedwater control system must~ prevent the water I
gqqc g g from reaching the main steam lines following full closure of MSIVs from high power.
, 1 14.2.12.3.24 Relief Valves
- a. Objectives
- 1. To demonstrate proper operation of the main steam relief valves and determine their capacity
- 2. To demonstrate their leaktightness following operation.
2 i
e 14.2-179 Amendment 5
.,w-- .4 --.-..m ,,.._m._ ,.,,.,.-._,._,,.m,_.,_,,,,__ _ _ _ . , _ . , . , , _ . , . . , _ _ _ , _ _ . . _ _ . _ _ . _ . . _ . _ , _ _ _ , _ , - .
. - .- . . - . . .. . . _ = . - . - .
30f(4
- b. Prerequisites ,
The reactor is on pressure control with adequate bypass ~
i or matt steam flow.
L c. Test Method A' functional test of each safety relief valve (SRV) shall be made as early in the startup. program as practical. This is normally the first time the plant reaches 250 psig. The test is then repeated at rated reactor pressure. Bypass valves (BPV) response is monitored during the low pressure test and the-electrical output response is monitored during the rated pressure test. The test duration will be about 10-seconds to allow turbine valves and tailpipe sensors to reach a steady state.
! The' tailpipe sensor responses will be used to detect
- the opening and subsequent closure of each SRV. The BPV and MWe responses will be analyzed for anomalies l indicating a restriction in an SRV tailpipe.
Valve capacity will be based on certification by ASME code stamp and-the applicable documentation being available in the onsite records. Note that the nameplate capacity / pressure rating assumes that the flow is sonic. This will be'true if the back pressure is not excessive. A major blockage of the line would
- not necessarily be offset and it should be determined that none exists through the BPV response signatures, l
i I Vendor bench test data of the SRV opening responses L will be available onsite for comparison with l Section 5.2.2. The acoustic monitoring subsystem will
.be monitored during the relief valve test program to determine that the setpoints do reflect valve open/ valve closed conditions.
! SRV opening and reclosure setpoint data will be obtained and evaluated-during each high power trip test at which an SRV actuation is anticipated.
14.2-180 Amendment 5 i
- d. Acceptance Criteria l,.2 ve i 0' '
- 1. There should be positive indication of steam
, discharge during the manual actuation of each See a #a aeA .> L dye \ ~,
- 2. /, Decay ratio for pressure control variables is as specified in the GE startup test specification. '
- 3. 2. The tenperature measured by thermocouples on the.
discharge side of the valves should return to the temperature recorded.before the valve was open as required in the GE startup test specification.
The acoustic monitors shall indicate the valve is closed after valve closure.
W.3. During the 250 psig and the rated pressure functional tests, steam flow through each relief valve as compared to average relief valve flow is
-as specified in the GE startup test specification.
"h5 h- 1.
^
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- v to an cc
.14.2.12.3.25 Turbine Trip and Generator Load Rejection
- a. Objective The test objective is to demonstrate the proper response of the reactor and its control systems following trips of the turbine and generator.
r
- b. Prerequisites Power testing has been completed to the extent necessary for performing this test. The plant is stabilized at the required power level.
l 14.2-181 Amendment 5
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.. l'- / 3 'd (ph "i HCGS FSAR 4/84 c .- ' Test-Method
~
The t ine is tr d-at the differe power els
. .the ghout-the wer ascen n pro . For e f)?e ep 5 bine tr , the main erator eakers ain.1 e for a so there no ris n turbi gener r
'I d54 speed;tiwhereas, i he gen tot trip the ma ge ator brea s open d residu turbin stea ill use a mor ary ris n the ge ator ed.
At test condition 3, a turbine trip will be initiated manually from the control room. At test condition 6, a generatoc trip (load rejection) will be initiated by
-simulating a condition that will cause the generator ocedpdCI~
breakers to open. During both transients it is expected that the reactor will scra ^- it is not expected the HPCI or RCIC will initiate. Reactor water NIM Y 000EEE/ e/
a level, pressure, and heat flux will be monitored. The action of relief valves will be monitored.
, A generator-trip will be performed at low power such that nuclear boiler system steam generation is just within bypass valve capacity. The purpose of this test l is to demonstrate scram avoidance. "
During all three transients, main turbine stop, control, and bypass valve positions will be monitored.
! Prior to the low power generator trip, bypass valve
! capacity will be measured.
l
- d. Acceptance Criteria Le ve l 1 *
- 1. For turbine and generator trips at power levels greater than 50%, the response times of stop, control, and bypass valves shall be as specified in the GE startup test specification.
l '
l 2. Feedwater control system settings must prevent flooding the main steam lines.
- 3. The reactor recirculation pump drive flow coastdown shall be as specified in the GE startup test specification.
14.2-182 Amendment 5 I
r-
, ny'GG .
l 1
INSERT # 3 This _ test is performed at three different power levels in the ,
power ascension program. For the turbine trip, the main gengrator remains loaded .for a time so there is no rise in turbine generator speed, whereas, in the generator trip, the main
' generator outputLbreakers open and residual steam.will cause a momentary rise in' turbine generator speed.
INSERT # 4 (add to the sentence) and the recirculation pump trip (RPT) breakers will open.
._ _ .-*-___m_.____m__
d bh5 I{ph HCGS FSAR '4/84 4.' The positive change in vessel dome pressure and heat flum must not exceed the limits specified in the GE startup test specification.
, 5. The total time delay from start of turbine stop valve motion or turbine control valve motion to complete suppression of electrical are between the fully open contacts of the RPT circuit breakers shall be less than the limit specified in the GE startup test specification.
Le ve l 2 :
6.; , The messitred bypass valve capacity snall be equal to or greater than that required by the GE startup test specification, which compares bypass valve capacity to the accident analysis.
1 7/.2. There shall be no MSIV closure during the first three minutes of the transient and operator action shall not be required during that period to avoid the MSIV trip.
JH3, For the generator trip within bypass valves capacity, the reactor shall not scram for initial thermal power valves within that bypass valve capacity and below the power level at which trip scram is inhibited.
, FN. Low water level recirculation pump trip, HPCI and RCIC shall not be initiated, l
>6.jf'Feedwater level control shall avoid loss of
, feedwater due to high level trip during the event.
'the temperature measured by thermocouples on the 1/.6, discharge side of the valves should return to the temperature recorded before the valve was open as required in the CZ startup test specification.
The acoustic monitors shall indicate the valve is closed after valve closure.
14.2-183 Amendment 5
3dW h' HCGS FSAR 4/84 14.2.12.3.26 Shutdown From Outside the Main Control Room
! a. Objective l The test objective is to demonstrate that the reactor can be brought from an initial steady-state power level to hot standby and that the plant has the potential for being safely taken to a cold sautdown condition from hot. standby from outside the main control room.
- b. Prerequisites The plant is operating at the required power level.
- c. Test Method The test will be performed at a low power level and i will consist of demonstrating the capability to scram and initiate controlled cooling from outside the -
control room. The reactor will be scrammed gunb
- 4eeteted from outside the control room'after a simulated control room evacuation. Reactor pressure and water level will be controlled using SRVs, RCIC, and RHR from outside the control room during subsequent l cooldown. The cooldown will continue until RHR shutdown cooling mode is placed in service from outside the control room. Alternatively, verification of i . satisfactory operation of RHR shutdown cooling mode
- from outside the' control room may be done at some other, more convenient time during the startup program, r In either case, coolant temperature must be lowered at least 500F while in the shutdown cooling mode. During
.the shutdown cooling mode demonstration, cooling to the RHR heat exchanger via the safety auxiliaries cooling
( system and the station service water system will be-
[ accomplished from the remote shutdown panel. All other operator actions not directly related to reactor vessel
( level, temperature, and pressure control will be
[ performed in the main control room. The plant will be i
maintained in hot standby condition for at least 30 t
minutes during the performance of this test.
- d. Acceptance Criteria Ieyel 2 :
During a simulated main control room evacuation, the ability to bring the reactor to hot standby and subsequently cool down the plant and control vessel 14.2-184 Amendment 5
'6 9f HCGS TSAR 4/84 pressure and water level shall be demonstrated using equipment and edntrols located outside the main control room.
)
14.2.12.3.27 Recirculation Flow Control l
- a. Objectives i l
! 1. To determine plant response to changes in the l recirculation flow l l
.2. To optimize the setting of the master flow controller ;
. 3. To demonstrate plant loading capability. !
- b. Prerequisites i
The reactor is operating at steady-state conditions at the required power level.
l l
- c. Test Method With the reactor plant at the 50% load line, the recirculation speed loops are tested using large plus and minus step changes and and the speed controller i gains are optimized. After the speed loops have been ~
optimized, the system may be switched to the master manual mode and the automatic load following mode loop shall be optimized.
1 I l When the plant is tested along the 100% load line, the l recirculation system shall be tested by inserting small l plus and minus step changes in the local manual and
( master manual modes. The automatic load following loop is also tested by means of small load demand changes.
l l During recirculation flow control testing at the 50%
l
. and 100% load lines no scrams due to neutron flux or heat flux changes' transients are expected.
14.2-185 Amendment 5
- d. Acceptance Criteria, Lcy"I II
- 1. The. transient response to any recirculation system related varicble to any test input must not diverge.
Le ve { 2.
- 7.,, A scram shall.not occur due to recirculation flow maneuvers. Neutron flux and heat flux trip avoidance margins are as specified in the GE startup test specification.
3'.Z The decay ratio of any oscillatory controlled variable must be less than that required by the GE startup test specification.
pdbI '
. 01;;.J .ad ep;n ;;;: ' lec; Mj"-tmente-eee-as specified-in-the GMt'iYtFp74 sit'spetittcei.iern 5'.3. Steady state limit cycles shall not produce turbine steam flow variations greater than the value of steam flow specified in the GE startup l test specification.
(f.'j, In the scoop tube reset function, if the speed demand meter has not been replaced by an error meter, the speed demand meter must agree with the speed meter within the GE startup test specifications.
14.2.12.3.28 Recieculation System l
l a. Objectives l
- 1. To determine transient responses and steady-state conditions following recirculation pump trips at l selected power levels i
- 2. To obtain recirculation system performance data i
l l
l 14.2-186 Amendment 5 L . _ _ _ _ _ . _ , _ _ . _ ~ _ - .
39 (,G s
- 3. To verify that cavitation in the recirculation system does'not occur in the operating region of the power / flow map.
3
- 4. To verify the adequacy of the recirculation
-runback to mitigate a scram upon loss of one feedwater pump.
- 5. To verify that the feedwater control system can control water level without causing a turbine trip / scram following a single recirculation pump trip.
- 6. To demonstrate the adequacy of the recirculation pump restart procedure at the highest possible power level.
- b. Prerequisites The reactor is operating at steady-state conditions at required power level,
- c. Test Method ,
Single pump trips are performed at test condition 3 and
- 6. Dual pump trip is demonstrated at test condition 3.
The one-pump trip tests are to demonstrate that water level will not rise enough to threaten a high level trip of the main turbine or the feedwater pumps. The dual pump trip verifies the performance of the RPT circuit and the recirculation pump flow coastdown prior to the high power turbine generator trip tests. Single pump trips are initiated by tripping the MG set generator output breaker.
Adequate margins to scrams and capability of the -
feedwater system to prevent a high level trip will be monitored. The two pump trip will be initiated by simultaneously tripping both recirculation RPT breakers using a test switch. _The recirculation pump restart demonstrates the adequacy of the restart operating procedure at the highest possible power level.
14.2-187 Amendment 5
. ~, .. . _ _ _. .. ,. _,
4 .
4 HCGS FSAR 4/84 At several power and flow conditions, and in.
. conjunction with sihgle pump trip recoveries, recirculation system parameters are recorded.
a
- At test condition 3 and at.near' rated recirculation flow, a loss of a feedwater pump is simulated. This is done prior to an actual feedwater pump trip to determine the adequacy of recirculation pump runback feature in preventing a scram.
While at test condition 3, it will be demonstrated that the cavitation interlocks which runback the
- recirculation pumps on decreased feedwater flow are adequate to prevent operation where recirculation pump
- or jet pump cavitation can occur.
- d. Acceptance Criteria Level 1'
- 1. During recovery from one pump-trip, the reactor shall not scram.
</ lJ y.<$ 2 ' -
i
- 2. Neutron flux, heat flux, and reactor water level
. scram avoidance margins are as specified in the GE
.p' 4 startup test specification.
i.
,/ ' ' ' /2 The two pump drive flow coastdown time following a dual recirculation pump trip is as specified in the GE startup test specification.
l
- .7 System performance parameters, including core L flow, drive flow, jet pump M-ratio, core delta-pressure, recirculation pump efficiency and jet pump nozzle and riser plugging criteria are as specified in the GE startup test' specification.
i F.3' Runback logic shall have settings adequate to prevent operation in areas of potential
. cavitation.
l J'.'/ The recirculation pump shall runback upon a trip i of the runback circuit as required by the GE l
startup test specification.
i l 14.2-188 Amendment 5
+ ,,- . , - - - . ------.,-.-,-..,_,,,,--n,,_.-, n. ,, ___.,,,,.___,_,,_.,._.,,,m..-,_,.___,,,_, .,_,,_,,,.-__,,,-n ,--
. 4/rj'GG ;
~
14.2.12.3.29 Recirculation System Flow Calibration
]
- a. Objective The test objective is to perform a complete calibration of the installed recirculation system flow instrumentation, including specific signals to the plant process computer.
- b. Prerequisites The reactor is operating at steady-state conditions.
The initial calibration of the recirculation system flow instrumentation has been completed.
- c. Test Method l
l During the testing program at operating conditions required for rated flow'at rated power, the jet pump flow instrumentation is. adjusted to provide correct flow indication based on the jet pump flow. The flow-biased APRM/RBM system is adjusted to correctly follow core flow based on drive flow. Additionally, the total core flow and recirculation flow signals to the process computer will be calculated to read these two process variables.
L
- d. Acceptance. Criteria l
Leve i 2__ -
- 1. Jet pump flow instrumentation shall be adjusted such that the jet pump total flow recorder
. provides core flow at rated conditions.
- 2. The APRM/RBM flow bias instrumentation shall be adjusted to function per design at rated conditions, as specified in the GE startup test specification.
0 42L -
l l
l 14.2-189 Amendment 5 i
. bl 2J hh te. .. .
The flow control system shall be adjusted to limit maximum core flow to the value specified by the GE startup test specification.
t e
1 l
W/M 1
, HCGS FSAR 4/84 14.2.12.3.30 Loss'of Turbine-Generator and Offsite Power ,
i
- a. Objective The objective of this test is to demonstrate the response of the reactor and electrical equipment and systems during loss of the main generator and offsite power.
Prerequisites b.
The SDGs are in the auto-start mode, and the plant is 1
operating at power.
- c. Test Method With the power plant synchronized to the grid between i 20% and 30% power, the main turbine generator will be tripped followed by manual trips of all offsite power j to the 13.8 kV ring bus. This will simulate loss of turbine generator and offsite power.
l Reactor water level and the operation of safety systems, including RPS, standby diesels, RCIC, and HPCI, will be monitored.
The loss of offsite power condition will be maintained for at least 30 minutes to demonstrate that necessary l equipment, controls, and indication are available following the station blackout to remove decay heat
, from the core using only emergency power supplies and l distribution systems.
r
- d. Acceptance Criteria l Level [:
- 1. All safety systems, such as the RPS, SDG, RCIC, l and HPCI, function per design without manual assistance. Reactor parameters are maintained l within acceptable design limits. Normal reactor cooling systems maintain adequate suppression pool water temperature, adequate drywell cooling, and l
l 14.2-190 Amendment 5
[
HCGS FSAR 4/84 prevent actuation of the automatic ;
depressurization system.
Levee 3' * -
- 2cf . Proper instrument display to the reactor operator ,
shall be demonstrated, including power mor.itors, pressure, water level, control rod position, suppression pool temperature, and reactor cooling system status.
7.Z. The temperature measured by thermocouples on the discharge side of the valve should return to the temperature recorded before the valve was open as required in the GE startup test specification.
The acoustic monitors shall indicate the valve is r closed after valve closure.
14.2.12.3.31 ::,c:11 Piping Vibration Tesd5
- a. Objective ,
. The test objective is to verify that steady state
! vibration and transient induced pipe motion of systems discussed in Section 3.9.2 are acceptable,
- b. Prerequisites The system piping to be tested is supported and i restrained properly. Instrumentation for monitoring i . vibration has been installed and calibrated, where l applicable.
I
- c. Test Method L This test is an extension of the preoperational test f program. During Ateady state operation, designated pipes as delineated in Section 3.9.2 will be monitored l for vibration. Dynamic vibration measurements will be
- made on applicable piping following various plant and I system transients as specified in Sections 3.9.2.1.2.3, l 3.9.2.1.3, and 3.9.2.2.4.
i l
l 14.2-191 Amendment 5 i
. - - - _ _ - . ~ . _ _ _ . . - _ - _ _ . _ _ _ _ _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . -
- d. Acceptance Criteria hev cl .7 '
The piping displacements at the established locations shallinot exceed the limits specified by the piping designer, which are based on not exceeding ASME Section III Code stress values or ANSI B31.1 values.
These acceptable vibration levels will be used as acceptance criteria in the appropriate piping vibration startup test procedures.
t 14.2.12.3.32 Reactor Water Cleanup System L
- a. Objective The test objective is to demonstrate the operation of the RWCU system.
- b. Prerequisites The reactor has been operated at a near rated temperature and pressure long enough to achieve a steady-state condition.
- c. Test Method With the reactor at rated temperature and pressure, process variables are recorded during steady-state
- t. operation in three modes of operation of the RWCU system blowdown, hot standby, and normal. The bottom i head drain flow indicator will be calibrated by taking l flow from the bottom drain only and using the RWCU system inlet flow indicator as a standard to ecmpare I
against.
- d. Acceptance Criteria l l l.d /cl 2.'
- 1. The data indicating operation in the listed modes l shall be acceptable as specified by the GE startup
, test specification.
- 14.2-192 Amendment 5
- 2. Recalibrate bottom head flow indicator against RWCU flow
- indicator if the deviation is greater
~than GE startup test specifications.
- 3. Pump vibration as measured on the bearing housing and coupling end shall be less than or equal to GE.
startup test specifications.
I 14.2.12.3.33 Residual Heat Removal System
- a. Objectives ,
- 1. To demonstrate the ability of the RHR system to remove residual and' decay heat from the nuclear system, so that refueling and nuclear system servicing can be performed >
- 2. To condense steam while the reactor is isolated from the main condenser, in conjunction with the RCIC system.
I
- b. Prerequisites Preoperational testing has been completed. The test procedure has been reviewed, approved, and released for testing. Instrumentation has been checked or calibrated as appropriate. The plant is at or near
[ normal operating pressure and temperature.
I
- c. Test Method Three modes are tested to verify system capability under actual operating conditions. The modes to be tested are suppression pool cooling, shutdown cooling and steam condensing. During the operations, the heat i transfer rate is controlled to maintain acceptable I cooldown rates. Data are recorded and reviewed to
- verify the satisfactory operation of the RNR system within design limits.
14.2-193 Amendment 5
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- d. Acceptance Criteria, L p l 2..'
- 1. The RHR system performance in the steam condensing mode, suppression pool cooling mode and shutdown cooling mode meets the requirements of the GE startup test specification.
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! 14.2.12.3.34 Drywel4 Cooling Oye;.e.
- a. Objective I
The test objective is to demonstrate, under actual operating conditions, satisf actory erfer r ce of the com/in#
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- b. Prerequisites W Artroptufe fneyesA'%A 5 s S *W I*S'k Airfl:e b;l;r.;i..; ef t.'.eyey;;;; hr: _::r ;;;&l :t:d. '
Power ascension testint is in progress. l l R een s ~ M '"c psee W** % 0e. seans%s%m Jab.
l c. Test Method
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i Drywell atmospheric 4 temperatures are monitored and j -
recorded during plant heatup and power operation up to ,
2/ rated power. C, t -r arrr.n = =r-f te r -t r = m.........,.
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-fl;we er.d/;; ;;;1in; "rter fl;;; cre ;;de, if l : quire t l p\ 1.e r.ria*=4a re;;pt;bl; t:;;;;;tese limit .
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- d. Acceptance Criteria ,
Lo ,d. L'. ,,4 sfe w W almosf:4* r e'c. l l t,Drywel ature control shall meet or exceed the i limits specified in the plant technical specifications. ,
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14.2-194 Amendment 5
, k A A mos h *S In cddition, drywall atmosphoric, and hot piping penetration con-crete temperatures are checked at various power levels, up rated, with minimum drywell cooling capacity in service. Design to W ((
temperature limits are verified to be met, and cooling system adjustments are made as required to maintain acceptable tempe ra-tures.
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14.2.12.3.35 Off; : Ts,,eetaeat System l a. Objective.
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The test objective is to demonstrate proper operation of the c'f;;; .. ...6 system over its expected operating range.
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- b. Prerequisites l
i Initial calibration of instrumentation has been completed. Power ascension testing is in progress.
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- c. Test Method , r> a '-
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TEI ;::f e. 22..;;, Adjustments will be made, if necessary, to meet acceptable system performance.
- d. Acceptance Criteria y2=q[ N Laul3:
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performance as verified by data analysis shal meet designrequirementsspecifiedinSection11.3.l.sPjj(aasd!lf, 14.2.12.3.36 Water Level Measurement l l
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This test was included in Section 14.2.12.3.14. l l
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14.2.12.3.37 Penetration Temperature Test t f Yis hesh was inc luef ed tw becle'o n 14*2.12.3.
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- . Objectiye To verifv that the devwell Denetratinne ==enciated with
.; .yst--. provide adequate protection Ior the h M 'ing "aarrata p surr;; .d h_ "reseyuisii.es -
- i. Power ascension _ testing-is-irrprogress.
A- Tnetet'=entatier 1:- ::libratad_ _
-c. Te.L i-iwUiva p wu6say numbup and yvwer operations, the Conv&cte temperatures surrounding hot penetrations will be }
monitored. -
- d. Acceptance Criteria -
m The- concr* 6 : . _ . . . m 60VEding ndt piping 9' pe g ns shall not exceed 2000F.
14.2.12.3.38 Safecy Auxiliaries Cooling System l l
f a. Objective l
l The test objective is to demonstrate that the safety auxiliaries cooling system (SACS) performance margin is I
adequate to support engineered safety features ;
equipment over their full range of design requirements. ;
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- b. Prerequisites.
Initial instrument calibrations have been completed.
The plant is operating at the required test condition.
- c. Test Method
, 1 During the performance of the RHR shutdown cooling mode test, the SACS will also be evaluated to determine the heat removal capacity of the systen and deconstrate the capability of achieving. cold shutdown within the time specified in the design specificiation. - Ou ir.;
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- d. Acceptance Criteria t bfyr I A -
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14.2.12.3.39 BOP Piping Vibration and Expansion l This test was included in Sections 14.2.12.3.15 and 14.2.12.3.31. l l
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- I ]k SAC.5 healexclaryr sLA nuuS_or exceed 2_.
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- 14.2.12.3.40 CONFIRMATORY IMPLANT TEST OF SAFETY-l RELIEF VALVE DISCEARGE l
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- a. OBJECTIVE The objective of this test is to confirm assumptions and methodologies used in the plant unique analysis (PUA) (see a summary report in Appendix 3E) and show that the loads and structural responses documented in the PUAR for SRV discharge related loads ,
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are conservative compared to the responses which occur during actual SRV discharges. ,
- b. PREREQUISITES
- 1. Power level should be sufficient to support steady steam flow, during the test duration, through SRV discharge line with normal plant operating pressure l at the SRV.
- 2. Instrumentation for monitoring loads and structural responses has been installed and calibrated.
- c. TEST METHOD A shakedown test will be conducted to verify the test set-up is functioning properly. The -
to' sting will consist of single valve actuations (SVA) and subsequent consecutive valve actua-tions (CVA) of the same valve. Selection of the SRV discharge line used for testing will be based on WUREG-0763, " Guidelines for confirm- .
atory Inplant Tests of Safety-Relief valve 1
Discharges for BWR Plants," recommendations.
Data will be collected and analyzed by computer code to verify design analysis. ;
- d. ACCEPTANCE CRITERIA l Level 1 i
The peak pool boundary pressure during air I
I clearing and steam discharge during the valve '
actuation is less than the predicted valve specified in the PUAR.
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