ML20094N697
| ML20094N697 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/13/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8408160348 | |
| Download: ML20094N697 (150) | |
Text
r; Putsc Service O PS G Cornpany Electnc and Gas 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 l- Robert L. Mitti General Manager Nuclear Assurance and Regulation August 13, 1984 Director of Nuclear Reactor Regulation "
U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:
HOPE CREEK GENERATING STATION DOCKET NO. 50-354 DRAFT SAFETY EVALUATION REPORT OPEN ITEM STATUS Attachment 1 is a current list which provides a status of the open items identified in Section 1.7 of the Draft Safety Evaluation Report (SER). Items identified as " complete" are those for which PSE&G has provided responses and no confir- -
mation of status has been received from the staff. We will consider these items closed unless notified otherwise. In order to permit timely resolution of items identified as
" complete" which may not be resolved to the staff's satis-faction, please provide a specific description of the issue which remains to be resolved.
Attachment 2 is a current list which identifies Draft SER Sections not yet provided.
In addition, enclosed for your review and approval (see Attachment 4) are the resolutions to the Oraft SER open items, and FSAR question responses 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,
, h/
8408160348 840813 i l PDR ADOCK 05000354 l E PDR ,
0Q\ -
( \ \
The EAttgpE%epts ww om u
77 Director of Nuclear:
Reactor _ Regulation 2 8/13/84
~C D..H.. Wagner USNRC'. Licensing-Project Manager
~
W. H. Bateman USNRC Senior-Resident Inspector l's FM05.1/2 i
4
?
-3 F-
,e . . - , . , - -,nw--s-m1.w- w--+.-v -y.. r n pq e,.-n-nm.-,-e- .,ee, , , ,mae , n. wen,, ,.,+e- e-4 -,-mn , ,,- w.,w ,a wa-g -
UNITED STATES OF AMERICA NUCLEAR REGULATORY. COMMISSION DOCKET NO. 50-354 PUBLIC SERVICE ELECTRIC ~AND GAS COMPANY s
Public Service Electric and Gas Company hereby submits the-enclosed Hope Creek Generating Station Draft Safety Evalua-tion Report open item responses and FSAR Ouestion responses.
The matters set forth in this submittal are true to the best of my knowledge, information, and belief.
Respectfully submitted, Public Service Electric and Gas Company
\
By: "A , fj]
' ~ ~ ~ ~
T .' J . ' M ai~t'i Vice Presi t- Engineering and Constr tion Sworn to and subscribed before~me, aNotaryPubjic of New Jersey, this /3 - day of August 1984.
s )) (
DAVig K. gURD NOTARYPUBLIC OF NEW JERSr v My Comm. Espires go.23.g5 GJ02/2
DPLTE8 '8/13/84 ATTAOMENT 1 DSER R. L. MITIL TO OPEN SBCTION A. SCHWENCER ITEM MJMEER SUBJECT STATUS LEITER DATED 1 2.3.1- Design-basis tengeratures for safety- Open related a1xiliary systems 2a 2.3.3 Accuracies of meteorological Caplete 7/27/84 measurements 2b 2.3.3 Accuracies of meteorological Cmplete 7/27/84 measurements ,
2c 2.3.3 Accuracies of meteorological Caplete 8/13/84 measurements (Rev. 1) 2d 2.3.3 Accuracies of meteorological Couplete 8/13/84 measurements (Rev. 1) 3a 2.3.3 Upgrading of cosite meteorological 02tplete 8/01/84 measurements progra (III.A.2) 3b 2.3.3 Upgrading of onsite meteorological couplete 8/01/84 measurements program (III.A.2) (Rev. 1) 3c 2.3.3 Upgrading of onsite meteorological Open measurements program (III.A.2) 4 2.4.2.2 Ponding levels Couplete 8/03/84 Sa 2.4.5 Wave inpact and runup on service Cm plete 6/01/84 Water Intake Structure Sb 2.4.5 Wave inpact and runup cn service Open water intake structure Sc 2.4.5 Wave inpact and runup on service Cmplete 7/27/84 water intake structure 5d 2.4.5 Wave impact and runup cn service Caplete 6/01/84 i
water intake structure i
( 6a 2.4.10 Stability of erosim protection Open structures 6b 2.4.10 Stability of erosion protection Open structures 6c 2.4.10 Stability of erosion protection Cm plete 8/03/84 l 5tructures l
M P84 80/12 1-gs l
I'
i i
ATTA09ENT 1 (Cont'd) l DSER R. L. iEZrL TO OPEN SECTICN A. ST WENCER ITEM NUMBER SUBT.ECT STATUS LETTEF DXTED 7a 2.4.11.2 Thermal aspects of ultimate heat sink Cmplete 8/3/84 7b 2.4.11.2 Thermal aspects of ultbnate heat sink Cmplete 8/3/84 8 2.5.2.2 Choice of maximum earthquake for New Open England - Piedmont Tectonic Province 9 2.5.4 Soil darrping values Complete 6/1/84 10 2.5.4 Foundation level response spectra Cmplete 6/1/84 11 2.5.4 Soil shear moduli variation Complete 6/1/84 12 2.5.4 Combinaticn of soil layer properties Cmplete 6/1/84 13 2.5.4 Lab test shear moduli values Cmplete 6/1/84 14 2.5.4 Liquefaction analysis of river bottm Cmplete 6/1/84 sands 15 2.5.4 Tabulations of shear moduli Cmplete 6/1/84 16 2.5.4 Drying and wetting effect cn Cmplete 6/1/84 Vincentown 17- 2.5.4 Power block settleent monitoring Complete 6/1/84 18 2.5.4 Maxin n earth at rest pressure Complete 6/1/84 coefficient 19 2.5.4- Liquefaction analysis for service Cmplete 6/1/84 water piping 20 2.5.4 Explanation of observed power block Cmplete 6/1/84 settlement 21 2.5.4 Service water pipe settlement records Cmplete 6/1/84 22 2.5.4 Cofferdam stability Cmplete 6/1/84 23 2.5.4 Clarification of ESAR Tables 2.5.13 Cmplete 6/1/84 and 2.5.14 M P84 80/12 2 - gs
ATTACIMENE 1 (Cont'd)
DSER R. L. MITIL 'IO '
OPEN SECTICN A. SCHWENCER I'1Di NJMBER SUBJECT STA'IUS IEITER DATED +
24 2.5.4 Soil depth models for intake Couplete 6/1/84 ,
structure 25 2.5.4 Intake structure soil modeling Complete 8/10/84 26 2.5.4.4 Intake structure sliding stability Open -
27 ,2.5.5 Slope stability Couplete 6/1/84 28a 3.4.1 Flood protection Complete 7/27/84 28b 3.4.1 Flood protection Conple'ce 7/27/84 28c 3.4.l' Flood protection Couplete 7/27/84 28d 3.4.1 Flood protection Couplete 7/27/84 28e 3.4.1 Flood protection Conplete 7/27/84 28f 3.4.1 Flood protection Couplete 7/27/84 28g 3.4.1 Flood protection Conplete 7/27/84 29 3.5.1.1 Internally generated missiles (outside Conplete 8/3/84 containment) (Rev. 1) 30 3.5.1.2 Internally generated missiles (inside Closed 6/1/84 containment) (5/30/84-Aux.Sys.Mtg.)
31 3.5.1.3 Turbine missiles Conplete 7/18/84 32 3.5.1.4 Missiles generated by natural phenomena Conplete 7/27/84 33 3.5.2 Structures, systems, and conponents to Conplete 7/27/84 be protected frca externally generated missiles .
34 3.6.2 Unrestrained whipping pipe inside Conplete 7/18/84 containment 35 3.6.2 ISI program for pipe welds in Couplete 6/29/84 break exclusion zone M P84 80/12 3 - gs
- ,-m=- ,w- g .- t. - 9 - w -*-p-s.3 - ---.
~
1 l
l l
l l
ATTACIMENT 1 (Cont'd)
DSER R. L. MITTL TO OPEN SECTIN A. SCIMENCER ITEM MJMBER SUBJECT STATUS IEITER DATED 36 3.6.2 Postulated pipe ruptures Conplete 6/29/84 37 3.6.2 Feedwater isolation check" valve Open operability 38 3.6.2 Design of pipe rupture restraints Open 39 3.7.2.3 SSI analysis results using finite Conplete 8/3/84 element method and elastic half-space approach for containment structure 40 3.7.2.3 SSI analysis results using finite Cm plete 8/3/84 element method and elastic half-space approach for intake structure 41 3.8.2 Steel containment buckling analysis Conplete 6/1/84 42 3.8.2 Steel containment ultimate capacity Couplete 6/1/84 analysis 43 3.8.2 SRV/IOCA pool dynamic loads Conplete 6/1/84 44 3.8.3 ACI 349 deviations for internal Couplete 6/1/84 structures 45 3.8.4 ACI 349 deviations for Category I Conplete 6/1/84 structures 46 3.8.5 ACI 349 deviations for foundations Conplete 6/1/84 47 3.8.6 Base mat response spectra Conplete 8/10/84 Rev.1 48 3.8.6 Rocking time histories Conplete 6/1/84 49 3.8.6 Gross concrete section Couplete 6/1/84 50 3.8.6 Vertical floor flexibility response Conplete 6/1/84 spectra 51 3.8.6 Conpariscn of Bechtel independent Conplete 8/10/84 Rev. 1 verification results with the design-basis results M P84 80/12 4 - gs
ATTACH >L7FP 1 (Cont'd)
DSER R. L. MITIL 'IO '
OPEN - SECTICN A. SCIMENCER ITEM M.1MBER SUBJa7 STA'IUS IEITER DATED 52 3.8.6 Ductility ratios due to pipe break Conplete 8/3/84 53 3.8.6 Design of seismic Category I tanks Conplete 6/1/84 54 3.8.6 Combination of vertical responses Conplete 8/10/84 Rev.1 55 3.8.6 'Ibrsional stiffness calculation Conplete 6/1/84 56 3.8.6 Drywell stick model developnent Conplete 6/1/84 57 . 3.8.6 Rotational tine history inputs Conplete 6/1/84 58 3.8.6 "O" reference point for auxiliary Conplete 6/1/84 building model 59 3.8.6 Overturning moment of reactor Conplete 6/1/84 building foundation mat 60 3.8.6 BSAP element size limitations Conplete 6/1/84 61 3.8.6 Seismic modeling of drywell shield Conplete 6/1/84 wall 62 3.8.6 Drywell shield wall boundary Cm plete 6/1/84 conditions 63- 3.8.6 Reactor building dome boundary Couplete 6/1/84 conditions 64 3.8.6 SSI analysis 12 Hz cutoff frequency Conglete 6/1/84 65 3.8.6 Intake structure crane heavy load Couplete 6/1/84 drop 3.8.6 Inpedance analysis for the intake Conplete 66 8/10/84 Rev.1 structure ,
i 67 3.8.6 Critical loads calculation for Conplete 6/1/84 reactor building dome 68 3.8.6 Reactor building foundation not Cm plete 6/1/84 contact pressures M P84 80/12 5 - gs
ATTACINENT 1 (Cont'd)
DSER R. L. MITIL 'IO '
OPEN SECTIOi A. SCHWENCER
, ITEM NUMBER SUBJECT STA'IUS IErrEP DATED 69 3.8.6 Factors of safety against sliding and Conplete 6/1/84 overturning of drywell shield wall 70 3.8.6 Seismic shear force distributicn in Couplete 6/1/84 cylinder wall 71 3.8.6 Overturning of cylinder wall Ccuplete 6/1/84 72 3.8.6 Deep beam design of fuel pool walls Ccmplete 6/1/84 73 3.8.6 ASHSD dome model load inputs Ccmplete ,6/1/84 74 3.8.6 Tornado depressurization Ccmplete 6/1/84 75 3.8.6 Auxiliary building abnormal pressure Ccmplete 6/1/84 76 3.8.6 Tangential shear stresses in' drywell Ccmplete 6/1/84 shield wall and the cylinder wall 77 3.8.6 Factor of safety against overturning Ccmplete 6/1/84 of intake structure 78 3.8.6 Dead load calculations Ccmplete 6/1/84 79 3.8.6 Post-modification seismic loads for Ccmplete 6/1/84 the torus 80 3.8.6 Torus fluid-structure interactions Ccmplete 6/1/84 81 3.8.6 Seismic displacement of torus Ccmplete 6/1/84 82 3.8.6 Review of seismic Category I tank Ccmplete 6/1/84 l design 83 3.8.6 Factors of safety for drywell Ccmplete 6/1/84 buckling evaluation j 84 3.8.6 Ultimate capacity of containment Ccmplete 6/1/84 (m-terials) 85 3.8.6 Inad ocnbination consistency Ccmplete 6/1/84 i
l I M P84 80/12 6 gs
ATTACNtDrr 1 (Cont'd)
DSER R. L. MITIL '10 ;
OPEN SECTIOi A. SCHWDiCER ITEM NUMBER SUBJECT STA'IUS IEITER DATED 86 3.9.1 Cm puter code validation Open 87 3.9.1 Information on transients Open 88 3.9.1 Stress analysis and elastic-plastic Cmplete 6/29/84 analysis 89 3.9.2.1 Vibratim 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 Cmplete 6/29/84 92 3.9.2.2 Triple flued-head containment Conplete 6/15/84 penetrations 93 3.9.3.1 Imd cmbinations and allowable Cmplete 6/29/84 stress limits 94 3.9.3.2 Design of SRVs and SRV discharge Cmplete 6/29/84 Pi ping 95 3.9.3.2 Fatigue evaluation on SRV piping Cmplete 6/15/84 and IOCA downcmers 96 3.9.3.3 IE Information Notice 83-80 Cmplete 6/15/84 97 3.9.3.3 Buckling criteria used for conponent Couplete 6/29/84 supports 98 3.9.3.3 Design of bolts Cmplete 6/15/84 l 99a 3.9.5 Stress categories and limits for Caplete 6/15/84 l core support structures l
99b 3.9.5 Stress categories and limits for Cmplete 6/15/84 core support structures 100a 3.9.6 10CFR50.55a paragraph (g) Cmplete 6/29/84 l
l l
M P84 80/12 7 gs
e l
l i
ATUOMENF 1 (Cant'd)
DSER R. L. MITFL '10 OPEN SECTIGi A. SCHWENCER ITEM NLMBER SUBJECT STA'1US LETTE3t DATED .
100b 3.9.6 .10CFR50.55a paragraph (g) Open 101 3.9.6 PSI and ISI progrand for punpa and Open valves 102 3.9.6 Leak testing of pressure isolation Conglete 6/29/84 valves 103a1 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103a2 3.10 Seismic and dynamic qualification of Open ,
mechanical and electrical equipment 103a3 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103a4 3.10 Seismic and dynamic qualificatim of Open mechanical and electrical equipment 103a5 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment 103a6 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment 103a7 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent ,
103b1 3.10 Seismic and dynamic qualificatim of Open mechanical and electrical equignent 103b2 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 1
103b3 3.10 Seismic and dynamic qualification of open mechanical and electrical equipment
- 103b4 ' 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent l 103b5 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent i
i l
l M P84 80/12 8 - gs l
I 4
ATPACMENT 1 (Cont'd)
DSER R. L. MITTL 10 '
OPEN SECTICN A. SCHWENCER *l ITEM NUMBER SURTECT STATUS IEITER DATED 103b6 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103c1 3.10 Seismic and dynamic qualification of Open m chanical and electrical equipment 103c2 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 103c3 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipment 103c4 3.10 Seismic and dynamic qualification of Open mechanical and electrical equipnent 104 3.11 Environmental qualification of NRC Action mechanical and electrical equipment 105 4.2 Plant-specific mechanical fracturing Cmplete 7/18/84 analysis 106 4.2 Applicability of seismic andd IDCA Cmplete 7/18/84 loading evaluation 107 4.2 Minimal post-irradiation fuel Ccmplete 6/29/84 surveillance program 108 4.2 Gadolina thermal conductivity Conplete 6/29/84 equation 109a 4.4.7 TMI-2 Item II.F.2 Open 109b 4.4.7 TMI-2 Item II.F.2 Open 110a 4.6 Functional design of reactivity Couplete 7/27/84 control systems 110b 4.6 Functional design of reactivity Conplete 7/27/84 control systems 111a 5.2.4.3 Preservice inspection program Cmplete 6/29/84 l (conponents within reactor pressure boundary) l M P84 80/12 9 - gs 1
l ATTAOMENT 1 (Cont'd)
DSER R. L. MITIL 10 -
OPEN SECTIOi A. SCHWENCER ITEM NUMBER SUBJECT STATUS LETTER DATED 111b 5.2.4.3 Preservice inspection program Complete 6/29/84 (ca ponents within reactor pressure boundary) 111c 5.2.4.3 Preservice inspection program Ccuplete 6/29/84 (ca ponents within reactor pressure boundary) 112a 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 leakage detection ,
112b 5.2.5 Reactor coolant pressure boundary Complete 7/27/84 leakage detection 112c 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 leakage detecticn 112d 5.2.5 Reactor coolant pressure boundary Cm plete 7/27/84 leakage detection 112e 5.2.5 Reactor coolant pressure boundary Cmplete 7/27/84 leakage detecticn 113 5.3.4 GE procedure applicability Cm plete 7/18/84 114 5.3.4 Compliance with NB 2360 of the Sumer Cmplete 7/18/84 1972 Addenda to the 1971 ASME Code 115 5.3.4 Drop weight and Charpy v-notch tests Caplete 7/18/84 for closure flange materials 116 5.3.4 Charpy v-notch test data for base Cmplete 7/18/84 materials as used in shell murse No.1 117 5.3.4 Capliance with NB 2332 of Winter 1972 Open Addenda of the ASME Code 118 5.3.4 Imad factors a .d neutron fluence for Open surveillanco capcules M P84 80/12 10- gs I
i 1
i i
t
\
ATTAC19ENT 1 (Cont'd)
- DSER R. L. MITTL TO OPEN SECTICH A. SCHWENCER ,
IT1M NUMBER SUa7ECT STATtJS IEITER DATED 119 6.2 TMI item II.E.4.1 Caplete 6/29/84 120a 6.2 TMI Item II.E.4.2 Open 120b 6.2 TMI Item II.E.4.2 Open 121 6.2.1.3.3 Use of NUREG-0588 Cmplete 7/27/84 122 6.2.1.3.3 Tenperature profile Caplete 7/27/84 123 6.2.1.4 Butterfly valve operation (post Caplete 6/29/84 accident) 124a 6.2.1.5.1 RPV shield annulus analysis Ca plete 6/1/84 124b 6.2.1.5.1 RPV shield annulus analysis Cmplete 6/1/84 124c 6.2.1.5.1 RPV shield annulus analysis Cmplete 6/1/84 125 6.2.1.5.2 Design drywell head differential Cmplete 6/15/84 pressure 126a 6.2.1.6 Redundant position indicators for . Open vacuta breakers (and control rom alarms) 126b 6.2.1.6 Redund&nt position indicators for Open vacuts breakers (and control rom alarms) 127 6.2.1.6 Operability testing of vacuum breakers Cmplete 7/18/84 128 6.2.2 Air ingestion Caplete 7/27/84 129 6.2.2 Insulation ingestion Cmplete 6/1/84 130 6.2.3 Potential bypass leakage paths Cmplete 6/29/84 131 6.2.3 Administration of secondary contain- Cmplete 7/18/84 ment openings M P84 80/12 11- gs
ATTACMENT 1 (Cont'd)
DSER R. L. MITIL TO OPEN SECTION A. SOiWENCER ITEM NI M SR SUBTECT STATUS IEITER DATED 132 6.2.4 containment isolation review Caplete 6/15/84 133a 6.2.4.1 Contairunent purge system Open 133b 6.2.4.1 Containment purge system Open 133c 6.2.4.1 Contairunent purge system Open 134 6.2.6 Containment leakage testing Cmplete 6/15/84 135 6.3.3 LPCS and LPCI injection valve Open interlocks 136 6.3.5 Plant-specific IDCA (see Section Conplete 7/18/84 15.9.13) 137a 6.4 Control room habitability Open 137b 6.4 Control .xxzn habitability Open 137c 6.4 Control room habitability Open 138 6.6 Preservice inspection program for Cm plete 6/29/84 Class 2 and 3 coroponents 139 6.7 MSIV leakage control systen Conplete 6/29/84 140a 9.1.2 Spent fuel pool storage Cmplete 7/27/84 140b 9 .1.2 Spent fuel pool storage Cmplete 7/27/84 140c 9 .1.2 Spent fuel pool storage Couplete 7/27/84 i 140d 9 .1.2 Spent fuel pool storage Conplete 7/27/84 141a 9.1.3 Spent fuel cooling and cleanup Cmplete 8/1/84
- system 141b 9.1.3 Spent fuel cooling and cleanup Cmplete 8/1/84 systen l 141c 9.1.3 Spent fuel pool cooling and cleanup Conglete 8/1/84 ,
system .
1 M P84 80/12 12- gs l
ATTAC19 TENT 1 (Cont'd)
DSER R. L. MITIL 'IO OPEN SECTION A. SCHWENCER ITEM NUMBER SUBJECT STA'IUS LETTER DATED 141d 9.1.3 Spent fuel pool cooling and cleanup Cmplete 8/1/84 systen 141e 9.1.3 Spent fuel pool cooling and cleanup Cmplete 8/1/84 systen 141f 9.1.3 Spent fuel pool cooling and cleanup Caplete 8/1/84 systen 141g 9.1.3 Spent fuel pool cooling and cleanup Conplete 8/1/84 system 4 142a 9.1.4 Light load handling system (mlated Closed 6/29/84 to refueling) (5/30/84-Aux.sys.Mtg.)
142b 9.1.4 Light load handling system (related Closed 6/29/84 to refueling) (5/30/84-Aux.Sys.Mtg.)
143a 9.1.5 overhead heavy load handling Open 143b 9.1.5 Overhead heavy load handling open 144a 9.2.1 Station service water system Cmplete 7/27/84 144b 9.2.1 Station service water system Cmplete 7/27/84 144c 9.2.1 Station service water system Caplete 7/27/84 145 9.2.2 ISI program and functional testing Closed 6/15/84 of safety and turbine auxiliaries (5/30/84-cooling systems Aux.Sys.Mtg.)
146 9.2.6 Switches and wiring associated with Closed 6/15/84 HPCI/RCIC torus suction (5/30/84-Aux.Sys.Mtg.)
147a 9.3.1 Capressed air systems Ccuplete 8/3/84 (Rev 1) 147b 9.3.1 Capressed air systems Caplete 8/3/84 (Rev 1)
M P84 80/12 13- gs
ATTACIMENT 1 (Cont'd)
DSER R. L. MITTL '10 OPEN SECTION A. SOMENCER ITEM NLMBER SUBJECT STA'IUS LETTER DATED 147c 9.3.1 Cmpressed air systems Ccuplete 8/3/84 (Rev 1) 147d 9.3.1 Ccmpressed air systems Ccmplete 8/3/84 (Rev 1) 148 9.3.2 Post-accident sanpling system Open (II.B.3) 149a 9.3.3 Equipment and floor drainage system Cmplete 7/27/84 149b 9.3.3 Equipnent and floor drainage system Ccmplete 7/27/84 150 9.3.6 Primary containment instrument gas Ccmplete 8/3/84 system (Rev. 1) 151a 9.4.1 Control structure ventilation system Ccmplete 7/27/84 151b 9.4.1 Control structure ventilation system Cmplete 7/27/84 152 9.4.4 Radioactivity monitoring elements Closed 6/1/84 (5/30/84-Aux.Sys.Mtg.)
153 9.4.5 Engineered safety features ventila- Cmplete 8/1/84 tion system (Rev 1) 154 9.5.1.4.a Metal roof deck construction Ccmplete 6/1/84 classificiation 155 9.5.1.4.b Ongoing review of safe shutdown NRC Action capability 156 9.5.1.4.c Ongoing review of alternate shutdown NIC Action capability 157 9.5.1.4.e Cable tray protection Open 158 9.5.1.5.a Class B fire detection system Ccmplete 6/15/84 159 9.5.1.5.a Primary and secondary power supplies Ccmplete 6/1/84 for fire detection system 160 9.5.1.5.b Fire water pung capacity Cmplete 8/13/84 M PO4 80/1214 gs
XETADMBtr 1 (Cont'd)
!' DSER R. L. MITTL 10 GWI SECTICBI A. SOfMENCER ITB4 NLMBER SUBJECT STATUS LETTER DEN lD 161 9.5.1.5.b Fire water valve supervision Ca plete 6/1/84 162 9.5.1.5.c Deluge valves complets 6/1/84 163 9.5.1.5.c Manual home station pipe sizing Complete 6/1/84 164 9.5.1.6.e Runote shutdown panel ventilation Cenplete 6/1/84 165 9.5.1.6.g anergency diesel generator day tank Complete 6/1/84 protection 166 12.3.4.2 Airborne radioactivity monitor. Caplete 7/18/84 positioning Portable continuous air monitors Caplete 7/18/84 l 167 12.3.4.2 168 12.5.2 Equipment, training, and procedures Cenplete 6/29/84 l for inplant iodine instrumentation i 169 12.5.3 Guidance of Division B Regulatory Caplete 7/18/84 (kiides 170 13.5.2 Procedures generation package Cmplete 6/29/84 nutanittal ,
171 13.5.2 1MI Itan I.C.1 Caplete 6/29/84 172 13.5.2 PGP Counitment Complete 6/29/84 173 13.5.2 Procedures covering abnomal releases Caplete 6/29/84 of radioactivity 174 13.5.2 Resolution explanation in FSAR of Couplete 6/15/84 TMI Items I.C.7 and I.C.8 .
175 13.6 Physical security Open 176a 14.2 Initial plant test progran ,
Caplete 8/13/84 i
M P84 80/12 15- gs i
j
ATTACIMENT 1 (Cont'd)
DSER R. L. MITTL 'IO OPEN SECTION A. SOiWENCER ITEM NLMBER SUBJECT STA'IUS LETTER DATED 176b 14.2 Initial plant test progem Caplete 8/13/84 176c 14.2 Initial plant test program Cmplete 7/27/84 176d 14.2 Initial plant test program Caplete 8/13/84 Rev. I 176e 14.2 Initial plant test program Caplete 7/J7/#f' 176f 14.2 Initial plant test program Caplete 8/13/84 176g 14.2 Initial plant test program Open l 176h 14.2 Initial plant test program Caplete 8/13/84 1761 14.2 Initial plant test program Cmplete 7/27/84 177 15.1.1 Partial feedwater heating Cmplete 7/18/84 178 15.6.5 IOCA resulting frm spectrum of NRC Action postulated piping breaks within RCP i 179 15.7.4 Radiological consequences of fuel NBC Action handling accidents
! 180 15.7.5 Spent fuel cask drop accidents NRC Action 181 15.9.5 'IMI-2 Item II.K.3.3 Cmplete 6/29/84 182 15.9.10 'IMI-2 Item II.K.3.18 Cmplete 6/1/84 183 18 Hope Creek DCRDR Open 184 7.2.2.1.e Failures in reactor vessel level Cmplete 8/1/84 sensing lines (Rev 1) 1 185 7.2.2.2 Trip syste sensors and cabling in Caplete 6/1/84 turbine building ,
186 7.2.2.3 Testability of plant protection Caplete 8/13/84 Rev. 1 systems at power M P84 80/1216- gs l
m.
4 ATTA09 TENT 1 (Cont'd)
DSER R. L. MITTL '10 OPEN SECTICH A. SOlWENCER ITEM NUMBER SUB7ECT STKIUS LEffrER DM1lD 187 7.2.2.4 Lifting of leads to perform surveil- Caplete 8/3/84 lance testing ,
188 7.2.2.5 Setpoint methodology Caplete 8/1/84 189 7.2.2.6 Isolation devices Caplete 8/1/84 190 7.2.2.7 Regulatory Guide 1.75 Ca plete 6/1/84 4 191 7.2.2.8 Scram discharge voltne Caplete 6/29/84 192 7.2.2.9 Reactor mode switch Caplete 6/1/84 193 7.3.2.1.10 Manual initiation of safety systens Caplete 8/1/84 194 7.3.2.2 Standard review plan deviations Caplete 8/1/84 (Rev 1)
- 195a 7.3.2.3 Freeze-protection / water filled Caplete 8/1/84 instrument and sanpling lines and cabinet temperature control j 195b 7.3.2.3 Freeze-protection / water filled Cmplete 8/1/84 j instrtment and sanpling lines and cabinet tenperature control I
196 7.3.2.4 Sharing of ommon instrtment taps Caplete 8/1/84 197 7.3.2.5 Microprocessor, multiplexer and Conplete 8/1/84 conputer systems (Rev 1) 198 7.3.2.6 TMI Item II.K.3.18-ADS actuation Open 199 7.4.2.1 IE Bulletin 79-27-Ioss of ncm-class Couplete 8/1/84 IE instnnentation and control power I
system bus during operation 200 7.4.2.2 Remote shutdown systen Conglete 6/1/84 l 201 7.4.2.3 RCIC/HPCI interactions Caplete 8/3/84 202 7.5.2.1 Imvel measurement errors as a result Caplete 8/3/84 of environmental tenperature offects on level instrumentation reference leg 3
( M P84 80/12 17- gs -
T' ATTACitENr 1 (Cont'd)
DSER R. L. MITIL '10 OPEN SECTIOi A. SCHWENCER ITEM NUMBER SUBJECf STATUS IErrER IATED ,
203 7.5.2.2 Regulatory Guide 1.97 Cm plete 8/3/84 204 7.5.2.3 TMI It m II.F.1 - Accident monitoring Couplete 8/1/84 205 7.5.2.4 Plant process cmputer system Conplete 6/1/84 206 7.6.2.1 High pressure / low pressure interlocks Cm plete 7/27/84 207 7.7.2.1 HELBs and consequential control system Cmplete 8/1/84' failures 208 7.7.2.2 Multiple control system failures Caplete 8/1/84 209 7.7.2.3 Credit for non-safety related systems Couplete 8/1/84 in Chapter 15 of the FSAR (Rev 1) 210 7.7.2.4 Transient analysis recording system Cmplete 7/27/84 211a 4.5.1 Control rod drive structural materials Cmplete 7/27/84 211b 4.5.1 Control rod drive structural materials Cmplete 7/27/84 211c 4.5.1 Control rod drive structural materials Cmplete 7/27/84 211d 4.5.1 Control red drive structural materials Cmplete 7/27/84 211e 4.5.1 Control rod drive structural materials Caplete 7/27/84 212 4.5.2 Reactor internals materials Caplete 7/27/84 213 5.2.3 Reactor coolant pressure boundary Caplete 7/27/84 material l I
214 6.1.1 Engineered safety features materials Cmplete 7/27/84 i l
, 215 10.3.6 Main stean and feedwater system Caplete 7/27/84 l materials 216a 5.3.1 Reactor vessel materials Caplete 7/27/84 M P64 80/12 18- gs
I l
ATTAOMENT 1 (Cont'd)
DSER R. L. MITTL TO OPEN SECTION A. SOMENCER ITEM NLMBER SUBJECT STATUS LETTER IRTED 216b- 5.3.1 Reactor vessel materials Ccmplete 7/27/84 217 9.5.1.1 Fire protection organization Open 218 9.5.1.1 Fire hazards analysis Ccuplete 6/1/84 219 9.5.1.2 Fire protection ackninistrative Open controls 220 9.5.1.3 Fire brigade and fire brigade Open training 221 8.2.2.1 Physical separation of offsite Ccaplete 8/1/84 transmission lines 222 8.2.2.2 Design provisions for re-establish- Ccuplete 8/1/84 ment of an offsite power source 223 8.2.2.3 Independence of offsite circuits Ccuplete 8/1/84 between the switchyard and class IE 1 buses t
224 8.2.2.4 Ctenon failure mode between onsite Ccmplete 8/1/84 and offsite power circuits 225 8.2.3.1 Testability of autcnatic transfer of Ccmplete 8/1/84 power frczn the normal to preferred power source 226 8.2.2.5 Grid stability Ccuplete 8/13/84 Rev. 1 227 8.2.2.6 Capacity and capability of offsite Ccuplete 8/1/84 circuits 228 8.3.1.1(1) Voltage drop during transient condi- Ccnplete 8/1/84 tions 229 8.3.1.l(2) Basis for using bus voltage versus Ccuplete 8/1/84 actual connected load voltage in the voltage drop analysis 230 8.3.1.l(3) Clarification of Table 8.3-11 Ccmplete 8/1/84 M P84 80/12 19- gs l
- - - , , , . , -, --,-c---- - , , - - , , , - - . - - , - - , . , - - - - --.-v ,, - -- ,,-, ,-,.-.-,-n, - , . - - ~ ~ ,
l ATTAC MENT 1 (Cont'd)
D6ER R. L. MITIL '1D
- OPEN SECTICN A. SOfWENCER ITM NUMBER SUBJECT STKIUS LETTER IRFED 231 8.3.1.1(4) Undervoltage trip setpoints Ca plete 8/1/84 232 8.3.1.l(5) road configuration used for the Conglete 8/1/84 voltage drop analysis 233 8.3.3.4.1 Periodic systen testing Caplete 8/1/84 234 8.3.1.3 Capacity and capability of onsite Couplete 8/1/84 AC power supplies and use of ad-ministrative controls to pmvent overloading of the diesel generators 235 8.3.1.5 Diesel generators load acceptance Caplete 8/1/84 test 236 8.3.1.6 Capliance with position C.6 of Caplete 8/1/84 IG 1.9 237 8.3.1.7 Decription of the load sequencer Caplete 8/1/84 j 238 8.2.2.7 Sequencing of loads on the offsite Caplete 8/1/84 power systen i 239 8.3.1.8 Testing to verify 80% mininun Open '
i voltage
! 240 8.3.1.9 Capliance with BrP-PSB-2 Caplete 8/1/84 241 8.3.1.10 toad acceptance test after prolonged Caplete 8/1/84 no load operation of the diesel generator 242 8.3.2.1 Ca pliance with position 1 of Regula- Cmplete 8/1/84 tory Guide 1.128 243 8.3.3.1.3 Protection or qualificaticr1 of Class Conglete 8/1/84 1E equipnent fran the effects of fim suppression systems 244 8.3.3.3.1 Analysis and test to demo *nstrate Caplete 8/1/84 adequacy of less than specified separation M P84 80/12 20- gs ,
l 1
( ;
XITACitlENT 1 (Cont'd)
DSER R. L. MTITL TO 0F58 SECTIGt A. SOllENCER ITBt IUlBER IKIILTECT STATUS LEITER DNTED 245 8.3.3.3.2 The use of 18 versus 36 inches of Ocuplete 8/1/84 separation between raceways 246 8.3.3.3.3 Specified separation of raceways by Ca plete 8/1/84 analysis and test-247 8.3.3.5.1 Capability of penetrations to with- Omplete 8/1/84 stand long duration short circuits at less than maximan or worst case short circuit 248 8.3.3.5.2 Separation of penetratian primary Complete 8/1/84 and backup protections 249 8.3.3.5.3 'Ihe use of bypassed thermal overload Caplete 8/1/84 protective devices for penetration protections 250 8.3.3.5.4 Testing of fuses in accordance with Omplete 8/1/84 R.G. 1.63 251 8.3.3.5.5 Fault current analysis for all Couplete 8/1/84 representative penetration circuits 252 8.3.3.5.6 The use of a single breaker to provide Complete 8/1/84 penetration protection 253 8.3.3.1.4 Cannitment to protect all Class 1E cmplete 8/1/84 "
equipment fran external hasards versus only class IE equipment in one division 254 8.3.3.1.5 Protection of class 1E power supplies Complete 8/1/84 fran failure of unqualified class IE loads 255 8.3.2.2 Battery capacity Couplete 8/1/84 256 8.3.2.3 Automatic trip of loads to maintain Ca plete 8/13/84 sufficient battery capacity M P84 80/12 21;- gs
l l
. I NrrA094ENr 1 (Cont'd)
DSER R. L. MrrrL '!O OPSI SBCFIGI A. Son etCER rret IR3etR SUE 7BCr STARIS LETTER DNFED ,
257 8.3.2.5 Justification for a'0 to 13 second Conglete 8/1/84 load cycle 258 8.3.2.6 Design and qualification of DC Caplete 8/1/84 system 1 =in to operate between ,
i minima and maxima voltage levels 259 8.3.3.3.4 Use of an inverter as an isolation Ca plete 8/1/84 device 260 8.3.3.3.5 Use of a single breaker tripped by Cap'.ete 8/1/84 a IDCA signal used as an isolation device i 261 8.3.3.3.6 Autcmatic transfer of loads and Caplete 8/1/84 interconnection between redundant divisions 262 11.4.2.d solid weste control program open f
263 11.4.2.e Fire protection for solid radwaste Caplete 8/13/84 l storage area 264 6.2.5 Sources of oxygen open 265 6.8.1.4 ESF Filter Testing Cmplete 8/13/84 266 6.8.1.4 Field leak tests Caplete 8/13/84 267 6.4.1 Control roan toxic chemical Ca plete 8/13/84 detectors 268 Air filtration unit drains Open 269 5.2.2 Code cases N-242 and N-242-1 Open 270 5.2.2 Code case N-252 oa TS-1 2.4.14 Closure of watertight doors to safety- Open l related structures ;
i t
I
{
M P84 80/12 22- gs
r "
- ~
ATTADMENT 1 (Cont'd)
DSER R. L. MITTL 'IO OR!N SECTION A. SOfWENCFA ITI!N NIMBER SUILTECT STA1tJS IEITER IRTED TS-2 4.4.4 Single recirculation loop cperation Open TS-3 4.4.5 Core flow monitoring for crud effects Complete 6/1/84 TS-4 4.4.6 toose parts monitoring systen Open TS-5 4.4.9 Natural circulation in nonnal Open operation I
j TS-6 6.2.3 Secondary containment negative Open
[ pressure l
TS-7 6.2.3 Inleakage and drawdown tilne in Open secondary containnent TS-8 6.2.4.1 IAakage integrity testing Open
! TS-9 6.3.4.2 ECCS subsystem periodic ccuponent Open l testing l
TS-10 6.7 MSIV leakage rate l TS-11 15.2.2 Availability, setpoints, and testing Open of turbine bypass system TS-12 15.6.4 Primary coolant activity II-l 4.2 Fuel rod internal pressure criteria Caplete 6/1/84 tr-2 4.4.4 Stability analysis subnitted before Open second-cycle operation l .
M P84 80/12 23- gs
ATTACHNENT 2 DATE: 8/13/84 r
DRAFT SER SECTIONS AND DATES PROVIDED 8,5CTION DATE SECTION DATE 3.1 3.2.1 11.4.1 3.2.2 11.4.2 5.1 11.5.1 5.2.1 11.5.2 6.5.1 13.1.1 8.1 13.1.2 8.2.1 13.2.1 8.2.2 13.2.2 8.2.3 13.3.1 l 8.2.4 13.3.2 8.3.1 13.3.3 8.3.2 13.3.4 8.4.1 13.4 8.4.2 13.5.1 8.4.3 15.2.3 8.4.5 15.2.4 8.4.6 15.2.5
- 8.4.7 15.2.6 i 8.4.8 15.2.7 l 9.5.2 15.2.8
- 9.5.3 15.7.3 j 9.5.7 17.1 i 9.5.8 17.2 l 10.1 17.3 ,
10.2 17.4 '
10.2.3 10.3.2 10.4.1 10.4.2 10.4.3 ,
10.4.4 11.1.1 11.1.2 11.2.1 11.2.2 11.3.1 11.3.2 CTidb MP 84 95/03 01
DATE: 8/13/84 i
ATTACNMBEL),
OPEN ITEM DSER SECTION SUBJECT i
I 2c 2.3. 3 Accuracles of meteorological t measurements.
1
! 2d 2.3.3 Accuracies of meteorological l measurements.
160 9.5.1.5.b Firewater pump capacity.
176a 14.2 Initial plant test program.
176b 14.2 Initial plant test program. i 176d 14.2 Initial plant test program. i l
176f 14.2 Initial plant test program. !
176h 14.2 Initial plant test program.
186 7.2.2.3 Testability of plant protec- L tion system power. :
226 8.2.2.5 Orid stability.
l 263 11.4.2.e Fire protection for solid ,
l Madweste storage area. l 265 6.8.1.4 BSF Filter system. l
! 266 4.4.1.4 Field leak tests.
t l 247 6.4.1 Control room toxic chemical detectors. ;
.n---- +
t j
)
Attachment 3 (cont'd)
QUESTION NUMBER FSAR SECTION 430.102 9.5.5 430.108 9.5.5 430.110 9.5.5 430.111 9.5.5 430.135 9.5.7 430.149 9.5.8 430.151 10.2 430.169 10.4.4 O
RSC: sal 8/13/84 '
e l
9 O
l I
I i
i I
ATTACHMENT 4 i
l i 1
fe L. -
and d DSEROpenItemNo.2c{(Section2.3.3) . \
!l 4
The meteorological measurements program, during plant operation, will include '
those parameters currently measured. Meteorological parameters are to be available for display through the radiation monitoring system central radia-tion processor (CRP), although the method of display has not been specified.
Calculations of atmospheric transport and diffusion are also to be available through the CRP, although the models and/or methodology have not been described.
Response :
i For theOpen DSER information requested item 3a and b. above, see the response to 1
+
l l
(T .. , .
HCGS DSER Open Item No. 160 ( DSER Se ction 9. 5.1. 5. b )
FIRE WATER PUMP CAPACITY ,
The source of water for the fire protection system is from two 350,000 gal fire water storage tanks. Of the 350,000 gal of storage capacity for each tank, 321,000 gal is dedicated to the fire protection water system, and the remaining amount is available for the fresh water system. Water is pumped by two 650 gpm deep-well water pumps, each of which is capable of filling the fire water portion of either fire water storage tank within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 15 minutes. Operating in parallel, both well pumps can fill the fire water portion of one tank in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The greatest water demand for the fixed fire suppression system has not been specified. The staff will require the applicant to verify that each fire water pump has enough capacity to meet Section C.6.b of BTP CMEB 9.5-1.
RESPONSE
Each fire water pump has enough capacity to meet Section C.6.b of BTP CMEB 9.5-1 as clarified in FSAR Sections 9.5.1.6.19 and 9.5.1.6.21. FSAR Sections 9.5.1.2.3.1 and 9.5.1.2.3.2 have been revised to clarify the fixed fire suppression system demand and the fire water pump capacity. _
l l K51/2-8
p._, ,.. . . . . ... . . .
t<
- \
1/11 ll FIRE PUMP DESIGN FLOW
SUMMARY
CALCULATION +
OB'JECTIVE: To show that the Hope Creek Generating Station Fire Protection System is capable of providing:
- a. 2228 gpm of water at a pressure of 68 psig to the {
largest hydraulic designed sprinkler system, IWS8
- j-t b.
And at the same time provide 'a total of 500 gpm at a minimum pressure of 65 psig to the following hose stations, which could be used to fight a ' fire in the area covered by lWS8: 1H-H109, 1H-HR106, and lH-HR105.
M ETHOD: 1. The hose station pressure was calculated at the interf ace between the standpipe and the hose station.
2 A flow diagram was drawn which showed the sprinkler interf ace and all standpipe-hose station interfaces.
- 3. Node numbers were assigned to the flow diagram.
- 4. By inspection of the piping isometric drawings, it was deteomined which hose station and standpipe had i the longest equivalent length of - pipe. The longest yard piping loop was used.
- 5. The equivalent lengths of pipe were dete rmine d using the criteria of NFPA- 13-19 7 8, Table 7-4. 2.
- 6. The pressure at the 1WA8 sprinkler interf ace was calculated twice: once with no flow to the hose stations and 2228 gpm to IWS8, and once with 750 gpm to the hose stations and 2228 gpm to lWS 8.
This was done to assure that the sprinkler inter-face pressure did not drop below 68 psig.
- 7. The pressure was calculated twice at the standpipe-hose station interf ace: once with 2228 gpm to lWS8 and 0 gpm to the hose station, and once with 2228 gpm to IWS8 and 750 gpm to the hose station.
- 8. The method used for calculating the pressure loss is outlined in NFPA 13-1978 and is based on the l Hazen-Williams formula:
P= (4.521 01.85 c l.85 d4.87 where Q = Flowrate in gpm c = Friction loss coef ficient d = Inside diameter of piping (inches)
P = Pressure loss per foot in psi DSMt OPml IEDt /6 O l 1
l
,-,m,,,,e ,.,_,wn, _,,, ,g m _ -. -.yy -,- g,-.- n
2/2
- 9. Pressure vs. flow was plotted for the sprinkler system and for.the hose station-standpipe interface on log base 1.85 graph paper. From the plot , the hose station flow rate was detemined for a pressure I-of 65 psig. ,
RESULTS: The largest ' sprinkler system is lWS8 at elevation 102'-0" .
in ' the turbine building. With a pressure of 65 psig at the hose station-standpipe interf ace, and 2228 gpm flow-ing to lW58, the maximum calculated fJowrate available i for the hose stations is 500 gpm. TMe calculated X ,
pressure at the sprinkler interface is 76 psig. There fore ,
the fire pumps are capable of delivering, over the
. longest route, dater for tne largest sprinkler system ;
plus a margin of 500 gpm for manual hose streams that can be brought to inace on the same fire.
M
}
+
I DSM OP H ITEM /f, O r i I
- i 9
HOSE STATION DESIGN,'PER NFPA 14
SUMMARY
CALCULATION OBJECTIVE: Show that the Hope Creek Generating Station Fire Protection- System is capable .of providing a 100 gpm _
flowrate to the highest fire hose station in each of i the power block buildings (turbine, diesel / control, reactor and. service radwaste) at a pressure greater than 65 psig.
METHOD: 1. From the fire protection drawings and the-pipe system isometrics, determine the highest hose station in each building with the longest equiva-lent le ngth.
- 2. Assume a flowrate of 100 gpm.
- 3. Determine the pressure at the highest hose station with a flowrate of 100 gpm using the Hazen-Williams formula and the method outlined in NFPA 13-1978.
RESULTS: The results are tabulated below:
l PRESSURE BUILDING ELEVATION HOSE STATION 100 GPM REACTOR 205'-4" l-CHR201 67 psig TURBINE 175'-0" l-CHR104 80 psig DG/ CONTROL 203'-6" l-VHR401 70 psig RAD / SERVICE 17 9'-0" 0-WHR201 80 psig 5
i xs2/2-n ossa opsu IT m /Go
- k. Nuclear Regulatory Commission's Appendix A to BTP APCSB 9.5-1 and 10 CFR 50, including Appendix R.
9.5.1.2.3 Fire Protection Water Supply Systems 9.5.1.2.3.1 Water Source Fire protection water supply is from two, 350,000-gallon, fire water storage tanks located north of the plant. Each tank feeds the FPS and the fresh water system. Of the 350,000 gallons of storage capacity for each tank, 328,000 gallons is dedicated to the fire protection water system, and the remaining amount is available for the fresh water system. Water is pumped by two deep-well water pumps, each of which is capable of filling the fire water portion of either fire water storage tank within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Inserb A x r
The fire pump suction piping and valve arrangement allows either fire pump to take water from either or both fire water storage tanks. With the present arrangement and normal valve line-up, a leak in the pump suction piping could cause the loss of water l from both tanks. However, low water inventory in the storage tanks is annunciated in the control room. Isolation valves have been provided in the storage tank supply headers and in the fire pump suction headers to prevent loss of reserved fire' water from both tanks. This combination of water level instrumentation and isolation valve arrangement provides adequate protection against losing the fire water inventory from one tank and/or both tanks. l 9.5.1.2.3.2 Pumps Two 100%-capacity, UL-listed, horizontal, centrifugal fire pumps are provided in accordance with NFPA 20, each with a rated flow and pressure of 2500 gpm and 125 psig, respectively. One fire .
pump is electric-motor-driven and the other is diesel-engine- !
driven. A jockey pump rated at 50 gpm and 125 psig is also provided to maintain the system pressure between 115 and 125 psig and to provide makeup for system leakage. The two fire pumps and i
the jockey pump are arranged so that each pump can take suction from either tank and pump water into the yard loop system.
Inserb R y The electric-motor-driven fire pump starts automatically at i 110 psig. If it fails to start or cannot meet the water flow l l demand, the diesel-engine-driven fire pump starts automatically l 1 l l
DSER OPEN ITD4 /hd 9.5-16 Amendment 4 l
i
.+
' n.s er-$ A . . . _ . .
- 7-he. dedica.ded Dre wadce a-forag e <~ n;w ei4y o[ 3.18, 000 g a // ens in ca ch. AanM w ://
pro vo de wa.4er fa meed the demand o[
a?a 78 3pm ofihe largesd .sprink/er s y.s f e m p/us .roo gpm for manua/ so.se s fream s
. For a-hours w nseed B -
Eo.ch fire pump is ca-pa b/c c> l pro v'o disy , ove r th e lo ng e s + roud-e o f -dh e i m /er s upply s y.s fe m, 6Ae design demand o f'aaa 8 gym for dhe larges-f sprin Kle r .s y.s k m o,. i the desi}n pressure a f to 8 psig and .s c o g p m for ma n u-a] A o.s e
.stesams. --n add:-//on , eac4 pamp i ., c.apdIc.
of pro v n'dia >y a minimum o f &> 3 p .s i J o.f f/> c h es t s+anop/ spe o u.tled wsWh s oc) g pn h 13 Flowing Sroe the ont/e t in a teorden t c wl% l NPPA 14 See S= clio n 9 5^. l.4. I ? > 9 .5" ! &>. 2 1 an&
Table 9. .S'- /8.
l DSER OPEN ITEM fg 1
l
9 DSER OPEN ITEM 176a (Section 14.2)
INITAL PLANT TEST PROGRAM FSAR FIGURE 14.2-5 , " Test Schedule and Conditions," should be modified to reference the subsection number for the various tests, or the test abstracts should be modified to include the " Test No." as denoted in the figure.
RESPONSE
FSAR Figure 14.2-5 has been revised to reference the subsection number for the various tests as requested above.
I '
I l
l l
1 o
i t
DSER OPEN ITEM 176b (Section 14.2)
INITIAL PLANT TEST PROGRAM The following FSAR Subsection 14.2.12 test abstracts should be modified as stated to provide adequate acceptance criteria:
Test Abstract Modification 1.5.d.1 A reference should be provided for acceptable closing times.
1.7.d.1 A reference should be provided for the design 1.15.d.2 specifications.
1.23.d.4 Reference should be 6.2.5.2.5 d.6 Reference should be 6.2.5.2.3 1.3 f. d . 6 A reference should be provided regarding sage levels of hydrogen buildup 1.41.d.1 A reference should be provided regarding the appropriate accuracy of response 1.47.d.4 A reference should be provided for the prescribed time.
i 1.52.d.2 1.60.d.3 The parameters in these tests should meet or exceed the 1.61.d.1 design values described in their respective references; 1.65.d.2 they should not simply "be comparable" or " compare 1.71.d.2 favorably.
3.24.d.5 1.68.e.1 A reference should be provided regarding the negative pressure specification.
~
Additionally, all startup tests should be modified to specify the appropriate level of acceptance criteria (Level 1, 2, or 3) as defined in FSAR Subsection 14.2.12.2.
RESPONSE i l
-FSAR Section 14.2.12.1 was revised in Amendment 6 to provide the information requested above. g, g,g g In addition, section 14.2.12.3.24.d.5 has been revised to reflect the new GE Test Specifications and all the startup tests in Section 14.2.12.3 have been modified to specify the appropriate level of acceptance criteria.
a- -
- w w - - -eu - - -, . . - - _<---w -,-
Re.v.1_
I HCGS DSER Open Item 176d (Section 14.2) l INITIAL PLANT TEST PROGRAM l
l The response does not address the concerns of IEE Information Notice Number 83-17, March 31,1983. The concern is that if a time delay prevents fuel from being supplied to the diesel generator following a shutdown signal, the air supply may be exhausted before the fuel supply is reinstated. The response to this item should be modified to address these concerns.
RESPONSE
The response to 0640.10 has been revised to provide the information requested abov.e.
I M P84 126/07 2-dh
___m__.___ _ ._ __ _ . _ _ _
r - '-- - - - -
-OUESTION~640.10 (SECTION 14.2.12)
Modify your FSAR submittal to address the following concerns regarding emergency diesel generator testing:
- 1. FSAR Subsections 1.8.1.108 and'14.2.13.5 state that Regulatory Guide 1.108 (Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants) is not applicable to Hope Creek. It is the staff's position that this guide is applicable to your facility.
Therefore, either delete or provide justification for this statement. -
l
- 2. FSAR Subsections 1.8.1.108 and 14.2.13.5 take exception to l Position C.2.a(5) of Regulatory Guide 1.108. These subsections state that testing of'the sequencing controls after the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test run does not subject the controls to i more severe conditions than testing accomplished under other circumstances. Provide technical justification for your position or perform this test in accordance with this guide.
Additionally, modify FSAR Subsection 14.2.12.1.30 (KJ-Emergency Diesel Generators) to perform a restart simulating loss of ac directly after the 24-hour run in accordance with your statement in the aforementioned FSAR subsections.
- 3. Modify FSAR Subsections 14.2.12.1.30 (KJ-Emergency Diesel Generators), 14.2.12.3.30 (Loss of Turbine-Generator and Offsite Power), or other test abstracts as appropriate, to:
- a. Perform the simultaneous, redundant diesel starts i specified in' Position C.2.b of Regulatory Guide 1.108.
- b. Include prerequisite testing to ensure the satisfactory operability of all check valves in the flow path of cooling water for the diesel generators from the intake <
to the discharge (see I&E Bulletin No. 83-03: Check Valve Failures in Raw Water Cooling Systems of Diesel Generators). ,
- c. Provide assurance that any time delays in the diesel generator's restart circuitry will not cause the supply of compressed air used to initially rotate the engine to be consumed in the presence of a safety injection
- signal (see I&E Information Notice Number 83-17, March l .
.31, 1983).
RESPONSE fSU 5*O* 3 /* 3* / /0!
- d A' #
- I' " '
be- -e v s's col a s reg ac sied a.bo ye
""C 0:;2_stery 0;id: 1. ^* i= ::t rgliedl; te ;;s. 7;,1; is %
- ry;;y,y t___
j;;tified er et-ted ir.1;pl::: tetien rectinn n nr n 640.10-1 Amendment 6
\ .- -- _. , - . - _ - _ _ - - . - . . - - - - _ _- -. -_- - - . - -
HCGS FSAR 6/84 Guid i. ige whicn providw. ihet th: ; ide is te be re:d in the # ~
e=s'erti;; ef eubaitt:1: f;r ;;n trer&ian par =i tr . o -_.
Section 14.2.12.1.30.c.6 has been revised to state that a restart simulating loss of ac power will be performed following the 24-hour run.
Upon restart, a sequencing check will not be performed since the 24-hour run test has no effect on the sequencing circuit. The sequencing circuits are located in the emergency load sequencer panels remote from the diesel generator room. The circuits will not have left their standby state since the 24-hour run is accomplished without a loss-of-power or loss-of-coolant accident condition, and is synchronized to the grid. However, the sequencing will be checked during the ECCS integrated initiation during loss-of-offsite power test described in Section 14.2.12.1.47.
Simultaneous redundant diesel starts are accomplished as described in Section 14.2.12.1.47.c.2.
Section 14.2.12.1.30 has been revised to include prerequisite component testing on all diesel generator cooling water check valves.
The diesel generator control design has a time delay relay which .
holds the fuel racks closed to allow the unit to come to a j complete stop. However, in the event of an emergency start signal due to ECCS requirements during the count down of the time )
delay relay, this relay is functionally overridden and the fuel i racks open to allow the diesel to continue to run or restart through the normal starting air sequence described in Section 9.5.6.
l l
9 640.10-2 Amendment 6 1
l
4 l
~
1.8.1.107 Conformance to Reculatory Guide 1.107, Revision 1, February 1977: 00alif:, cations for Cement Groutino for Prestressina Tend ons .n Containment Structures
~
L.
Regulatory Guide 1.'107 is not applicable to HCGS.
1.8.1.108 Conformance to Reculatory Guide 1.108, Revision 1, Auaust 1977: Per:. odic Testina of Diesel Generator Units <
Used as Onsite E:.ectric Power Systems at Nuclear Power Plants Altheegh Regulatery Guide ? ? O" ie uvi applicabl= iv J 'COS, vv6 1_ 1--,____..--
-- - - HCGS complies with with the ,
following exception: gg , ole 1187 Position C.2.a(5) requires that the accident loading sequence to design load requirements be performed directly after the 24-hour run. This does not test the sequencing controls under a more
~
severe condition than if sequentially loaded at an earlier or later period. A restart simulating loss of ac power can be performed directly after the 24-hour run. Sequencing, however, ,
will be performed when the loads can be lined up for operation and all four diesels are available.
1.8.1.109 Conformance to'Reculatory Guide 1.109, Revision 1,
! October 1977: Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluatina Compliance with 10 CFR Part 50, Appendix I HCGS complies with Regulatory Guide 1.109.
For further discussion, see Chapter 15.
1.8.1.110 Conformance to Reculatory Guide 1.110, Revision 0,
, March 1976: Cost-Benefit Analysis for Radwaste Systems For Licht-Water-Cooled Nuclear Power Reactors -
HCGS complies with Regulatory Guide 1.110.
i 1.8-67 Amendment 1
l HCGS FSAR 1/84 calibration completed prior to performing the preoperational test. j l
)
14.2.13.5 SRP II.e, Reaulatory Guide 1.108, Revision 1, August l 1977: Periodic Testina of Diesel Generator Units !
Used as Onsite Electric Power Systems at Nuclear Power i Plants l
..Itherch Daen1=tery Ceide 1.!O* i: not :pplic:ble t: SCCO, per -
i-t: i=vlear2tstien :::ti:n/-HCGS complies withytt, with the following clariff. cations: [ g'cj e 1, j o?
- a. Position C.2.a (5) requires that the accident loading sequegce to design load requirements be performed directly after the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run. This does not test the sequencing controls under a more severe condition than if sequentially loaded at an earlier or later period.
A restart simulating loss of ac can be performed directly after the 24-hour run.
Sequencing, however, will be performed when the loads' -
can be lined up for operation and all four diesels are available.
~ ' -~
14.2-206 Amendment 4
[
l DSER OPEN ITEM 176f (Section 14.2) t INITIAL PLANT TEST PROGRAM (0640.14 Items 1&2)
Commit to ensuring that a neutron count rate of at least 1/2 count per second registers on the startup channels before startup begins and that the signal-to-noise ratio is greater than two, or modify FSAR Subsection 14.2.13-1 to include an exception to Regulatory Guide 1.68, " Initial Test Programs for Water-cooled Nuclear Power Plants," Appendix A, Paragraph 3.
RESPONSE' FSAR Section 14.2.12.3.6.d has been revised to provide the requested information.
G
- - - - --c, - - , - , --,,-r , - - -,wy - - --- - - - - - - ,. , ,- .. .c > v- --s~-, ,- - n
. - ~ _ . . _ . . - . .- .- . -
+
t i w
\
l 1\
. ~DSER OPEN' ITEM 176h (Section 14.2)
, ' INITIAL PLANT TEST PROGRAM
-Q 640.20 i, PREOPERATIONAL TESTS
- 32. .The Reactor Building Ventilation Test (FSAR Subsection 14.2.12.1.68) should state that the RBVS will isolate on a high radiation signal,
.not on LOCA. Additionally, Subsection "e" of this test abstract should be incorporated into subsection "d".
- 37. The Cranes and Hoists Test (FSAR Subsection 14.2.12.1.59.b.2) prerequisite static load tests should be accomplished at 125% of rated
, load in accordance with NUREG-0612.
f
- 40. A test abstract should be provided for the Makeup Domineralizer regardless of whether construction is compic:ted prior to or af ter initial fuel load. .
- 47. Refer to 640.20, Item 37.
POWER TESTS
, 4. A test abstract should be provided which describes the conditions under
) which baseline data for the Loose Parts Monitoring System is obtained.
- 5. No response has been provided.
8.
FSARSubsection9.4.5.l.cstatesthatthemaxgmumdesigntemperaturefor concrete structures within the drywell is 150 F. The acceptance criteria j of.the Penetration Tempe.ature Test (FSAR Subsection 14.2.12.3.37)
.i should p modified accordingly, or justification should be provided for l the 200 F limit.
l l- RESPONSE
- The information requested above for preoperational tests 32, 37, 40 and 47 was provided as part of Amendment 6 to the HCGS FSAR. <
l The response to Question 640.20 has been revised to provide the information requested above for Power Tests 4 and 5.
7--
l l
I l
176h cont'd The response to Power Test No.8 is as follows:
(a) The maximum design temperature for concrete structures within the drywell are specified in Section 3.8.2. Subsection 9.4.5.1 lists the design bases for drywell air cooling system. Therefore Subsection 9.4.5.1.c has been revised to reference the correct section for concrete design temperatures.
(b) Section 14.2.12.3.34 has been revised to reference Section 3.8.2 instead of specifying the maximum design temperature.
o (c) FSAR Section 3.8.2.3.4 states the maximum design temperature of 150 F for concrete structures in drywell for normal operation. Civil-Structural Design Criteria, Appendix A, Secton 3.5 specifies this and includes, "except for local areas which are allowed to have increased temperatures not to exceed 200 F." Concrete surrounding hot piping penetrations is one of these " local areas" (as defined by ACI-349) and is what is tested in Power Ascension Testing.
e
HCGS FSAR 10/33 OUESTION 640.17 (SECTION 14.2.12) 1 Modify FSAR Subsection 14.2.12.3.24 (Relief Valves) to describe or reference any confirmatory in-plant tests of safety-relief valves to be performed in compliance with NUREG-0763 Guidelines for Confirmatory Inplant Tests of Safety-Relief Valve Discharges for BWR Plants."
RESPONSE
Adescriptionoftheconfirmatoryin-planttestsfb=b=prfern'"-
et ugg_g mili h,available hu 3===_.=ry tagg7__
) !
in FsAR S wbsecA-io n 14 212.34O.
1 i,
I l
I l
i 640.17-1 Amendsent 2
~_ _ -- _._. ___-_______ _______._ ___._._. .___.___ _ _ _ _____._._.__ _ ____.__. _ _ _
6/84
) -
- d. Acceptance Criteria j
, 1. _All valves, alarms, controls, interlocks, and logic shall function in accordance with the eyotem ,
e4ectrtcai schemeWes for core spray.
GC Preoperabs onal -res f speciAca- - .
- 2. For the core spray test mode and core spray injection mode, the pump head / flow requirements, i the NPSH requirements, and the system design flow requirements meet the GE preoperational test specification acceptance criteria.
- 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 i 14.2.12.1.8 BF-Control Rod Drive - Hydraulic ) ;
I
- a. Objective .
The test objective is to demonstrate that the control rod drive (CRD) system is fully operational, and that all components, including the hydraulic drive mechanism, manual control system, rod position indicator system, and all safety and control devices, function per design.
5
- b. Prerequisites ,
- 1. All component tests have been completed and approved.
- 2. AC and de power are available.
- 3. All instrumentation has been calibrated and l
instrument loop checks completed. )
14.2-40 Amendment 8--
' ~ -
F .. _ _ _ _ _ _ _ _ _ _ . _ . _ . _ . - . . _ . _ . _ . _ _ _ _ _ _ . _ . _ _ _ _ -
/o$
HCGS FSAR 1/84 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 transient performance. Because overall system performance is a mathematical function of its individual components, one component whose petformance is slightly worse than specified can be accepted 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 I .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 i viewed as highly desirable rather than required to be satisfied.
Good engineering judgement is necessary in the application of these rules, i
During performance of startup tests, technical specifications
- override any test in progress if plant conditions dictate.
14.2.12.3 Startup Test Procedures 14.2.12.3.1 Chemical and Radiochemical Monitors and Sample Systems i
- a. Objectives i 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 2
i
~
14.2-153 Amendment 4 1
--..--------,_-,,--,----,,-,,-,w-,,,.---,,,yme-,- - ,,.,v,,-,_,--,.,,.m mg,y-.--,w,-,- -,,,ww,w,,w,w-p-,,
20([0[
HCGS FSAR 1/84 data required *to demonstrate that fluids meet quality specifications and process requirements 1
- 2. Monitor fuel integrity, operation of filters and domineralizers, condenser tube integrity, operation of the offgas system and steam separator-dryer, and tuning of system monitors.
- b. Prerequisites s
Intrument calibration and preoperational testing of chemical, radiation, and radiochemical monitors have been completed.
- c. Test Method Prior to fuel loading, a complete set of chemical and radiochemical samples are taken to ensure that all sample stations are functioning properly and to ]
I 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 peccess radiation
, monitoring subsystem: 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.
d.- Acceptance Criteria Level 1 : l The chemical and radiochemical, and water of quality factors are maintained within the technical specifications and fuel warranty requirements. Gaseous gediculd e' and liquid effluents' activities shall conform with '
Technical Specifications.
f 14.2-154 Amendment 4
, . ,...-.-c.. - , - . , . ,,-,.----,----,~.--,.--r-wc,- ,_- , , . . . , - -,,--,--,r..----,,m,m_-~_...------m-+w-wy., -
No hh l
HCGS FSAR 4/84 14.2.12.3.2 Radiation Measurements
- a. Objective The test objective is 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.
- b. Prerequisites Prior to fuel loading, a survey of natural background radiation is made at selected locations throughout the plant site.
4
- c. Test Method During reactor heatup and at selected power levels subsequent to fuel loading, gamma dose rates, and where appropriate, neutron dose rate measurements are made at specific locations around the plant including all potentially high radiation areas.
f d. Acceptance Criteria i Lael 12 j 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 L I l
40 &
- b. Prerequisites -
Section.-14.2.10 (initial fuel loading) describes the prerequisites for commencing fuel loading.
- c. Test Procedure
. The~ fuel loading procedure includes any tests performed
, during the fuel loading evolution, including subcriticality checks, shutdown margin verifications, and control rod functional checks.
- d. Acceptance Criteria Lo ve i l '
i The core shall be fully loaded in accordance with established procedures and the core shall be suberitical by at least 0.38% AK/K with the i
analytically determined strongest rod withdrawn.
14.2.12.3.4 Full Core Shutdown Margin
- a. Objective 1
l The test objective is to demonstrate that the reactor 1 will remain subcritical throughout the first fuel cycle l with the most reactive control rod withdrawn.
- b. Prerequisites i
l The core is fully loaded at ambient temperature in the
! xenon-free condition.
- c. Test Method i
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
i ;
- d. Accepi;ance Cri,teria Luci 1; .
The shutdown margin measurements shall verify that the core remains suberitical with.the most reactive control rod withdrawn and all other control rods, fully inserted O ?22Ri . 21,,'3riticality477 %
-g 'y gshould by at least 0.38% AK/K.
dd critical.
occur within tl.0% AK/K of the predicted 14.2.12.3.5 Control Rod Drive System '
- a. Objective e The test objective is to obtain the baseline data for i
the CRD system, and to demonstrate that the system .
1 operates over the full range of primary coolant conditions, from ambient to operating.
- b. Prerequisites Preoperational testing of the CRD system has been 3
completed and the system is ready for operation.
- c. Test Method k
The startup tests performed on the CRD system are an i
extension of the preoperational tests. Initial post i 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 14.1-5. f,, Ad/i/fon prope respon se of & '
CRD plow conhel ehe wilt & y ,,.,f- g ,
- d. Acceptance Criteria i
The insertsrffd' withdrawal' times, scram-t'imes, and l
frictio(Eest results, sihall meet the requirements of -
the Gi startup test' specification.Iimits.
e CRD .e /7 / d '
t, system flow requfrements and f}ow control alve
/
14.2-157 Amendment 4
._ - _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ ._ ~ , . _ _ . - _ _ _ _ _ _ _ - - - - , _ - - - _ , _ _ _ _ . _ _ . _ , _ . . . - _ , . _ _ . . _ _ . , - .
l l Goh{d i l i
I iLevel I*
i The wintv.uaJ soeets a sera 4; to sk B mut % ret"I'"*d.s a4 A SE shkp +esh
! s pe'J.'em b s.
i U.e ve l A :
lThe .Pr;ch rest cesuB.s skal d %
re7utte d s d% GE sh<bp +es+ speciheahns.
! Leve l 3 :
- Tks. CRD sph m 3faw retutre-ed.s a J #l w 5
l con 4.Ivalee. res. pace $$eek & regutveds j et +Ls G E s 4.<Le +e.s + spathca.k1 .
a i
I L
i i
l e
L h
i 1
T
~ ODYL
res s h e GE p)mWI e mee ,
e req cifica .
14.2.12.3.6 - ' Source Range Monitor Performance and Control Rod
! Sequence ,
- a. Objective i
i
- 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. = :t2 of ^
DM '
rrrrter ;;1__ _ _ _ ___ J - _^ _
'__1_f. ;
l b. Prerequisites ,
I Fuel loading is complete, neutron sources have been installed, and all control rods have been inserted. l The CRD system is operational.
- c. Test Method i
With the neutton sources installed, source range monitor count rate data is taken and compared to the required signal count and signal count-to-noise count j d4C3' S.
ratio. Source range data is taken during rad - Se5 withdrawals to the point of criticality.d"During heatup to rated temperature, critical rod patterns are 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 l RSCS which prevents out of sequence movement. -dumho, i et^rrr:1 :" :r' re! :: : tr :::;1:t;: ..L.,, r;;;;
I)cOUDI ;;:---'-- t h e 21 : :t r i ;; ' , _ . , , . ' __ E l _ _ ,
. f f..* :: +
r rrr: .ill ^- .::::' '.
l i d. Acceptance Criteria An d. A y l The neut ignal e o-noise co atto a 15"f mini counts e SRMs shal t the r ements e GE s up test spect ation. 9 M GI+'
14.2-158 Amendment 5
1 ATTACH. A k ve L 1_: _
_Then._.wJ ba__m.._wh sty l_and -4. - mise
_ _ - . ._. CO M radis aE_.M_. l easy _.h a M_et ___
__ e aitmwm_. aendrea-.. cout k_ al_% coui/sec R
. _ e &_ . ap ired . o ge ra ble_. S R M s . __... . __.._
m wh g.h wwm e-me membaMeN- *M=* 6 6*
e em . _ey. e A Trn GM.
B.
... . ._ d_ f . O _b D. _ . #. .[ _.N I _ .b 4[ MO . ~
. . . a per-iod y caker % Bo _ .se conds.
- e. . . - .*- .-.=m. -- . - - . . . - . .. - = ...- -
., - - . e ---
ae= =m.-- e me e =--wm* m_ --w we .
e ep-- m 4-- .. e & *=
e
-- p y-asm.m =- ww.
a mee- -e- e
_ w-m eme ma ,.6se -mm-. On -w*e e
.Ah . e ems.. a ee--a, m.
.e we g gy -em. .
I
.m. .-mi .- . . e. W .beim. eke-
-+ . m *D . -
-g- . & -e.
-- -- -4e- -
-.w .my g m - e . . w-.m>- p_, e. -a e =+6 4-m e 6 e . ae
-e .g e
_p . g . -
.w.
,-.es. - me - e.. -, . .-. e e---
4.
l
. .__ . _ .__..._. . . __ _. . 4 . . _ . _ . .
1
--. .. - l L .. .. - - . . - - - - - . --- - . - .- -- ..-. - - ... . . _ . . - . . . . . . ...
l
00E9h l
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 ,
- a. Objective .
The test objective is to determine IRM system response to neutron flux and to optimize the IRM overlap with the SRMs and APRMs.
- b. Prerequisites The reactor is critical and the TWM gains have been set 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,
- d. Acceptance Criteria 12.. e i IL Each IRM channel must be on scale before the SRMs exceed their rod block setpoint. Each APRM must be on scale before the IRMs exceed their rod block.setpoint.
LM .44aa,diach IRM should be adjusted for half decade overlap with SRMs and one decade overlap with APRMs.
14.2-159' Amendment 5
.,-- -,--~~ ---,--,, - - - - - - . .,-s--+ . - . - . . , - - - , - , - , , - , - . - + - - - - - - - - - , - - , , , - -
IO[{ph l
l l
HCGS FSAR 4/84 14.2.12.3.9 Local Power, Range Monitor Calibration
- a. Objective 1
i The test objective is to calibrate the LPRM. l 1
- 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 -j procedure.
- d. Acceptance Criteria ,
Lcvei l '-
Each LPRM reading should be within 10% of its calculated value.
1 i
l l
14.2-160 Amendment 5
.w--- y . i-,,-,..--_,_y--,,, -, , _ , . _ , ,w-,.--.-,,w-y#---, ,,.,,--w---,,,-,-,,--,--,,...ey -- m ,- -,4
// 4-Gd HCGS FSAR 4/84 14.2.12.3.10 Average Power Range Monitor Calibration
- a. Objective The test objective is to calibrate the APRM'. l
- 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 APRM 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 Luai d.;
- 1. The APRM channels must be calibrated to read equal to or greater than the actual core thermal power.
- 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 equir:d L, in. Gi. sr.arv.up ww=t
- - - - ' ' ' - -
- i a a o r -&4. m i ni mum v aAut ce pire A bas *4 on TPP , M t-H6R. , ee -fradt.n c4 rafee pwar b wi4ki 6 Ma. IT mi4 s specIEec0 's a 4de. ;
$C shy fug desh specific,dia w . ,
r 14.2-161 Amendment 5 i
1
0
. HCGS FSAR 4/84 14.2.12.3.11 NSSS Process, Computer
- a. Objective The test' objective is to verify the performance of the
. process computer under plant operating conditions. l
- 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 vari. ables sensed by the computer j become available. The validity of these variables is -
verified and the results of performance calculations of the NSSS and the balance-of-plant (BOP) are checked for accuracy.
- d. Acceptance Criteria Le ve l d '-
- 1. The process computer performance codes calculating 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 i
14.2-162 Amendment 5 t
- _ . . , _ . . - _ .-_ _.. _ . . , . . _ - , . . ~ _ . _ , . _ . . _ . _ , . _ _ _ - - . _ . . . _ , _ _ _ . . - . - - ~ - . . _ . _ _ _ - , - . - - .
I$rl(t&
- 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.
- 2. s < 2 . J,... .a , , d --- A --a
) nn! A T!! 7 .i X I,~,,.,-. 74,'i E - Q- v3'3'4he sw_ _ . .: p .ox wi . ..: x- =. --- = x. 1. g - ~
. f 14.2.12.3.12 Reactor Core Isolation Cooling System
- a. Objective 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
- 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 i
l
)
14.2-163 Amendment 5 l
hh
. HCGS FSAR 4/84 above the reactor. pressure to simulate the largest
! expected pipeline' pressure drop. This CST testing is done to demonstrate general system operability and for making 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 high steam flow isolation trip setpoint while' injecting 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.
Two consecutive reactor vessel injections starting from cold conditions in the automatic mode must satisfactorily be performed to demonstrate system i reliability. Following these tests, a set of CST l injections are done to provide a benchmark for
! comparir Ja 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 flow 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 l conditions to span the RCIC operating range.
A demonstration of extended operation of up to two
- hours (or until pump and turbine oil temperature is i stabilized) of continuous running at rated flow
- conditions is to be scheduled at a convenient time i during the startup test program, i
Depressing the manual initiation pushbutton is defined as automatic starting or automatic initiation of the RCIC system.
l
- d. Acceptance Criteria Lc fel L_*.
- 1. 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.
I I
14.2-164 Amendment 5
b d
. I
- 2. The RCIC turbine shall not trip or isolate during automatic or manual start tests. ,
Le ve l 2 *,
{.l. The turbine gland seal system is capable of preventing steam leakage to the environment.
g.2 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
- t. margin, the transient start speed peaks must not 4 exceed the requirements of the GE startup test specification.
p.M. The speed and flow control loops are adjusted to meet the decay ratio specified in the GE startup test specification.
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.
- b. Prerequisite
! Fuel loading has been completed and sufficient nuclear.
I heat is available to operate the HPCI' pump.
Instrumentation has been installed and calibrated.
I c. Test Method 1 The HPCI system is designed to be tested in two ways:
l
- 1. By flow injection into a test line leading to the l condensate storage tank (CST), and
- 14.2-165 Amendment 5 l
,, , - . . - - . . . . - . . _- . ~ - ~ - - . - . . . , . _ , , . . . , , - - . , . - - - . - , - , , , , . - - . , - , . . . . - . . .- ,. - . . - - , . - , . . , - - . . - -
l Y 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 making 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 te demonstrate system reliability. Following these tests, a set of CST 1
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 i
operating, small step disturbances in speed and flow ,
command are input (in manual and automatic modes ;
respectively) in order to demonstrate satisfactory j
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 l
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.
4
, 14.2-166 Amendment 5
-, ..n.. ,- , . , , . , . . . . - . , , ~ . - - , , _ - ~ . , , - . . , , , , _.n . - - - . _ . - . . _ - - , , - - - - - _ , - . . nn-.
- d. Acceptance Criteria Leve l 1 * -
t 1. Following automatic initiation, the pump discharge i flow must be equal to or greater than the rated l
_ 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.
14 vei 2 '.
J.,, 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.
p(?, 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.
4 14.2.12.3.14 Selected Process and Water Level Reference Leg Temperatures
- a. Objectives
- 1. To establish low speed limits for the
! recirculation pumps to avoid coolant temperature i stratification in the reactor pressure vessel l (RPV) bottom head region l
l l 2. To ensure that the measured bottom head drain temperature corresponds to bottom head coolant temperature during normal operation.
I
- 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 l
l
d
- b. Prerequisites .
The. plant is in a hot standby condition. System and test instrumentation have been installed.
- c. Test Method I During initial heatup at hot standby conditions, the bottom drain line temperature and applicable reactor
' parameters are monitored as the recirculation pump speed is slowly lowered to determine the proper setting
, of the low speed limiter. The parameters above are also monitored during planned recirculation pump trips to determine if temperature stratification occurs in the idle loop (s) and to assure that idle loop-to-bulk 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.
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 }
calibration. Recalibration will be performed if -
necessary.
- d. Acceptance Criteria
- Le ve l d '
i
- 1. The reactor recirculation pumps shall not be started unless the loop to loop delta-temperatures 1
and steam dome to bottom drain delta-temperatures are within the technical specification limits.
4 Le vel Z '.
) .f, During two pump operation at 100% core flow, the j difference between the bottom drain line j thermocouple and recirculation loop thermocouple i is within the delta-temperature required in the GE startup test specification.
l 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.
i f
14.2-168 Amendment 5
,-- ,---nm.-c . ,. m.,..--ww-.wn,-,,,--,,,,--,a r.-.,,n-e.n-----,e. .w, .,n-,---r,-..~----,m. -
-.w,~n--an~,w,---,n-e
A HCGS FSAR 4/84 14.2.12 4 15 System Espansion
, 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.
1 i -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 -
- During heatup, observations and recordings of the-i 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.
I
{ d. Acceptance Criteria ,
1 L3tet "1
- 1. There shall be no evidence of blocking of the i
displacement of any system component caused by thermal expansion of the system.
- 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 j temperature.
i
- 4. The piping displacements at the established transducer locations'shall not exceed the limits 14.2-169 Amendment 5
..,.,,---w~, - - , , , - , , - , ,. . , . . - , - - -, , , - . - - - - - , , . .,,.-,-v,,-w,---.-.--,w.w---,-- w w-,,m ,,- . ,w..--~ ,,c ,.. ,y-.-.-- ,..- - , - - -,
i
specified by the piping designer, which are based on not exceeding ASME Section III Code stress
. values. These specified displacements will be 4
ysed as acceptance criteria in the appropriate startup test procedures.
C;.. F- .c utstrinution 14.2.12.3.16 TlP theerb k
- a. Objective The test objective is to demonstrate the j reproducibility of the TIP system readings.
- b. Prerequisites The core is at steady-state power level with equilibrium xenon, so as to require no rod motion or change in core flow to maintain power level during data acquisition by the TIP system.
- c. Test Method
- 1. Core power distribution da.A are obtained during the power ascension test program. Axial power i distribution data are obtained at each TIP
. , location. At intermediate and higher power ,
levels, several sets of TIP data are obtained to j determine the overall TIP uncertainty.
- 2. TIP data are obtained with the reactor operating i with a symmetric cod pattern and at steady-state
, conditions. The total TIP uncertainty for the l test is calculated by averaging the total TIP t
uncertainty determined from each set of TIP data. ;
The TIP uncertainty is made up of random noise and '
- geometric components.
i l
14.2-170 Amendment 5 e
_ . . . . - . . _ - . _ . _ , _ _ . . _ . _ - ~ _ . _ _ . - - _ . . - - . _ _ . , , . - ~ . , - , . . _ _ _ . - _ - - . , , . -
< i
4/84
- 3. Core power symmetry is also calculated using the TIP data:' Any asymmetry, as determined from the analysis, will be accounted for in the calculations for MCPR.
- d. Acceptance Criteria tevel Z :
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.
f
- b. Prerequisites j
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. MAPLHGR.
I 14.2-171 Amendment 5
5 bGl
- d. Acceptance Criteria Levai d_ '-
Core f. low rate, core thermal power level, MLHGR, MCPR, and MAPLHGR not exceed the limits specified by the plant technical specifications.
14.2.12.3.18 Warranty Test
- a. Objective The test objective is to demonstrate the reliability of the NSSS and to measure the steam production rate and plant heat rate.
- b. Prerequisite 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 rated conditions. During the 100-hour run, the steam production rate and plant heat rate is measured.
- d. Acceptance Criteria levei 1:
The reliability of the NSSS and the ability of the NSSS to develop rated output shall be demonstrated to be within warranty specifications.
Core Power - Void Mode 14.2.12.3.19
- a. Objective The objective of this test is to measure the stability of the core power void dynamic response, and to 14.2-172 Amendment 5 l
,-e---.-.y,. --m-,- -, - - - , . - - - -. - - - - , . , - - . - . - - - - - - , . - - - - - - - - - - - * ~ - . -
r-
N
- l l
HCGS FSAR 4/84 demonstrate that its behavior is within specified design limits. -
- b. Prprequisites 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
. ley 2l 1(
The transient response of any system-related variables
- to any test input must not diverge. System related variables are heat flux and reactor pressure.
1 14.2.12.3.20 Pressure Regulator l
- a. Objectives i
I
- 1. To determine optimum pressure regulator setting to
! control transients induced in the reactor pressure l control system.
- 2. To demonstrate the takeover capability of the t
backup pressure regulator via simulated failure of the controlling pressure regulator and to set the regulating pressure difference between the two j regulators and an appropriate value.
i i l .
14.2-173 Amendment 5 i
~ - .- - -.-- --- _ .- -- - -.__ - --
- 3. To demonstrate smooth pressure control transition i between the tirbine control valves and bypass
! valves.
- b. Prerequisites Instrumentation has been checked and calibrated. The plant is at a steady-state power level, i
- c. Test Method i 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 l operating pressure regulator. The bypass valve is
! tested by reducing the load 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
- power - void mode test report.
- d. Acceptance Criteria l Level I'
- 1. The transient response of any pressure control
- cystem related variable to any test input must not
- diverge.
! Levei J2 -
l y W Ere recirculation manual mode the response time from initiation of pressure setpoint change to the i
turbine inlet pressure peak should be less than I that specified in the GE startup test ,
specification. !
l J /t Pressure control system deadband should be small l enough that steady state limit cycles shall
! produce steam flow variations no greater than specified in the GE startup test specification.
l f. D*For all pressure regulator transients the peak
! neutron flux / peak vessel pressure should remain below the scram settings by the margins specified in the GE startup test specification.
14.2-174 Amendment 5
N '
HCGS FSAR 4)S4 5'.4,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.
Lede1 73' ~
F.l. Contro1 or bypass valve motion responds to pressure input with deadband no greater than that required in the GE startup test specification.
- 7. ;;r--4-- af han ; ::: = mv.':t::: --- -4=i'- -
62th 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 and to verify that the maximum feedwater flow during pump runout does not exceed the flows assumed in Section 15.1.2.
- 4. To demonstrate response to feedwater temperature loss
- 5. To demonstrate acceptable reactor water level control.
- b. Prerequisite Instrumentation has been checked and calibrated as appropriate. The plant is operating at steady-state conditions.
14.2-175 Amendment 5 I
- c. Test Method
- 1. Reactor water level setpoint changes of several inches are used to evaluate and adjust the feedwater control system (FCS) settings for'all power and feedwater pump modes. The level setpoint change also demonstrates core stability to subcooling changes.
. 2. From near 100% power, one of the operating j feedwater pumps is tripped. The automatic 1 recirculation runback circuit will reduce
. 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.
I
- 3. The condensate /feedwater system will be subjected 1
to a loss of feedwater heating. The initial power level will be approximately 80% prior to the start of the test. It is expected that the feedwater temperature decrease will be less than 1000F.
4 j 4. Feedwater pumps and turbine parameters are i monitored during the power ascension to i demonstrate operability within specifications.
i This test includes initial calibration of the
- speed controllers, and verification that maximum
l f d. Acceptance Criteria L.c v c l A *'
i 1. The transient response of any level control system related variable must not diverge. ,
L e ve l 2 .'
y.j
- Level control system oscillatory modes of respons3, open loop dynamic response, response to step disturbances, and steady state operation shall meet the requirements specified in the GE startup test specification.
14.2-176 Amendment 5 i-I j
h2 DV
.p \ J .7. For feedwater heater loss, the maximum feedwater
\ \
.p temperature' decrease due to single failure is less
'J & than that specified in the GE startup test Q' ,V specification, and the resultant ~MCPR rust be
\ ~ greater than the fuel thermal safety limit specified in the FSAR.
&['Y :
q f _1 f.1 On the trip of one feedwater pump, the reactor t shall avoid low water level scram by the margin specified by the GE startup test specification.
D4 3d I' ) .3 Maximum speed attained shall deliver flows l h ? [2 consistent with the requirements specified by the jk,f GE startup test specification limits. l 14.2.12.3.22 Turbine Valve Surveillance
- a. Objective I
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, t
- b. Prerequisite
- The plant has been stabilized at the required power level.
- c. Test Method Individual main stop, control, and bypass valves are <
j 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 i this test are determined. The rate of valve stroking and timing of the closed-open sequence are chosen to
, minimize the disturbance introduced. ,
l l
14.2-177 Amtndment 5 l
- d. Acceptance Criteria, Leve l ' 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
- 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
j b. Prerequisites
- The plant has been stabilized at the required power level.
- c. Test Method ;
~i
- 1. Individual closure of each MSIV is performed at l 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 l MSIVs is performed at about 100% power. Operation
- of the RCIC system and the relief valves is i 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 )
1 14.2-178 Amendment 5 l
- j
h l
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 Leve l 1 : 1
startup test specification.
Level 2 :
- 7. /, Peak neutron flux, vessel pressure, and steam flow l, shall remain below scram or isolation trip settings by a margin consistent with design requirements when individually testing the MSIVs.
+
J.1. Following the full closure of all MSIVs, vessel pressure and heat flux level shall be as specified
! .in the GE startup test specification.
, , .3 v 5- f
{ g
\ ,
/.2. The RCIC system and relief valves shall function 3' y. ,. in accordance with the GE startup test p 1. ' ';.<' ~3 specification following the MSIV closure from high power.
-- > I.3 The reactor must immediately scram and the C"{e.hA' sg feedwater control system must prevent the water g q .4 0 i 4
, from reaching the main steam lines following full
" closure of MSIVs from high power.
1 14.2.12.3.24 Relief Valves i
- a. Objectives 1
1
- 1. To demonstrate proper operation of the main steam relief valves and determine their capacity
- 2. To demonstrate their leaktightness following operation.
t 14.2-179 Amendment 5
---r-.r- -
--,,-,vr_,-. -..-,. ,wr - . . vwe --~,w ~.v,- -y---wy-,..-- ,m--
l l
l
- b. Prerequisites ,,
l The reactor is on pressure control with adequate bypass or malt steam flow.
- c. Test Method i
A functional test of each safety relief valve (SRV) !
shall be made as early in the startup program as l 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.
4 The tailpipe sensor responses will be used to detect i 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, i
i Valve capacity will be based on certification by ASME I code stamp and the applicable documentation being ,
i available in the onsite records. Note that the l
nameplate capacity / pressure rating assumes that the flow is sonic. This will be true if the back pressure ,
l 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.
i !
I Vendor bench test data of the SRV opening responses will be available onsite for comparison with 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.
i SRV opening and reclosure setpoint data will be obtained and evaluated during each high power trip test at which an SRV actuation is anticipated.
I l
14.2-180 Amendment 5
- d. Acceptance Criteria Lc ve i .l' ~
- 1. There should be positive indication of steam
, discharge during the manual actuation of each valve.
Level Z'
- 2. /, Decay ratio for pressure control varisoles is as specified in the GE startup test specification.
f.2.The temperature 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.
MG Tii. . . Jw s. vnpadsid'5.5Asnv s y. cii.y swiiiv.6 .. -
f} hk' " " {5. favorably-with Section 5.2.2 and the accident analysis.
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.
- b. Prerequisites 1
Power testing has been completed to the extent necessary for performing this test. The plant is stabilized at the required power level.
i 14.2-181 Amendment 5
\
l
. _ _ . . _ _ . . _ . . _ _ . _ . . . ._ . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . _ , _ _ _ , . , _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ ~ . _ . . . _ . . , , - _ . ~ , . - .
t i
e l1. 44. <aaa passn -as+ s s4.-
Dart ,%,.4 a <elt 4udve , a.s
% meauraa by M W e., aAu+ he \w < 4% %.
p wevay. vaIve anpsse hg w ra %
4L4 specihaa t. % QE sk4up 4, +
r s pe h areas, n
i s
P
- n I
n i
e
. l e o e
- c. Test Method l
The t ine is te d at the iffere power els the ghout the wer ascen n pro . For e
[yP6 bine tr , the main erator eakers main 1 ed
<5 for a et so there no ris n turbt gener e i dSC speed whereas, i he gen tor tri the ma ge ator brea s open d residu turbin stea ill use a mom ary ris n the ge ator ed.
4 At test condition 3, a turbine trip will be initiated manually from the control room. At test condition 6, a generator trip (load rejection) will be initiated by simulating a condition that will cause the generator otobf*C ~
breakers to open. During both transients it is expected that the~ reactor will scra J tt is not M I M '/
expected the NPCI or RCIC will initiate. Reactor water ftfE E/2/
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 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.
- d. Acceptance Criteria le ve l l ' '
- 1. For turbine and greater than 50%, generator the response tripstimes at power of stop, levels control, and bypass valves shall be as specified in the GE startup test specification.
J
- 2. Feedwater control system settings must prevent floading 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
. , . , _ _ . . . . , -_ . - _ - . , . , . _ , , . _ . - - . . , - - ._._-..___._m. .c..... - . - . .
r b//[8h INSERT $ 3 This test is performed at three dif ferent power levels in the power ascension program. For the turbine trip, the main gen 9tator remains loaded for a time so there is no rise in turbine generator speed, whereas, in the generator trip, the main generator output breakers 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.
f
~
l .
i e
l l
l l
e
. i
[d l
- 4. The pewitive change in vessel dome pressure and heat faut 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 vel 2
- 6.; The measured bypass valve capacity shall be equal to or greater than that required by the GE startup test specification, which compares bypass valve capacity to the accident analysis.
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.
A.3, 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, j
F//. Low water level recirculation pump trip, HPCI and RCIC shall not be initiated.
kC.jf' Feedwater level control shall avoid loss of feedwater due to high level trip during the event.
The temperature measured by thermocouples on the It'.6, 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.
14.2-183 Amendment 5
-1 l
l HCGS FSAR 4/84 4 14.2.12.3.26 Shutdown From Outside the Main Control' Room
- a. Objective 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 shutdown 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 will consist of demonstrating the capability to scram and initiate controlled cooling from outside the control room. The reactor will be scrammed end 4eektod from outside the control room af ter a simulated control room evacuation. Reactor pressure and water level will be controlled using SRVs, RCIC, j and RHR from outside the control room during subsequent cooldown. The cooldown will continue until RHR shutdown cooling mode is placed in service from outside the control room. Alternatively, verification of 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.
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 RRR 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 maintained in hot standby condition for at least 30 minutes during the performance of this test.
- d. Acceptance Criteria l
_ Ley e l l '
i During a simulated main control room evacuation, the I
abil'ity to bring the reactor to hot standby and subsequently cool down the plant and control vessel 1
14.2-184 Amendment 5
'd h!
HCGS FSAR 4/84 pressure and water level shall be demonstrated using equipment and cdntrols located outside the main control room.
14.2.12.3.27 Recirculation Flow Control
- a. Objectives
- 1. To determine plant response to changes in the recirculation flow
- 2. To optimize the setting of the master flow controller
- 3. To demonstrate plant loading capability.
- b. Prerequisites The reactor is operating at steady-state conditions at the required power level.
- 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 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.
j When the plant is tested along the 100% load line, the recirculation system shall be tested by inserting small 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.
During recirculation flow control testing at the 50%
and 100% load lines no scrams due to neutron flux or heat flux changes transients are expected.
l 14.2-185 Amendment 5 l l
- d. Acceptance Criteria, Lcycl $.?
- 1. The transient response to any recirculation system related variable to any test input must not diverge.
Leve( 7 '
74, 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.
7.Z The decay ratio of any oscillatory controlled variable must be less than that required by the GE startup test specification.
e ?. 01; .J ...d ep;.. ;;;:* la^7 Mjr-tre.t: cro-as s pec i f i ed-t rr the' GEvaFEUp~ Tbs t' spe'cif f c = i. i chs r.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.
8'.d. 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 Recirculation System
- a. Objectives l 1. To determine transient responses and steady-state conditions following recirculation pump trips at selected power levels
- 2. To obtain recirculation system performance data 14.2-186 Amendment 5 y - . , .
- - - - , ,..-,..-....y _.,,....,----r -
.-..,_----.--,,e-- ,e,
- 3. To verify that cavitation in the recirculation ,
system does' not occur in the operating region of l the power / flow map.
- 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.
l
, 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 I
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.
- p. 14.2-187 Amendment 5 1
HCGS FSAR 4/84 At several power and flow conditions, and in conjunction with sihgle pump trip recoveries, recirculation system parameters are recorded.
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 2'
- 1. During recovery from one pump-trip, the reactor shall not scram.
/ L wl 2 '
/
>/' 2. Neutron flux, heat flux, and reactor water level scram avoidance margins are as specified in the GE 7
startup test specification.
4 c J2 The two pump drive flow coastdown time following a dual recirculation pump trip is as specified in ;
the GE startup test specification, j t.7 System performance parameters, including core
- 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.
l 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 of the runback circuit as required by the GE startup test specification, l
14.2-188 Amendment 5
.,...,_._.--,,--,-,_.-,-,..,-,.--,.-,-m.. , , ,
ofGG HCGS FSAR 4/84 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 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,
- d. Acceptance Criteria
, level 2
- 1. Jet pump flow instrumentation shall be adjusted such that the jet pump total flow recorder-provides core flow at rated conditions.
I
- 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.
3 4etM '
1__
14.2-189 Amendment 5 l
.-- , , - , - - . _ . .-.-,..,-.,.n., ,.,-m_.,.. .._,-,-..e.,,.,. n- - , , ,,._,,n.w. va, y,. , - , . , . ,,-
l i
i NZ' hh es -- : !
The flow control system shall be adjusted to limit maximum core l flow to the value specified by the GE startup test specification.
t
-e m
.l 1
i l
I s _ _ _ _ _ - - - - - -
14.2.12.3.30 Loss of_ Turbine-Generator and Offsite Power
- 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.
- b. Prerequisites The SDGs are in the auto-start mode, and the plant is operating at power.
- c. Test Method With the power plant synchronized to the grid between 20% and 30% power, the main turbine generator will be tripped followed by manual trips of all offsite power l to the 13.8 kV ring bus. This will simulate loss of turbine generator and offsite power.
Reactor water level and the operation of safety systems, including RPS, standby diesels, RCIC, and l HPCI, will be monitored.
The loss of offs'ite power condition will be maintained for at least 30 minutes to demonstrate that necessary equipment, controls, and indication are available following the station blackout to remove decay hect from the core using only emergency power supplies and distribution systems.
- d. Acceptance Criteria l Lcvel {:
- 1. All safety systems, such as the RPS, SDG, RCIC, l and HPCI, function per design without manual assistance. Reactor parameters are maintained within acceptable design limits. Normal reactor cooling systems maintain adequate suppression pool water temperature, adequate drywell cooling, and 14.2-190 Amendment 5 l
[
HCGS FSAR 4/84 prevent actuation of the automatic depressurization system.
Leve2_2.*
2;f . Proper instrument display to the reactor operator shall be demonstrated, including power monitors, pressure, water level, control rod position, suppression pool temperature, and reactor cooling system status.
J.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 closed after valve closure.
I 14.2.12.3.31 2:7;;11 Piping Vibration Tes-b5
- 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 j restrained properly. Instrumentation for monitoring .
vibration has been installed and calibrated, wnere !
applicable.
l c. Test Method This test is an extension of the preoperational test program. During steady state operation, designated pipes as delineated in Section 3.9.2 will be monitored for vibration. Dynamic vibration measurements will be !
made on applicable piping following various plant and system transients as specified in Sections 3.9.2.1.2.3, ,
3.9.2.1.3, and 3.9.2.2.4. I 14.2-191 Amendment 5
- d. Acceptance Criterip Lev s( .1 '
The piping displacements at the established locations.
shall:not 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.
14.2.12.3.32 Reactor Water Cleanup System
- 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 l
With the reactor at rated temperature and pressure, process variables are recorded during steady-state operation in three modes .of operation of the RWCU system: blowdown, hot standby, and normal. The bottom head. drain flow indicator will be calibrated by taking flow from the bottom drain only and using the RWCU system inlet flow indicator as a standard to compare j against.
l
- d. Acceptance Criteria l l
tevez z.-
- 1. The data indicating operation in the listed modes 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, a
- 3. . Pump vibration as measured on the bearing housing and coupling end shall be less than or equal to GE startup test specifications.
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 4
- 2. To condense steam while the reactor is isolated from the main condenser, in conjunction with the RCIC system.
- 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 i normal operating pressure and temperature.
- 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 transfer rate is controlled to maintain acceptable i
cooldown rates. Data are recorded and reviewed to verify the satisfactory operation of the RHR system within design limits.
i 14.2-193 Amendment 5
r:
47g46
- d. Acceptance Criteria, L a o di 2 .'
- 1. The RHR system performance in the steam condensing i mode, suppression pool cooling' mode and shutdown J cooling mode meets the requirements of-the GI '
startup test specification, utsk u.Tmd ,
14.2.12.3.34 Drywel4 Cooling .,, m..
)
- a. Objective 1
The test objective is to demonstrate, under actual operating d gu:11 :t erP conditions, satisfactory
=-i< raniin; y;te; [erfer e,f A rgce rf dMthe coo /l I i
$6eewe+unsi, inslading Constufe. surrouast sg swf p~p, ,, y tvsls'oas,
- b. Frerequisites W ~
- comple+eQ Approptake preopd**A +esh ,
D ,z e" Airfih 5;1:ncing of O.e,,;,st;; '.2:~ sy:r ;c.,;picted.
Power ascension testing is in progress. l Represa We peach =4tm have beent%s b e*4,4.
- c. Test Method M iua.el M h psm e d a=
]
Drywell atmosphericatemperatures are monitored and
- recorded during plant heatup and power operation up to
, -f' rated power. Ds9msen. t -- ,-- w,A rce 2 ;;ns tt-.t : f to be--at-e- hal ow tha A-s i r'" 1*mi*?- Adpri=ar** t e r i.r
+1e.;s end,';; :::lin; " ster f1 :: : e sede, if ::q.:ireb pg h eiania sc::pt:bic t;;p:::tuce limits.
- d. Acceptance Criteria L%d. 1 : J s4em W almosf:4,v e'c. l 1,Drywel rature control shall meet or exceed the limits specified in the plant technical specifications.
i
/
[ M 4
14.2-194 Amendment 5
, %d Anosh%
In addition, drywell atmospheric, and hot piping penetration con-c'ete temperatures are checked at various power levels, rated, with minimum- drywell cooling capacity in service.
up to Design YT ((
temperature limits are verified to be met, and cooling system adjustments are made as required to maintain acceptable tempera-tures.
l 42:
% % une rete. 49erxkm su rroQ pping pere % A,g.,a p._su n+ ecceee A akmWe locaR area Iin.ib b aw,=2 ogr3 s pe c t-Cich 7 Seeb a,s.a.
I e
r ---,,----------.,--,,-,n,--- . ,-,--.. . _ _ _ , ~ , _ , , - , , , , , . , _ - - , - - , - - , , - . . - - , - . , - . . _ , , . , , - - , - , , - , , , .
&aseous Rac0maOL 14.2.12.3.35 Of f;;; Ts,,eatasat System
- a. Objective The test objective is to demonstrate proper operation of the eff;;; Lu. mwau system over its expected operating range.
be secas Pn Acas$ e
- b. Prerequisites Initial calibration of instrumentation has been completed. Power ascension testing is in progress,
- c. Test Method t
c ,_ __ -- ,
- r> n w a s Lt~e mor4-- - - t _ , ,, , ; _ - - - . - ;; ;;--. ,.nnaer- ,,2; 6 ie --------A '- : eff
. f _.
. Dee,cg pc ke = c--eica testir.;.[:t steady-st:tb_c_e_;=di
- 4 aae
/kb system paramwLocs ef LL, effges tieetzent ey-*== =re monite::d end m ded fc c e. ale-me., sr! synem IEI -;; r f ; _ ;;. . :: . Adjustments will be made, if necessary, to meet accepte.ble system performance.
- d. Acceptance Criteria ne I ,
_[ etoelQ*
-(g*6+i#w='=2 h - @ t'.t,s-g h v- - oi.g.r t 'E: -ZII;;r.t it;;
th: _f:; r c r r il_ - . -r a -- Y -I ::::f '__-_2 s; ci .
J..- ...:...;- ; ovosificcti:.:. System performance as verified by data analysis shall meet design requirements specified in Section 11.3.I,8 g M k, 14.2.12.3.36 Water Level Measurement l This test was included in Section 14.2.12.3.14. l 14.2-195 Amendment 5 / ;
1 i
. - - . . , - . - . . . , - - . . - _ , , , . , ---.-,,---.--.,c
'i i
5""[GG
$\
D arla3 poww ascersin lesl:q af s4ea}y -sMe e eJ;b r , b * *'* " 5 '"#" * * # '
- P #'
- P'"d'*' '
- A lJa+, k a y s le ~ Claw, p re ss u r< , heuperah r e., .
aaD dewpa un rece a a.
I l
i 4
i r
i, e
i li
- i. . ._ _. . - _ . . . - _ , . . . . _ . . , . . . , - . . . _
51 % l l
. HCGS FSAR 4/84 14.2.12.3.37 Penetration Temperature Test t.Sf Yis he s h t,o ns l a c luc0 c8 N beclio n 14'2.12.3.
- l
. Objectiye To varify that the devwell penetratinne ===nciated with o
hoj_.pi.pi.y .y=6. provide -adequate ~ protection ror the~
surs;;r.dir.g canerera 4
- b. "re s eyuis i Lwas
- 1. Power ascension testi.ng -is-in progress.
- 2. Yner" er_tstier 1: ::libret=_d m
- c. Te.i. i-iwLiwd a wusing nessup anu wwer operations, tne conucate temperatures surrounding hot penetrations will be l m ontcored. _
JL_ Acceriance Criteria m 4
1 ne he- cenew> - ; __ _.i. .. TJnMning ncirpiping pe g ns shall not exceed 2000F.
14.2.12.3.38 Safety Auxiliaries Cooling System l
- a. Objective The test objective is to demonstrate that the safety auxiliaries cooling system (SACS) performance margin is adequate to support engineered safety features equipment over their full range of design requirements.
14.2-196 Amendment 5 J
82 Q
- b. Prerequisites.
Initial instrument calibrations have been completed.
The plant is operating at the required test condition.
- c. Test Method During the performance of the RHR shutdown cooling mode test, the SACS will also be evaluated to determine the heat. removal capacity of the system and demonstrate the capability of achieving cold shutdown within the time specified in the design specificiation. Our ir.r, c0eratie ef ether ES? quig.;i.t, u. s.g--t-ilit, ef S;.00 t: ::i.t:i. *ha ** quired ==i-ca e=t -ill M
- 1:L wa.
J. Acceptance Criteria
, L/vr I /2 :
q '9 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 Se.s AHach'"end y haas-A_Aesly lu 4 < m w && e sp a Q s b te 9. a - 3 77 2-197 Amendment 5 l
f i
/fffc]e bm en } $--
14.2.12.3.40 COIBFIRMATORY INPLAlrf TE8T OF SAFETY-RELIEF VALVE DISCRARGE
- - 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 33) and show that the loads and structural responses documented )
in the PGAA for SRV discharge related loads , , i are conservative compared to the responses l :
which occur during actual SRV discharges. t l
- b. PREAEQUISITES
- 1. Power level should be sufficient to l support steady steam flow, during the !
test duration, through SRV discharge i l
line with normal plant operating pressure at the SRV.
- 2. Instrumentation for monitoring loads and structural responses has been installed and calibrated. l l
j
- c. TEST METHOD )
A snakedown test will be conducted to verify i l
tho' test set-up is functioning properly. The
- te' sting will consist of single valve actuations (SVA) and subsequent consecutive valve actua- ,
tions (CVA) of the same valve. Selection of I
the SRV discharge line used for testing will be l based on NUREG-0783, " Guidelines for confirm- 1 l
i atory Inplant Tests of Safety-Relief valve Discharges for BWR Flants," recommendations.
Data will be collected and analyzed by computer code to verify design analysis.
- d. ACCEPTANCE CRITERIA l
Level 1 The peak pool boundary pressure during air clearing and steam discharge during the valve actuation is less than the predicted valve specified in the PUAR.
I .
--.-,--.-----__,-,,_,,-------,-a,-- ,,e--- ,w-_,-,_ _ ,,- -..,--- , - - - - - - - ,,m.ww-wwm-e,.-,nw-- --o---
- HCGS FSAR 4/84 14.~ 2.13 SRP RULE REVIEW ..
14.2.13.1 SRP 14.2, II, Reculatory Guide 1.68 Revision 2, Auoust 1978: :nitial Test Procrams for Water-Cooled Nuclear Power Plants HCGS complies with Regulatory Guide 1.68, with the following exceptions and clarifications:
- a. Position C.1 provides the criteria for selection of plant features that are tested during the initial test program. At HCGS, testing is conducted on structures, systems, components, and design features as described in this section based on their safety-related functions.
The objective of Regulatory Guide 1.68 is to describe the scope and depth of a test program required to ensure that plant structures, systems, and components perform satisfactorily in service. The basis for this Regulatory Guide is Appendix B to 10 CFR 50, which l l specifically applies only to testing the performance of -
safety-related functions. Therefore, this Regulacory Guide is applied only to plant structures, systems, and
- components that have safety-related function, defined as those plant features necessary to ensure the integrity of the RCPB, the capability to shut down the rcactor end maintain it in a safely shutdown condition,
, or the capability to prevent or mitigate the 1
consequences of accidents that could result in offsite exposures comparable to the guideline exposure of l 10 CFR 100.
i l Safety-related structures, systems, and components are identified as such in this section and are tested to meet the requirements of Regulatory Guide 1.68. Other systems and components within the plant are not safety-related may or may not be tested in accordance with the i Regulatory Guide. Since the plant units that are not safety-related by definition do not compromise the safety-related aspects of the plant, it is not planned to test them to the Regulatory GJide, i-l b. Position C.7 and Section 1.h of Appendix C of Regulatory Guide 1.68 state that one of the objectives 14.2-198 Amendment 5
,,-.--,s- - - - --- --_, -~~-- - _ %,_-re-- -v --.r.,.-------g-----..,_.m ..,.m---r,., ---,--,,-~.----w,,--.-~,....w -.- .-.,
5Cd/s0 s;
d'a 8 :-
4 Y
~ 1.c I E
% .C $3 j .
Eh '
= g Il -! 53 j i d 3 ^8 5 !
l$ t 5
]
d ic ! A - ri 3
!z .
a e .
- ,, g.t- c . 3 . , g I 3 9$ 3 - ! t; l 8 ~-{
y 1
_-l 1.j 8 e 1 J!
- i g< gvv i.
2
. e4 i v
gu g t u y B R u LM .
"I. 3 w
., e55.8 1* E $ ,E ,E IuE 3 s"3 W 1
-s 8 a 1 Mj 85 3 g 3 g8 !
1 a h h :3 '.
- = -
.i [5ett51==l*3]!3
]uw98 : : aij1-11 JjI!
3 if ja
, 843
! e , sa
,g %l t ,aaa 68 3 J j I g I
- 4' .
i 3 )
- w 3.
3e 2 '
lili3 - - 3 li g,' :e.:1!.a321 8 8 8i ,! 8 8 8 s !, 81 ! !1 "I;ae e e
,"- } .3 ,
ks t *. $N2 xg 5}-- E} cs
=
w n
w
=
w E, 5
=2 w w -w a===:
w w <w w w w w w w w
- : -n-}
w w
^
w w
{
N ww
~ A er a ww ww www,w =w%%%% w% ; w w ww %w ww
=
R$ 5 ww = www %% % )
~ww w www t
- $ a -
I ww % www wwwwww ww % %% -
_ w c ie w n ww w w4 j
~
k = = w= k k wk w w & 6=
- = e ww wwww%w w w ww % Xw ww =w h!i ww wwwwwwwwwww w w w (= x = g ll wwwwww - -
1
- a 5-i l45=. i I a!
d<If:,4as1eis
! ..l*; '
g .g 3 , s3
_a ,1- s-! g 3 gi s 3] 3 !53 a,2N g 3 F Q
- !gil8ill dE '
E
-2ityj13e 3!]
el3dj 3s 8lii" j ge$ g8
! 3 D. 1- !gjs :l"iligjs j ! i gg y c l
l!Hlyijijig;{i!{a_3:3;.fihl
- .,. .s. 1 3,,23 J 5EE N $ 3 5 I'- $$$ .EMEEIM$}}}$ - 3 03i!d -~~-~umd19hwdHb4145 u11Xttvund a l c} * ==vms :e w r.; n g m n a m a g u p 3
. p l
tb b
b hh HCGS FSAR 10/83 ( . QUESTION 640.14 (SECTION 14.2.12) For compliance with Regulatory Guide 1.68, Appendix A.3, modify FSAR Subsection 14.2.12.3.6 (Source Range Monitor Performance and Control Rod Sequence) to ensure:
- 1. A neutron count rate of.Et least 1/2 count per second registers on the startup channels before startup begins. -
l i
- 2. The' signal-to-noise ratio is greater than two. l
- 3. Initial criticality will be approached on a startup rate of less than 1 decade / minute.
I
RESPONSE
The ::: ;ttre: criteri: fer th: =ini==m-n :t ;n count-rat: i: ::
;ecified in the C" St:: tup T::t Op ification in Oe:tien 10.2.12.2.5.d. f.dditicn:ll1 , th: =inirr :: nt : t: will :::t the s % ic;;;nte of Ch:pter 15, Technic:1 Specificatiene.
/ The acceptance criteria fer the signal ceent-te-neize retic le --
U entifiM in the M Startu; Test Specificatien in 5:0tien t [ 10.0.12.2.5.d. Section 14.2.12.3.6.c has been revised to indicate a precaution that initial criticality will be approached at a period greater , than 30 seconds (equivalent to starting rate less than 0.91 decades per minute).
$W IC U P1
.. J Id Jd a
/% 3. )) . 3. f.,ji ygg gegy gg vi_$q
~ !% CNCffY /*Cc EPreem Mt R=L's A fe, ,z g g ,y y, y M 72U U Co a di hkTE QF gND j 'l2. 000 Nn f Er2 5 ccon D I
N & GbN AL - TD - h36is.f pAno 62Qux pt/w my, ( 640.14-1 Amendment 2
,_y._._, _.. ..
.,,,_..,_,._..y *, ,_ ," , _ , __ , . . . * ,m, _ . , , , , , . . ..m.m,,_,._,_,_m,, . . , . ,
f)9(otr /o& HCGS FSAR 10/83 drive startup test, described in Section 14.2.12.3.5, i during initial heatup and just after fuel load. Also, the reactor protection systes is verified to operate following scheduled transient tests such as MSIV isolation and turbine trip.
- 2. Leak Detection: Although there will be no startup test procedure designated Leak Detection, portions of leak detection governed by Technical Specifications will be functionally checked just prior to fuel load using
, station surveillance and calibration procedures.
, Setpoints related to leak detection high steam flow in , 4 HPCI and RCIC are verified and set as stated in 1 Sections 14.2.12.3.12 and 14.2.12.3.13. Normal operation of leak detection systems, such as the drywell equipment drain sump pump will be accomplished using station operating procedures. , j
- 3. Equipment and Floor Drainage: Although there will be no l startup test designated Equipment and Floor Drainage, ;
these systems will be functionally checked using station operating procedures. Any portions of equipment and floor drainage systems governed by
~
Technical Specifications will be functionally checked prior _to fuel load using station surveillance and 4 calibration pcocedures. l,l l 1 l 4. Leos:e Parts Monitorir.g: Alt.hcagn there will be no I { startup test procedure designated Loose Parts ; i Monitoring, additional data to supplement the ' j preoperational program on loose parts monitoring will j b aky 3gs ed in revised Section 14.2.10.
- 6. Hotwell Level Control: Although there will be no startup test' procedure designated Hotwell Level i
Control, operation of the hotwell level control system will be verified using station operating procedures and monitoring hotwell level during Phase III startup
- testing.
1 j 7. Leak Detection System - Refer to response for item 2 1 above. l
, 8. Penetration Coolers: Addressed in Amendment 1, Section 14.2.12.3. .
- 9. ATWS Test: Although there will be no startup test procedure designated ATWS, the.ATWS subsystems are thoroughly checked out logically and functionally l during the preoperational test program, as described in Sections 14.2.12.1.2.c.6, 14.2.12.1.3.c.3.,
14 2.12.1.4.c.4., 14.2.12.1.8.c.9, 14.2.12.1.9.c.7., 640.20-8 Amendment 2 l
~ fgiik-a: L page sg o fH I
_ ,_ . ._~. . _ . . .....
.. * .- k . .N .- - .. .
. _. ._. ._ T ,Ed_Q_%cunA 100.To. ._ of _.s ysdew low _O.r__. syd ea_ pi ain .. o n.w lic1, j
._ 4AL._se+tsors_ _ nre. loc e E. .In. addibon,
.. not s t . _ Sign <tkuves _ oS seIec$eA . .
_Yramsleds__ (e.g. MSi\/ elosuces # reeinu.I hon _ f o p _ A :p t ) _ % + m e g _ cause 1werease8
._ Mois,e - leveA.S _ ad . N S tara . will loe. .. .
_ . . ...... _ hke.n 4._.a.11h__.E*:1YR sko.4
..._..._ _ . _ _ _ _ (Yl _._44 6 l'S C_.. FM(tf n .1__ .k 4 Y .__I H_ __ . _ . .
- .__..(__. f= _. - - ._ ___ . _ _ _ _.__ __
- l A_'. E leL b a/ ya L b u_U L [cr> hll . 3 olds e3 Lea' i n . _
_ . _S.e e 'No a 3 l. Y_:.2. 2.2 . 3. 3 0 .O n d.. / 4': A.) 2. 3. A A
.p 6w_ ew- __"..N 'h- _ _-WG.4e - pWeue . _ w a emmehh' 6
- 48 -*.'ON' * +8**6. .m e e
.e game . n_e N .eque -eau. . _ e. . .N= .' _ -e . e .-
_w.Oe_. .e u.-9. g . e _.w _. e ee.. . ... m_. , am.w . h
- 1-. em . . em e. -.e h .w.e 4 e. _ .h e e eaa._ _ .p , , ,,,,g, egm,mgyamee e, m 6 6 em m. @ . enemi.-ie -en.wy m. e - -
i l - . . . . - - - . . _ . . . .. ._-.- ._ . ._ .. . . ... . . . l e@e . . @ .-O
$ e G emom gm.e em e . -e.gmu..- - = . e o e., e es e e
$9 (ph? , HCGS FSAR 10/83 OUESTION 640.23 (SECTION 14.2.12)
To help facilitate approval of future changes to the Hope Creek Initial Test Program, list and provide technical justification for any startup tests or protions of startup tests which you believe should be exempted from the license condition requiring L prior NRC notification of major test changes to tests described ig FSAR Chapter 14. Such a list should include those tests not necessary to verify the proper design, construction, or performance of systems, structures, or components important to safety (fulfill General Design Criteria (GDC) functions and/or are subject to 10 CFR 50 Appendix B Quality Assurance , requirements)., RESPONSE - i cgf U;i..; ris_:: ' * : ; ;;ide,-the following tests are exempted
- from the license condition requiring prior NRC notification of major test changes
- -
- 1. N - Steam Production [SeeUca /S 2. / 2. 3. /s)
Justification: The sole purpose of this test is to demonstrate the nuclear steam supply system provides sufficient steam to satisfy all appropriate warranties as defined in the contract between General Electric and PSE&G.
- 2. .TN Pressure Regulator ( 5ec 6 N.7. D . 3. 2o) 1 Justification: The purpose of the test is to tune the
- pressure regulator control ayutes, to' demonstrate the backup pressure regulator, and to demonstcate smooth i pressure centrol transition between the bypass valves and the turbine contrcl valves. This system is classified as a. power generation system and is not a safety-related system, does 'not fulfill a general design criteria, and is not subject to 10 CFR 50 l Appendix B requirements. ,
i
- 3. Tee M er-etb - Feedwater System - Water Level Setpoint Changes ( 6e< ho u / 4, Z . j L 3,2 Q Justification: The purpose of the test is to tune the feedwater control system for all feedwater pump and
- valve configurations. This system is classified as a
; power generation system and is not a. safety-related I
system, does not fulfill a general design criteria or is not subject to 10 CFR 50, Appendix B requirements. i 640.23-1 Amendment 2
h0 hl
. HCGS FSAR
- 4. N- Feedwater Pump Trip [6ec hc.i /4 7.12. 3.21)
Justification: The purpose of this test is to verify that the reactor recirculation runback circuit activated by a feedwater pump trip will act to drop power within the capacity of the remaining feedwater pumps. The acceptance criteria for the test is simply
, that there is an avoidance to scram due to the runback e circuit, thus providing a capacity factor improvement.
This is not a safety-related circuit, does not fulfill general design criteria, and is not subject to 10 CFR 50 Appendix B requirements.
- 5. T N Turbine Valve Surveillance [b'CNCH IN'2'II+3*A2 Justification: The purpose of the test is to demonstrate acceptable procedures and maximum power le'vels for recommended periodic surveillance testing of the main turbine control, stop, and bypass valves without producing a reactor scram, thus providing a capacity fac. tor improvement. This test does not prove a safety-related system or circuit, does not fulfill general design criteria, and is not subject to 10 CFR 50, Appendix B requirements.
6. Test *NtM 94 - Recirculation Flow Control [.See Noe Ni2. I2 e 3
, Justification: The purposes of thic test are to adjust and demonstrate flow control capability and to determine that the electrical compensators and controllers are set for desired system performance and stability. Thic system is considered a power generation system and is not considered a safety aystem. No portions of this test fulfill a general design criteria, nor is this system subject to 10 CFR 50, Appendix B requirements.
- 7. Test =Her@9p- Recirculation Pump Runback hec /so w /4s 2.12, 3,2 g)
Fert6-deo A p Tre Tes t. Justificati his test is accomplished in conjunction with . The justification is the same as that given for N r-f m y ,f u ,,
- 8. TesMosa=20D- Recirculation System Cavitation [$ c b,, /4,7, J ,7, U2, Justification: The purpose of this test is to show that l the recirculation system flow will be runback to 1 prevent operation in areas of potential cavitation to l protect installed plant equipment. The test does not )
l address any nuclear safety-related concern, does not fulfill a general design criteria, and the runback g l , 6,40.23-2 Amendment 2 l i
,-s --,,---,-g ---vn --,,
-,-w., u---,,m w -
, -.-,,,-,_p------,n , , , --
h HCGS FSAR 10/83 ) I circuit is-not subject to 10 CFR 50 Appendix B requirements.
- 9. N Reactor Water Cleanup (RWCU)[$ec4ico 84.2.113 32j Justification: The purpose of the test is to ,
demonstrate specific aspects of the mechanical l l operability of the RWCU system, including NPSH to the RWCU pumps, non-regenerative h'at e exchanger j performance, and bottom head flow indication. The test does not prove any safety-related aspects of the RWCU l system, such as system isolation. Additionally, the ; test does not fulfill general design criteria, nor do ) functions of the test fall under 10 CFR 50 Appendix B requirements. i 9 j l T I i ( 640.23-3 Amendment 2
.=. .- . - - _ - . - - _ - - - . -. _ _- ._. . . _ _ _ .
HCGS FSAR 4/84 RESPQNSE CH, Juing Figure 44.2-3 .. . vuidw, the following tests are exempted from the license condition requiring prior NRC notification of
. major test changes:
'j 1. Test:4ME3h-Steam Production [$eA N,7,/2,3./d l Justification: The sole purpose of this test is to demonstrate the nuclear steam supply system provides sufficient steam to satisfy all appropriate warranties as defined in the contract between General Electric and PSE&G.
i ] 2. Tesh Pressure Regulator [6ech,, /4,2./z. 3.2d g ) Justification: The purpose of the test is to tune the pressure regulator control system, to demonstrate the backup pressure regulator, and to demonstrate smooth pressure control transition between the bypass valves and the turbine control valves. This system is classified as a power generation system and is not a safety-related system, does not fulfill a general 1 design criteria, and is not subject to 10 CFR 50 Appendix B requirements. 1 3. TeaMENurmeedhs- Feedwater System - Water Level Setpoint
- Changes ( See/7m /4,2. J2,3.2/)
, Justification: The purpose of the test is to tune the ! feedwater control system for all feedwater pump and l vo.1'/e configu- at
- ons. This system is classified as a
- power generation system.and is not a safety-related system, does not fulfill a general design criteria or is not subject to 10 CFR 50, Appendix B requirements.
- 4. N Feedwater Pump Trip (Sec/re, /4,2. / 2. 3. 2/) l Justification: The purpose of this test is to verify
- that the reactor recirculation runback circuit 4
activated by a feedwater pump trip will act to drop power within the capacity of the remaining feedwater pumps. The acceptance criteria for the test is simply j that there is an avoidance to scram due to the runback circuit, thus providing a capacity factor improvement. 3 This is not a safety-related circuit, does not fulfill 1 general design criteria, and is not subject to j 10 CFR 50 Appendix B requirements. 4 SRAI (7)-5 Amendment 5
a HCGS FSAR 4/84 12.*
- 5. N Turbine Valve Surveillance [Sechu. l'/.2,43 Justification: The purpose of the test is to demonstrate acceptable procedures and maximum power levels for recommended periodic surveillance testing of the main turbine control, stop, and bypass valves ,
's' without producing a reactor scram, thus providing a ,
capacity factor improvement. This test does not prove a safety-related system or circuit, does not fulfill general design criteria, and is not subject to 10 CFR 50, Appendix B requirements.
- 6. N Recirculation Flow Control (Sec// . F/,2.2./233.27)
Justification: The purposes of this test are to adjust and demonstrate flow control capability and to determine that the electrical compensators and
- controllers are set for desired system perf'rmance o and stability. This system is considered a power generation system and is not considered a safety system. No portions of this test fulfill a general i design criteria, nor is this system subject to 10 CFR 50, Appendix B requirements.
- 7. Tesl ANQ,,Jh Recirculation Pump Runback [Sec//m /4.2./2 3J 2.3) fenf. + A Justification: This tes is acco.nplib N inTm reJ.
conjunction with T m L . ::!. The ustification is the same as that given for . . J/, 4/ dge ,
- 8. Tgrf4FA40EW Recirculation System Cavitation (5,ch% W.2.l / 2..g 2 d j Justification: The purpose of this test is te show that the recirculation system flow will be runback to
, prevent operation in areas of potential cavitation to protect installed plant equipment. The test does not i address any nuclear safety-related concern, does not fulfill a general design criteria and the runback circuit is not subject to 10 CFR 50 Appendix B l requirements. !
- 9. Reactor Water Cleanup (RWCU)( Seell* N,7. /2.l 3. 32) '
! Justification: The purpose of the test is to , . demonstrate specific aspects of the mechanical i operability of the RWCU system, including NPSH to the RWCU pumps, non-regenerative heat exchanger performance, and bottom head flow indication. The test r does not prove any safety-related aspects of the RWCU
- system, such as system isolation. Additionally, the g l test does not fulfill general design criteria, nor do I
SRAI (7)-6 Amendment 5
~
HCGS FSAR F l containment prepurge cleanup system (CPCS), and the reactor l building ventilation system (RBVS). The drywell air cooling l system removes heat.from the drywell during normal plant l operation, plant shutdown, and certain abnormal conditions. The capability to purge the drywell and torus is provided by the l CPCS and the RBVS, as described in Section 9.4.2. 9.4.5.1 Desian Bases The design bases for the drywell air cooling system are as follows:
- a. During normal reactor operation, limit the average air temperature inside the drywell to 1350F maximum, with no location over 1500F, and 1280F maximum around the recirculating pump motors.
- b. ~During scram, but without loss of offsite power (LOP),
limit the maximum ambient temperature in the area under the reactor vessel to 1650F or lower, for up to 30 minutes.
- c. During normal reactor operation, prevent concrete structures within the drywell from exceeding their maximum design temperature N sped erdin Shekom 23. 8 . 2
- d. During normal s.hutdown operation, limit the air temperature inside the drywell to 1040F maximum and 600F minimum by use of the RBVS purge mode.
- e. In the event of LOP and reactor scram, limit the ambient temperature inside the drywell to 1850F.
- f. The single failure of an active or passive component in the system cannot result in a complete failure of the system.
5
- g. The drywell-airscooling system is not a safeguard
( system for a loss-of-coolant accident (LOCA), but can l 9.4-70
F HCGS FSAR 10/83 i Each SACS loop is located in a different room. Power is supplied from four independent divisions. Failure of either a motor-
. operated valve (MOV), standby diesel generator, electrical division, or pump does not prevent the system from removing the full heat load. This arrangement ensures that the full heat removal capacity required is available after a postulated single
- active failure.
The TACS has no safety-related function. Failure of the system does not compromise any safety-related system or component, nor does it prevent a safe shutdown of the plant. In the event of a LOCA, LOP, or a pipe break, TACS is isolated from the SACS. f 9.2.2.4 Test and Inspection C l Inservice Inspection and functional testing of the safety-related portions of the system and components will be in accordance with the examination and testing criteria of Articles IWA, IWD, IWP and IWV of Section XI, ASME Code, 1977 Edition and addenda through Summer, 1978. - l The specific examination and tests of the system and components will be listed in the Station Inservice Inspection (ISI) and Inservice pump and Valve Test (ISI) Program Administrative Procedures. 9.2.2.5 Instrumentation Applications The SACS is designed for remote operation from the main control room. In addition, one loop of the SACS and its associated valves can be operated from the remote shutdown panel. Local and remote indications are provided to monitor process parameters of the system. The following conditions are annunciated in the main control room:
- a. High-high/ low-low level in the expansion tank 9.2-15 Amendment 2
Y F T he. 5 A C S i s Nesde0 la hN O puoP"*0~0 {esh p lese J fower asceadm her$ f =N~t* -
& Ds4 h a[oslvacls SY Nse $*sbs a v e-
) bects%$ Im soUrc$rees N.2. 82. l. lG uA.
14* 2e il 3 38-6 '
- l.
t I R e f i i
- - - - - - % w w w- w w-s e,.w- - - -eaw,wr-wa v -,-sv - , - - . < e -*- vm we.,.v- -w+w-w-smy----w -- y w --
agmw4 w w
l piv 1- ! HCGS DSER Open Item No. 186 (DSER Section 7.2.2.3) TESTABILITY OF' PLANT PROTECTION SYSTEM POWER. We will require that the applicant demonstrate the capability of the design for on-line testing of each instrumentation channel, logic, actuation device and actuated equipment in the ECCS and BOP ESF systems. All actuated contacts and devices should be considered and those which cannot be tested on-line should be identified and justification provided. 1
RESPONSE
The response to Question 421.22 has been revised to provide the requested information concerning on-line testability. l i 186-1 l l 1 I
e V T BCGS FSAR 4/84 ' * . l pUESTION 421.22 (SECTIONS 7<t, 7.3, 7.4, 7.5, 7.d, & 7.7) l The design of the instrumentation channels, logic and actuation devices of nuclear plant safety systems should include provisions for surveillance testing. Guidance is included in Reg. Guide 1.118 and IEEE Standard 338 for implementing the requirements of IEEE Standard 179, whicts requires in part that systems be designed to permit periodic tesh ng during reactor operation. . Section 3.1.2.3.1 and 7.2.2.3,2 includes a brief description of the at-power testing capability of the reactor protection system. Bowever, sufficient information has not been provided to determine the acceptability of the at-power testing capabilities provided in the Bope Creek design. Provide a detailed discussion with illustrations from applicable drawings on the at-power testing capability of the reactor trip system, engineered safety features actuation system and auxiliary supporting features, the actuation instrumentation for the reactor core isolation cooling system, and the inslirumentation and controls that function to prevent accidents (i.e., high pressure / low pressure interlocks) or terminate transients (i.e., level 8 - turbine trip). This discussion should include the sensors, signal conditioning circuitry, voting logic, actuation devices and actuated components. Include in the discussion those design features that will initiate protection systems automatically, if required during testing, upon receipt of a valid initiation signal. i RESPONSE i As required by IEEE Standard 279, capability for at-power testing has been provided in the design of the HCGS safety systems. ' Conformance to the guidance specified in Regulatory Guide 1.118 and correspondingly, IEEE Standard 338, is as stated in Section 1.8.1.118. The analysis portions of the various system descriptions in , < Chapter 7 for the safety-related systems referenced in the ' question describe the methods by which the safety system designs satisfy the testability requirements of IEEE Standard 279. The l specific sections covering the testability of these systems are listed below: RPS - 7.2.1.2 I ECCS - BPCI 7.3.1.1.1.1(c)
- ADS 7.2.1.1.1.2(c)
- CORE SPRAY 7.3.1.1.1.3(c)
- RHR-LPCI 7.3.1.1.1.4(c)
PCRVICS 7.3.1.1.2(d) RHR-CSCM 7.3.1.1.3(c) RER-SPCM 7.3.1.1.4(c) . 421.22-1 Amendment 5 DSER OPEN ITEM /b f"3 o ./)
- O e ae e
' .. ossa ceau ream /g/, BCGS FSAR 4/84
~
DSER OPEN ITEM / Ide l PCIS CACS - Supp. Chambec to 7.3.1.1.5(j) 7.3.1.1.6.1(c ) Drywell Press. Relief . .
- RB to Supp. Chambec 7.3.1.1.6.3(c)
Press. Relief Sys. .
- BOAS 7.3.1.1.6 5(c)
- Outs 7.3.1.1.6.4(6.
DICRMIS 7. 3.1.1.7( j )) 7.3.1.1.s(c IISIVSS FRVS . 7.3.1.1.9 . RBVIS 7.3.1.1.10(h) EAS - SSWS 7.3.1.1.11.1(c)
- SACS 7.3.1.1.11.2(c)
PCIGS 7.3.1.1.1.11.4(c) 7.3.1.1.1.11.5(c) CACWS EACS - RBEAC 7.3.1.1.11.6.1(c)
- ABDA 7.3.1.1.11.6.2(c) l
- ABCA 7.3.1.1.11.6.3(c)* )
- SWIS 7.3.1.1.11.6.4(c) l RCIC 7.4.1.1.3 SLC 7.4.1.2.3 RRCS 7.6.2.7.2(b) 7.6.2.7.2(n) ;
7.6.2.7.4.1 Design drawings in the form.of elementary diagrams, P& ids, logic diagrams, instrument location drawings, and electrical drawings that describe this capability are listed in Tables 1.7-1, 1.7-2, and 1.7-3. In response to the NRC's request for additional information during the meetird of January 11, 1984, review of the systems identified above, with the exception of the reactor protection system (RPS), reactor core isolation cooling (RCIC) system, standby liquid control (SLC) stem, and redundant The review w&Ms.g_ _3jp reactivity control
'rterrin:
system (RRCS) r!!! 5the capability for # performed.the at-power test and sentogs used in these systems. All actuated contacts and ster., suosyst t., or ompo t1 devices"p" . !! 5: considered. f'Any ing wi be i ntif anos cas e capa u ty e at-p er te 1 be vided. The ults 11 be C da tifi tion w suhri 'ed by J umen in a evisio to th resee ' e in i Ju 1984 J w w ,,,,.,la ,,a s 4. a 4 w. e,. ;/_9.,,,,,?In,amed ,,,,gs., es > 3 p s,L. es J tesd %4wds I C.c us: p'O 5 .ss s p a.pe , A < gII p e n, pas. % .ld.sA. r a A ca .=de.s &' P's-'" M ! 3.s.a.a I i ewikkes, b lf.. 4 sf wik&.i, =J n.cm.l yeM epp=J, p
& p b .l % 4 . p 4 f 4 3 ,* 4 3
da A Juhhi%a
.l. .
,.i;,t. g a. .i
.e 9 4 dalil. A eve J & .afaq .w. 4.p .
pA*.h"d*Q'd u 4 5tr A -->sSu% pla. t, =P k.J **LA.o f.4 A m .s 4. see'. e A a tu
- InstRT A. -
During the review, the at-power testability of an item uns established ( i if on affirmative response could be verified for/'bnfallowingthree
~
questions: .
- a. Is the item sufficf6ntly accessible to conduct the test during normal operation? - .
,~.
i Is the item sufficiently isolatable to permit its !
- b. '
safety-related function to be verified or is a safety-related system or subsystem encospassing the item isolatable and testable? Does any bypassing method that must be used to accomplish the [ c. test conform to position C6 of Regulatory Guide 1.118T fj- gr@ g&tyg5M y these criterialtwo ADS SRVs. itemscause which would weredepressurization judged to be untestable if tested, andatthepower. the i steam-tunnel tangerature elements, which are inaccessible. The reliability and redundancy of the ADS instrumentation, logic, and actuation devices and the multiplicity of the $RVs adequately justify the lack of ADS at-power testability. Adequate element multiplicity and comparison tests of at-power output signals and electrical characteristics preclude the need for change-of-state testability of the steam-tunnel temperature elements.
~
i l i . 1 l l DSER oPEN ITEN /g[
nm om xm /84 /gffr ( 1 0
} -/$e.se cs-;lwi m 1 er* S L" M0 & *AITS'S S Cy s fe**ts fis 0 floaig _ Hems wat(IPAf fo be an1%ts'lEth/e af fea*Y.*'
\
- a. TcIS +k LOCA 5;ydahd* l**""*I Dewt b drysuef( & p r.essare ; Or vsAQ ^ LAA) \
in , McA',on eri l so ste% re. lac pi&g ;vovw. s RieAE.b do noi so--is-]core 5&f" l ' Io 'aloove. . TkTs a- b+s yciY' $ Iaccp+sh'on n+5 n sip s t % close 4^3 edrcGsm-t i so ch;on. voIves, +c hip f(a M c c. 6-cenV.ey g edSihi"5cie. Co stko ) reew-isolafiom. TL, ce'Med:ectqdgmm' 7s id. w h h f o a <e '7.s-2c (ske<+3 2-s-) a d Rpn '7.6 -27 (skicfs 2 el U . All o+k c m+hds
-G< actaeion o? % t 1 ~ogi(>msu'k caa. Ice var;kaa+poacy on\p%s pac-ica(c<v- a c k c4 A s ,- g w =A p= : eg= a* en v o+-
s ,. t < s + e a . I h, 3C. D - i k 00sMCi Cf. C't(C[iff V' k ( C G C '.b r b cA' d d.i 4(dA Agad (4.[tA4, s M8 FICOF 1 M( ar<a h,A-khp rat.n% si tot san %w .u.som ul[ Jo
- b ai, u . wa l om om im n' Md.ividM' kih-ki[ rediaflow sipls
== : := ==2 L __a
. caw 'oe. ver; '<c4 9 to h hpd 6-43 lepc wodu.ks 6+ me
% be oS one o+ a.%e since e.acA up(
l
+<t+ed.
h %uauthd (-tLG%le% Jourusy, r;n.J6ae.ul. ci 4 ciounds of % ?crs he. Ryve 2 3 -2L (sLu.-% (.-9) .rk o
+
AEM 9% Logi c cinuitry o2 & 'Tels "Nc/ 4 ad. cfoes adf 'in hi ki+ 1% -Wd,'n.g t, o4' 6 ac h l ae % 4rox s;gx*(s Som
% PCIS %:, fu idivickaA. adatec/
co m w c h s, 4
\
na o,- na in l - ---__.m.... .
' ge, y 1 DSER Open Ites No. 226 (DSER Section 8.2.2.5)
GRID STABILITY In regard to the grid stability analysis presented in Section 8.2.2 of the FSAR, it is the staf f concern, due to the close proximity of the Sales and Hope Creek Generating Station, that simultaneous trip of Hope Creek Unit 1, Salen Unit 1 and Salem Unit 2, should be considered. In response to this concern, the applicant by Amendment 4 to the FSAR stated that the Bope Creek Station will remain stable with the loss of both Salem Units 1 and 2, clarification and basis for this statement will be pursued with the applicant.
RESPONSE
'7~h e Pesponse -fo fuesfion V30 8 ho3 heen te.vued
-tu pro vs d e. 9-h e. r-egu e sted in form o Hon.
2 e l i
.j
_ _ _ _ - _ . _ _ _ . - _ . . _ _ . _ _ _ . . _ . _ _ _ _ _ _ - . - _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ . _ _ _ . - _ ._]
/ HCGS FSAR 1/84 I
4 OUESTION 430.8 (SECTION 8.2) In regard to the grid stability analysis presented in Section 8.2.2 of the FSAR, it is the staff concern, due to the close proximity of the Hope Creek generating stations, that simultaneous trig.of Hope Creek Unit 1, Salem Unit 1, and Salem Unit 2 should be considered. Either provide the results of a grid stability analysis that demonstrate grid stability assuming simultaneous failure of these three units or provide the results of analysis that demonstrates that trip of both Sales Units I and 2 will not cause trip of Hope Creek.
RESPONSE
(Section8.2.2hasbeenrevisedtoprovidethisresponse}' < In accordance with these " Reliability Principles and Standards for Planning Bulk Electric Supply Systems of l the Mid-Atlantic Area Coordination Group" (MAAC), grid 4 stability analyses have been performed as indicated in Section 8.2.2. Additionally, analysis of the most severe multi-phase fault with' delayed clearing (stuck 500KV i i Breaker 60X) on the Hope Creek - Keeney 500KV line at f Hope Creek, shows that Salem No. 1 and 2, and Hope Creek No. 1 Units will loss synchronism and trip. However, the 500KV system remains transiently stable. i . j i 1 0 e i DSER OPEN ITEM 7)h _ _ . - - _ _ _ _ . . . - . _ , _ , . . _ _ _ _ . _ _ . _ _ _ _ _ _ _ - _ _ . _ , _ _ _ _ . _ . _ , _ _ . . , _ , _ . _ , _ _ . . _ _ , . , . _ _ _ . _ ~ , _ _ _ . _ . _ . . . . _ _ . _ _ _ _ . _ _ -
u .
.. ... ~
- 1....
SCGS DSER Open Item No. 263 ( DSER Section 11.4.2.e) FIRE PROTECTION FOR SOLID RADWASTE STORAGE AREA , Insuf ficient information has been provided regarding the fire protection features for the solid waste equipment processing the asphalt and also the solid waste product storage area. RESPONSE , FSAR Section 9.5.1.2.31 provides a description of the fire pro-taction features for the radwaste building. This section has been revised to indicate that the preaction sprinkler system protects the filled radwaste drum storage area. e e 9 4 9 e l l l e e K53/17 ,
---.-,.--..~--,-~----w.. - -
, - - - , - .------~-,-,,---<----..--,--..--_..----,,.c-,---,-..-,-,..-...----,-,em.
> AU9 -7 '8d i} P 6 91. 5 5 HCGS FSAR 1/84 i
9.5.1.2.29 New Fuel Area Fire Protection Portable extinguishers and hose stations are provided in the vicinity of the new fuel area. Automatic smoke detection is also provided by photoelectric-beam-type smoke detectors installed on
! the reactor building wall just above the polar crane for alarm and annunciation bottr locally and in the main control room.
A 4-inch curb is provided all around the top edge of the new fuel l vault. The new fuel vault is provided with a steel plate cover, i Thus, water inadvertently spilled on the refueling floor, which i is at floor elevation 201 feet, is not likely to drain into the Furthermore, a 6-inch floor drain is located at the
~
vault. . bottom of the vault at floor elevation 181 feet 4 inches to
- preclude accumulation of water.
l 9.5.1.2.30 Spent Fuel Pool Area Fire Protection A hose station and portable extinguishers are provided in the l vicinity of the spent fuel pool. Automatic smoke detection is provided by photoelectric-beam-type smoke detectors installed on m the reactor building wall just above the polar crane for alarm
) and annunciation both locally and in the main control room.
7
- .. .w l
{ 9.5.1.2.31 Radweste Building Fire Protection I - i Most of the radwaste area is separated from the control area, f reactor building, and turbine building by 3-hour fire barriers l with Class A fire doors.
; The radweste area ventilation system can be isolated.
i I All drainage in the radweste area is directed to the liquid i radweste sumps. Wet pipe sprinklers are provided for the solid radwaste area. A proaction sprinkler system is provided for the radwaste truck loading area and solid radweste3drum storage area. JCLLi0 I
. )
i DSER OPEN ITEM R(p 3 9.5-30 ^:'r'eO
----m------M-a-w m*y. cyp. m _ _ _ +% wewwe rm _ tww __ __ww-3-+ >
)
i 1 5 of 7 j HGCS DSER OPEN ITEM NO. 265 (DSER Section 6.8.1.4) ESF FILTER TESTING Regarding ESF filter testing, FSAR Table 6.8-6, page 6, table note b, states charcoal filter leakage testing is acceptable at a penetration of 0.25%. This value is inconsistent with Regulatory Guide 1.52 which requires in-place leakage testing for both HEPA and charcoal filters with air acceptance of less than 0.05% penetration. Please correct this statement or provide justification for departure with Regulatory Guide 1.53.
RESPONSE
FSAR Table 6.8-6, page 6, table note b has been revised to state that the downstream concentration is less than 0.05 % penetration. L t M P84 126/19 03-az
-"-'*"-""'--'------*--""v'----'---'-'--"vT- ---'~v--T--"-"- ~- '"
-~*--"'-C'"" M " - *e* - - --'" -- "- * -" ' ' " -'-"*"-
HCGS FSAR 12/83 l i TABLE 6.8-6 (cont) Page 6 of 8 l
- b. Inplace Testing of Adsorber l
- 1. Refrigerant (R-11 or R-112) is introduced into the
- upstream side of the adsorber at a concentration of approximately 20 ppe at rated airflow. The downstream concentration is less than(eve 64 of the upstream 20 ppe. No more than four tests are conducted on any given carbon adsorber. No o eUF radioactive isotopes are used in the efficiency tests performed on the carbon adsorbers. Each charcoal adsorber is tested for leakage using the test method presented in ANSI N510.
- 2. The installed carbon adsorber filter bank is visually and dimensionally checked for conformance to the design specifications.
- 9. FILTER HOUSINGS 4
1 l In addition to the housing manufacturer's shop tests, a field
; performance test is conducted for each housing. The housings are ! designed to withstand pressures ranging from 6 to 23 inches w.g.
I
- 10. FILTER INSERVICE TESTS AND INSPECTIONS
- a. The air filtering systems are subject to inplace testing before initial startup and after each HEPA filter or adsorber change, with the test interval not to exceed 18 months, in accordance with the recommendations of Regulatory Guide 1.52. ;
i l L l b. Periodic testing of the HEPA filter banks ensures that j the filter bank performance is not degraded through
- normal use, or during standby, to a level below that i
assumed in the accident analyses. Test methods and I sensitivities are the same as or equal to those for a initial acceptance of the system components. If the C test results indicate that performance of a component
" has fallen to the level assumed in the accident analyses, the component is replaced.
The following filter inservice tests and inspections l c. are performed at regular intervals during plant life to 1 Amendment 3 l (_. -__.__- - - . _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _
v _ . _ , - _ _ - 1 of 7 HGCS DSER OPEN ITEM NO. 266 (DSER Section 6.9.1.4) FIELD LEAK TESTS Regarding FSAR Table 6.8-6, Page 7, note 3, change to read,
" Field leak tests are conducted after each change of HEPA or charcoal filters in a system."
RESPONSE
FSAR Table 6.8-6, page 7, note 3 has been revised to include Field Leak Testing after charcoal filter change. 4 l JEStaz M P84 126/19 01-az i
J i l HCGS FSAR 12/83 l TABLE 6.8-6 (cont) Page 7 of 8 l determine that the filtration systems are functioning correctly: I- 1. With the fan running, readings on the differential pressure gauges, which are mounted on the filter plenum, are observed and recorded.
- 2. HEPA filters are replaced when the pressure drop across them reaches 3.0 inches w.g. Where there are two HEPA filter banks in series, the second one is changed at 4 inches w.g.
- 3. Field leak tests are conducted after each change of HEPA filters in a system or cho reo& ,
. 4. Field leak tests of HEPA filter banks are
- conducted with cold-generated dioctylphtholate, and a light-scattering aerosol photometer is used for measuring percentage pe'netration. An efficiency of less than 99.95% requires corrective action, as stated previously l .
l 5. Corrective action after a leak test may consist of increasing the contact pressure on a seal or replacement of a cell or cells. After corrective action is taken, an add (tional leak test is made
- 6. Tests of successive canisters of charcoal in the i airsteam of the charcoal adsorbers are made every 12 months after the charcoal adsorber bank is installed. Test procedures are the same as those
- used during initial batch qualification for
- elemental iodine and methyl iodide attenuating capacity. Tests for hardness, ignition
+ temperature, and radioactivity are not made on these samples.
- 11. ELCIWORK s
) a. Leakage tests on all ductwork are conducted during l
/ construction ossa opsu ITsu J/s(p Amendment 3 l
c-4 of 7 HGCS DSER OPEN ITEM NO. 267 (DSER Section 6.4.1) CONTROL ROOM TOXIC CHEMICAL DETECTORS Where are the Toxic Chemical detectors and associated instrumentation included in the Control Room Air Supply System?
RESPONSE
No toxic chemical detectors and associated instrumentation is required (and therefore not included) in the Control Room Supply System. Evaluation of accidents relating to the release of toxic chemicals is addressed in FSAR Section 2.2.3.1.3. Also, per DSER Section 6.4 page 6-3:
"With respect to toxic gas protection, the staff's evaluation in accordance with SRP Section 6.4, RGs 1.78 and 1.95 indicated that there is no danger to control room personnel from toxic chemicals, including chlorine, stored onsite or of fsite, or transported nearby (See Section 2.2.3)." ,
9 M P84 126/19 02-az
j ., j . . 1/34 NCGS FSAR l l 00ESTION 430.102 (SECTION 9.5.5)* fouling in the diesel engine cooling water system th degrade system cooling performance, and the compatability of any corrosion inhibitors or antifreeze compounds used with theIndiate if materials of the system.conformance with the engine manufacturers re om (SRP 9.5.5, Parts I & III) on jacke wateryrs.L . in therinstruc-Colt tion has included recommendation)fij, A g ..mA W , book. &
- 2 i detir . Jhe
- n zeer wat4r cooling I.i_:, r _-- r_;;tur-- ;;c::s softened to a total hard than 50 ppe or 3 GR/ GAL., and "NALCO 41" corrosion inhibitor is added and maintained in the ratio of 3-3.75 pints These per 100 gallons measures will of water (0.5-0.6 fluid ounces per gallon).f.- 2 .- i... . e l ; _ ; f - t : r- -
h- M e t ' - e4 ::;;ais-preclude long term corrosion and scale.-irrel:t wwlin, v wfin; .. Mm .: - G T '.. :
.,ei.. :re 9455 6 4--+.':;11..g,ill net b: 2 ; e51:r t; Section 9.5.5.2 has been revised to show that ' the
- do water is treat 9d in ac.cordan l
gee:r : d tivi... P + ge w -, l IN b . .... i i i, 1 i i Amendment 4 430.102-1 . . .
INSERTA M30.10L ., 4 M / w J ,= = l a -- 4 n =m A %
. s. .-(
6 w ,. 4a Aug ,v ~a 4 &-# EA .ft. & swN6 A M J zd Hess ja.t. 7 HC43 A safy y A mAJd xts Y $ ,A ' 22s $A' u 1 A L E -
,P ' ^ 2 7 "
M f -a x mk p ew! y 4 1 A _- ^- M A_ AA _ l I G l I l
- gup .
l
/N.SEET g
# m./ar_ .
? & .SPG. & M-:#==-_w f k lle=
.A __J. M 3. 4...A_ _n_i_,,A .- 2_ ~_d.
....T
~
t% y e an,p% - A,- .x _- + r s .. _. s a
=-
, ,3--_u ._=w---+- _- ,. ,
l . . . -. . ...-. .. 2 6 . d . M L ~ A e A : ' _ m = - A A-l.. .--... J A - l a ~yneh - LAA. - j . _ . .- . . .. . . . . .
-. g - n+L. O!=4j. _/
4_& 3 - _ _ d 4 .__ @ at;_. M 1.- r61;4/M _ d idenA+W .A e/x,u,L- - 1 1M.- e . ( A) N t 4 ,_4 4 4. A a y A 1.JA7 : 1 1
, - _ _._ A _. _ _
-- J :A A_-__,w X ,au a o i
nA . _d 1. .& A,Lx. _ ,- - A*^ ^ 4h A UQ f & .='i l
. _ = _ .
l s **. .
. . Nao.e a.Lca o - ,
%;fA.e-.A M*.
fAA . Aa fx k Mf _
. . "- ^_A -- -
L__ a p : x a_ , -_ x_.>s. . -. _...
==
.v. ,y_- _ 4 em y - - - . .
.- ===
m A_M241 W A__ ,L_.A_fA ~ '~ '~ s.,wvA A ,~A* t ym. .
^
f AA M A Aa 1
- __ , -~ - #__
I _ . . . $.. __w__ A ' A L 4 .4 u;L/L _ v d' nw5 s 4
* . n, gW l4W hh : _-- --
- ^
1_-- p _/ _ - -
~
A A A AAA -
/ 7 /
A A k, ) A. A_ - -
-A
. .. . .. ._ _ . !L?.
.......... _.=.
a
. . .. . . . . . . . . . . . . . . . . . . . . . ~ . . .
- e s . s st is 840268 0 69 eso. m (a .t] -
= 1 _c- eY <= Y OM
... oL A'. --. .
oOe gg g ge,, ON Wh@ G me . T MN ee e.e 4 i m m M gm e e mO N N _ Ge -h em qmee ^ emmlupm enus Me __ ee em e em em Gene m@ 6 & 9 6 e Og9 -- -- 666
-m.e Se . emm- ,- - . _
M9 4 64 4Ne W4 No. - emuDG & -_- DN penG$W em m M N+ O M M N _ G 8
.s e e-em. 6 6
aume e e em
@@ me r 4 HEW m
Y J l
==
9 w -
7-- esitgq
- BCGS FSAR 1/94 4
~
f p GUESTION 430.108 (SBCTIN 9.5.'5) 1 ' Recent licensee event reports have shown that tube leaks gre t i Eoing esperienced in the heat enchangers of diesel engine ,1acket f cooling water systems with resultant engine failure to start on 1 demand. trovide a discussion of the means used to detectInclude 1! l 1eakage and the corrective measures that will be taken. tube 1l l > jacket water leakage into the 1mbe oil system (standby model, jacket 1 Aube all leakage into the jacket water (operating mode), 1 l rater leakage into the engine air intake and governor systestrovide2lt l Ioperating or standby mode).or outleakage in each of the above gondi 3' ! 2' tolerated without degrading engine gerformance or causing engine 3 i failure. Zhe discussion should also include1SRP the effects g.S.5, of Parts II ; ja water / service gater systems leakage. ;
; aI1 l
l nas m s:
~
l : a - - 1--a,..,...-- 1= x m m m M g i m n g s m c..i , - ... The heat enchangers are procured to A9E Section !!! design and quality
. requirements, and are seismically qualified. e-- NStar ,$, .
Leds at teos in heat enchange equipment are very difficult to discern j by any means short of removing the heat anchanger fra the syste and Instruments te determine lute etl ) subjecting it to hydrostatic testing.in water er water in lute oil are gener the lee oil level er the cooling water level is not re11 ele 81n a mucn c,.sef as there are se many influences other than the heat enchange equipment. ) d ' l rA c s a e a r.g'& J -
-l a - e ,- a &m &j w- +J.be cool
. l g
. Generally, late oil la the water tyste has ne detrimental effect a the j gnaine. Neuever. wate* in the luhe oil % of concerny i f & & - & 1 .= di _ ,a aus '
m m.a \ lean ***' S baul=-f M spmeeRYm"e=wf &y 'b! AWS *::! %$ & ( percedment. j afA M O=~h / Ande, ool, de}M% q 3l> 9MM*4 u* !3-
'The rocker are lubrigstia syeta is separated free the sete Idefeatten l systs because of the greatetty of the rocker systa to seerses of water (splinder heads rester asseelies, etc). Additia of meter to that /,, .
,I ) "
syste. des to heage, would be detected by the high rester are tant l 1 m et alare, 430.100-1 Amendment 4 8*01/10/848'
r
%.ma fp.WRT $. - . . . . .
eth. s (= 4 L - a xx , e, t n ,- At4z :er& sie .x4
,pa.2 isexu & % .
(
&@ A z6. _
.JJ<
d b.xa.4 y.66 y44/
. 1 A J 6C M A f A y, J
/A .f 4 eM ara A A MA/
6
&. 1A :r- s y A 4AGW G MLL. & \
. 6 2-Jy A a A-s-/f A
.4C a./. s1A 44 , AM 2k
. - . ,s , a a,ma>u - w. _ _ , , , ,. m m sa u. _ s-sa, - ._-----aa, a a ma--m- a--mn m- -,,
t __ And ,. .. 2-' b J2_ _ p- - _, -
. ye e-k /> Y $
hMt4 =yw Ep % /0 &,7'& . m :i1 - ~
' =
kYen .d>1ehw 4 , z 'L '" ' gb abfga!,yw:nWU
.24 40 y .drf4 A&+ ,
_LAn n. Th. A4/ = ,4Aar .L< af J u 4 pd qJ4 ep;a AM sn de&;ur % . d s/m 4<hy 4 4 a p u .,< rnA r.
- ^ % . J f h 4 l J + - y :+ s - - j l -- L _-
I w 2 :^ ^- M!f f'.' Y A P_^' r -
^ ^
\ yE '
. A+!rg & Abew7+
2f am
/
s e4:s .4 W -- w - - W s. 3/r
. - ----_--m--.._
a. g
- p. < k
- Ha/M d v -
L L >$ A ,,- / ,. fje=.. l2-- l*, n'. _ n - k , \ j. _ k s Y)n 2' xx .-, L ,12,. , L-,4 , _ Q-,1 y1_. M M AS aAf
.. ~) . f p ~ , M4 _~ L-- % & 4 ,..
.. & ='=? M s :_?_'+8 =.5-W. 2 s.8su, -a s = .
. . ._,: 3/~...,-
1 A L. .--. - - . . . y 4_ . ., ,r - -
& A. & EA d _^$ $65l5.] ' '
. . . . - ._. . ..... .. - ... . . - . _ . . .. . ,1+ b ] . . . .. . . . . . . - . .
m . aI r u fr.' S X ,< . & }d. A .XA<. _ .
= W % . m .; t We- - e = = * : ,= E ng x - AA -..
g , +A . J e. A A . A -A
~~
1 5
.rA .
?
1 4 . * . _ A. ' n ' .
& A ' } ^;, I a {: =. = y- A > f - 2 JJ V, -
-s . AA a > L- -' ., p. : .y . .a.:
w-
,_ s . -w___,
aj NA " w y ,: = l ? x l = ,' y jk = Y"
-. k .
______.___- - - ~- ~~ ~ ~~ l
__ _._ _ _ _ _ m _ _ _ .._ - _ _ _ _ ___-._m__ _. _. _ e ea
^
A ^ mAA A Aj & 3
&^
1 An=M't p (9c-m;r) n1A/ n A,e-- A A A At.n 1--- < a z4 d7
.A-L. .
/
l ap.xsi xa l y7"-1 - f- .Ak 2f, Nk .^ =[==A . 1 1 ! W - ^~, ? y -- , 2 i pla - p A =.. yam. n pA A h :., - ._
^ ,f^K y .Jr -
__ m_ Yh ) '- eA / - /g &
*M M w
A w a . - -_ __.
. a - m-- , s . s.
4 N'D HCGS FSAR 1/84 QUESTION 430.110 (SECTION 9.5.5) Figure of the FSAR shows a three-way thermostatic valve, labelled 5 connecting the jacket' water system and the intercooler water system. The FSAR states that both cooling water systems are self contained and closed loop systems. Describe the purpose of this valve, its size, and its mode of operation. (SRP 9.5.5,. Part III)
RESPONSE
The three-way thermostatic valve is a 1-inch valve. It's function is to temper the injector cooling water (i.e., that portion of the intercooler water used to cool the fuel injection nozzles) by mixing the hotter water from the jacket water system
' pqSegeT' with the cooler intercooler water. Th: rized '?:ter i: :: turned R:turnin; th: ;;t;; t; th;
- the j chet '?:ter :: incier terh.
j::h:t ;:t:: :nd 'nt;;;;;;;; a:t;; ;ystea; is ecceap;.;hel ... th. ;; pectic; ;;;p :::;; lin::. 4At: fel : d;;; :t ;-= van * *ke j h:t :t:; ,.t.a and the i=*=rcaa1=e ==*=e -fete- frem crer tin; indep;nd:ntl/. See re peace t: 0.;; tion 120.'!? fer-rdditirn:1 i.. w.m-Lien. 4 I
)
s . ' /g 430.110-1 Amendment 4
4 I O
. /usetr 430.//0
&~ AAL
_ %$ f . . . . Oh E We b l l
-$M .
pqm a s l A p.z2s>ag /4 NN D za x na a P' M Y #$ " ; a e4 Y + -- + < f b MA J x4-,s MA s AM f . 13 1 _ - . _ . . . _ -__-o
l*. $ sao.no ~ + f gjailure of the t'hermostatic valve at either position woulb be of little c6nsequence. The nozzle cooling water could not get colder than the inter-cooler water nor hotter than the jacket cooling water. The nominal spread in these temperatures is not sufficient to cause a problem in the injection system. . M his valve does not prevent the [operatingindependently. See response jacket water to Question system 430.113 for the intercoo and additional information. e G S e e 3/3
, , , , - - - . + ~ , , , - . - - ,,-,-.+=-----,,--=,-e--,,...
i l l-l HCGS FSAR 1/84 7 dnueTI y 30.111'g(SECTION 8.3 s 9.5.5) The diesel generators are required to start automatically on loss of all offsite power and in the event of a LOCA. The diesel generator sets should be capable of operation at less than full load for extended periods without degradation of performance or reliability. Should a LOCA occur with availability of offsite d A o diesel generator / running in an unloaded (standby) power condi 2ionforanextendedperiodoftime,shouldnotresultin degradation of engine performance or reliability. In Section 9.5.5.1 ce the FSAR you state that the diesel generator should " remain operational after 8 hours of no-load operation, provided that the SDG runs up to a minunum of 25% of full load for 1 hour ismediately after such no load operation." Verify the following:
- a. Verify that the statement conforms to the manufacturer's recommended no-load operation for this diesel or justify non-conformance.
- b. Verify that the conditions for no-load operation will be included in the plant operating procedures.
(SRP 8.3.1, Parts II and III and SRP 7.5.5, Part III)
RESPONSE
- a. In conformance with the manufacturer's la test l recommendation, Section 9.5.5.1.e. has been revised to state t that the diesel generators should remain operational after l b / hours of no-load to a minimum of operation, 50% of full provided load for I that hourthe SDG runs up immediately j after such no load operation. j
- b. The conditions for diesel generator no-load operation will be included in OP-SO.KJ-001(O), Emergency Diesel Generator Operation. Available January 1985.
$ ': LA$ A -
^ '
A~j'~')f &b . Q Obe % \ y..& M, JN &A % 11,19 W , y . 3.o-* . (tplh1+h!) I 430.111-1 Amendisint 4
.~
l BCGS FSAR 4/84 l l 9.5.5 STANDBY DIESEL GENERATOR C90 LING WATER SYSTEN The standby diesel generator (SDGf cooling water system provides cooling water to the SDGs and is safety-related. 9.5.5.1 Desion Bases The design bases of the SDG cooling water system are as follows:
- a. Cool the engine cylinder jackets, turbocharger, combustion air, generator outboard bearings, speed governor oil, and the lubricating oil sufficiently to permit continuous operation of the SDG at full load
- b. Maintain the jacket coolant in a warmed condition while '
the diesel engine is in normal standby status to promote reliable starting
- c. Ensure that the single failure of any active component f will not affect the operation of more than one SDG
- d. Remain functional during and after a safe shutdown earthquake (SSE) 12,
- e. Remain operational af terkhours of no-load operation, provided that the SDG ruhs up to a minimum of 50% of full load for I hour immediately after such no-load operation
- f. Permit testing and inspection of active system components during plant operation ,
- g. Withstand wind, tornadoes, floods, and missiles.
~
l The SDG cooling water system is designed to Seismic Category I l requirements and complies with IEEE Standard 387. The quality j group classification and corresponding codes and standards that 1 apply to the design of the system are discussed in Section 3.2. j 5.5-s2 am.nda.nt 5
HCGS FSAR 6/84 nzspons: - resere A : i am ~~ Qhe250gallonlubeoilmake-uptankisprovided I insv .sii - aarts, one upper and one 1 algae Insert 8 : growth is detMW - is the ogE'Se a b_t; w make up minate tanks the a algae lube oil ad event further growth.
....-d": 1_ -- og shAp canxs will be ,
M uring each refueIIM ::t=7-_ l 1
- b. The standby diesel generator lube oil make up tank material is carbon steel, SA 515 GR. 70. The exterior of the tank is coated using Colt Industries standard protection system. The system consists of a primer of Gordon Bartells 13409, yellow, and a finish coat of Gordon Bartells 14-811, suede grey, both applied according to the paint manufactures recommendations.
The interior of the tank is not coated because the lube oil is non-corrosive.;nd th: t:nh i: ::;erted te be l ddet t. Leistein;d in th; f;11 ;;nditi;n. p J j
- c. The vent and emergency pressure " relief vent pre terminated indoors, directly above the tank. The fill line is routed to the outside (west) of the auxiliary -
buildin grade.' gTheat elevation 105 feet line is capped and 0 inches, has 3 feetclosed a normally above isolation valve located in the building to prevent water from entering the liner It is not protected from missiles and tornadoes because it is not safety-related. .
- d. The lube oil makeup tank bottom is hemispherical. The line to the diesel generator sump is approximately 1.75 inches above the bottom of the dish. Should there be any carry over into the transfer line, it would be trapped in the strainer and/or filter before entering the engine sump.
A normally closed drain valve is provided at the low point of the tank, reference Figure 9.5-27. The drain valve will ed in_accordance w}th plant operating procedure ele _terious/to remove anWsediment, water or other material tna N y accumulate in the bottom of the tank. concerns will be addressed by July, 1984:
- a. Description o e rote l ,ggeve ;
- b. Effects of a ion l ee% og c44rel to -tb tow oi ( ** * *
. -i l .
- F 430.135-2 Amendment 6 3' f
y . -
- l. . . _ . Insert A . . .. . . . . .
. .. .. ... ~.
. . . - _ < ~ . . . . . . . . - ~ . . . .
F
. _ . _ . _ Deleteribus matercal 'ss prevented from enterin$ the . .
.. _ diesel engine. lube, oil maka-up tank by: . . . .
- i. Procur/In3 b'igh quality, high purity tube oil w'ith p lubricatAg properties : requ H
~
i
'n i accordanc.e with the vnanufacturerb .
re c.o mm endations.
- a. Insuring that addihre filling operations to increase make-up tank \evel are. perforae4 throgh the instded basket strainer in the Viu We..
The. \u.W. ou make-up tank conservahon vent perdih to.nk. vewting when regu el and prekibits aie loorne.
.imporWes f rom contiavous\3 en+ena3 the to.nK.
Mave.-u.p +.aak, Outg su be a.cco.aptiskd in a.cce< dance wi% a. y;rtten y.c.aoc , A e.ea+< ned c.q d t.he. precedure sti lbe peste.A in tke vacini+3 of t.ke lolse o'il fill be .
.The luime .it M\ he. w;u %e \ableJ to ide.n6f)
[the fin \ine. conneetan .pargese and a refren.u:e to lthe. appuc lele procedu.re. l l I l l
I 1 Ivi s ert B .. _ .. _ _ .. - _ . _ . . _ . _ . . _ . _ _ . _ _. .. .. a, A1ga e format %. may .oecar.. Aue .A. .. condensate. accumulation in the. make.-up tu.be. .od bank. Prior to diesel erv$6e. oper bibh 4:.eshh3 he lu.be ei\ make-up to.nK drain _will . 7 he sped *- opened to. remove any .Waterj sediment; algae or other de\e.tenous m4+evi=I._Tf tube. oil puri+3 is degraded any of ttee followin3 1N/ can be imp \emen+ed to ces hee labe, oil pu.ci h in %<. make-u.p ta.n k: ..
- 1. Att dele +e<&s ma+e<(al ma$ be. nuovec\ by 6caisin3 \ v he o\ \ theou.9 % % e. a <a i n \ in e..
- 2. The iube. dil make -up te.oK can be. drained, clea.ned a.od re4 died with fresk tube. oil, l 3. A chewiica\ adchWe. can be a.dded to remove.
alp <, or o%er biologic t growth if advised b3 a, Tribde33 spec.ialtst. l
-6e
_-x . - _ _ - _ _ _ - -
HCGS FSAR 1/34 QUESTION 430.149 (SECTION 9.5.8) . Figures 1.2-35 through 1.2-39 show the routing of the diesel engine exhaust system from the diesel generator room to the roof of the auxiliary building. The figures show that the exhaust mufflers for all the diesel generators are located in a common corridor (Elevation 102'-0") and that the exhaust stacks pass through the following areas:
- 1. Remote D/G control and vital switchgear areas (elevation 130'-0")
- 2. Vital battery control rooms (elevation 137'-0")
- 3. Switchgear HVAC Area (elevation 163'-0")
- 4. Diesel and control rooms HVAC area (elevation 178'-0")
The exhaust system is considered a high energy system by virtue of temperature. A exhaust system pfpe break in any one of these areas and a single active failure in one of the other diesels or just pipe break in the exhaust system in the muffler corridor, switchgear HVAC area, or diesel and control room HVAC area could i result in an inability to shut down the plant., The figures referenced above do not clearly show or decribe the diesel engine exhaust stack enclosures. Describe the stack enclosure in each of the areas noted above and show that an exhaust stack break in any one of these areas will not result in the inability to shut down the plant or result in failure or unavailability of all, diesel generators. (SRP 9.5.8, Parts II and III) RESPONSE - As discussed 15 response to Question 430.82 the SDG combustion air exhaust system is not classified 3s high energy system. Therefore a high energy pipe break # not considered. The exhaust stack which passes through the areas mentioned in the l, above question is designed to Seismic Category I requirements, as discussed in Section 3.7. It '_: = ' -- - :--i f : f -!
- h db-188887~
l 2.in::.' :rtiti:n p;;.1. as shown on Fig'ures 1.2-35 through 1.2-39s> *-*4 G30,M94j .to minimize heat rejection and noise in the areas through which it passes, j l 1 430.149-1 Amendment 4
- 1[3
woo.nr I "I . dA9AMj=~L x A r- s M i p -_- 3 3 ~ W- ,A . - _ _ - -
- n._
- _n l
,jf s - __A_>!r, Q M & ,
1 i M .xL ,r 2 J .X4u D ; 1
< l xA /NA . . . !
)
1 l 1 I 23
, ,-_, .., , . , .-.,. a ,.-. ,, ., . - . , . . - . .
" I 4
i 4, :1 ... I: l_j, q
,,, / ,
y f !!
/c
=
8 JF W ,y s l
!s
$I I
l I ! li i
=i g l1 jt,8 --
ii 2 1 L , yA *. -
<e mi ,
! 4 '
' ~
;2
_ t e.g e 1 ai c m
, +~
g c c 2 ,e s _pr,a
@ j I' M..- ! !.e g h O
- ul
! I@ 1 ,
.g e-*yTE G e[, %e11 'l O-*
eT j ] {u..;l]... i I a y [L.3 @ ;{ dg e-c.2.....L ki o i . t .. , s "T~1rel P " T e~1 y@v!!g t ;yd @ ! l 8 i - A, g n..i .!n i As @i I f s _e
.. 1 g_f.,.3,.,.,, a3 i ra .
e3FW :_ .._... i , i 1 gL ~ O~~T33 0 ;t -k ,1 l3 f g
<F-- I I ~ ~M i
. i,i),!l!3pisj lls -
We Gio.~ = la la 1 ma_Ir *
!- {
' VG~l
. e r a.Tr m'iN . T -
<, ! Cu .
g* i s - i g @' 5 r[s i;h'h .y
! D1 f I li
- . u ~-
i
. u. t
. 1 t
I [ WI e-l l t gl l i g it - , l e- 5 l i f_ l l 3 g.k e, s- f!!gy,r-la
}. w'k ' ]l i .
in. .
; E'y ~ Hg 1. ,, g (
c' 1
, } l 1
f. 19 r i.
" "3 it
't h {y 14 l ' ii J$
f g} I ()
* 'I '
y A ~ k))
-T ! l luf
- w ,
y : ,
;5_.;n:k.
3 C
..r-
. 5 .'- -G.
,l l
,a -
I .1\ [ -1 'g 4' f
" -E I, M] Q g N
7 g%5 i lff i @y, g g '
~ '
hh g@,p,ll j ' Pe
!r
_3,g,,"!3 a i uw 4
%:, m . . .
m 2- 2 i trp
&, l
.- _- 7ll- m
-- - + --. i g
5 61
'1,ky 1 31, "2-g
.T 83 jl h*
'l' [ [
lj f l] lj . - . .r: - ( . y G)
IICGS FSAR ifgg 00ESTION 430.151 (SECTION 10.2) Espand'your discussion of theadditional Provide turbine speed control and explanation of overspeed the turbine protection system. and generator electrical load following capability for the turbine speed control system with the aid of the system ' schematics (including turbine control and extraction steam valves to the heaters). Tabulate the individual speed control
- protection devices (normal, emergency and backup), the design speed (or range of speed) at which each device begins operation to performs its protective functionIn(in orderterms of percent to evaluate theof normal turbine operating speed). 1 l
adequacy of the control and overspeed protection system provide l schematics and include identifying numbers to valves and mechanisms (mechanical and electrical) on the schematics. l Describe in detail, with references to the identifying numbers, l the sequence of events in a turbine trip including response times, and show that the turbine stabilizes. Provide the results ! l of a failure mode and effects analyses for the overspeed protection systems. Show that a single steam valve failure (SRP l l cannot disable the turbine overspeed trip from functioning. ; 10.2, Parts II & III) M 10.2.2.5 and 10.2.2.6 have been revised to inel ion except for a failure rotecti . Such an analysis analyst ~ iles generated by a failure is not necessary ' y a low probability of the overspe on s safety-related systems; ile. analysis of af ussed in Section 3.5.1.3. See the response 430.152. , i 430.151-1 Amendment 4 ,
--- fk
)
RESPONSE ; Sections 10.2.2.5 and 10.2.2.6 have been revised to include the i requestedinformation.3 G Ag,, p ganggasstatues g overspeed ases and effects protection system.analysis has not been preeered a ainssum et two independent lines of eerense as Mgggem et fatture
,,g ,,e ser protection against overspeed and that no single of ear devise er steam volve een disable the turbine overspeed trip Gro Sumetiendag.
J,.dddeh jturbine missilies generated by a failure of the overspeed protection systems would have a low probability of affecting any safety-related systems; turbine missile analysis i 10 discussed in Section 3.5.1.3. See the response to Question 430.152. N _f 7 .,i 1
- 76f d e ,s fi y a f d a ,e e .s e/n ~, m 7 f 4 M .;1 - / , \
4,,3 cwVA /Ae a /./ k ,, a f Ta /d so. y ~)*f)
\
3-6 e
-. Ylo
1/04 HCGS FSAR Rapid closure of the control valves or stop valve closure initiates an input signal to the reactor protection system (RPS) to initiate reactor shutdown. . 10.2.2.6 Overspeed Protection h lvner4(2) a9* To protect the turbine-generator against overspeed, when the turbine speed begins increasing, the EHC system will rapidly throttle the control valves and the intercept valves. If the speed continues to rise, the main stop valves and the intermediate stop valves will be closed by one of the following trip devices:
- a. A mechanical overspeed trip that is initiated if the turbine speed reaches approximately 10% above rated ,
speed
- b. Electrical overspeed trip that serves as a backup to
! the mechanical trip and is initiated at approximately l 12% above rated speed. - The mechanical overspeed trip device (Figure 10.2-9) is an . unbalanced ring mounted on the turbine shaft and held concentric ; by a spring. When the turbine speed reaches the trip speed, the l centrifugal force acting on the ring overcomes the tension of the I spring, and the ring snaps to an eccentric position. The ring then strikes the trip finger, which actuates the mechanical trip valve. This three-way valve feeds hydraulic fluid at 1600 psi to the lockout valve. When tripped, this valve blocks the hydraulic fluid supply system and releases the emergency trip system i pressure, which causes the main stop valves, control valves, and combined intermediate valves to close. Failure of the hydraulic portion of this trip results in the closure of main stop valves, i control valves, and combined intermediate valves. Failure of the ! normal turbine speed control system will not prevent the turbine overspeed control system from shutting down the turbine. The electrical overspeed trip receives its signal from a 112% speed trip relay (VCS 840, Figure 10.2-10) operated by the signal from a magnetic pickup, through a magacycler and a voltage comparator (Figure 10.2-11). u The signal from the speed relay energizes the master trip relay I XKT 1000 (Figure 10.2-11) which then energizes the mechanical 10.2-8 Amendment 4 _. Y(,
..__ ___ _ _ _ ~ _ _ _.
HCGS FSAR 1/84 I trip solenoid (NTS) and de-energizes the master trip solenoid b valves MTSV-A and MTSV-B which removes the emergency trip system ! pressure causing the turbine valves to close. Loss of either signal or hydraulic function of this trip results in a main stop valve closure.
\ ewAG ->
When the mechanical overspeed trip is being tested, using the overspeed governor lockout device, the electrical overspeed trip protects the turbine against overspeed. An additional feature of the protective system that will minimize the likelihood of an overspeed condition is the power / load unbalance circuitry (Figure 10.2-12). Generator load is sensed by means of three current transformers and is compared with the turbine power input which is sensed by the turbine intermediate pressure sensor. Control valve action will occur only when the power load unbalance is approximately 40 percent or greater while the generator current (load) is lost at a rate equivalent to going from rated to zero in approximately 35 msec. or less. There are four steam lines at the high pressure stage. Each line is provide with one stop valve in series with one control valve. Steam from the high pressure stage flows to the moisture i separators and then to the three. low pressure stages. Each of i
'he six low pressure lines has a combined intercept valve that consists of a stop valve in series with a control valve, in one housing. All of the above valves close within 0.2 seconds on turbine trip. Assuming a single failure within the above system of 20 valves in case of a turbine overspeed trip signal, the j turbine will be successfully tripped. l The diversity of devices shown on Table 10.2-1 ensures that l l stable operation following a turbine trip proceeds from the l requirement that both the stop valves and the combined intercept valves close in a turbine trip, thereby preventing steam from the l
l main steam line from entering the turbine and preventing the i expansion of steam already in the high pressure stage and in the moisture separator. An additional provision is made to
- automatically isolate the major steam extraction lines from the turbine by power-assisted check valves. Closure times of the i check valves will be in accordance with the turbine manufacturer's recommendations.
< Any postulated accident, including the effects of high or moderate energy pipe failures, that results in a loss of
- hydraulic pressure or loss of the electrical signal to the
! 10.2-9 Amendment 4
~~
Yb
I-+@ l 9Alhah. % 7each up p +> A > nM a g nd Q M & w y)-Ay n.f l Q'uk 4e y #e & Q a >u-M % 4 % wk.ih w sy AzGLtd, a u ua p m p1 4
@m
~
6d,-e q Ac rw-la+ @ l 9 R L he g A de
- A A-
;% csua.
n%un.:u ad 6 a gi 9 a Qa w m L tr.& 4 g .7k4 4 A 1. A gM w I s w-A v - --
%.e &dua4
_a n , a va n p ., S YQ r 9, a aa+ r n y I 4P
- W *Y *'ti-e l
h
T Af 8 4 g s o . A -j p;r 2,. ~ - Tc /' 7B/.rs' orsq.stric PRWEcrias,<
/l;, pAe f.as a y .,.,.s f n e rp ee f a ,,,i,,;.,,,,, a f ale f<.,s e
. p. /op ed. 74 <.
f.,, s,,/p a ,4.,t' /. sr.a z . f . e s e d.'
/;/s n s...,y ,e d,,,4,,e,e.s. an
//.,, / ,e , - Aae fr -. p
. Co.,7% / V r /re,.y
/e />p, , , , , , ~r/cy. -
c . , 4 ,.a l* /d s A,o f cos, fro / /C,efio.
~4. . Ze A > ee ,of sne /s e,s ~
3 l Co frv/ ~ Ove *el 7.+/fo f B ec /r-ap. t;?h r+ y ee d /J 3,<a re e 3 v v.// so,..,fr+ f ~h a=/
- 9. 7 ,.y E.A., . / tr. /*e s ~ 19.Mhs f aGesses..*w/s3e,, ia.c i/ .cxa.,,e -s.r.,, ,t / .p s./a,,,;.,$y C. ,/>) e c A , e / 7. ~ p S a e l7 y . E Ac f. n e r / Yr-f
(,. . /% 7' ,, ;t /, ,9 .c-/ , ..- c/ <~ 4
- a - A-</r ya: 4;'< b~ /a R"d T> .y sp c/,m En adNtism, esos destures are useds 6 mfE" made of spor tian of all volve. If hydraulic pressure is Best, all to sias selves will class.
Stuer/Emed Whalanes to reduce overspeed se less of hi$ leads. spress easses amarseuse ensam velve. tin .ir .i.e. 4t** d Co /~ / O <;t c.,,/4 h . reaf ,, 4 ,, i es' e ,,,.g , 64 m _._--..----.i,---__,__,..,__,-,_ , , . , - - , - .,.,--.,-,,,,-.-,_.--..---..--_w, -
, , , , , . , ,.~.. .,.,-,r----
i RCGS FSAR 1/34 . ! i 00ESTION 430.161 (SECTION 10.4.4) l I: Provide the results of a failure mode and effects Jnalysis to l determine the effect of malfunction of the turbinf bypass system '~
,, including controls on the operation of the reactor and matn tuttilne generator unit. (SRP 10.4.4, Parts II & III).
i
RESPONSE
As discussed in the response to Question 430.166, the failure ' mode of the bypass valves is closure of the valves. The offects ?- of this malfunction on the operation of the reactor is summarized in Table 15.0-1 and discussed in Sections 15.2.2 and 15.2.3. The ' ' effect of this malfunction on the turbine generator is discussed in the response to Question 430.166.
- i
~%c, g fMMI[ ' fN '
dus., to 'o ce M *
. o.
"'". V ,
. M a ,a y a L u a ~ ~ h s** k b u c ' 1 as.
- a. . _ u r.
ms y.... .. y _g '
- p. The bypass valves would move to full open position causingg;;;;:1 h y,,i; n::: :$:; the consequent increase in coolant voids would cause the vessel water level to increase.
The pressure regulation system, sensing the pressure reduction, would cause movement of the turbine control valves to reduce turbine steam flow so as to maintain the pressure. [Then,therearetwopossiblescenarios: l l 1. If the water level swell were large enough to cause a high water level (L8) turbine trip, then the remainder of the event would be slatlar to - and the consequences would not be worse than - the , transient caused by " Pressure Regulator Failure - Open," as described in 45A2 Section 15.1.3.
- 2. If the pressure regulation systas can gain control soon enough to !. '
prevent the LS turbine trip, then the turbine control valves would settle to a position corresponding to about 755 MR steam flow through turbine control valves and 25K WR steam flew through f/ , 7t . i ! turbine bypass valves, with no appreciable effect on the reacter. If the tuttine bypass misoperetten cannot be readily corrected, the ! l reacter operater would take appropriata action, which could include !, reacter shutdesaw , 1 l}}