ML20092C762
| ML20092C762 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/15/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8406210315 | |
| Download: ML20092C762 (102) | |
Text
, _ _
Putsc Servce C-l PS G Company E'ectnc and Gas 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation June 15, 1984 Director of Nuclear Reactor Re2ulation 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 Draf t 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 s.aff. 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 Attac hment 4) are the resolutions to those Draft SER open items listed in Attachment 3.
Should you have any questions or require any additional information on these open items, please contact us.
Very truly yours, 4
Ip j /
B406210315 840615 PDR ADOCK 05000354 E PDR 300l The EAtrtg3cmae;dts Ip 1
% 4112 (3M) 4 84
Director of Nuclear Reactor Regulation 2 6/15/84 C D. H. Wagner USNRC Licensing Project Manager W. H. Bateman USNRC Senior Resident Inspector FM05 1/2
DATE: 6/15/84 ATTACHMENT 1 DSER R. L. MITIL TO OPEN SECTION A. SCHWENCER ITEM NUMBER SUBJECT STATUS LETTER DATED Sared 2.4.5 Wave impact and runup on service Complete 6/1/84 water intake structure 7b 2.4.11.2 Thermal aspects of ultimate heat sink C m plete 6/1/84 9 2.5.4 Soil damping values Cmplete 6/1/84 10 2.5.4 Foundation level response spectra Cmplete 6/1/84 11 2.5.4 Soil shear moduli variation Cmplete 6/1/84 12 2.5.4 Cmbination 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 bottom 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 on Complete 6/1/84 Vincentown 17 2.5.4 Power block settlement nonitoring Cmplete 6/1/84 18 2.5.4 Maximum earth at rest pressure Complete 6/1/84 coefficient 19 2.5.4 Liquefaction analysis for service Cmplete 6/1/84 water piping i 20 2.5.4 Explanation of observed power block Complete 6/1/84 settlement 21 2.5.4 Service water pipe settlement records Cmplete 6/1/84 22 2.5.4 Cofferde stability Cmplete 6/1/84 23 2.5.4 Clarification of FSAR Tables 2.5.13 Cmplete 6/1/84 and 2.5.14 24 2.5.4 Soil depth models for intake Car $lete 6/1/84 structure 27 2.5.5 Slope stability Cmplete 6/1/84 M P84 80/12 1-gs Page 1 of 5
ATTACHMENT 1 (Cont'd)
DSER R. L. MITIL TO A. SCHWENCER OPEN SECTICN NUMBER SUR7ECT STATUS IffrER DATED ITEM 3.5.1.2 Internally generated missiles (inside Ccruplete 6/1/84 30 containment) <
3.8.2 Steel containment buckling analysis Ccunplete 6/1/84 41 3.8.2 Steel containment ultimate capacity Cmplete 6/1/84 42 analysis 3.8.2 SRV/IDCA pool dynamic loads complete 6/1/84 43 3.8.3 ACI 349 deviations for internal Cmplete 6/1/84 44 structures 3.8.4 ACI 349 deviations for Category I Cmplete 6/1/84 >
45 structures 3.8.5 ACI 349 deviations for foundations Ccrnplete 6/1/84 46 3.8.6 Base mat response spectra Ccunplete 6/1/84 47 3.8.6 Rocking time histories Ccunplete 6/1/84 48 3.8.6 Gross concrete section Complete 6/1/84 49 3.8.6 Vertical floor flexibility response Cmplete 6/1/84 50 '
spectra s
3.8.6 Design of seismic Category I tanks Cmplete 6/1/84 53 3.8.6 Cmbination of vertical responses Complete 6/1/84 54 3.8.6 Torsional stiffness calculation Cmplete 6/1/84 55 3.8.6 Drywell stick nodel develogrnent Complete 6/1/84 56 3.8.6 Rotational time history inputs Ccrnplete 6/1/84 57 3.8.6 "O" reference point for auxiliary Complete 6/1/84 58 building nodel Overturning m ment of reactor Ccunplete 6/1/84 59 3.8.6 building foundation mat ,
BSAP element size limitations Cmplete 6/1/84 60 3.8.6 Seismic modeling of drywell shield Ccrnp1'ete 6/1/84 61 3.8.6 wall M P84 80/12 2-gs Page 2 of 5
ATTACHMENT 1 (Cont'd)
DSER R. L. MrITL 'IO OPEN SECTION A. SCHWENCER ITEM NUMBER SUR7ECT STA'IUS IEITER DATED
'62 3.8.6 Drywell shield wall boundary Ccmplete 6/1/84 conditions ,
- 63. 3.8.6 Reactor building dame boundary Complete 6/1/84 conditions 64 3.8.6 SSI analysis 12 Hz cutoff frequency Comp'lete 6/1/84 65 3.8.6 Intake structure crane heavy load Couplete 6/1/84 drop 67 3.8.6 Critical loads calculation for , Cmplete 6/1/84 reactor building dame , ,
68 3.8.6 Reactor building foundation mat Ccmplete 6/1/84 contact pressures 69 3.8.6 Factors of safety against sliding and Cmplete 6/1/84 overturning of drywell shield wall 70 3.8.6 Seismic shear force distribution in Conglete 6/1/84 cylinder wall 71 3.8.6 overturning of cylinder wall ,, Complete 6/1/84 72 3.8.6 Deep beam design of fuel pool walls Conglete 6/1/84 73 3.8.6 ASHSD dczne model load inputs Conglete 6/1/84 74 3.8.6 Tornado depressurization Conglete 6/1/84 75 3.8.6 Auxiliary building abnormal pressure Complete 6/1/84 76 3.8.6 Tangential shear stresses in drywell Cmplete 6/1/84 shield wall and the cylinder wall 77 3.8.6 Factor of safety against overturning Cmplete 6/1/84
- of intake structure i
.78 3.8.6 Dead load calculations Complete 6/1/84
- . 79 3.8.6 Post-modification seismic loads for Couplete 6/1/84 the torus 80 3.8.6 Torus fluid-structure interactions Cmplete 6/1/84 81- 3.8.6 Seismic displacement of torus Conglete 6/1/84 t
M P84 80/12 3-gs Page 3 of 5
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pTTlCHMENT 1 (Cont'd)
DSER R. L. MITTL TO OPEN SECTION A. SCHWENCER ITEM NUMBER SUR7ECT STATUS LETTER DATED 82 3.8.6 Review of seismic Category I tank Ccanplete 6/1/84 design 83 3.8.6 Factors of safety for drywell Complete 6/1/84 buckling evaluation 84 3.8.6 Ultimate capacity of containment Complete 6/1/84
'(materials) 85 3.8.6 Load combination consistency Complete 6/1/84 !
92 3.9.2.2 Triple flued-head containment Complete 6/15/84 penetrations .
95 3.9.3.2 Fatigue evaluation on SRV piping Complete 6/15/84
. and LOCA downccmers 96 3.9.3.3 IE Information Notice 83-80 Ccinplete 6/15/84 98 3.9.3.3 Design of bolts ccinplete 6/15/84 99 3.9.5 Stress categories and limits for Ccunplete 6/15/84 core support structures 110b 4.6 Functional design of reactivity Ccuplete 6/1/84 !
control systems 124 6.2.1.5.1 RPV shield annulus analysis Ccznplete 6/1/84
- 125 6.2.1.5.2 Design drywell head different.ial Ccznplete 6/15/84 pressure 129 6.2.2 Insulation ingestion Ccuplete 6/1/84 132 6.2.4 Containment isolation review Ccinplete 6/15/84 134 6.2.6 Containment leakage testing Ccunplete 6/15/84 141g 9.1.3 Spent fuel pool cooling and cleanup Ccanplete 6/15/84 system 145 9.2.2 ISI program and functional testing Ccrnplete 6/15/84 of safety and turbine auxiliaries cooling systems 146 9.2.6 Switches and wiring associated with Ccznplete 6/15/84 <
HPCI/RCIC torus suction M P84,80/12 4-gs Page 4 of 5
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i ATTACHMENT 1 (Cont'd)
DSER R. L. MITTL TO ,
OPEN SECTION A. SCHWENCER ITEM NUMBER SUILTECT STATUS LETTER DATED
) 152 9.4.4 Radioactivity monitoring elements Cmplete 6/1/84 154 9.5.1.4.a Metal roof deck construction Caplete 6/1/84 classificiation ;
158 9.5.1.5.a Class B fire detection system Complete 6/15/84 159 9.5.1.5.a Primary and secondary power supplies Ccmplete 6/1/84 for fire detection system ,
161 9.5.1.5.b Fire water valve supervision Complete 6/1/84 [
.162 9.5.1.5.c Deluge valves Cmplete 6/1/84 1 163 9.5.1.5.c Manual hose station pipe sizing Ccmplete 6/1/84 164 9.5.1.6.e asmote shutdown panel ventilation Cm plete 6/1/84 165 9.5.1.6.g Emergency diesel generator day tank Conglete 6/1/84 ;
protecton j i
174 13.5.2 Resolution explanation in FSAR of Cmplete 6/15/84 IMI Items I.C.7 and I.C.8 182 15.9.10 TMI-2 Item 'II.K.3.18 Cmplete 6/1/84 185 7.2.2.2 Trip system sensors and cabling in Complete 6/1/84 turbine building 190 7.2.2.7 Regulatory Guide 1.75 Cmplete 6/1/84 -
193 7.2.2.9 Reactor mode switch Cmplete 6/1/84 194 7.3.2.2 Standard review plan deviations Complete 6/1/84
! 197 7.3.2.5 Microprocessor, multiplexer and Complete 6/1/84 !
cenputer systems
(- p 200 7.4.2.2 Remote shutdown system Conglete 6/1/84 205 7.5.2.4 Plant process computer system Cmplete 6/1/84 .
' 209 7.7.2.3 Credit for non-safety related systems Complete 6/1/84 -
in Chapter 15 of the FSAR I i
210 7.7.2.4 Transient analysis recording system Complete 6/1/84 ;
i 218 9.5.1.1 Fire hazards analysis Complete 6/1/84 r i
TS-3 4.4.5 Core flow monitoring for crud effects Complete 6/1/84 l t
14-1 4.2 Fuel red internal pressure criteria cmplete 6/1/84 JSags !
.M P84 80/12 5-gs Page 5 of 5 ,
ATTACHMENT 2 DATE: 6/15/84 DRAFT SER SECTIONS AND DATES PROVIDED SECTION DATE SECTION DATE 3.1 3.2.1 11.4.1 3.2.2 11.4.2 t 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 8.2.4 1.3. 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 8.4.8 15.2.7 9.5.2 15.2.8 9.5.3 15.7.3 9.5.7 17.1 9.5.8 17.2 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-
ATTACHMENT 3 DSER' OPEN SECTION ITEM NUMBER SUBJECT 92 3.9.2.2 Triple Flued-Head Containment Penetrations 95 3.9.3.2 Fatigue Evaluation of SRV Piping and LOCA Downcomers 96 3.9.3.3 IE Information Notice 83-80 98 3.9.3.3 Design of Bolts 99 ,3.9.5 Stress categories and limits for core support structures 125 6.2.1.5.2 Design drywell head differential pressure 132 ' 6.2.4 Containment Isolation Review 134 6.2.6 Containment Leakage Testing 141g 9.1.3 Spent fuel pool cooling and cleanup system 145 9.2.2 ISI program and functional testing of safety and turbine auxiliaries cooling systems 146 9.2.6 Switches and wiring associated with HPCI/RCIC torus suction 158 9.5.1.5.a Class B fire detection system 174 13.5.2 Resolution. explanation in FSAR of TMI Items I.C.7 and I.C.8 ,
M P84 80/13 1-mw
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t HCGS DSER Open Item No. 92 (DSER Section 3.9.2.2)
TRIPLE FLUED-HEAD CONTAINMENT PENETRATIONS Additional information is needed to address the design considerations used tor the triple-flued head containment penetration.
RESPONSE
For the information requested above see response to Question l- 210.37.
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HCGS FSAR 6/84 QUESTION 210.37 (SECTION 3.9.2)
In FSAR Figure 3.8.6, a Type A triple flued head containment penetration with bellows is shown for those cases where thermal expansion is to be accommodated. Provide a list of those systems and the penetration identification numbers where these triple l flued head containment penetrations ha'fe been used. Describe the design considerations applicable to those penetrations and specifically address how fatigue, torsion, and stiffness modelling in piping. analyses we.re considered.
RESPONSE
The bellows used in Type A triple flued head containment penetrations are designed, fabricated, tested, and examined in accordance with the requirements for Class 2 c~om'p onents of ASME B&PV Section III Code.
Non-NSSS :
The list of non-NSSS systems and their penetration identification numbers that use a Type A triple flued head are shown in Table
! 210.37-1.
The design considerations for Nuclear Class 1 flued heads consist
\ of evaluation of the loads transmitted to the flued head by the l
i piping from both sides due to
- a. Thermal expansion
- b. Seismic r'eactions
- c. Dead weight loads
! d. Internal pressure
- e. Dynamic loads
- f. Thermal gradient effects through the flued head body
- g. Thermal transient effects as a result of temperature and pressure changes in the system
- h. Discontinuity effects resulting from dissimilar metal welds, if any l 1. Fatigue analysis using cumulative usage approach (NB-3653.5 of Section III).
DSER OPEN ITEM 95l 210.37-1 Amendment 6
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i Nuclear Class 2 flued heads are evaluated to the loads listed ;
above with the exception of items f., g., and i.
The Type A triple flued head containment penetrations are [
anchored to the building steel as shown in revised Figure 3.8-6.
In the connecting piping analyses, the flued head is considered a rigid anchor. Piping reaction loads (forces, Fatigue bending moments i is considered j and torsion) are evaluated as stated above.
per item (i) above and includes evaluation of the flued head and the butt weld between the flued head and the process pipe.
l NSSS:
The main steam piping and the head fittings are designed and f abricated to the requirer.ents of the 1971 edition of Section III for the ASME B&PV Code with addenda through and including those of Summer 1972. The main steam head fittings are analyzed to the i
requirements of NB-3200 and the 1977 editions of Section III of the ASME B&PV Code and are evaluated to more restrictive stress limits of BTP MEB 3-1 in SRP 3.6.2. The design report for main 1 ,
steam head fittings includes the evaluation of fatigue and the . r r
effect of pipe rupture loads. The head fitting is modelled as a pipe element with rigid stiffness, and its effect on the main l steam piping is evaluated to the requirements of NB-3600 of ;
j Section III of the ASME B&PV Code. .
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MCGS FSAR 6/84 5 TANLE 210.37-1 8 l IC TYPE A TRIFLE FLUID HEAD CONTAINMENT PENETRATIONS '
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- F-1A 2 107'-0" Main Steam 26" NSSS F-IN 107'-0" hain Steam 26" NSSS 2
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F-1C P-1D 107'-0" Main Steam 26" NSSS 2-[
113'-2" Feed Water- - - 24 Non-NSSS 1: (
F-2A 2I 113'-1" Feed Water 24" Non-NSSS -
F-2N 2' '
20" 106'-0" RNR Shutdown Cooling From RFV Non-NSSS P-3 1; 106'-0" RNR Shutdown Cooling Return 12" Non-NSSS
- F-4A 2:
' 106'-0" RMR Shutdown Cooling Return 12" Non-NSSS F-4N Non-NSSS 3 I
100'-9" Core Spray to Reactor 12" F-SA i 108'-9" Core Spray to Reactor 12" Non-NSSS 3 F-5N 106'-0" LPCI 12" Non-NSSS 3 F-6A ,
Non-NSSS 3:
106'-0" LPCI 12" F-6N i F-6C 106'-0" LPCI 12" Non-NSSS 3l 106'-0" LFCI 12" Non-NSSS 3[
F-6D 106'-0" HFCI Turbine Steam Supply 10" Non-NSSS 3 F-7 114'-0" Chilled Water From Drywell Coolers 8" Non-NSSS 3 P-SA 3, 114'-0" Chilled Water From Drywell Coolers 8" Non-NSSS P-SN 3, 140'-0" RWCU Supply 6" Non-MS$$
F-9 4' 148'-0* RMR Shutdown CLG to Reactor Kladspray 6" Non-NSSS F-10 106'-0" RCIC Turbine Steam Supply 4" Non-NSSS 4t F-11 103'-0" Main Steam Drain 3" Non-NSSS 4 F-12 110'-0" Chilled Water to Dryve11 Coolers 8" Non-NSSS 4 I F-38A P-38N 114'-0" Chilled Water from Dryvell Coolers 8" Non-NSSS 4 4i i
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HCGS DSER Open Item No. 95 (DSER Section 3.9.3.2)
FATIGUE EVALUATION ON SRV PIPING AND LOCA DOWNCOMERS The staff requires a fatigue evaluation be performed on the SRV piping and LOCA downcomers in the torus suppression pool.
4
RESPONSE
i- The response to this item has been provided in the response 3
to Question 210.42. .
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HCGS FSAR 6/84 QUESTION 210.42 (SECTION 3.9.2)
The safety relief valve discharge piping and downcomers are ASME Class 2 and 3 components. A through-wall leakage crack in these l lines resulting from f atigue caused by SRV actuations and small !
LOCA conditions would allow steam to bypass the pressure l supprescion pool. This could result We in an unaccep, table over-
, there fore , require that pressurization of the containment.
the applicant perform a fatigue evaluation on these lines in accordance with the ASME Class 1 f atigue rules and include the '
ef fects of pressure, moment, and thermal gradient loadings. The staf f will require that the results of the fatigue evaluation be documented in the FSAR.
RESPONSE ,'
i Fatigue evaluation of the safety relief valve discharge piping is discussed in Section 5-3.4.3 of the HCGS Plant Unique Analysis Report (PUAR) . The Mark I Owners' Group prepared and submitted a generic f atigue evaluation report to the NRC on November 30, The HCGS wetwell SRV piping is 1983, which applies 'to HCGS.
adequate for f atigue based on this generic evaluation.
The vent system, which includes the vent lines, vent header and downcomers, is ASME Class MC. A fatigue evaluation has been performed for the vent system. This evaluation was performed in accordance with the ASME Class MC fatigue rules which are identical to the ASME Class 1 fatigue rule s. PUAR, Sections 3-2.4.5 and 3-2.5.1 de scribe the f atigue analysis methods and results.
1 DSER OPEN ITEM 9J
- 210.42-1 Ame ndme nt 6
HCGS DSER Open Item No. 96 (DSER Section 3.9.3.3)
IE INFORMATION NOTICE 83-80 s The staff's review of Section 3.9.3.4 of the applicant's FSAR relates to the methodology used by the applicant in the design of ASME Class 1, 2, and 3 component supports. The review includes assessment of design and structural integrity of the' aupports. The review addresses three types of supports: plate and shell, linear, and component standard types. More information regarding the design and construction of ASME Class 1, 2, and 3 component supports is required.
The applicant should address its actions taken in response to IE Information Notice 83-80'. '
RESPONSE ,
i For the information requested above see response to Question 210.53.
r M P84 95/06 3-dh
C HCGS FSAR 6/84 QUESTION 210.53 (SECTION 3.9.3 )
Describe what actions have been taken to address the staf f concerns regarding stif f pipe clamps as described in IE Information Notice 83-80.
RESPONSE
The . applications of stif f pipe clamps on HCGS will be reviewed based on IE Information Notice 83-80.Section III of the ASME B&PV Code does not provide rules for evaluating stresses due to loadings from nonintegral attachments such as clamps; howsver, clamp-induced stresses will be evaluated by me thods consistent with the intent of the Section III of the ASME B&PV Code. The procedure will include the following: .
- 1. Identify the locations of "stif f" clamps installed on ASME Section III Nuclear Class 1 piping systems.
- 2. Identify the types of clamps, the loads acting on the clamps and the bolt pro-load values used in their installation. In piping stressos due to all loading conditions at the locations of stif f clamps will also be identified and reviewed.
, 3. Add the primary membrane and bending stresses caused by the snubber load being transmitted to the pipe through the clamp to the stresses caused by internal pressur.o and bending computed by equation 9 of NB-3652 Clamp-induced stresses caused by the constraint of the expansion of the pipe due to the internal pressure will be added to other secondary and peak stresses by calculating the ef fective increases in the C1 and KL stress indices in accordance with NB-3681. Clamp induced stresses due to dif ferential-temperature and dif ferential-thermal-expansion coef ficients will be accounted for by computing the ef fective C3 and K3 stress indices.
Clamp-induced stressos on elbows caused by the constraint of pipo wall ovalization will be accounted for by computing the ef fective increases in C 2 and K2 bending indicos. The fatigue usage from clamp-induced plus other stresses will be calculated at governing locations.
Although bolt proloads are not addressed under the ASME B&PV Code rules for piping , bolt proloads could result in camage to a pipe if a clamp we re poorly designed. Calculations will be made to ensure that bolt proloads could not result in plastic do formation of the pipe walls.
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t HCGS FSAR 6/84 RESPONSE (Cont'd) .
A brief summary of the criteria used and the results of the analysis will be submitted in August, 1984.
HCGS clamps were not used to meet stif fness criteria. They were designed to meet the requirements for strength and load distribution using a minimum of space.
The clamp design utilizes a double nut arrangement to prevent the nuts from backing of f. The low temperature ( <600 'F) and stresses in the bolt from preloads will not cause a relaxation of the material. Consequently, no lif t of f from the piping will occur.
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HCGS '
DSER Open Item No. 98 (DSER Section 3.9.3.3)
DESIGN OF BOLTS The staff's review of Section 3.9.3.4 of the applicant's FSAR relates to the methodology used by the applicant in the design of ASME Class 1, 2, and 3 component supports. The review includes assessment of design and structural integrity of the supports. The review addresses three types of supports: plate and shell, linear, and component standard types. More information regarding the design and construction of ASME Class 1, 2, and 3 component supports is required.
The staf f needs further information on the design of bolts.
RESPONSE
For the information' requested above see response to Question 210.47. .
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M P84 95/06 2-dh
QUESTION 2k0.47 (SECTION 3.9.3)
The staf f's review of your camponent support design finds that additional information is required regarding the design basis used for bolts.
- a. Describe the allowable stress limits used for bolts in equipment anchorage , component supports, and flanged connections.
- b. Provide a discussion of the design methods used for concrete expansion anchor bolts used in component supports.
RES PONS E Non-NSSS: , , ,
- a. The allowable stress limit used for bolting in equipment anchorage and in pipe support components is 0.5Su but shall not exceed 0.9Sy under all service le ve ls .
For flanged connections, the bolt allowable stress used in the piping analysis are ASME Subsection III, 1979 Summe r Addenda, Subsections NB, NC and ND for Class 1, 2 and 3, re spe ct ively.
- b. The capacities of concrete expansion anchors are based on actual testing of anchors to f ailure. The f ailure loads are divided by the factor of safety (typically 4 in accordance with NRC Bulletin 79-02) to establish the allowable design loads. Baseplate flexibility is, considered in the design of concreed expansion anchor bolts in accordance with IE Bulle tin 79-0 2.
NSSS:
- a. 1. Equipment Anchorace Equipment anchorage is not in the NSSS scope.
- 2. Component Support Bolt ing The following bolting design limits are typical of components mounted directly on base plates. .
- RWCU Pump The support bolting of this pump, which is not sa fe ty-re lated , is designed for the ef fects of pipe load and SSE load to the requirements of the ASME B4PV Code ,
Section III, Appendix XVII. The stress limits of 0.41 Sy for tension and 0.15 Sy for shear are used.
210.47-1 Ame ndme n t 6 93
RESPONSE (Cont'd)
- RCIC/SLC Pumps and RCIC Turbine The equipment-to-base-plate bolting satisfies the following design criteria: For normal and upset conditions,1.0 S is used for primary membrane (or te nsion) , and 1.5 8 for primary membrane plus bending (if applicable), where S is the allowed stress limits from the ASME B&PV Code,Section III, Appendix I, Table I-7. 3. For emergency and f aulted conditions, stresses shall be less than 1.2 times the allowed limits for normal and upset conditions.
The allowed stress limits used for bolting in pipe supports and pipe-mounted equipment supports are as per ASME B&PV Code,Section III, . Subsection NF.
For service level A and B, *.he bolts meet the criteria of Paragraph NF-3 280. For service level C and D, Article 2460 of Appendix XVII, with the f actors indicated in Article 2110 of Appendix XVII, is the applicable design requirements for bolting. The stresses calculated under these criteria do not exceed the specified minimum yield stresses at tempe ra ture .
- 3. Flanced Connections Flanged connections are not in the NSSS scope.
- b. Expansion Anchor Expansion anchors are not in the NSSS scope.
210.47-2 Ame ndme nt 6 i
- HCGS i
1
(, DSER Open Item No. 99 (DGER Section 3.9.5)
STRESS CATEGORIES AND LIMITS FOR CORE SUPPORT STRUCTURES Further information has been requested on the stress categories and limits for core support structures and the applicable codes i i
used for evaluation of the f aulted condition. This is an open !
item.
r
,I RESPONCE For the information requested above, see the responses to Que stions f
[
210.55 and 210.48. .
s e s O
G e
O 4
K53/1-6
NCGS FSAR 6/84 QUESTION 310.48 (SECTION 3.9.3)
Describe 7 hose short-term and long-term actions being taken to preclude the occurrence of cracking in jet pump hold down beams as described in IE Bulletin 80-07.
RESPONSE
. The preload on the hold-down beams will be reduced from 30 to 25 kips in accordance with General Electric recommendations. This will increase the expected life of the beams to 19-40 years. The need for inservice inspection will be based on a lead-plant experience and GE testing, and will be conducted such that any crack initiation will be detected prior to beam failure.
Section 3.9.5.1.2.1 has been revised te include this information.
a 1
DSER OPEN ITEM 7/
210.48-1 Anondrsant 6 i l
NCGS FSAR 3.9.,5.1.'t.4 Control Rod Guide Tubes ,
The control rod guide tubes, located inside the vessel, entend from the top of the CRD housings up through holes in the core plate. Ea:h tube is designed as the guide for a control rod, as well as being the vertical suppert ior a four-lobed orificed fvel support and the four fuel assemblies surrounding the coptrol rod.
The bottom of the guide tube is supported by the CR0 housing, which an turn transmits the weight of the guide tube, fuel support, and fuel assemblies to the reactor vessel bottom head.
A thermal sleeve is inserted into the CRO housing from below and is rotated to lock the control rod guide tube in place.' A key is insected into a locking slot in the bottom of the CRD housing to hold the thermal sleeve in position. .
3.9.S.1.2 Reactor Internals , ,
The reactor internals consist of those items listed in Section 3.9.5.1.b. Those that involve coolant flow paths are described in the following paragraphs.
3.9.5.1.2.1 Jet Pump Assemblies The :et pump assemblies are located in two semicle'cular groups in the clowncomer annulus between the core snroud and the RPV wall.
The design and performance of the jet pumps are discussed in References 3.9-14 and 3.9-15. Each stainless steel jet pump consists of a driving nossle, a suction inlet, a throat or utning section, and a diffuser, as shown on Figure 3.9-5. The driving nossle, suction inlet, and throat are joined together as a High-removable unit, and the diffuser is permanently installed.
pressure water from the recirculation pumps is supplied to each pair of jet pumps through a riser pipe welded to the recirculation inlet nettle thermal sleeve. A riser brace consists of cantilever beams welded to a riser pipe and to pads '
on the RPV wall.
The nossle entry section is connected to the riser by a metal-to-metal, spherical-to-conical seal The Firm contact is joint.throat maintained by a holddown clamp. section is' supported laterally by a bracket attached to the riser.TheThere is a slip-diffuser is a fit joint between the throat and diffuser.
gradual conical section changing to a straight cylindrical section at the lower end.
l .G L1 M ' Q ostn orrN ITEM 9/ 3.9-102 L
,o -
l l
NOPE CREEK Mas - sgR 210.48 INSERT !
The proleps: en the hold-down beams will be reduced f rom 30 This to 25 kipe in asserdance wLth General Electric recommendations.
will increase the espected life of the beams to 19-40 years. The I need for inservice ingpection will be based on a lead-plant l esperience and 08 testing, and will be conducted such that any l initiation will be detected prior to beam f allue.
O i
l l
l 09tp OPRM ITRM $7 l i
l
IOCGS FSAR 4/04 atmation 2YMis (SBCT!0ff 3.9.9) erif thsh. h deelen and analysts of your reacter internals is tv lent to guteettien M.
588t95E tepe Creek reacter internals were designed and procured prior to the Leevance af Subsectten Wei, doelen criteria M ofinSectten !!! ofdraft an earlier the ASMS of the MPV ASNS Code.
Mpv was used as a guide in the desten of the reactor
- ' inter le. Zhese criterna are presented in Section tubsequent to 3.9.5.3 the 3 and issuance were used in Atou of tubsection M.
of Subsection M, comparteene were ma3e to assure that the pre-M doelen meets the equivalent level of saf3ty as presented by Suteoctien M.
Y 4
1 1
i 9
l 210 55 1 Mendment 6 esen ortw it:M 9y
l HCGS DSER Open Item No. 125 (DSER Section 6.2.1.5.2)
DESIGN DRYWELL HEAD DIFFERENTIAL PRESSURE ,
The applicant has performed a pressure response analysis for the head region postulating the rupture of the 6-inch head spray line. ;
The initial conditions of 15.45 psia, 135'F, and 30 percent ,
relative humidity were used in the analysis. The maximum differential pressure between the head region and the drywell was ' calculated to be 15.64 psi. However, the applicant has not provided the design drywell head -
differential pressure. The staff requires this information and will perform a confirmatory analysis to verify the ,
applicant's results. We will report on this matter in a supplement to this SER.
- RESPONSE Section 6B.3 has been revised to provide the information requested above.
e I
t ,
i l
M Pb4 95/05 1-dh i
'\
l HCGS FSAR 6B.3 DRYWELL HEAD REGION SUBCOMPARTMENT ANALYSIS sue de=.tvu vasis pcessure diffecencial veuween cue acy-wil uwod ,
- nd ;ntein;;at cegion le . 66umous 1 u-ywi&mmmuu vi Lie dcy-ell i heed. A pressure analysis of the drywell head region for a postulated head spray line break was performed. The effects of a 6-inch residual heat removal (RHR) head spray line break bound l those of a 2-inch RPV head vent line, which is the only other i line that runs through the drywell head region.
Figure 6B-12 illustrates the basic arrangement of the head region. Venting from the head region is accomplished through ,
ventilation openings as shown on Figure 6B-12. These vent j openings provide a total vent area of 7.722 square f,eet, with an '
equivalent orifice discharge coefficient of 0.75 to relieve !
l pressure buildup caused by the postulated break. }
( e
[
t e
. I l Figure 6B-13 is the schematic flow diagram with vent flow areas ;
and discharge coefficients used in the drywell head venting analysis.
l t
To determine the peak pressure in the drywell head, all insulation was assumed to remain in place. Initial conditions of ;
15.45 psia, 1350F, and 30% relative humidity were used in this analysis.
The pressure transient of this analysis is presented on Figure 6B-14. It can be seen that the maximum pressure in the drywell head region is 31.086 psia and occurs 1.7534 seconds ;
after the head spray line break. Considering the containment f pressure to be atmospheric (no drywell air displaced into the ;
suppression chamber), a drywell head to containment pressure i differential of 15.64 psid occurs at approximately 2.0 seconds l after the break.
_ILvsrR7 k ;
v l 6B.4 REFERENCES ,
r 6B-1 "Subcompartment Pressure Analyses," BN-TOP-4, Revision 1, November 1977, Bechtel Power ;
Corporation, San Francisco, California. ,
l DSER CPEN ITEM /dd 6B-5 l
INSERT The manufacturer's design differential pressure (between the drywell head and containment region) for the water seal plate is conservatively defined as 20.0 psid. For the refueling bellows, the manufacturer's design differential pressure is defined as 15.0 psid. However, an evaluation of the bellows manufacturer's design report revealed that sufficient margin exists between the calculated stresses snd the allowable stress such that the bellows can easily -
withstand the increased stress due to an applied 15.64 psid.
r DSER OPEN ITdM 125 MP84 93 06 1-vw r
r e HCGS DSER Open Item No. 132 (DSER Section 6.2.4) i CONTAINMENT ISOLATION REVIEW The basis for the staff's acceptance has been the conformance of the centainment isolation provisions to the Commission's ,
regulations as set forth in general design criteria and applicable regulatory guides, staff technical positions, the acceptance criteria of NUREG-0800, and industry codes and standards.
[However, the applicant has not provided the staff with the information requested to conduct the containment isolation review.
We will require the applicant to provide for the containment isolation figures contained in the FSAR proper valve identifica-tion and the location of the test, vent, and drain lines on these figures. We will also require large enough P&ID drawings in order to conduct our review.] 'We will rep ~ ort on this matter in a supplement to this SER.
RESPONSE l The penetration details in Figure 6.2-28 have been revised to show proper valve identification and the available test provisions.
F K53/3
i I
.T* .
l 1
FROM MAIN STEAM PRIMARY SEAL SYSTEM KP V032 .
RPV CONTAINMENT g j i i KP V033 i I
P1Aj .
AO j D 1 i
- OM 1 8 KP-V016 l 8>< MSiv > D C-- !
_ MSiv MsSv !
p AB-V028 A8 V032 .
AB V003 MAIN STEAM DRAIN l
! l
?
f
- AB V145 A8_V146 m_m m a m j j ISOLATION VALVES f P1A P-15 P 1C P1D .. !
A8 YO2B V029 V030 V031 ey n ;
As-v032 v033 V034 v035 1 fg,,yo,3 As w 3 Mm f
d AB V060 V000 YO61 V062 -
l MAIN q r KP VO10 V000 V008 V007 ,: STEAM , AB V064 +
DRAIN a g !
Lj AB V132 !
vantmaans watves I'ts i
p.ga PS PJC P30
+
w vess vert was mi d g V !
KP V010 i o.vuse vess vens vess
[1C s .s, v.us _ , , :
- * * ** * ** As V026 KP V025 r f evne vist veas visi d k a 1D "V" V's we v* ,
KP V024 . :
s .- s. , ,
TO M AIN ~ !
em we. vess use l f
M mese m eans vesi wie via vise wr viss l FROM MAIN STEAM j SEAL SYSTEM A8 V027 g me vses van taas ;
A- mese vist via viss !
HOPE CREEK i l g DETAIL 1 GENERATING STATION {'
,,,, ,,, ,,,,,,,,y,ty,,
, 8 ,. , ,e , ,e , , , FINAL SAFETY ANALYSIS REPORT
,' as.vens ves. was vens EP-vete weis vote vois t MAIN STEAM LINES :
FIGURE 6.225 e(SEE LEGEND) SHEET 1 OF 48 AMENDMENT 6,064 f
l i
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BD V046 l w_a ;
v, ;
1 P PRIMARY BD V013 '
> BD-V047 i RPV CONTAINMENT s L l AE V015 g l a FROM i Ym ~~55i'5 l BD V006 (TYPICAL AE V016 l go FOR HPCI)
, 4
~
4 4, I FEEDWATER f y( , , -- LINE AE V004 AE V003 - AE V002 AE V001 l p.2A l AE-V075 9 P J L
> AE V020 d L 1 P i
' AE V074 d b 1 P
< l AE V019 d L LJ L .J k *
- ISOLATION VALVES P 2A P 25 (
, AE V003 AE-V007 pq d#
AE V002 AE V006 q r TO THE OTHER FEEDWATER LINE AE V021 AE V021 4 k e
80 V006 BJ V058 9 r !
i BG V072 :
AE V001 AE V006 J L AE V128 TEST / DRAIN VALVES P2A P-28 DETAIL 2 i AE V020 V018 -
l AE V019 V017 .!.. !
q :
Q AE V015 V014
.. 4 4. .....j... .'.... .....---.AEyV005 d --
AE VD16 V013 l* AE V008 .'-* ., _,_,
- . . p.2B '
FROM I 6---- J s
s BG V072 -
RWCU z BG V071 - .
$ MOPE CREEK o AE V075 -
GENERATING STATION AE V074 - FINAL SAFETY ANALYSIS REPORT h
BO V047 BJ VOIS BD V046 BJ V017 FEEDWATER LINES AE V004 AE V008
- ., FIGURE 6.23
, ,,. r u n , AE V127 AE V128 W J4F67 R @ D G7 nmemndcuTann.-
a -
ryt s . . . . -
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l' PRIMARY'
- LOOP) CONTAINMENT (RECl l
l i sCav-442s !
l A _ J kA 9
-- - , -, a ;
l eC-v167 aC-vlsa ,
l l
j
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L BC-V078 BC V071 - BC V164 1 V i BC V303 J L l P.3 l -
l % 1 V I 1 P i i BC-V165
> BC V302 d i d L 1 V LJ <
i BC V166 d L 1P DETAIL 3 I
i MOPE CREEK GENERATING STATION FINAL SAFETY ANALYSIS REPORT l
l
? RHR SHUTDOWN COOLING SUCTION Ll FIGURE S.225 i $5EE LEGErgc@ SHEET 3 OF 48 AMENOMENT 6.0f
i M .
I r
l r
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.- [
l t
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l RPy PRIMARY f (RECIRC LOOP) CONTAINMENT 4 . f i i
! ! CLOSED SYSTEM I i
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,, P4A ' BC-V346 i
d L BC V118 ;
1 F !
' BC V355 l
' d b '
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' BC V170
% DETAIL 4 ISOLATION VALVES ;
TEST /DR AIN V ALVES P4A P48 (
l V189 j 8C 172 8C 170 V171 f SC-V346 V334 ;
BC V355 V335 ;
SC V074 V183 HOPE CREEK OENERATING STATION l 8 11 V117 FINAL SAPITY ANALYSIS REPORT l DSER OPEN ITEM /3d OTHER VALVES '
t
' RHR SHUTDOWN P4A P 48 COOLING RETURN LINES l 8C V014l V111 l
FIGURE 8.228 l
$SEE LEGEND) SHEET 4 OF 48 AMENDMENT 6.067 j
6.t. .
l t
PRIMARY L RPV CONTAINMENT
. OTHER VALVES l P5A BE V002 f
i P4A BC-V005 I l* t
- CLOSED SYSTEM P4B BC V017 !
l s
g P4C BC-V114 ,
l
- I
%, j P40 BC-V102 hg ,
f.
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BE VODI BE V002 e _\
BE V003 ' ', ' ',
w- w- .a 1
j
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l P4A l q
g BE V079 8E V080
' BE V078 l
d k i
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9 d L i P
BE V072
. i
' d b BE V070 ;
1 P l7, TEST / DRAIN VALVES d L BE V069 P-5A BE V070 V069 V078 V077 V072 V001 P4A BC-V173 V177 V320 V321 V119 V076
] g '
- P48 BC V174 V178 V324 V325 V120 V075 P4C BC V175 V179 V353 V354 V121 V182 ;l ISOLATION VALVES P40 BC V176 V180 V332 V333 V122 V181
. P4A BE V003 ,
ir P4A BC V004 CONTAINMENT
!f PRIMARY .
i P48 BC-V016 P4C BC V113 OR - - - - - - - - - - - - --
)!
P40 BC V101 .l P4B l -l i
RPV !
CLOSED SYSTEM j
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$e ! :
- BE V007 :
I l: qM e
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BE V005 BE V006 ' ', !e !
w, w- .'a c
$= 2 ' !
~ B E V075 .B..E..V.0.76.... ...i ;
BJ-V018 BE V074 l
BE V071 NOTE,_ , =_s ,
w, J
. 1 P "BE4073 U I7
' ' i BE V068 ;
J L l
HOPE CREEK
! p 3
GENERAN STATION ,t DSER OPEN ITEM / 3 42 BE V067 d k FINAL SAFETY ANALYSIS REPORT LJ DETAIL 5 RHR LOW PRESSURE COOLANT INJECTION AND NOTE: CORE SPRAY DISCHRGE LINE j FOUND ON CORE SPR AY
$ SYSTEM ONLY. FIGURE 6 2 28 (SEE LEGENO) SHEET 5 OF 48 AMENDMENT 6.06/84 l
_ ._.- ._-_ - - .- - _.,-__.,-.- -- , . ~ . -
l i l
fxt .
i L
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! i RPV PRIMARY CONTA INMENT j I
I .
I
@FD V001 _
D C
$>@<FD-V002 , r FD V057 j d
FD-V058 j ( FD V006 i
' FC-V024 ) f 1 F J k f FD V067 ;
- >FD V059 d k l J L >
9 r i LJ < > FD V023 LJ
) f (, - J L l J L FD V000 7
y l 1 , .
FD-V051 ,
k FD-V061 ,
I
- LJ ,
R {
i
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PRIMARY RPV i CONTAl NMENT FC V037 FC-V038 f m_m m_m ,
r w, .a G !
@FC V001 ~
FC@2
- D<
1 F -
z :::
FC V021 l!
l FC V053 r- &_e u _2 t q FC V052 a g l
L r, r, m
N l fFCV054 FD-V024 FC V025 J L 5 , LJ e P
- , L FC VOSS 4,
r '
g3 q HOPE CREEK O
/V$ FC V048 GENERATING 8TATION .
$ J k FC V056 FINAL SAFETY ANALYSIS REPORT g
LJ HPCI AND RCIC STEAM SUPPLY LINES l DETAll 8 !
FIGURE 6 2 28 l *(SEE LEGEND) 8HEET 6 OF 48 AMENDMENT 6. 06/B4
l
- u. . . . . . . . . - . . . .
l l
l l
l 1
l l
9 i
PRIMARY U
RPV CONTAINMENT
~
l* 1 V BC V174 ' '
l d k l
l 9 P !
8C V173 i > !
. . i BG N161 NN BG V001 Y
SG V002 J L G V166 p.g - .
<: 1 V
' BG V020 ;
J L '
l G V166 LJ l t l \
l l
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i DE TAIL 7 ;
I HCPE CREEK ,
GENERATING STATION j FINAL SAFETY ANALYSl3 REPORT DEER OPEN ITEM /3d RWCU LINE I
i FIGURE 8.2 28 i
,(SEE LEGEND) SHEET 7 O_F 48.____A@@M@NT @,06/j
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PRIMARY f RPV CONTA NMENT :
r1 ;
1 r r
' BC V098 1 J L 1 r !
' BC-V077 ,
J L
~
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, 6 ,
s yy y M k
- sC.Vo21 _ sc-Vo20 :
1 P ,
1 i BC V299 !
(. J L P 10 q r i ' BC.V079 d
9 r l
' BC.V300 :
J L 9 r L J NOTE j [BC V080 i
I l
LJ
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t l
l DETAIL 8 i
?
NOTE: THE TEST LINE IS ATTACHED ,
TO THE REMOVABLE BLIND l' FLANGE !
HOPE CREEK f GENERATING STATION i FINAL SAFETY ANALYSl3 REPORT DSER OPEN ITEM /3A HEAD SPRAY LINE [
r l
FIGURE 6.2 28 NRFF t FnFMn) . _ _ . ._ _ __ _. ._ _ _ _ _ _ flHVEM_C QC9@Q0@lM@%T 0 0 .
gt: . . . . . . - -
- t.' f. , .
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<m.
l -
l l ,
PRIMARY RPV CONTAINMENT
- r" "I i
9 P i AB V041 J L A8 VO42
, . 1 F g J L g N AS-V039 AB-VO40 AB V12E
)
1 P i A8-V123 \ 3 g
AB V018
- l (/ J L lP 12l
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- 1 P
< > A8 V127 j [ AB V019 J k LJ i
l.
!, . DETAIL 9 MOPE CREEK GENERATING STATION FINALSAFETY ANALYS13 REPORT l
DSER OPEN ITEM /3d MAIN STEAM DRAIN LINE
, FIGURE 8.2 28 (SEE LEGEND) SHEET 9OF48 AMENDMENT 6.C
64 - - . . . . . . . . . .
j l
l I
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f" "l !
PRIMARY >
RPV. CONTAINMENT ] BB V059 f J L l 3 r l
< l 88 V060 '
d L h
- .1, l
v , ;
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88.V061 35-SV4310 gg.sy4311 (
,\ 1 r !
! f BB V194 l p.ty l i BB V197 :
s L i L .. !
f f 88 V195
) BB V196 j g J L t gg . LJ l
i DETAIL 10 I i
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l l
[
MOPE CREEK GENERATING STATION i FINAL SAFETY ANALYSIS REPORT DSER OPEN ITEM / 3,2 L REACTOR REClRCULATION PUMP !
WATER SAMPLE LINES .
l i
' l FIGURE 8.2 28 l
,(SEE LEGEND) SHEET 10 OF 48 AMENOMENT 6.06 ,
t
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l PRIMARY l~ RPV CONTA' NMENT r, r1 l l l
1 r 1 r ;
1
' BH V032 . BH.V038 ' l
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! BH V031 l BH V037, r' 1 r d ' ., s ' E l
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BH V030 BH V029 E*M
.\ f I
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> BH V033 ..
J L [
t 9 P SH. N I BH V034 i J k LJ [
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DETAIL 11 1 t
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I l
r r
HOPE CREEK i CENERATING STATION i l'
FINAL SAFETY ANALYSIS REPORT DSER OPEN ITEM /JA !
STANDBY LIQUID CONTROL LINE !
I FIGURE S 2 23 l
- (SEE LEGEND) SHEET 11 OF 48 AMENOMENT 6.06 '
l
, we . ... - , .
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i
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FRIMARY ;
RFV CONTA NMENT l
Il '
i 9 r :
i '
BB V201 '
d k j 1 r - -
.i
' BBv200 d k -
1 y& 2 _,_.
BB V042 BB-V043 - BF V008 [
. .i 1 P q r i EB-V190 , BF V155 ;
J k J k -
l 1 P 3 r :
i B3 V198 '
' BF V156 i d k J L
~
LJ LJ t l
I DETAIL 12 i I
l ISOLATION VALVES TEST VALVES l F 19 9 20 F 19 P 20 f t
BB V043 BB V047 BB V'e00 V204 l BF V008 BF V099 BB V201 V205 BB V190 V203 .
BB V198 V202 l l
HOFE CREEK i l
BF V156 V772 GENERATING STATION ['
BF V156 V773 FINAL SAFETY ANALYSIS REFORT DSER OPEN ITEM /,3 A BB V042 V046 !
REACTOR RECIRCULATION !
PUMP SEAL WATER LINES ;
e FIGURE B.2 28 (SEE LEGEND) SHEET 12 OF 48 AMENOMENT 6.0t 'l
r '1 GS V001 1 r ..ru .- . _
< \
d L PRIMARY PPV CONTAl,NMENT 8A A *L
, A T TO PRIMARY CONTAINMENT r N 7' HYDROGEN RECOMBINEH SYSTEM GS V004 5 r os.V006 j ( GS-V114 dL l fGs.y134 AO 1 l ,
c -
GS V000 GS V021 l
AO GS V067 GS V023
' GS V066 -
% GS V066 m_m 1 P J L GS VOSS GS V023 9
1 r GS VD22 h
es TO SUPPRESSION CHAMBER GS V020 PURGE INLET LINE
- '/
i I
l MOPE CREEK GENERATING STATION DSER OPEN ITEM /32 DRYWELL PURGE INLET VENT LINE FIGURE 8.2 28 8(SEE LEGEND) SHEET 13 OF 48 AMENDMENT 6.06
_ . . _ _ _ .__ _ _4 ._
\
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t t
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RPV CONTAINMENT l
g I ! !
I
-P 23l M GS-V025 AO j l l' I !
., i ei, i
~
GS.V024 GS V026 -
GS V073 8' A d ... . .TO PRIMARY CONTAINMENT f l 7 7 7' HYDROGEN RECOMBINER SYSTEM GS-V002
, GS V003 !
l* , y
' v-
- . GS-V113 <
J L
GS V000 !
1 1 P :
GS V183 4 i i
4 L ,
LJ I I
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DETAIL 14 l
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1
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i i
f HOPE CMEEK ,
GENERATING STATION ,
FINAL SAFETY ANALYSIS REPORT DSER OPEN ITEM /3A DRYWELL PURGE I VENT LINE FIGURE 6.2 28
- (SEE LEGEND) SHEET 14 OF AS AMENDMENT 6.
\
i
.. - _ - - . . .-- - - . . - - . _ - _ . . . = . . _ - _ _ _
i PRIMARY .
RPV CONTAINMENT t
l l
@SC V019 $
O i l
~$w n a u
- a. ,,
- . . . ~ . .
. r 1 P 1 P 1 r
' BC V073 i i BC-V077 i l P 24A i ' BC V318 J L J L
[
J L '
1 F 1 P 9 r
' BC-V072 ' '
BC V095 ';
' BC V319 '
J L d 6 J L ,
p o
LJ LJ LJ l
i i
OR ,
l ,
PRIMARY i
CONTAINMENT ll l RPV l [
1 P ,
L BC V352 d L 1 P
< BC-V351 n s k (")' j
$'- BC V115 l
~$' BC V118 9
1 l
1 V 1 P t BC V184
' BC V334 1 r k d k P248 ' '
8C V330 d
}
J k !
1 F 1 P I r ' '
SC.V199 <
> BC V335 l
' BC V331 J L 4 L d L LJ LJ LJ DETAIL 15 l l
HOPE CREEK l GENERATING STATION !
FINAL SAFETY ANALYSIS REPORT !
l DSER OPEN ITEM /3,1 [
l I CONTAINMENT SPRAY LINES i i
FIGURE 6.2 28
- (SEE LEGEND) SHEET 15 OF 48 AMENDMENT 6.06/84 f l
' .e .
l i
l I
I I
, F ;
l 1 I PRIMARY l
[
R,PV CONTAINMENT ;
H8 V004 I- '
H8 V002
- r1 i 1 P HB V123 i
~
M I .
- Q & y 5 HB V005 HB V006
~
_\ H3.V'25 -
H8 V124 p.25
( y r V .
P d I t
E DETAIL 16 ISOLATION VALVES TEST VALVES j P 25 P 26 P 25 P 26 j 5 H8 V000 HB V045 H8 V002 H8 V042 L l
- f. H8 V006 HB V046 HB V004 H8:V044
! l
. H8 V124 H8 V127 5 H8 V123 H8 V126 HOPE CREEK !
i GENERATING STATION j HB V125 HB V128 FINAL SAFETY ANALYSIS REPORT ,
DSER OPEN ITEM / 35L LIQUID RADWASTE i
[: COLLECTION LINES l
FIGURE 6 2-28 1
- (SEE LEGEND) SHEET 1C OF 48 AMENDMENT 6.06
e, s . . . _ . . . _ _ . . .
PRIMARY RPV CONTAINMENT J
t -. .
_N
! 9 P l P 27 j > KA V7177 J L KA-V061 gg L.!
l l
OR l
PRIMARY RPV CONTAINMENT T1 i r
' KG V151 d L
..".I.E,,,,, %_m % _2 WTE
= ,, ,,,,,
, KG V016 - KG V034 KG V009 I -
! P 31 NOTE:
THERE IS CAPA8ILITY TO VENT THROUGH THE BREATHING AIR STATIONS l
l HOPE CREEK GENERATING STATION MTAIL U FINAL SAFETY ANALYSIS REPORT DSER OPEN ITEM /38 SERVICE AIR AND BREATHING AIR TO DRYWELL g FIGURE 6 2 28 ISEE LEGEND) SHEET 17 OF 48 AMENOMENT 6. 06/B4
l j
,=
l I
i L RPV PRIMARY C. ONTAINMENT .
l g j
( '
w-
', J
!, KL V035 ;
l
- l f 8 - i KL-V026 <- KL V025 P 28A KL V039 l
l RETURN t
@ ~ @ -+44Es KL V003
+34 y *) 4 KL V002 l
KL-V001 -
KL V049 O :
l P-a l ->4-E]$3 KL V004
,q 7 j g KL V051 ,
SUCTlON ll I
$ DETAIL 18 i N ISOLATION VALVES g '
g P3A P 288 -
E KL V026 KL V028 a-MOPE CREEK KL.V027 GENERATING STATION KL V025 FINAL SAPETY ANALYSl3 REPORT ,
N y TEST VALVES P 28A P 288 KL V037 INSTRUMENT GAS LINES ,
KL V035 '
KL V039 KL VO40 PlGURE 8.2 28
- (SEE LEGEND) SHEET 18 OF 48 AMENDMENT 6. 06.5 l-
t 1-l 1
l 1
i i . !
9 P34A I w
. . - 3 g.. ~ .... ..
i i ,
1 l E-De(--
SE V016 ,
5 l' t
I '
i b i !
7-4 '*
i SE V026 SE-V021 I l SE V011 [ ! i 1,
INDEXING :. .~ ~.--- ~ .
MECHANISM ,
P-34G AO L
e I
, v, y .
I
SE V004 i i SE V006 SE V000 SE V005 r
i i
l [
TEST VALVES P 34A SE V016 SE V011 ;
P448 P 34C SE V018 SE V013 DETAll19 SE V019 SE V014 l P 34D ISOLATION VALVES SE V015 f P 35E SE V020 i t
s NOPE CREEK P 380 SE V029 SE V024 GENERATING STATION j ,
P-34E SE V030 SE V025 FINAL SAPETY ANALYSIS REPORT I U
-l I ,
I g TIP PROSE El GUIDE TUBE LINES i a: '
I M *(SEE LEGEND) FIGURE 4 2 28 i , SHEET 19 OF 48 AMENDMENT 6.06/8_4 ,
7 i
l
I i
l I
i e
t
)
l r"I 1 r
' FD V013 J L PRIMARY RPV CONTAINMENT
{
! a , @-
i t;
i
! I hNFD V006 20 pp.ygog l
.: q p ,
LJ j i
" 5 ..J !
..._I- I v
wa na y ,
DETAIL 20 i ;
r ISOLATION VALVES TEST / DRAIN VALVES P.201 P 207 P 201 P 207 FD V006 FC V005 FD V062 FC V058 FD-V007. FC V006 FD V063 FC.V057 i FD V004 FC V003 FD V029 FC V020 FD V013 FC V040 HOPE CREEK ,
GENERATING STATION !
DSER OPEN ITEM /3A FINAL SAFETY ANALYSIS REPORT FD V065 FC-V062 HPCI AND RCIC EXHAUST LINES !
FIGURE 6.2 23 i
- (SEE LEGEND) SHEET 20 OF 4B AMENOMENT 6.06!54 l
,. h 9
PRIMARY RPV CONTA NMENT m . . - ..
'I ~
__ __ _ t _ _ _ _ _ f __ _
CLOSED SYSTEM q
$ P P 202
&a .,
l , a g g l sJ vose sJvoos sJ V021 l
DETAIL 21 ISOLATION VALVES P 202 P 208 BJ voos sD voo3 TEST VALVES P 202 P 208 sJvose so vo35 l M CREEK sJvo21 sD v015 GENERATING STATION FINAL SAFETY ANALYSIS REPORT l
l DSER OPEN ITEM / 3.2.
1 HPCI AND RCIC PUMP SUCTION LINES FIGURE 6.2 28 SHEET 21 OF 48 AMENDMENT 6,0614
- (SEE LEGEND) l
l
., . ... . i l
e u vois pensawy u veso i arv confasemesw? ;
= ==2.!ma ;
i 4__ _
A !
s, -,
l
' "" um ,
i !
,M
/ ;
1
_Y. .
i OR m
s ,, arv enwrasmmert ,
eumso svmu
^
l
- s. _'_
! N- '
' so vues j esNass , ,,,,,
< b l 5 ' so.vtes '
i l
A U
~7
/
i M
i l
\ .
)
.T.
l HOPECREEK GENE R ATING STATION FINAL SAFETY ANALYS15 REPORT l
DSER OPEN ITEM / 3 4. DETAIL 22 RCIC AND HPCI MINIMUM FLOW RETURN LINES FIGURE 6.2 28 O(SEE LEGEND) sHE ET 22 OF 48 AMFNDMENT 6 06%4 t
p FD V011 T1 3 , ,
FD V041 d b d L PRIM ARY RPV CONTAINMENT FD V025 9
3 M
ac V268 h FCV043 BC V256
[j 8C V266 8C V267 CLOSED SYSTEM FC V007 FC V000 FC V000
%. U
~
=
P 204
\.1...__-- --- :_- :::_---'
DETAll 23 port catEK GENERATING 8 TAT 10N FINAL SAFETY ANALYll5 REPORT DSER OPEN ITEM / 3 .2.
HPCI AND RCIC VACUUM BREAMER NETWORK LINE o(SEE LEGEND) IE I '
?
?
l
' i t
I l
PRIMARY l
, RPV CONTAINMENT l
l P1 Il I ;
! FC V008 I J k
- x9 .
4 :
l FCV010 FC.V000 FGV011 ,
N P 210 l FC VOIS I
t I
( . i
_Y_;
\ ____________ .
- f II l
\
l oETAIL 24 l t
More cmEn MMRATING STATION FINALSAFETY ANALYSl4 REPORT DSER OPEN ITEM /Jg - -
RCic VACUUM PUMP LINE Flount s.t.se andfNOWf NT 8',0$k SHEE7 24 OP 4
- (SEE LEGENO) --
3 .
,t
-- - '4- .
I I
I i
, I l l i PRIMARY CONTAINMENT !
l RPV 1
\
g a y g 's
"\ g e--
\I '
. s .
f l' I i
i : i 3 :
- . t l
I
' h :' }
i ". T t, c
......._.___._.._._...I.
8 I
t i
- I a CLOSED SYSTEM ,
i
- :: l x
K 1 P P 211 A l e .
Ug gQ , a m_2
- l 7 j
8C V326 BC-V327 : j
\ T SC V001
[
ISOLATION VALVES t
's P-211 A SC-V001 i
P 2118 BC V006 8C-V2SS l g I P-211C BC-V103 = !
DETAIL 25 v .J i
' 's P211D 8C-V088 2*
-) - TEST VALVES BC V288 I BC V289 !
P-211 A SC V326 8C V327 b P2118 8C V328 BC V329 SC-V282 SC V283 90lWE CREEK l
- BC-V294 GEMRATING STATION <
P 211C 8C V341 BC V340 BC V293 z FINAL SAFETY AflALYSIS REPORT :
$ P 2110 BC-V342 8C V343 BC V292 8C V291 l O !
a:
T4 m' i RHR PUMP SUCTION LINES :
o , i
% FIGURE 6.2 28 ,
o(SEE LEGEND) SHEET 25 OF 48 AMENDMENT 6.06/84 f
-a - -- . .
f r '1 ;
1 V PRIMARY
' BC V095 RPV CONTAI N M E N,T_,_,,,,__,,_,,,,,,,,,, l
- , r 7
BC-PSV-F025D j g l 4 B j
.g, -
c.V,s........ ....,
BC V028
- r@ ! :
e
- 2 -
1 1:---i:
BC V027
, c l
l l
I W
>0 BC V316 i
BC V026 : ,
,,, k a BC V260 :
7' i '
l : :
l . .
t l
k e : f P 212A j y y j
. .. . .. . _ .. . .... t
t
- l CLOSED SYSTEM
\ -
.1 . . . ._._..__._.._.._._......
U TEST VALVES ,
ISOLATION VALVES P-212A P 2128 P 212A P2128 BC V360 BC V338 {
BCMV-F025D BCMV F025C BC V361 BC V339 -
SC PSV F0258 BC-PSV F025A DETAIL 26 BC V077 BC V334
'V ' DC VO28 BC V124 BC V095 BC-V335 cn
% SC WWE7 BC V125 BC V316 BC V317 x SC WW26 BC V126 ra U BC VO34 BC V128 HOPE CREEK i GENERATING STATION {
$ BC V031 BC V131 FINALSAFETY ANALYSIS REPORT l
$. BC V260 BC-V206 m
$ NOTE: FOR THE PROPER VENTING OF RHR SUPPRESSION POOL BACKPRESSURE, FLOW CONTROL VALVES WATER COOLING BC LV F053A AND 8 (NOT SHOWN) MUST BE & SYSTEM TEST LINES OPENED DURING TESTING.
FIGURE 6.2 28
- (SEE LEGEND) SHEET 26 OF 48 AMENDMENT 6. 06%4
., . . . . . . t I
f i
l l
RPV !
r.. ,
i
- :. i i i-
- {
- l
,.C..L.OSED SYSTEM ! .
,.._CL.O.S.E.D
.. S..Y.S..T.EM, ,
i '
i ! ! !
i IP 2138
- I
! / i. i. -
' ?
ia j P 2isAl (
- i5 V T :
eincVo37 i
- .i W l
._._ _._. ,_. _.:_ ,y . y:::_._. . _.::: . ._ .:. 4 i
\
6 v 9
l
. I
- 1 a B : i 8 '
i
- ]',_. :
i ! *db e
[
- l l CLOsEo SYSTEu ,i l
l senv.
F066A d "C"V "3'^ i I
i
= aC PSV-l t
t :. :. Fosse t l
.. a I !. I 3C.PSV F007 I
BCMV4431s l
l BC-V253 BC-V256 f i
DETAIL 27 I
L 3
HOPE CREEK GENERATING STATION ;
FINAL SAFETY ANALYSIS REPORT [
DSER OPEN ITEM /3J (
l
! RHR RELIEF TO ;
SUPPRESSION CHAMBER LINES l l
I
! FIGURE 8 2 28
$SEE LEGEND) SHEET 27 OF 4B AMENDMENT 6,06/84 i
n a .- . . . . - . . _ . _ .. -
j I
l
~
t J b 4
, , --; l l BC-V325. >
d b f
l 1 P l'
., PRIMARY > BC V324 ;
d L RPV CONTAINMENT -
t I
I i t !
I i
- j
- i
?
YBC-V314 - ,
BC V015 !
BC V322 l j lP 214Al : ;
b :
i !
, i !
sc-V323 i :
\ .. '
/l\ ! t h ! !
_I...______._______.__. _ _ _ ___ __ _
j j
- l-I I
- i t I 8
4 I . . _ .i !
CLOSED SYSTEM i
. i
! ISOLATION VALVES DETAIL 28 l
P-214A P 2148 j i
scV015 BC-V112 1 (Y
D i N ,'W AND DRAIN VALVES l l
g F214A P 2148 e
- 900PE CREEK [
" BC V322 BC V336 GENERATN00 STATION a FINAL SAFETY ANALYSIS REPORT ,
h BC-V323 BC V337 o
BC V324 BC V353 RHR TO SUPPRESSION CHAMBER h BC V325 BC V354 SPRAY HEADER LINES l E
BC V315 f BC V314 FIGURE 5 2 28 7
- (SEE LEGEND) SHEET 23 OF 48 AMENDMENT 6.06'B4 I
I l
i t
i L
PRIMARY L RPV CONTAINMENT -
t I i
- I I.
I
- t. i. ,
I l
I
!. I i
l l CLOSED SYSTEM i ! -
I 3 _ _ ._ _ . . _ _ _ _ I _. ._ _ _ _ _ Y J l- [
l _f. I. -
l i
l :::
s .. :::
1 I
[
4 i.
P 216A l ,
r 1 r j P
%, j [ BE V110
.2 -
wB E V105 e[
BE V019 9 i t
ISOLATION VALVES T P 216A SE V019 DETAll 29 i M !
N P 2188 SE V020 i P-210C BE V018 g
H 8E V017 MOPE C3tEEK ,
P 2100 '
- GDERATING STATION
- FINALSANTY ANALYSls REPORT TEST AND DRAIN VALVES I a: P 216A BE V105 BE V110 CORE SPRAY PUMP l
$ P 2168 8E V104 BE VIII SUCTION LINES l P 216C BE V106 BE V109 P 2160 BE V107 8E V108 t FIGURE 6 2 28 !
l[ *(SEE LEGEND) SHEET 29 OF AB AMENDMENT 6.0614 . l c
- L re .. . . . . . . . . . . . . . . . . . \
I I
f I
l I
i f
i
, PRIMARY '
RPV CONTAINMENT l
,e. -. .. ~.C..L.O..S.E..D..S.Y..S.T...-...~...
E.M. !
V
- i
! l
. l l l- !
l BE-PSV F0128 ! ! t
%3 ,
9 r i : c
- i
' BE V117 : : !
d L j BE V101 l [
i :
i* .
BE V026 l t BE V102
( a
!. if :.
7 BE 012A SE V036 BE V127 .
f, P 217A l % ,j,~ i i O @ .. i !
\ Y BE V103 g' ! I
', .j
! .. i
. v -- . >
k P-2178 l BE V025 !
! I
% SE V118 . ..- :
I BE V128 :
.Y . ... ... ..... .. ..
y o :
BE V035 j q
r 8 J L :.
BE 150 l V : ,
........... J [
i DETAIL 30 !
I i
i MOPE CREEK GOGERATING STATION !
FINAL SAFETY ANALYSIS REPORT DSER OPEN ITEM /32 CORE SPRAY TEST TO SUPPRESSION CHAMBER LINES i
FIGURE 6.2 28
- L
- (SEE LEGEND) SHEET 30 OF 48 AMENDMENT 6. 06/B4 I
I I
I j
l h
- i
- i 4
- I !
PRIMARY g
! , RPV CONTAINMENT ! i i
GS V076 j GS V077 :
GS V082 d L '
$ . . GS V028 73 AO GS-V088 l AO AO = =
9 r ,
4 d k
- r1 ;
d Qu Qu FROM PRIMARY 2 L J ' CONTAINMENT GS PSV 5030 '
7 ' F' f
- HYDROGEN -
GS V080 V00p as.V006/ RECOMBINER d L V SYSTEM
- l f GS V116 d L i i r i d
i GS V186 i d L ;
LJ
.___.._.____________Y_
t E
DETAIL 31 (
i l
l F
HOPE CREEK -'
GENERATING STATION FINAL SAFETY ANALYS15 REPORT [
DSER OPEN ITEM /32 !
$UPPRESSION CHAMBER PURGE
& SUPPRESSION CHAMBER l VACUUM RELIEF LINE l FIGURE 6 2 28
- (SEE LEGEND) SHEET 31 OF 48 AMENDMENT 6. 06/84
I i
i I
V067 PRIMARY RPV CONTAINMENT [
. ; y )..... l GS V020 GS-V006 ,
i TO ORYWELL PURGE VENT LINE (DETAIL 13)
AO ' ,
b IP 2201 Gs v022 ,, y,,, ;
o
---De(-3 r
r, 1
j [GS-V115
%g , $g 2 FROM
, PRIMARY GS V010 MTAINMEM {
l HYDROGEN RECOMBINER I
3-.----..._______Y__
%, a $ 3 3 1 V )
GS.V038 Gs PSV 5032 ,
i GS-V068
, l l
i DETAIL 32 l i
i s
l !
l HOPE CREEK GENERATING STATION l FINAL SAFETY ANALYSIS REPORT ,
DSER OPEN ITEM /3A '
SUPPRESSION CHAMBER PURGE INLET & SUPPRESSION CHAMBER
- VACUUM RELIEF LINE l
FIGURE 6.2 28
- (SEE LEGENO) SHEET 32 OF 48 AMENDMENT 6. 06fB4 l l
t
i r"I t q
r EE V017 t
@' '@ lr
%k 2 $m a j w , ,,
EE V002 EE V001 l t
lP 222 ] EE V016 EE V020 !
i
_____ -::_- :_ ::?_-
I i
3:. .E. ,
i P 223 U
i e- '2 9 I EE V003 EESO'18 I b f l :
i g
(([EE V019 h EE V004 t 6
EE V011 i
[
DETAIL 33 i NOPE CREEK T GENERATING STATION FINAL SAFETY ANALYSIS REPORT DSER OPEN IT"M / 3 2. ;
SUPPRESSION POOL CLEANUP SUPPLY AND RETURN LINES ,
FIGURE 6.2 28
,(SEE LEGEND) SHEET 33 OF 48 AMENDMENT 6.06S4
p;, , .
l 1
l
. 1 l A.. l
- PRIMARY i
mpV CONTAINMENT
- i i
r a t .
/
. T . !
l G8 V082 G8 V046 "A !
8 G8N275 G8 V231 [P4Al l NOTE
. r .-
l I G8 V081 ,- G8 V048 8 V233 8 V232 } fGBV274 4 lp.gg l -
i a L m '
1 f NOTE 5
Cno m
e ra
$ s 5 i E i ISOLATION VALVES ISOLATION VALVES DETAIL 34 P48 G OV081 '
G8 V048 r4A l G8.V082 G8 V046 PMA G8 V083 G8 V070 G8 V071 Pm l G8V084 l
' ' i- WST VALVES
,' P4A G8 V231 G8 V230 GB V275 . MOPE CREEK i GENERATING STATION D488 G8 V237 G8 V236 G8 V241 FINAL SAFETY ANALYSl5 REPORT i TEST VALVES P48 G8 V233 ' G8 V232 G8 V274 CHILLER WATER SYSTEM LINES :
P 38A G8 V215 GB V234 G8.V240 NOTE: FIGURE 8.2 28 SHEET 34 OF 48 l THIS l$ A TEST TAP ON P 38A AND PJ88 ' AMENDMENT 6.06/B4
- (SEE LEGENO)
l 1
-e ,
- i. . . .
l l
PRIMARY l HPV CONTA NMEN'*
r ED V141
- -p4-ED V023 g Q ""
g pc pc ...... . !
--ED V024 % F \ j [yED V066 '
[ED V027 l P 29 l LJ LJ ;
ED V095 l
l ----D4- ;
l
\
ED V025 (m/
so ,,
\ ED V068 i
d f ED V069 L
d 1r DETAIL 36 !
(
l l \
HOPE CREEK i OENERATiteG STATIDN :
I DSER OPEN ITEM /3A FINAL SAFETY ANALYSIS REPORT i
REACTOR AUXILIARIES COOLING LINES FIG'JRE 6.2 28 i
- (SEE LEGEND) SHEET 36 OF 48 AMENDMENT 6 06/54 L
.~.. .
PRIMARY MPV CONT /.INMENT j
WITH0 R AWAL CONTROL - HYDRAULIC MOD III CONTROL DRIVE l P 36 l UNIT INSE RT (1)
P 36 l
1 DETAIL 36 l
l l
l (1) TYPICAL OF PENETRATIONS A THRU D
- HOPE CREEK OENERATING STATION DSER OPEN ITEM /3A FINAL SAFETY ANALYSIS REPORT CMD SYSTEM i
FIGURE 6.2 28
- (SEE LEGEND) SHEET 38OF48 AMEN 0 MENT 6. 0691
- - . . _ - - . . - . - . _ _ - . - . . . . - - ~ - - _ _ - - _ - .
i i
i i
t t
- I l
i
, i i
i PRIMARY
, RPV CONTAINMENT I
f
- I l
4
+ r I
i TYPICAL OF NUMEROUS !
INSTRUMENT PENETRATIONS l
. i DETAll 37 i l
i l i
f t
I I
t MOPE CRSEK OENERATING STATION DSER OPEN ITEM /3A ' I'8 TYPICAL OF INSTRUMENT LINES PENE.
TRATING PRIMARY CONTAINMENT AND ,
CONNECTING DIRECTLY TO REACTOR COOLANT PRESSURE BOUNDARY :
FIGURE 6.229 l
- (SEE LEGEND) SHEET 37 OF 4B AMENDMENT 6. 0644 l
l . ~. . . . . . . . . . . . .
1 .
l-2 ,
' . j t
PRIMARY RPV CONTAINMENT GS-V103 ,
- i GS V032 GS V102 gg.V031 l
,\ GS V061 GS V062 ..
t l
l DETAIL 38 ISOLATION VALVES 1 38 GS-V031 GS V032 i
J4A SK V008 SK.V000 q 44C
- SK.V006 SK.V006 [
M 44E GS V045 GS.V046 s '
AMC GS VN7 GS V048 fl b -
. TEST VA4.VES ,,,,g ,,,g '
4 38 GS V103 GS V081 GS V062 GS V102 GENERAT1000 STATION FINALSAFSTY ANALYSIS REPORT o J4A SK.V016 SK.V017 SK v024 SK.V007 J4C SK.V013 SK.V014 SK.YO25 SK V004 TYPICAL OF INSTRUMENT LINES !
h o PENETRATING PRIMARY CONTAINMENT I
Jef GS V006 GS V074 GS-V078 GS VDD4 AND CONNECTING DIRECTLY TO A10C GS V007 GS V075 GS-V079 GS V006 THE CONTAINMENT ATMOSPHERE FIGURE 5.2 28 -
SHEET 38 OF 48 AMENDMENT 6. 06M4
- (SEE LEGEPc)
_ _ _ _ _ - , _ _ _ . . . _ - _ . _ _ . - ______.__-_-[
. . . . . . . . . . . - .-. r i
- n. ,
I i
?
5 "- . , . i,
- v. -
i; 9
i l
l PRIMARY l RPV CONTAINMENT l
L
- . INSTRUMENTATION t
t g q== cpC .
l usAa / !
l i
i I
' DETAll 39 i
J4A se v563 J.7A se v565 i Aeo es v564 f A100 as.v566 r HOPE CREEK GENERATING STATION FINAL SAFETY ANALYSl3 REPORT _
DSER OPEN ITEM /3g TYPICAL OF INSTRUMENT LINES FOR SENSING CONTAINMENT ATMOSPHERE
- - FIGURE s.2 25 l
SHEET 30 OF 45 AMENOMENT 6.06/94
- *(SEE LEGEND)
L l
t L
i I
l
~
i GS V136 l f l d '
e, G&vos3 !
If L I
PRIMARY J CONTAINMENT ,
G&V064 i
J10 coo--h:
GSV104 GSV033 sc GSV034
(
r GS V066 ;
i l
l 1 P LGSV064 j (GS V106 3 7 ,
e LJ i
~
! LJ GSVMGJ L GSV066 !
LJ . >
PRIMARY ,,
ppy CONTAINMENT GS V112
,> d G5V111 GS V042 -
~\lJ 202 l cs.yo71 "
f f d k GS-V043 LJ r- m_a ,
% w- , I
. GS VC72 ,
DETAIL 40 HOPE CREEK ;
,SEE LEGEND) GENERATING STATION i FINAL SAFETY ANALYS18 REPORT ,
t HYDROGEN / OXYGEN i DSER OPEN ITEM /32 ANALYZER LINE l I
t FIGURE 6.2 23 SHEET 40 OF 48 AMENDMENT 6.0644
,_------y_ - , , . - __ _ - , _ , , , _ _ _ , _ , _ _ , . _ _,, _ , _ _ . , , , _ _ _ _ _ , , . , , , . . _ , , . - , _ , , , _ , , , , ,
, , , , _ _ _ .,_,.-,__,y7_.-.-__, , . , -
=
pp l
[
c PRIMARY CONTAINMENT ppy GP-V121 O w _a
- n_A -.. .
Ym y,
1 Y g NOTE 1 - t LJ
\ .
DETAll41 ISOLATION VALVES l
GP.V120 GP.122 J 36C NOTE 1: UNNUMBERED VALVE GP.V001 GP.V002 J-36D TEST VALVES J-30C GP 121 HOPE CREEK GENERATING STATION J360 GP.V132 FIN AL SAFETY AN ALYSi$ REPORT DSER OPE!! ITEM / 3 4. PRIMARY CONTAINMENT LEAKAG RATE TESTING LINES FIGURE 5.2 28 0, SHEET 410F 48 AMENDMENT 6.,,
- (SEE LEGEND) - - - . _ _ - _ _
4 ** e - ee e e
e s
r
[
I PRIMARY RFv CONTAINMENT I
e e t J 201 l I_
< .DC --
v100 GS v061 GS.V052
\ .W ... .
1 cS voe4s P
1 r GS V101 [ GSv006j [
LJ ga l
DETAIL 42 ,
( ISOLATION VALVES !
l J-201 GS V051 GS v052 l
l J 2io GS vo40 Gs vo41 i 1 J 212 GS vo49 GS v060
( en s . ,
t 5 TEST VALVES 2 Sani os viot os v0s4 as v0e6 cs viOO ,,,,,,,,,,
5 5250 os.v110 Gs v069 Gs v070 GS v109 OENERATING STATION ,
8 J212 as v0ee os vos3 cs vos5 cSv0es riNAL sArm Analysis REroRT 5 l lg TYPICAL OF INSTRUMENT LINES PENETRATING THE SUPPRESSION CHAMBER l
FIG U RE e.2 28 (SEE LEGEND) SHEET 42 0F 48 AMENOMENT 6. 06/84 i
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. . . . . . . . i J 207 GS V044 [
tamammmmu ;
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SENERATING STATION !
FINAL $AFETY ANALYSIS REPORT DSER OPEN ITEM / J.2 {
TYPICAL OF LINES FOR SENSING !
SUPPRESSION CHAMBER PRESSURE F l
I FIGURE 6.2 28
-E LEGENO) SHEET 43 0F AB AMENDMENT 6. 06/84 [
_ - -~ _ _ _ _ _ _ _ .
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l GENERATING STATION DSER OPEN ITEM / 3 .2 FINAL SAFETY ANALYSIS REPORT INSTRUMENT AIR TO SUPPRESSION CHAMBER LINES l FIGURE 8.2 28 e(SEE LEGEND) SHEET 44 OF 48 AMENDMENT 6. 06/84
I NOTE 1 PRIMARY RPV CONTAINMENT
- 1 Y i GP V005 d b m_2 ,
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e3 p GP V133 .
GP V004 .
i e BJ V500
]\ l J 209 l l
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NOTE 1: UNNUMBERED VALVE l
l DETAIL 45 l
MOPE CREEK OENERATING STATION l FINAL SAFETY ANALYSIS REPORT DSER OPEN ITEM /J4 SUPPRESSION POOL LEVEL LINE t
FIGURE 6.2 28 ,
- (SEE LEGEND) SHEET 46 OF Ab AMFNOMENT 6. 06/Pt
t 9
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I MOPE CREEK SENERATING STATION DSER OPEN ITEM /Jg FINAL SAFETY ANALYSIS REPORT ,
TYPICAL OF INSTRUMENT {
LINES FOR SENSING SUPPRESSION POOL LEVEL l
FIGURE 6.2 28 ;
E LEG ENO) SHEET 48 0F 40 AMENOMENT 6.0694
. -_-. .. - - . .. -,-- --.- -._--__, _-.-._ -.-.. . . - _ . - .--._...)
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3 7 T1 PRIMARY j g 3 g RPV CONTAINMENT RC V032 RC V0474 L j I,
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_ _ _ _ tTm _ x' 'J2 RC4V 0643A RC4V 0643B I P 227 j OR r-
- Il pq :
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RC V033 1 F 1 P
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ISOLATION VALVES 990LATION VALVES i
J206 RC4V4728A RC4V47298 l P 227 RC4V4643A RC4V46438 J 220 RC4V 0707A RC4V4707s F7E RC4V 4730A RC4V 07308 DETAIL 47 J 221 RC4V 0729A MC4V47298 RC4V473tA RC4V473tg y A10E
- M *
! s TEST VALVES P 227 RC V084 RC V032 RC V047 RC V024 J7E RC-V066 RC V036 RC V066 RC V004 GENE IN TION g J 10E RC V057 RC V037 RC V068 RC V005 FINAL SAFETY ANALYSIS REPORT i a.
O i g TEST VALVES ~
lg J 20s RCV049 RC V033 MC V060 RC.V001 l
l J 220 RC v051 RC V034 RC vos2 RC V002 FlouRE s.2 N' J221 RC V063 RC V035 RC VD64 RC V003 SHEET 47 OF 48 AMENDMENT 6,06/84_ ,
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PRIMARY RPV CONTAINMENT ] fRC V041 l
- Jm
- Jm
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. RC V040 RC4V 8803A RC4V 88038 i
RC V029 C V042 l
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_\ XV.3734A l J17Fl DETAIL 48 I
l Hortenatx OENIMATING STATION FINAL 8AFETY ANALYSIS REPORT DSER 0PE!I ITEM /34 POST. ACCIDENT L10UID SAMPLING AND REACTOR INSTRUMENTATION
- tSEE LEGENOl SHEET 48 0F 48 AMENDMENT 4. 06.34, l
tv, . . .- . . . . . . . . .. . . . . . .
i i
l M -NORMAL OPEN GLOSE VALVE !
I M -NORMAL CLOSED GLOSE VALVE f
l l
- NORMAL OPEN GATE VALVE I
N - NORMAL CLOSED GATE VALVE
- CHECK VALVE M
- MOTOR OPERATOR g/ - AIR OPERATOR
- RESTRICTION ORIFICE S'W *
-CHECK VALVE (WITH HAND OPER ATOR)
~
, - REl.lEF VALVE T
5
. - SOL 3NOID OPER ATOR I
% -STOPCHECK VALV8!
MOPE CREEE
' . GENER Aftes4 STATION I FINAL SAFETY ANALYSil AtiCNT
.l DSER OPEN ITEM /Jg LEGEND F64URE 62.ame SMEET1OF2 AMENOMENT4.644
[
en . . .
4 h4 - EXPLOSlVE VALVES .
$ -INDICATES CONTAINMENT ISOLATION VALVE dh -SPECTACLE FLANGE O -
- BALL VALVE y .
- 3:
- SCREEN ,
O-
-CHECK VALVE WITH MANUAL LEVER i
- SPRAY NOZZLE
/IN
-SUPPRES$10N CHAMBER i
l r1
! T - UNVALVED TEST CONNECTION M
l N HOPECREEK h.
w - GENERATING STATION l
FINAL SAFETY AN ALYlls REPORT h - D RAINPOT o v LEGEND F10URE 412Se SHEET 2 0F 2 AMENDMENT 6,644
E HCGS DSER Open item No. 134 (08ER Section 6.2.6)
CONTAINMENT LEAKAGE TESTING Theapplibant has provided information to demonstrate com-pliance with the testing requirements of Appendix J to 10 CFR 50. As a result of this review, the staff finds that the information provided by the applicant is not complete.
We are awaiting information from the applicant outlined in i
Section 6.2.4, in order to properly conduct our review of containment leak testing. We will evaluate this information when it becomes available and report on this matter in a supplement to this SER.
l RESPONSEi ;
, For the information requested above see the response to DSER l Open Item 132. ,
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HCGS DSER Open Item _No. 1410 (Section 9.1.3)
SPENT FUEL POOL COOLING AND CLEANUP SYSTEM Additionally, the information provided through Amendment 3 was not sufficient for the staff to complete its evaluation of the spent fuel pool sampling and monitoring. To complete the review, the following information is needed (1) Describe the sampling proceduro, analytical instrumentation, and sampling frequency for monitoring spent fuel pool purity.
(2) Stato the radiochemical limits for initiating correctivo action.
O ,
The applicant's responso should consider permissiblo gross gamma and iodine activities and the domineralizer decontamination factor.
HE8PON8E FSAR Section 9.1.3.2.2.4 has boon revised to provide the requested information. ,
t M PU4 95/05 2-dh
HCGS FSAR The stainless steel filter-demineralizer vessels are of the >
pressure precoat type. A tube nest assembly consisting of the >
tube sheet, clamping plate, filter elements, and support grid is ;
inse~ted c as a unit between the flanges of the vessel. The filter elements are stainless steel and are mounted vertically in the i vessel. Air scour connections are provided below the tube sheet, I and vents are provided in the upper head of each vessel. The ;
filter elements are installed and removed through the top of each vessel. The holding elements are designed to be coated with powdered ion exchange resin as the filtering medium. ,
The fuel pool filter-demineralizers maintain the following !
effluent water quality specifications:
l I
Specific conductivity at 250C, micromho/cm 50.1 i-pH at 250C 6.0 to 7.5 l Heavy elements (Fe, Hg, Cu, Ni), ppm 0.05 Silica (as SiO,), ppm <0.05 ;
- t Chloride (as Cl-), ppm <0.02 l
Total insolubles, ppm 90% removal to a k i minimum of 0.01 ppm i l- Pn s er-t A y The filter-demineralizers are designed to be backwashed periodically with water to remove resin and accumulated sludge from the holding elements. Service air pressure loosens the ,
l material from the holding elements and the backwash slurry drains l through the gravity drainline to the waste sludge phase separator ,
! in the solid waste management system.
l The resin tank provides adequate volume for one precoating of one filter demineralizer vessel.
The resin eductor transfers the precoat mixture of resin to the holding pump suction line at a flow rate of 4 gpm.
The holding pumps are designed to recirculate a uniform mixture of resin through the filter-demineralizer vessel being precoated at a flow rate of 1.5 gpm/fta of filter element surface area, and ,
to automatically start and maintain the precoat material on the l filter elements when the system flow rate falls below the value ;
necessary to keep the precoat on the elements. !
DSER OPEN ITEM / 9.1-21 w-W -m--- - - - * - - - - - - - - _ _ _ -
INSERT A The influent and effluent water of the FPCC is continuously monitored by on line PH and ' conductivity instrumentation.
In addition, grab samples of the influent H2O will be analyzed once per week for chloride and for gamma isotopic and once per month for heavy metals (Fe, Cu, Hg, Ni) . Grab samples of ef fluent water will be analyzed weekly for chloride, silica, suspended solids, tritium, and for gamma isotopes.
Decontamination factors (df) of greater than 10 are expected for any chloride present and greater than 5 for isotopes of Iodine and Cobalt. Resin bed (s) will be regenerated and/or replaced when these dfs are not achieved.
O DSER OPEN ITEM 1419 M P84 95/05 3-dh
i i
t HCGS !
DSER Open Item No. 145 (DSER Section 9.2.2)
ISI PROGRAM AND FUNCTIONAL TESTING OF SAFETY AND TURBINE !
AUXILIARIES COOLING SYSTEMS [
I RESPONSE ,
This item is not an open item per telephone conversation .
2 between J. M. Ashley (PSE&G) and John Ridgely (NRC-ASB) on i April 23, 1984. i F
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M P84 95/05 4-dh !
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TELEPHONE NOTES PSE&G Hope Creek Licensing (Bethesda)
DATE: April 23, 1984 FROM: J.M. Ashley l
TO: Dave Wagner
SUBJECT:
HCGS DSER Open Hem 145 Discussion (This telecon corrects my telecon of 3/22/84 on HCGS DSER Open Items - that telecon referred to Open Item 145, FSAR Section 9.2.2.
The reference should have been Open Item 146, FSAR Section 9.2.6.)
Ashley called to find out what NRC staf f concerns existed with respect. to FSAR section 9.2.2 (Item 145). ;
Wagner said-he had checked the section and that the NRC has no ,
, outstanding concerns with the section. The item was inadvertently l listed.
3 I A+
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l DSER oPEN ITEM /YO
e HCGS DSER Open Item No.146 (DSER Seetion 9.2.6)
SWITCHES AND WIRING ASSOCIATED WITH HPCI/RCIC TORUS SUCTION 8
RESPONSE
Tais item is closed per the March 22, 1984 telecon between J. M. Ashley (PSE&G), and J. Ridgely (NRC).
- 4 e O
~ ' "' M*v w v.v-+ _ _
e TELEPHONE NOTES PSE&G Hope Creek Licensing (Bethesda)
Date: March 22, 1984 From: J.M. Ashley .
To: D. Wagner, J. Ridgely (ASB)
Subject:
HCGS DSER Open Items Discussion Ashley called to find out what NRC concerns existed with y 8py ,f,q
- l respect to FSAR Sections 3.5.1.2 (Item 30), 9.2.2 (Item 1 and$ .e metspwone 9.4.4 (Item 152). u.$ es.
Ridgely explained that these' items were inadvertently listed as open items in the listing of open items at the front of the DSER. The NRC has no outstanding concerns with the sections.
- n-DSER OPEN ITEM f4/6 i
t i
TELEPHONE NOTES PSE&G Hope Creek Licensing (Bethesda) :
I DATE: April 23, 1984 ,
FROM: J.M. Ashley TO: Dave Wagner i
SUBJECT:
HCGS DSER Open Hem 145 Discussion .
1 (This telecon corrects my telecon of 3/22/84 on HCGS DSER Open r Items - that telecon referred to Open Item 145, FSAR Section 9.2.2. '
The reference should have been Open Item 146, FSAR Section 9.2.6.) ,
Ashley called to find out what NRC staf f concerns existed with respect to FSAR section 9.2.2 (Item 145).
Wagner said he had checked the section and that the NRC~has no };
outstanding concerns with the section. The item was inadvertently listed.
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DSER 01EN ITEM /d/h ,
.. . . . . . t l
HCGS DSER Open Item No.158 (DSER Section 9.5.1.5.a) !
CLASS B FIRE DETECTION SYSTEM I The applicant is providing a detection system in accordance f with NFPA 720. The portion of the system that protects the ;
reactor building is a Class A supervised system. The portion i of the system that protects the balance of the plant is provided with Class B supervision. This does not meet staf f guidelines. ,
A Class B system will not continue to provide detection capability l if a fault occurs in the system. l 1
The staff will require the applicant to provide a system that complies with NFPA 72D for a Class A system, with detectors ;
installed in accordance with NFPA 72E. J
RESPONSE
A class A early warning fire and smoke detection system, designed l in accordance with NFPA 72D, is provided in the reactor building (
l and in the following safety-related equipment areas which can be -
used for the first four hours of hot shutdown:
- a. 250 Vdc battery and inverter room for the RCIC system, j
- b. 250 Vdc battery and inverter room for the HPCI system, i
l c. Cable spreading room,
- d. Switchgear Room B, f l !
I e. Control equipment room.
I In addition, a Clasr. A system is provided in the remote shutdown i panel room. f The Class A system will initiate a fire alarm even with a single break or single ground f ault of the detection circuit I between the local control panel and the associated detectors. t L
The above design was discussed with the NRC at the DSER fire f l
protection open item meeting held April 18, 1984. The staff ;
indicated that this design would be acceptable. As justified !
be low , we have not provided a Class A system for two areas ,
used for initial hot shutdown. j The remainder of the plant is provided with a Class B early !
warning fire and smoke detection system designed in accordbnce l with NFPA 720. This includes two areas used for initial hot l
shutdown : >
t
) a. Control equipment mezzanine,
- b. Main control room.
1 h
158-1 ,
- - + . . . - _ _ _ _ -_ _ _ - - ~ _ ,,.___._-..m-,,,y....,_m-._.,_,._,._,,-__,y_. ,-_________,_,,.___,._.._.,,...--__,,._-.m_, - - _ - _
. I i
l REPSONSE (Cont'd)
The control equipment mezzanine detection system is backed up by a separate fire detection and alarm system for the automatic ,
CO 2 system and the control room is continuously manned, thus, The -
providing the equivalent of a Class A system in both areas. j
. safety-related equipment areas with Class a systems are located -
outside the reactor building and are accessible during nonmal plant operation.
i, The Class B system is electrically supervised to detect circuit breaks, ground f aults, removal of a detection device from a l detector circuit, and power failure. If any of these problems occur, fire detection system trouble is annunciated locally and i in the main control room. This would alert the operators to begin repairing the system immediately to bring it back to working order without jeopardizing plant safety.
j Except for the dif ferences stated in FSAR Section 9.5.1.6.15, j
the fire and smoke detectors are installed in accordance with [
NFPA 72E. Section 9.5.1.6.14 has been revised to clarify the e
extent of Class A and B systems at HCGS. r r
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- + ~ ~.-----mr, -. - , , , , , - - - - . ,-,,,-.n--,-,--,n_,-----,,n--, , , - _ , - . ~ . - - - ~ - - , - - , - , , - - _ , , , . . - - - - . - . , . . , . - - , - - , - , - - . - - - - - - , - -
areas, switchgear rooms, and other areas where potential exists for heavy smoke conditions.
As stated in Section 9.5.1.1.15.1, smoke and corrosive gases are generally discharged to the outside using the normal ventilation.
system and a separate smoke removal system provided for the l control area. Since the switchgear rooms and diesel generator l
' control rooms, at floor elevation 130 feet, and the heating and i ventilation equipment rooms, control equipment room and inverter !
rooms, at floor elevation 163 feet 6 inches of the diesel area, l are not provided with direct exhaust to the outside, portable ;
blowers will be provided for the fire brigade to use as smoke !
ejectors. l All battery rooms are provided with air exhaust systems that can l discharge smoke and gases directly outside.
9.5.1.6.13 Paragraph C.5.g.(3)
Paragraph C.S.g.(3) states that a fixed emergency communication ,
l system independent of the normal plant communication system !
I i
should be installed at preselected stations.
l >
l HCGS has no fixed emergency communication system that is solely provided for emergency situations. The plant has an intraplant, i five chaanel, page-party, public-address (PA) intercom system. l The in-plant communication system is provided with reliable ;
uninterruptible power supply (UPS) for uninterrupted j communiciations between all areas of the plant. At least one !
, channel of the PA system will be assigned for emergency use. In i addition, portable radio communication units will also be j j available during an emergency situation. ;
i l f 9.5.1.6.14 Parag'raph C.6.a.(2)
! Paragraph C.6.a.(2) requires that the fire detection systems !
comply with the requirements of Class A systems, as defined in i NFPA 7 2D G r="=t' s"'vedc'a'n P"nel **** *n p Zdg& )% l a the nd che solch rels= led e-g u.'p><wi I ar eas use.d 5-or in;WI ho+ Jhut down . ,cly w eniaj [
At HCGS, tne reactor buildingA)( provided withj a Class Aj. fire and sm arc t a re detecticn system in accordance with NFPA 72D.A The balance of the plant, however, is provided with a Class Bydetection system, '
which also complies with NFPA 72D. The Class B ystem is i electrically supervised to detect circuit breaks, ground faults, '
.nser4 o e,,1y u.,w-n,my D t' ;
DSER OPEN ITEM 883 8'"#
/[$' 9.5-51 Amendment 5 i
i
--. ,,_---.-. - ,n,.n,v, _n-,, _ , . _ _ _ . - _ - . . . . _ , , , , - . - - - . , - - ,
INSERT A The areas outside the reactor building are the 250V DC RCIC and HPCI battery and inverter rooms at Elevation 54 ft., the cable spreading room at Elevation 77 ft., the control equip-ment room at Elevation 102 ft., the switchgear room B at elevation 130 ft. of the auxiliary building SDG area., and the remote shutdown panel room at elevation 137 ft. The Class A system will not prevent the detection system from initiating a fire alarm due to an occurance of a single brea'k or single ground fault of the detection circuit between the local control panel and the associated detectors.
INSERT B This includes two areas; the control equipment mezzanine and the main control room for initial hot shutdown. The control equipment mezzanine detection system is backed up by a separate fire detection and alarm system for the automatic CO2 system, and the control room is continuously manned.
The safety related equipment areas with Class B system are located outside the reactor building and are accessible during normal plant operation.
DSER OPEN ITEM 158 MP84 93 06 2-vw
I i
removal of a detection device from a detector circuits, and power !
failure. If any of the above problems occur, a fire detection 2- system trouble is annunciated locally and in the main control :
room. A Plant operation will periodically test the system for i proper functioning, similar to inservice testing of other plant systems. l J;nsert C. !
At HCGS, the fire and smoke detection system is in compliance with NFPA 72D except that the operation and supervision of the ,
system is not the sole function of the plant operator. The plant operator's duties cover operation of the generating station and [
monitoring and supervi, sing the fire protection systems. l Paragraph C.6.a.(3)'
9.5.1.6.15 l !
l i
Paragraph C'.6.a.(3) requires that the fire detectors be installed !
in accordance with NFPA 72E. t I - t I,i At HCGS, the location of early warning fire and smoke detectors !
'l was determined and performed under the direction of a registered ;
- fire protection engineer. The location of the fire and smoke !
,, detectors complies with the guidelines of NFPA 72E except for the t i location of ionization and photoelectric detectors in high-bay l _' areas. The detectors are not located in each bay formed by deep :
t beams. NFPA 72E allows detector locations to be determined based ,
on engineering judgement considering ceiling shape, ceiling
- l.
- surfaces, ceiling height, configuration of contents, combustible -
characteristic and ventilation. j At locations in areas where composite construction is used, the diffusion of combustion particulates throughout the compartment f volume produced during the incipient and smoldering stages of the !
fire will negate the effect of beam depth and result in !
l acceptable levels of detection coverage.
9.5.1.6.16 Paragraph C.6.a.(6) l P
Paragraph C.6.a.(6) requires primary and secondary supplies be [
', provided for electrically operated control valves conforming to ,
NFPA 72D. ,
l ;
1 i
e DSER OPEN ITEM /31f l 9.5-52 Amendment 4 1 i !
INSERT C This would alert the operators to begin repairing the system immediately to bring the early warning fire and smoke detection system back to working order without jeopardizing plant safety.
DSER OPEN ITEM 158 MP84 93 06 3-vw l
IL
HCGS ,
DSER Open Item 174 (Section 13.5.2)
RESOLUTION EXPLANATION IN FSAR OF TMI-2 ITEMS I.C.7 and-I.C.8 The FSAR should be revised to provide a brief explanation of how Task Action Items I.C.7 and I.C.8 have been resolved, as described in 13.5.2c herein.
RESPONSE
FSAR section 1.10, items I.C.7 and I.C.8, have been revised to provide the requested in, formation.
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M P84 95/05 5-dh l
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- I.C.6 VERIFY CORRECT PERFORMANCE OF OPERATING ACTIVITIES ,
f Position i
It is required (from NUREG-0660) that licensees' procedures be L l
reviewed and revised, as necessary, to assure that an effective I
' system of verifying the correct performance of operating l activities is provided as a means of reducirg human errors and i j
improving the quality of normal operations. This will reduce the '
frequency of occurrence of situations that could result in or l contribute to accidents. Such a verification system may include i automatic system status monitoring, human verification of -
operations, and maintenance activities independent of the people ,
performing the activity (see NUREG-0585, Recommendation 5), or i both.
i
Response
l Verification of 6perating activities to provide a means of !
reducing human errors and to improve the quality of normal !
l' operations will be incorporated in Operations Department Administrative Procedure OP-AP.ZZ-020, Operations Management Audit Program (available March 1, 1985).
I
!.
- I.C.7 NSSS VENDOR REVIEW OF PROCEDURES
?
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Position r
t Obtain nuclear system supply system vendor review of power- t ascension and emergency operating procedures to further verify ;
their adequacy.
i
Response
L All startup test procedures from core load through power ,
ascension will be reviewed by GE. This review, as well as vendor review of test results, will be documented prior to commercial i operation. l
[
i lrW3cRT A w
DSER OPEN ITEM f7[ ,
1.10-21 Amendment 1 j r
\.
INSERT "A"
-The HCGS Emergency Instructions will be developed based on the NRC - approved BWR Owners Group Emergency Procedure Guidelines (EFGs). Due to GE's involvement in the development of the EPGs, it has been determined that an additional NSSS vendor review of the plant specific Emergency Instructions is not necessary.
DSER OPEN ITEM 174 MP84 93 06 4-vw
_ _ . .- - . _~ - . __.
r l l
- I.C.8 PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES FOR NEAR-TERM OPERATING LICENSE APPLICANTS (
,! Position Correct emergency procedures as necessary based on the NRC audit ;
of selected plant emergency operating procedures (e.g., small- r break loss-of-coolant accident, loss cf feedwater, restart of engineered safety features following a loss of ac power and i steam-line break). j
. i
. Response -g e,,lg,$6 ,
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INS EM b 3 rgency procep res /t11 be itten oil ng th Guid guide nes !
'.' All e of I 82- 7, Emprgenq:/ Operat ng Pr edur Writin ine, l L
! an the gu delines of the BWR nersproup ocedur Co tee, l' a long e the uidefines do not c trad t exis ng N '
ese rocedur will e av ilable y Mar 1, 985. l li irecti s.
'l ,
wil be mad , as cess y, bas on y aud s Cor ectio o these roc ures. p/ l
- I.D.1 CONTROL ROOM DESIGN REVIEWS i
Position ;
i
' ' Licensees and applications for operating licenses are required to conduct a detailed control room design review to identify and l
correct design deficiencies. This detailed control room design
}
review is expected to take about a year. Those applicants for i l
. operating licenses who are unable to complete this review prior to issuance of a license shall make preliminary assessments of 7 their control rooms to identify significant human factors and instrumentation problems and establish a schedule approved by us for correcting deficiencies. These applicants will be required l
to complete the more detailed control room reviews on the same
' ' schedule as licensees with operating plants. l I
. L j . !
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l DSER OPEN ITD{ /7p/
1 1.10-22 Amendment 1 ,
(
L _ _.-_ _ . _ _____ _ _ _ __ 1 _ _
'_ , l INSERT "B" A Procedure Generation Package (PGP) will be prepared in accordance with Supplement 1 to NUREG-0737. The PGP and plant specific Emergency Instructions will be based on the NRC - approved BWR Owners Group Emergency Procedure Guidelines (EPGs). As a result, it has been determined that
- a NRC review of the plant spe'cific Emergency Instructions is not necessary.
DSER OPEN ITEM 174 MP84 93 06 1/5-vw
.