ML20096G849
| ML20096G849 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/07/1984 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.25, TASK-TM NUDOCS 8409110175 | |
| Download: ML20096G849 (274) | |
Text
_- -
Pubhc Senace O PS G Company Electnc and Gas 80 Park Plata, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570. Newark, NJ 07101 Robert L. Mitti General Manager
' Nuclear Assurance and Regulation September 7, 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 wnich provides a status of the open items idertified in Section 1.7 of the Draft Safety Evaluation Report (SER). Items identified as " complete" are those for which PSE&G has provided responses and no confir-mation of status has been received from the staff. We will consider these items closed unless notified otherwise. In order to permit timely resolution of items identified as
" complete" which may not be resolved to the staff's satis-faction, please provide a specific description of the issue which remains to be resolved.
Attachment 2 is a current list which identifies Draft SER Sections not yet provided.
In addition, enclosed for your review and approval (see Attachment 4) are the resolutions to the Draft SER open items, and FSAR question responses listed in Attachment 3.
Please note that the proposed change to FSAR Section 12.3.4 in response to D5ER Item No. 166, covers radiation protec-tion items discussed via a telecon between Charles Hinson (NRC-RAB) and Russell Lovell (PSE&G) regarding supplon. ental radiation monitoring. PSE&G is making a commitment tc pro-vide the additional information requested by July 1, 1985.
It is our understanding that identifying the location of supplemental monitors may be handled as a comfirmatory item pending radiation protection / health physics inspections.
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- Also,< enclosed for;your review'(see Attachment 5) is a copy lof the, revised FSAR'..Section-l'.10 , Item II.K.3.25 as
, re'queated.by'G..ThomasJof?the_ Reactor Systems Branch.
In' reviewing.DSER-Appendix A, PSE&G.has identified the following'discrepanciest La . :LThe tag' number;of the Reactor Water' Cleanup Filtar.
-.Demineralizer Hoist'is incorrectly' entered in
" Table ' 2. l'~ of DSER: Appendix A . as 10H203. The cor-rect tag: number-was.10H213. As described.in the
~
' attached DSER responses, two new hoists, lAH220 and
.lBH220,:have
- replaced 10H213.
- b. The _ tag numbers of the Diesel Generator Underhung Crane, lAH400 through.1DH400, are incorrectly.
entered in Table'2.1 as OAH301 through ODH301.
- Should you have any questions._or_. require any additional
' - - information on these open items, please contact us.
1 --Very truly yours, Q[ 4'hd c.
I
- - }
Attachments / Enclosure
-, C - D. H. Wagner
~
'USNRC Licensing' Project Manager W. H. Bateman
.USNRC Senior' Resident Inspector 1
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DATE: 9/7/84 ATTACHMENT 1 DSER R. L. MITIL TO
' OPEN - SECTION A. SGWENCER ITEM ~ NUMEER SUh1ECT STATUS LETTER DATED 1 2.3.1 Design-basis temperatures for safety- Cmplete 8/15/84 related auxiliary systes 2a 2.3.3 Accuracies of meteorological Cmplete 8/15/84 measurements (Rev. 1) 2b 2.3.3 Accuracies of meteorological Cmplete 8/15/84 measurements (Rev. 1) 2c 2.3.3 Accuracies cf meteorological Cmplete 8/15/84 measurements (Rev. 2) 2d 2.3.3 Accuracies of meteorological Cmplete 8/15/84 memurements (Rev. 2) 3a 2.3.3 Upgrading of onsite meteorological Cmplete 8/15/84 measurements progra (III.A.2) (Rev. 2) 3b .2.3.3 Upgrading cf onsite meteorological Caplete 8/15/84 measurements program (III.A.2) (Rev. 2) 3c 2.3.3 Upgrading cf onsite meteorological NRC Action measurements progran (III.A.2) 4 2.4.2.2 Ponding levels Cmplete 8/03/84 Sa 2.4.5 Wave impact and runup on service Cmplete 9/7/84 water intake structure (Rev. 2)
Sb 2.4.5 Wave impact ard runup on service Cmplete 9/7/84 water intake structure (Rev. 2)
Sc 2.4.5 Wave impact and rurup on service Cmplete 7/27/84 water intake structure
- Sd 2.4.5 Wave impact ard runup on service Cmplete 9/7/84 water intake structure (Rev. 2) 6a 2.4.10 Stability cf erosion protection Cmplete 8/20/84 structures
.6b 2.4.10 Stability cf ermion protection Cmplete 8/20/84 structures 6c 2.4.10 Stability cf erosion protection Cmplete 8/03/84 structures M P84 80/121-gs
J' AFUOBENT 1 (Cont'd)
DSER R. L. MITTL 1D CPEN SBCTIGi A. SORENCER ITBt IE.BSER SUR7BCT STATUS 12TTER DMD 7a 2.4.11.2 Thermal aspects of ultimate heat sink. Caplete 8/3/84 7b 2.4.11.2 Thermal aspects of ultimate heat sink Coglete 8/3/84 8 2.5.2.2 Choice of maximum earthquake for New Ocuplete 8/15/84 England - Piedmont Tectonic Prtwince 9 2.5.4 Soil damping values Complete 6/1/34 10 2.5.4 Poundation level response spectra Complete 6/1/84
-11 2.5.4 Soil shear moduli variation Complete 6/1/84 12 2.5.4 Combination of soil layer properties Ocuplete 6/1/84 13 2.5.4 Iab test shear moduli values Ccaplete 6/1/84 14 2.5.4 Liquefaction analysis of river bottcza Couplete 6/1/84 sands 15 2.5.4 Tabulations of shear noduli Couplete 6/1/84 16 2.5.4 Drying and wetting effect on Ca plete 6/1/84 Vincentown 17 2.5.4 Power block settlement monitoring Ca plete 6/1/84 18 2.5.4 Maximum earth at rest pressure Caplete 6/1/84 coefficient 19 2.5.4 Liquefaction analysis for service Cm plete 6/1/84 water piping -
20 2.5.4 Explanation of observed power block Complets 6/1/84 settlement 21 2.5.4 Service water pipe settlement records Caplete 6/1/84 22 2.5.4 Cofferdam stability Cm plete 6/1/84 ht78480/122-gs
A11200ENT 1 (Cont'd)
R. L. MF11L 10 DSER A. SOMNCER CPSI SECTIG4 SLR7ECT S'DGts IEr11lR DN!ED ITEM NLDSER 2.5.4 Clarification of FSAR Tables 2.5.13 Complete 6/1/84 23 and 2.5.14 2.5.4 Soil depth models for intake Camplete 6/1/84 24 structure 25 2.5.4 Intales structure soil modeling CJIplets 8/10/84 2.5.4.4 Intake structure slidirg stability Cceplete 8/20/84 26 2.5.5 Slope stability Ccaplete 6/1/84 27 28a 3.4.1 Flood protection Conglete 8/30/84 (Rev. 1) 28b 3.4.1 Flood protection Caglete 8/30/84 (Rev. 1) 3.4.1 Flood protection Conglets 8/30/84 28c (Rev. 1) 3.4.1 Flood protection Canplete 8/30/84 28d (Rev. 1) 3.4.1 Flood protection Ccaplete 8/30/84 28e (Rev. 1) 3.4.1 Flood protection Complete 7/27/84 28f 28g 3.4.1 Flood protection Complete 7/27/84 3.5.1.1 Internally generated missiles (outside Canplete 8/3/84 29 (Rev. 1) contairunent) 3.5.1.2 Internally generated missiles (inside closed 6/1/84 30 contairunent) (5/30/84-Aux.Sys.Mtg.)
31 3.5.1.3 Turbine misslies Complete 7/18/84 32 3.5.1.4 Missilius generated by natural phencuena Ccaplete 7/27/84 3.5.2 Structurve., systems, and canponents to Couplete 7/27/84 33 be r,cotxted from externally generated missiles N P84 80/12 3 - gs
JGT4099fr 1 (Cont'd)
DSER R. L. 4r11L 1D GWI SBCTIGI A. SOBEN3R ITWI Nt3eER stb 3BCT STARE IETNR UGED 34 3.6.0 thitestrained whipping pipe inside Ccaplete 7/18 /84 containment i 35 3.6.2 ISI program for pipe welds in Ccuplets 6/29/84 hreak exclusion sons 36 3.6.2 Postulated pipe ruptures complets 6/29/84 37 3.6.2 Feedwater isolation check valve Ccuplete 8/20/84 cperability 38 3.6.2 Desigt of pipe rupture restraints complete 8/20/84 39 3.7.2.3 SSI analysis results using finite Complete 8/3/84 element method and elastic half-space approach for contairunent structure 40 3.7.2.3 SSI analysis results using finite Complete 8/3/84 element method and elastic half-space -
approach for intake structure 41 3.8.2 Steel contairment buckling analysis Ccaplete 6/1/84 42 3.8.2 Steel contairunent ultimate capacity Canplete 8/20/84 analysis (Rev. 1) l 43 3.8.2 SRV/IDCA pool dynamic. loads Ccaplete 6/1/84 44 3.8.3 ACI 349 deviations for internal Ccaplete 6/1/84 structures 45 3.8.4 ACI 349 deviations for Category I Ccuplete 8/20/84 structures (Rev. 1) 44 3.8.5 ACI 349 deviations for foundations Complets 8/20/84 (Rev. 1) 47 3.8.6 Base mat response spectra Ccaplete 8/10/84 (Rev. 1) 48 3.8.6 Rocking time histories Co plete 8/20/84 (Rev. 1) r M 704 80/12 4 - gs
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ATU O M Nr 1 (Cont'd) ,
DSER R. L. MITIL 10 '
GSI SBCTIGI A. SCMENCER ITBE MMIER SUBJECT STATUS IErNlR UGED 49 3.8.6 Gross concrete section C a plete 8/20/84 (Rev. 1) 50 3.8.6 Vertical floor flexibility response Caplete 8/20/84 spectra (Rev. 1) 51 3.8.6 Camparison d Bechtel independent Cm plete 8/20/84 verification results with the design- (Rev. 2) basis results 52 3.8.6 Ductility ratios due to pipe break Ccaplete 8/3/84 53 3.8.6 Desip d seismic Category I tanks Ccmpleta 8/20/84 (Rev. 1) 54 3.8.6 Ccabination d vertical responses ccuplete 8/10/84 (Rev. 1) 55 3.8.6 Torsional stiffness calculation Ccaplets- 6/1/84 56 3.8.6 Drywell stick model development Ccmplete 8/20/84 (Rev. 1) 57 3.8.6 Rotational time history irputs Complete 6/1/84 58 3.8.6 "0" reference point for auxiliary Ccaplete 6/1/84 building model 59 3.8.6 Overturning mment d reactor Ccmplete 8/20/84 building foundation mat (Rev. 1) l 60 3.8.6 BSAP elenent size limitations Ccuplete 8/20/84 (Rev. 1) ,
i 61 3.8.6 Seismic modeling d drywell shield Ccaplate 6/1/84 M1 62 3.8.6 Drywell shield wall boundary coplete 6/1/84 conditions 63 3.8.6 Reactor building dme boundary Ccsplete 6/1/84 ;
conditions ,
! M 784 80/12 5 - gs
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l A2'DOBENF 1 (Cont'd)
R, L. MITIL *10 DSER 9 96 SECTICBI ,
A.SONEMER IU SER SUWBCf S'th2tB IATIER DATED FITM 64 3.8.6 SSI analysis 12 Hz cutoff frequency Cceplete 8/20/84 (Rev. 1) 65 3.8.6 Intake structure crane heavy load ccuplets 6/1/84 66 3.8.6 Dupedance analysis for the intake Caglete 8/10/84 structure (Rev. 1) 67 3.8.6 Critical loads calculation for Ca plete 6/1/84 reactor building &me 68 3.8.6 Reactor building foundation met Complete 6/1/84 contact pressures 69 3.8.6 Factors of safety against sliding and Ccaplete 6/1/84 overturning of drywell shield wall 70 3.8.6 Seismic shear force distribution in Caplete 6/1/84 cylinder wall 71 3.8.6 overturning cf cylinder wall Canplete 6/1/84 72 3.8.6 Deep besa design cf fuel pool walls Ccuplete 6/1/84 73 3.8.6 ASHSD done nodel load irputs Complete 6/1/84 74 3.8.6 Tornado depressurization Canplete 6/1/84 75 3.8.6 Auxiliary building abnommal pressure Ccaplete 6/1/84 76 3.8.6 Targential shear stresses in drywell Complete 6/1/84 shield wall and the cylinder wall 77 3.8.6 Factor cf safety against overturning caplete 8/20/84 of intake structure (Rev. 1) 78 3.8.6 Dead load alculations Conglete 6/1/84 79 3.8.6 Post-modification seisnic loads for Complete 8/20/84 the torus (Rev. 1)
M PG4 80/12 6 - gs
JG"DCFJENT 1 (Cont'd)
DSER R. L. MITIL TO CPBI SBCTIGE A. SODENCER ITEN IUSER SWL7ECT STMUS IJITNR DEED 80 3.8.6 Torus fluid-structure interactions Complete 6/1/84 81 3.8.6 Seismic displacement d torus Caplete 8/20/84 (Rev. 1) 82 3.8.6 Review d seismic Category I tark Ccaplete 8/20/84 design (Rev. 1) 83 3.8.6 Factors d safety for &ywell C<mplete 6/1/84 buckling evaluation 84 3.8.6 Ultimate capacity d containment Ccaplete 8/20/84 (materials) (Rev. 1) 85 3.8.6 toad canbination consistency Complete 6/1/84 86 3.9.1 Ca puter code validation Complete 8/20/84 87 3.9.1 Information on transients Ccmplete 8/20/84 88 3.9.1 Stress analysis and elastic plastic Ccmplete 6/29 / 84 analysis 89 3.9.2.1 Vibration levels for NSSS piping Ccaplete 6/29/84 systems-90 3.9.2.1 Vibration nonitoring program during Ccaplete 7/18/84 testing 91 3.9.2.2 Piping supports and anchors Ccaplete 6/29/84 92 3.9.2.2 Triple flued-head containment Ccaplete 6/15/84 penetrations 93 3.9.3.1 Iced combinations and allowable Ccaplete 6/29/84 stress limits 94 3.9.3.2 Desi@ of SRVs and SRV discharge Ccmplete 6/29/84 Pi ping ,
M P84 80/12 7 - gs
e 1 .
ATm at err 1 (Cont'd)
DSER R. L. MITIL 'IO SECTIGf A. SODENCER Wel SDmE IATIER [RTED DEBSER SUBJECT JTDI 95 3.9.3.2 Fatigue evaluation cm SRV piping Complete 6/15/84 and IDCA downcomers 96 3.9.3.3 IE Information Notice 83-80 Complets 8/20/84 (Rev. 1) 3.9.3.3 Buckling criteria used for ocuponent Ccaplete 6/29/84 97 supports 3.9.3.3 Design cf bolts Ccuplete 6/15/84 98 99a 3.9.5 Stress categories and limits for. Complete 6/15/84 core support structures 3.9.5 Stress categories and limits for Canplete 6/15/84 99b core support structures 3.9.6 10CFR50.55a paragraph (g) Ccaplete 6/29/84 100s 100b 3.9.6 10CFR50.55a paragraph (g) Ccmplete 8/20/84 101 3.9.6 PSI and ISI programs for pungs and Ccuplete 8/20/84 valves 3.9.6 taak testing of pressure isolation Ccmplete 6/29/84 102 valves 3.10 Seismic and dynamic qualification of Ccaplete 8/20/84 103a1 mechanical and electrical equipment 3.10 Seismic and dynamic qualification cf Ccmplete 8/20/84 103a2 mechanical and electrical equipment 3.10 Seismic and dynamic qualification of Complets 8/20/84 103a3
' sechanical and electrical equipment 3.10 Seismic and dynamic qualification cf Ccmplete 8/20/84 103a4 nwchanical and electrical equipment I
M PG4 80/12 8 - gs
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JtFUGBeff 1 (Cont'd)
DSER R. L. MFITL E WWI SECTim A. armuMNR ITBt IU SER S R7BCT SDmE IATIER DmD 103a5 3.10 Seismic and dynmaic qualification of Quplete 8/20/84 i mechanical and electrical equipment i 103a6 3.10 Seimic and dynande qualification of Omplets 8/20/84 mechanical and electrical equipment 103a7 3.10 Seismic and dynmaic qualification of Ocuplete 4/20/84 mechanical and electrical equipent 103bl 3.10 Seimic and dynamic qualification of Caplete 8/20/84 mechanical and electrical equipment 103b2 3.10 Seismic and dynamic qualification of Ocuplete 8/20/84 mechanical and electrical equipment 103b3 3.10 Seismic and dynande qualificatim of Complets 8/20/84 mechanical and electrical equipent 103b4 3.10 Seimaic and dynamic qualification of Ccaplete 8/20/84 mechanical and electrical equipment 103b5 3.10 Seimic and dynamic qualificatim of. Complete 8/20/84 :
mechanical arri electrical equipment 103b6 3.10 Seimaic and dynmaic qualification of Complete 8/20/84 mechariical and electrical equipment i
103c1 3.10 Seismic and dpuumic qualification of Caplete 8/20/84 mechanical and electrical equipment 103c2 3.10 Sei mic and dynmaic qualification of Complete 8/20/84 '
mechanical'and electrical equipment 103c3 3.10 Seimmic and dynmaic qualification of Ocuplete 8/20/84 mechanical and electrical equipment j 103c4 3.10 Seimaic and dynamic qualificatim of Omplete 8/20/84
[
' mechanical' and electrical equipment '
! 104 3.11 Dwircrunental qualification of teC Action unchanical and electrical equipment e
n see 8c/12 9 - gs
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ATDOBENT 1 (Cont'd)
DSER R. L. MITIL 10 SECTIGi A. SODENCER CFBI SUN BCT STh2tJB IATIER Dh2E0 r!EN NLDSER 4.2 Plant-specific andanical fracturing Complete " 8/20/84 105 analysis (Rev. 1) 106 4.2 Applicability d solmic andd ICCA Complets 8/20/84 loading evaluation (Rev. 1) 107 4.2 Minimal post-irradiation fuel Ccaplete 6/29/84 surveillanos program 108 4.2 Gadolina thennal conductivity Ccaplets 6/29/84 equation 4.4.7 1MI-2 Item II F.2 Complete 8/20/ 84 109a 109b 4.4.7 1MI-2 Item II.F.2 Complete 8/20/84 110a 4.6 Functional design d reactivity Ccaplete 8/30/84 control systans (Rev. 1) 110b 4.6 Functional design cf reactivity Ccmplete 8/30/84 control systems (Rev. 1) 111a 5.2.4.3 Preservice inspection gregram Ccaplete 6/29/84 (ccuponents within reactor pressure boundary) 111b 5.2.4.3 Preservice inspection progran Conglets 6/29/84 (cruponents within reactor gressure boundary) 111c 5.2.4.3 Preservice inspection progran Ccaplete 6/29/84 (ocuponents within reactor gressure boundary) 5.2.5 Reactor coolant pressure boundary Ccaplete R/30/84 l 112a (Rev. 1) leakage detection 5.2.5 Reactor coolant pressure boundary Complete 8/30/84 112b (Rev. 1) leakage detection l
M PG4 80/1210 - gs L
ATIACINENT 1 (Cont'd)
DSER R. L. MITIL TO OPEN SECTICN A. SCIMENCER ITYM NL.MBER SUELTECT STA111S IfrIER IATED 112c 5.2.5 Reactor coolant pressure boundary Ccmplete 8/30/84 leakage detection (Rev. 1) 112d 5.2.5 Reactor coolant pressure boundary Canplete 8/30/84 leakage detection (Rev. 1) 112e 5.2.5 Reactor coolant pressure boundary Otmplete 8/30/84 leakage detection (Rev.1) 113 5.3.4 GE procedure applicability Otmplete 7/18/84 114 5.3.4 Canpliance with NB 2360 of the Sumer Caplete 7/18/84 1972 Addenda to the 1971 ASME Code 115 5.3.4 Drop weight and Charpy v-notch tests Caplete 9/5/84 for closure flange materials (Rev. 1) 116 5.3.4 Charpy v-notch test data fcr base Ocuplete 7/18/B4 naterials as used in sholl courso No. I 117 5.3.4 Otmpliance with NB 2332 of Winter 1972 Caplete 8/20/84 Addenda of the ASME Code 118 5.3.4 Imad factors and neutron fluence for Ccmplete 8/20/84 surveillance capsules 119 6.2 1MI itan II.E.4.1 Ocuplete 6/29/84 120a 6.2 1MI Item II.E.4.2 Caplete 8/20/84 120b 6.2 TMI Item II.E.4.2 Caplete 8/20/84 121 6.2.1.3.3 Use of NURDG-0588 Omplete 7/27/84 122 6.2.1.3.3 Temperature profile Cmplete 7/27/84 123 6.2.1.4 Butterfly valve operation (post Ccnplete 6/29/84 accident)
M P64 80/12 11 - gs
ArtmOperr 1 (Cont'd)
DSER R. L. MIT!L X)
WWI SECTIGI A. SOBEN(El FIBI IERGER E R7BCT SIRIUS I2rtWt DRIED 124a 6.2.1.5.1 IWV shield annulus analysis Omaplete 0/20/84 (nov. 1) 12e 6.2.1.5.1 W V shield annulus analysis complete 8/20/94 (Rev.1) 124c 6.2.1.5.1 prV shield annulus analysis Ocuplete 8/20/84 (Rev. 1) 125 6.2.1.5.2 Design drywell head differential Complets 6/15/84 pressure 126a 6.2.1.6 modundant position indicators for ocuplete 8/20/84 vacuum breakers (and control rean alarms) 126b 6.2.1.6 modundant position indicators for ocuplete 8/20/84 vacusan breakers (and control room alarms) 127 6.2.1.6 Operability testing of vacutan breakers Casplete 8/20/84 (Dev.1)
L 128 6.2.2 Air ingestion Complete 7/27/84 129 6.2.2 Insulation ingestion C g lete 6/1/04 130 6.2.3 Potential bypass leakage paths Ccaplete 6/29/84 l
131 6.2.3 Administraticm of secondary contain- Ocuplete 7/18/84 l
ment openings l
l, 132 4.2.4 containment isolation review Cesplete 4/15/84 l^ Czplete 8/20/84
!,- 133a 6.2.4.1 Contalruent purge system 133b 6.2.4.1 Containment purge system Cceplete 8/20/v4 133c 6.2.4.1 Containment purge system Complete 8/20/84 i
i n se4 Sc/12 12- os i
I L
M , si.
s-L i
e ATTA00Wfr 1 (Cont'd)
DBER R. L. MIT1L 10 ;
, WWI SBCr1GI A. Sone lNCER '
ITBI ' " IRBIER SUR7BCr STA11Ni LETTER DAT8D
- 1 34' ,
6.2.6 .Contairument leakage testing . Ccaplete 6/15/84 i 136 6.3.3 170B and IPCI injection valve C(riplete 8/20/84 i g interlocks :
I
.136 6.3.5- Plant-specific IDCA (see Section Caplete 8/20/84
. 15.9.13) (R w . 1) )
147a . 6.4 Control ross habitability complete 8/20/84 [
- 137b '
6.4 Control roan habitability- canplete 8/20/84 t
137c -6.4 control roan habitability Complete 8/20/84 l 138 - 6.6 Preservice inspection program for Complete 6/29/84 Class 2 and 3 caponents 139 ~ 6.7 MBIV W4 age control system Complete 6/29/84 ;
s 140s 9.1.2 spent finel pol' storage cmplets 9/7/84 l (Rw. 2) 140b 9.1.2 Spent fuel pool storage Complete 9/7/84.
(Rev. 2) 4 140c 9.1.2 8 pent fuel pool storage Caplete 9/7/84 ,
(Rw. 2) 1400- 9.1.2 8 pent fuel pool storage Complete 9/7/84 i m -(Rev. 2) 141a 9.1.3 Spent fuel cooling and cleanup Caplete 8/30/84 system (Rw. 1) 141b. 9.1.3 spent fuel cooling ard cleanup. Canplete 8/30/84 system- (Rev. 1) i 3 , 141c 9.1.3 spent fuel pool cooling ard cleanup Coplete e/30/84 systen (Rw. 1) ,
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M PG4 80/12 13 - gs_
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ATTAGMENT 1 (Cont'd)
DSER R. L. MITIL TO
,: OPIN , SECTION A. SOMNCER ITEM NUMEER SUBTECT STATUS LETTER DATED 141d' 9.1.3 Spent fuel pool cooling ard cleamp Cmplete 8/30/84 system (Rev. 1) 141e 9.1.3 Spent fue) pool cooling ard cleamp Cmplete 8/30/84 system (Rev. 1) 141f 9.1.3 Spent fuel pool cooling and cleamp Cmplete 8/30/84 system (Rev. 1)
- 141g. 9.1.3 Spent' fuel pool cooling and cleamp Cmplete 8/30/84 system ~(Rev. 1) .
- '142a 9.1.4 Light load handling system (related . Cmplete 8/15/84 to refueling) (Rev. 1)
.14'2b 9'. l .4 Light load handling system (related Ccaplete 8/15/84 4 1 to refueling) (Rev. 1) 143a 9.1.5 Overhead heavy load handling Cmplete 9/7/84-
- 143b' 9.1.5 . Overhead heavy load handling . Open Ll44a .9.2.1 Station service water system Cmplete .8/15/84 (Rev.~1)
~
144b 9.2.1 Station service water system - Cmplete 8/15/84.
~ .
(Rev. 1) 144c. .9.2.1 Statilon seNice water system Cmplete 8/15/84 c ., (Rev. 1) 145 9.2.2 ISI program and functional testing . Closed 6/15/84 of safety and turbine mxiliaries (5/30/84-
- l. coolirrf. systems Aux.Sys.Mtg.)-
l146: '9.2.6 Switches and wiring associated with - Closed 6/15/84 HPCI/RCIC torus suction (5/30/84-
,e # Aux.Sys.Mtg.)
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ATMOttENT 1 (Cont'd)
I DSER R. L. MITIL 10 l GWi SECTIGi A. SO M N3 R '
ITDE NLDBER _
SUBJECT SD2US IETNR INND {
' 147a 9.3.1 - Ccagressed air systens Caplete 8/3/84 (Rev 1) 147b 9.3.1 Ccupressed air systems ccuplete 8/3/84 (Rev 1) 147c 9.3.1 Ccupressed air systens Ccaplete 8/3/84 (Rev 1) 147d 9.3.1 Ccagressed air systens Ccuplete 8/3/84 (Rev 1) 148 9.3.2 Post-accident sampling system Ccaplets 8/20/84 (II.s.3) 149a 9.3.3 Equipment and floor drainage system. Ccaplete 7/27/84 149b 9.3.3 Equipment and floor drainage systen Ccaplete 7/27/84
-150 9.3.6 Primary cxmtairinent instnment gas Ccmplete 8/3/84 system (Rev. 1) 151a 9.4.1 Control structure ventilation system Ccaplete 8/30/84 (Rev. 1) 151b 9.4.1 Control' structure ventilation systen Ccmplete 8/30/84
' (Rev. 1) 152 9.4.4 Radioactivity nonitoring elements Closed 6/1/84
' (5/30/84-Aux.sys. Meg.)
153 9.4.5 Engineered safety features ventila- Ccaplete 8/30/84 tion system (Rev 2) 154 9.5.1.4.a Metal rotf deck ccmstruction Ccuplete 6/1/84 t
classificiation 1.
155 9.5.1.4.b ongoing review cf safe shutdown NRC Action capability .
i
! 156 9.5.1.4.c Ongoing review cf alternate slutdown NRC Action L capability M P84 80/12 15 - gs b-- -_...- ._.=.- - _ -.--- _ _ -_ -
v_, - .. ,
.mi, ATTACHMENT 1 (Cont'd)
E6ER R. L. MITIL TO OPEN SECTION A. SGWENCER ITEM- NUMEER SUBTECT- STATUS LETTER DATED 157 9.5.1.4.e ' Cable tray protection C m plete 8/20/84 158.- 9.5.1.5.a Class B fire detection systen Cmplete 6/15/84 159 9.5.1.5.a Primary and _ secondary power supplies Canplete 6/1/84 for fire detection system 160 9.5.1.5.b Fire water punp capacity Cmplete 8/13/84 161 9.5.1.5.b Fire water valve supervision Canplete 6/1/84 162 9.5.1.5.c Deluge vanes Cmplete 6/1/84 163 9.5.1.5.c Mamal tose station pipe sizing Cmplete 6/1/84
- 164 .9.5.1.6.e Renote shJtdown panel ventilation Canplete 6/1/84 165 9.5.1.6.g Emergency diesel generator day tank Canplete 6/1/84 protection 166 12.3.4.2- Airborne radioactivity nonitor Complete 9/7/84
~
positioning (Rev. 1) 167; 12.3.4.2 Portable contimous air nonitors Canplete 7/18/84
~ 168 12.5.2 Equipment, training, and procedures Canplete 6/29 /84 for inplant iodine instrumentation 169 12.5.3_ Guidance of Division B Regulatory Cmplete 7/18/84 Guides 170 13.5.2 Procedures _ generation package Canplete 6/ 29 /84 subnittal 171 13.5.2 ' TMI Item I.C.1 Cmplete 6'29/84 f
172 13.5.2 PGP Comnitment Caplete 6/29/84 173 13.5.2 Procedures covering abnormal releases Cmplete 6/29/84 of radioactivity M P84 80/1216 - gs
l ATDCIMENT 1 (Cont'd)
DSER R. L. MITTL 'IO OPEN SECTICN A. SOMENCER ITEM NLMBER SUBJlrf STARIS IETIER IWIED 174 13.5.2 Resolution explanation in FSAR of Canplete 6/15/84
'IMI Items I.C.7 and I.C.8 175 13.6 Physical security Open 176a 14.2 Initial plant test progra Ctmplete 8/13/84 1766 14.2 Initial plant test progra Ctaplete 9/5/84 (Rev. 1) 176c 14.2 Initial plant test program Ccuplete 7/27/84 176d 14.2 Initial plant test program Otmplete 8/24/84 (Rev. 2) 176e 14.2 Initial plant test progra Conplete 7/27/84 176f 14.2 Initial plant test progra ocuplete 8/13/84 176g 14.2 Initial plant test program Ccanplete 8/20/84 176h 14.2 Initial plant test progre Ctmplete 8/13/84 176i 14.2 Initial plant test prograt Ctaplete 7/27/84 177 15.1.1 Partial feedwater heating Ccmplete 8/20/84
., (Rev. 1) 178 15.6.5 IOCA resulting frcm spectrum of NRC Action postulated piping breaks within RCP I~
l 179 15.7.4 Radiological consequences of fuel NRC Action
(
handling accidents l 180 15.7.5 Spent fuel cask drop accidents NRC Action 181 15.9.5 'IMI-2 Iten II.K.3.3 Ocuplete 6/29/84 182 15.9.10 'IMI-2 Item II.K.3.18 Couplete 6/1/84 li 183 18 Hope Creek DGDR Ccmplete 8/15/84 l
M P84 80/12 17 - gs u
ATDOBENT 1 (Cont'd)
DSER R. L. MF11L 10 CHN SECTIm A. SODEN3R NIBGER StBJECT STMUS ETIER DPGED FITM 184 7.2.2.1.e Failures in reactor vessel level Cauplete 8/1/84 sensing lines (Rev 1) 185 7.2.2.2 Trip system sensors and cabling in Ca plete 6/1/84 turbine building 186 7.2.2.3 Testability d plant protection Caplete 8/13/84 systems at power (Rev. 1) 187 7.2.2.4 Lifting d leads to perform surveil- C aplete 8/3/84 lance testing 188 7.2.2.5 Setpoint nethodology Ccaplete 8/1/ 84 189 7.2.2.6 Isolation devices Cmplete 8/1/84 190 7.2.2.7 Regulatory Guide 1.75 Cmplete 6/1/84 191 7.2.2.8 Scram discharge volume Cmplete 6/29/84 192 7.2.2.9 Reactor node switch Cmplete 8/15/84 (Rev. 1)
L.
193 7.3.2.1.10 Manual initiation d safety systems Caplete 8/1/84
- l. .
194 7.3.2.2 Standard review plan deviations Ccaplete 8/1/84 (Rev 1) 195a 7.3.2.3 Freeze pwi.ection/ water filled Ccaplete 8/1/84 l
instrument and sampling lines and cabinet temperature cxmtrol i
L 195b 7.3.2.3 Freezer protection / water filled Caplete 8/1/84 instrument and sappling lines and cabinet temperature control 196 7.3.2.4 Sharing cf cannon instrument taps Cmplete 8/1/84 197 7.3.2.5 Microprocessor, multiplexer and Ccuplete 8/1/84 computer systems (Rev 1) --
l M P84 80/12 18 - gs .
RITAOBENT 1 (Cent'd)
DSER R. L. MITIL 70 SBCTIGi A. SOMMER WEN NLMER SLBJECT STRTUS IETIER IRIED ,
ITEM 7.3.2.6 - 1MI Ita II.K.3.18-ADS actuation Ccaplete 8/20/84 198 199 7.4.2.1 IE Bulletin 79-27-Ioss d_ nort class Complete 8/24/84 IE instrumentation ard control power (Rev. 1) system tais during cperation 200 7.4.2.2 Renote s'utdown systen Ccaplete 8/15/84 (Rev 1) 201 7.4.2.3 RCICAIPCI interactions Ccaplete 8/3/84 202- 7.5.2.1 Imvel measurement errors as a result Ccaplete 8/3/84 of envirornental temperature offects en level instrumentaticm reference leg 203 7.5.2.2 Regulatory Guide 1.97 Ccaplete 8/3/84 204 7.5.2.3 1MI Iten II.F.1 - Accident nonitorirg Canplete 8/1/84 205' 7.5.2.4 Plant process canputer systen Canplete 6/1/84 206 7.6.2.1 High pressure / low pressure interlocks Canplete 7/27/84 207 7.7.2.1 HELBs and cmsequential control systen Cauplete 8/24/84 failures (Rev. 1) l ,
208 7.7.2.2 m itiple control systen failures Ccs plate 8/24/84 (Rev. 1) 209 7.7.2.3 Credit for non-safety related systens Canplete 8/1/84 in Chapter 15 d the FSAR (Rev 1) 210 7.7.2.4 Transient analysis recording system Ccaplete 7/27/84 4
211a 4.5.1 Control rod drive structural mterials Ccaplete 7/27/84 4.5.1 Control red drive structural materials Ccmplete 7/27/84 211b 7/27/84 211c 4.5.1 Control red drive structural notarials Ccaplete f
9 M P84 80/1219 - gs 1
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l l
i I
1 ATDOte!Nr 1 (Cont'd)
DSER R. L. MITIL 10 GWI SBCTIQi A. SO M HCER T111M IUSER SUE 7BCT STATUS IJIT115t DA21!D 211d 4.5.1 Control rod drive structural materials Ccuplete 7/27/84 211e 4.5.1 Control rod drive structural materials Ccaplete 7/27/84 212 4.5.2 Reactor internals materials Otmplete 7/27/84 213 5.2.3 Reactor coolant pressure boundary Caplete 7/27/84 material 214 6.1.1 Engineered safety features materials Canglete 7/27/84 215 10.3.6 Main steam and feedwater systen Ocuplete 7/27/84 materials 216a- 5.3.1 Reactor vessel materials Ocuplete 7/27/84 216b 5.3.1 Reactor vessel materials Ocuplete 7/27/84 P
217 9.5.1.1 Fire protection organization Ccmplete 8/15/84 218 9.5.1.1 Fire hazards analysis Ccaplete 6/1/84 219- 9.5.1.2 Fire protection a&ninistrative Ccaplete 8/15/84 controls 220 9.5.1.3 Fire brigade and fire brigade Couplete 8/15/84 training
- 221 8.2.2.1 Physical separation of offsite Ccuplete 8/1/84 transmission lines l
222 8.2.2.2 Design provisions for re-establishr- Ocuplete 8/1/84
!j j
ment of an offsite power source l! -'
223 8.2.2.3 Independence of offsite circuits between the switchyard and class IE Couplete 8/1/84 buses Ocuplete 8/1/84 224 8.2.2.4 re-nnn failure mode between malte and offsite power circuits i'
1 .s
' y M.P84 80/11 20- gs 1
[; 8 b- i i
~
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1 A2'D OBert 1 (Cont'd)
DSER R. L. MF11L 10 ,
N 8ECfM . .... .
A. SQROGR S'A1BCr Styms IRITER ORTED fl198 NLDEER 225 8.2.3.1 Testability d stamatic transfer of Caplete 8/1/84 power from the nonnal to preferred power source 2 26 8.2.2.5 Grid stability Complete 8/13/84 (Rev.1) 227 8.2.2.6 Capacity ard capability of offsite Ccaplets 8/1/84 circuits 228 8.3.1.l(1) Voltage &cp & ring transient condi- Complete 8/1/84 tions 8.3.1.l(2) Basis "fcr using bus voltage versus Ccaplete 8/1/84 229 actual connected load voltage in the
. voltage drop analysis 8.3.1.l(3) Clarificaticn d Table 8.3-11 Ccuplete 8/1/84 230
~231 8.3.1.1(4) Undervoltage trip setpoints Ccmplete 8/1/84 8.3.1.l(5) Icad configuration used for the Caplete 8/1/84 232 voltage &cp analysis 8.3.3.4.1 Periodic systen testing Ccaplete 8/1/84
! '233 8.3.1.3 C@acity and capability cf onsite Canplete 8/1/84 234 AC- power supplies and use af ad-ministrative controls to prevent overloading cf the diesel generators
-8.3.1.5 Diesel generators load acceptance Ccaplete 8/1/84 235 i ' test Ccapliance with position C.6 of Ccaplete 8/1/84 2 36 8.3.1.6 IG 1.9 Decription cf the load sequencer Ccuplete 8/1/84 237 8.3.1.7 Sequencing of loads on the offsite Ccaplete 8/1/84 238 8.2.2.7 power systen l
l L
M P84 80/12 21 - gs l
l l
KPDOBENT 1 (Ccnt'd)
DSER R. L. MITIL '10
@EN SBCTICM A. SOMNCER SUBJECT STATUS IEITER DPSED ITDI NLBEER 2 39 8.3.1.8 Testing to verify 80% mininnsa Caplete 8/15/84 voltage
-240 8.3.1.9 CWpliance with BIP-PSB-2 Complete 8/1/84 241 8.3.1.10 Ioad acceptance test after prolonged Ca plete 8/20/84 l no load cperation cf the diesel (Rev. 1) generator 242 8.3.2.1 Ccapliance with position 1 cf Regula- Ccaplete 8/1/84 tory Guide 1.128
'243 8.3.3.1.3 Protection or qualification cf Class Ccuplete 8/1/84 lE equipnent fran the effects cf fire mappression systems 244 8.3.3.3.1 Analysis and test to demonstrate Cmplete- 8/30/84 adag_=_ q cf less than specified (Rev. 1) separation 245 8.3.3.3.2 The use cf 18 versus 36 inches cf C mplete 8/15/84 separaticn between raceways (Rcv. 1) 246 8.3.3.3.3 Specifi,ed separation cf raceways by Ccuplete 8/1/84 analysis and test 2 47 8.3.3.5.1 Capability cf penetrations to with- C mplete 8/1/84 stand long duration short circuits at less than maximum or worst case short circuit 248 8.3.3.5.2 Separation cf penetration primary Ccmplete 8/1/84 l
l- and backup potections I
249 8.3.3.5.3 The use cf bypassed thermal overload Caplete 8/1/84 protective devices for penetration protections 1
250 8.3.3.5.4 Testing cf fuses in accordance with Complete 8/1/84 l
R.G. 1.63 i
i M P84 80/12 22 - gs 5
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MTAQsert 1 (cont'd)
DSER R. L. MITTL 1D GWI SBCTIGE A. SOBBICER T113t NtBIBER SUBJECT SrMUS IEFIER DKEED 251 8.3.3.5.5 Fault current analysis for all Ocuglete 8/1/84 representative penetration circuits
! 252 8.3.3.5.6 The use of a single breaker to provide Otmplets 8/1/84 penetration p ut.ction
! 253 8.3.3.1.4 Chuaitment to protect all Class 1E Otmplete 8/1/84 i equipment fram external hazards versus only class lE equipment in one division
(
254 8.3.3.1.5 Protection of class lE power supplies C g 3ar.e 8/1/84 i
from failure of unqualified class lE l loads
_255 8'.3.2.2 Battery capacity Otmplete 8/1/84 256 8.3.2.3 Automatic trip of inada to remintain Ocuplete 8/20/84 sufficient battery capacity l
f 257 8.3.2.5 Justification'for a 0 to 13 second < Conglete 8/1/84 i load cych 258 8.3.2.6 Design and qualification of DC Ccuplete 8/1/84 systen loads to operate between mininnan and maximum voltage levels I -259 8.3.3.3.4 Use of an inverter as an isolation Ccaplete 8/1/84 device
! 260 8.3.3.3.5 Use of a single breaker tripped by Ccaplets 8/1/84 l -a IOCA signal used as an isolation
~ device 261 8.3.3.3.6 Automatic transfer of loads and Otmplets 8/1/84 interconnection between redundant divisions 262 -ll.4.2.d Solid waste control program ocuplete 8/20/84 l
M P84 80/12 23- gs
F Argosert 1 (cont'd)
DSER R. L. MITIL TO '
A. SQiNENCER WEN SBCrIGI . SDEUS LETTER DKEED ITBI IUSER E R7BCT 11.4.2.e Fire protection for solid raduaste Ocuplete 8/13/84 263 storage area 264 6.2.5 Sources of oxygen 0@ 8/20/84 Otsplete 6.8.1.4 ESF Filter Testing 8/13/84 265 6.8.1.4 Field leak tests Ocuplets 8/13/84 266 6.4.1 Control roaa toxic chemical Ocuplets 8/13/84 267' i
detectors Air filtration unit drains Ocuplete 8/20/84 268 5.2.2- Cbde cases N-242 and N-242-1 Ocuplete 8/20/84 269 5.2.2 Cbde <mse N-252 Complete 8/20/84 270 TS-1 2.4.14 Closure of watertight (bors to safety- Open related structures 4.4.4 Single. recirculation loop operation Open TS-2 4.4.5 Core flow monitoring for crud effects Ocuplete 6/1/84 TS-3 TS-4 4.4.6 r- parts monitoring system Open Natural circulation in normal Open l TS-5 4.4.9
- operation Secondary contaimnent negative Open TSr-6 6.2.3 Pressure 6.2.3 Inleakage and draudoun time in Open TS-7 i secondary containment TS-8 6.2.4.1 Isakage integrity testing Open BCCS subsystem periodic ocuponent Open TS-9 6.3.4.2
- testing M P84 80/12 24- gs ww- e.-- , - ~ , n- - - ,., , , m-we,,.,,,.s,.,,,,,,,,-,-,-,ew .-,-_...,---ma_--.- _ ,,-ve,awan_- _n,,,,,-r .-
ATDOBENT 1 (@nt'd)
D6ER R. L. MITIL 10 A. SOBENCER GWI SBCTIQi STATUS LETTER DfG1B IU SER SUETBCr JTEN TS-10 6.7 MSIV leakage rate TS-Il - l$.2.2 Availability, setpoints, and testing Open of turbine bypass system TS-12 15.6.4 Primary coolant activity IC-1 4.2 Fuel rod internal pressure criteria Complets 6/1/84 II-2 4.4.4 Stability analysis submitted before Open second-cycle operation 9
O 6
M P84 80/12 25- gs enm-- , . - , -,+wm,.g-,_,-- ,-,.---sn,-,.,+m-,,mn,, ,,, ,-- -
,,e m_, . .w--,,,---,.,,,-. ,-,,. , - ,,, - - , - ,,- ._.,w-,-m---o,, - , - - ,.,,,oa- ,wn,w, ., ~
-ATTACHMENT.2 DATE: 9/7/84-
. DRAFT SER SECTIONS AND DATES PROVIDED SECTION DATE SECTION DATE 3 .1 -
3.2.1 11.4.1 See Notes 1&5 3.2.2 11.4.2 See-Notes 1&5 5 .1. 11.5.1 See Notes 1&5 5.2.1 11.5.2 See Notes 1&5 6.5.1 See Notes 1&5 13.1.1 See Note 4 8.1 See Note 2 13.1.2 See Note 4 8.2.1 See Note 2 13.2.1 See Note 4 8.2.2 See Note 2 13.2.2 See Note 4 8.2.'3 See. Note 2 13.3.1 See Note 4 8.2.4 See Note 2 13.3.2 See Note 4 8.3.1 See Note 2 13.3.3 See Note 4 8.3.2 See Note 2 13.3.4 See Note 4
~8.4.1 See Note 2 13.4 See Note 4.
8.4.2- See Note 2 13.5.1 See Note 4 8! . 4 . 3 See Note 2 15.2.3 8.4.5 See Note 2 15.2.4 8.4.6 See Note 2 15.2.5 8.4.7 See Note 2 15.2.6 8.4.8 See Note 2 15.2.7 9.5.2 See Note 3 15.2.8 9.5.3 See Note 3 15.7.3 See Notes 1&S 9.5.7 See Note 3 17.1 8/3/84 9.5.8 See Note 3 17.2 8/3/84 10.1 See Note 3 17.3 8/3/84 10.2 See Note 3 17.4 8/3/84 10.2.3 See Note 3 10.3.2 See Note 3 10.4.1 See Note 3 10.4.2 See Notes 3&5 10.4.3 See Notes 3&5 10.4.4 See Note 3 11.1.1 See Notes 1&5 Notes:
11.1.2 See Notes 1&5 11.2.1 See Notes 1&5 1. Open it' ems provided in 11.2.2 See Notes 145 letter dated July 24, 1984 11.3.1 See Notes 1&5 (Schwencer to Mittl) 11.3.2 See Notes 1&5
- 2. Open items provided in June 6, 1984 meeting
- 3. Open items provided in April 17-18, 1984 meeting CT:db
- 4. Open items provided in May 2, 1984 meting 5.-Draft SER Section provided in letter dated August 7, 1984 (Schwencer to Mittl)
MP.84-95/03 01
r ,
v DATE: 9/7/84
= ATTACHMENT 3
.:Open. DSER::
o ' Item Section : Subject
~
-5a :2.4.5 Wave . impact _ and runup - on service water intake structure
-5b - 2.4.5 Wave; impact and runup on service water intake
~ structure ~
5d. 2.4.5 Wave . impact and runup on service water intake structure
'140a 9.1.2 Spent. fuel pool storage
-140b Spent fuel pool storage 9.1.2 140c 9 .1. 2. Spent fuel pool . storage 140d 9.1.2 Spent fuel pool storage
^
143a- 1, ~ ' 9 .1. 5 ' overhead heavy load handling.
- - 166 12.3.4.2 Airborne radioactivity monitor' positioning LQuestion 421.10-k
.M P84 95/03 02 re. - .
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SEP.-S'840270006 HCGS iDSER Open Item No. 5 ( DSER Section 2.4.5) s WAVE- IMPACT ' AND RUNUP ON' SERVICE WATER INTAKE STRUCTURE
- The ' applicant has analyzed the wind waves - that. would traverse plant grade ' coincident with the PMH surge hydrograph and runup
- on safety-related facilities. These calculations were based on the assumption that' wind waves would be. generated in the Delaware Estuary and progress to the . site. As the surge level would begin to rise, resulting .from the approaching eye of the postulated hurricane, the wind speed would . progressively change direction from the southeast clockwise to the west. Wave s Lencroaching on the southern end of the Island would be depth-
-limited ( i.e. , the waves would " feel" bottom and thus become shallow water waves) by plant grade elevation' on both .the Salem and '. Hope Creek sites. These depth-limited ( shallow wa ter) waves will impact and runup on the southern and western faces of the safety-related structures in the po'wer block. The
. applicant has stated that the southern face of the Reactor
- Building f and- the Auxiliary Building are designed for a flood protection level of 38.0 f t ms1 or 3.2 f t above the maximum calculated wave runup height of 34.8 f t msl and the other exposures of safety-related structures have a flood protection level of 32.0 f t mal or 1 f t above the maximum calculated wave runup height of 31.0 f t ' m sl .
The staf f has requested the applicant to provide additional information on the waves that impact on the river face of -
service water . intake s truc tur e. .The waves impacting ,on this face of the structure are not -reduced in height (d ept h-limited )
qas those that traverse plant grade.
>' . As indicated in Section 2.4.1, the applicant states that all D- accesses to safety-related structures (doors - and hatches) are provided with water-tight seals designed to withstand the head I of water associated with the flood protection levels. But, the L .
applicant has not indicated whether the water-tight doors are l designed to withstand. either the conbined loading effects of
. both static water level and the dynamic wave impact or , as ci ted in Sections 3.4.1 and 3.5.1.4 of this report, the impact
- of a barge propelled by winds and waves associated with a L hydrologic event that floods plant grade.
Based upon its ' analysis according to SRP 2.4.5, the staf f concludes that the flood protection level of El. 38.0 f t mal for the southern face of the Reactor Buidling and . Auxiliary
, ~ Building and El. 32.0 f t msl for the remaining safety-related
'L .
structures within the power block meets the requirements of Regula tory Guide 1.59. Until addi tional information and analysis L
K51/2-15 5-1 E
Il
SEP -5 '84 0 2 7 0 0 0 0 DSER Open Item No. 5 (Cont'd) are available, the - staf f cannot conclude that the flood pro-tection ' level of El. 32.0 f t ms1 for the Service Water Intake Structure meets the requirements of Regulatory Guide 1.59.
Based on its analysis, the staf f cannot conclude that the plant
-meets the requirements of GDC 2.with respect to the hydrologic aspects of Probable Maximum surges and Seiche Flooding.
RESPONSE
The requested information for the service water intake structure has been provided in the responses to the following NRC questions:
Information Provided Question No.
Wave runup elevations 240.8 Wave bnpact loads 240.9 Flood protection 240.8 and 410.69 As a result of discussions with the NRC staf f, the response to Question 410.69 has been revised and summary calculations for wave overtopping of the west and south walls have been submitted under separate cover.
Information on the ability of the doors and hatches to withstand the combined loading ef fects of static water level and the dynamic wave impact is provided in the response to FSAR Question 240.14.
K51/2/16- 5-2
SEF -5 aa g if 70 0 2 6 j HCGS FSAR QUESTION 410.'69 (Section 9.2.1)
Provide a figure (s) in . the FSAR which shows the protection of the ' station service water system from the flood water (includ-ing wave effects) of the design basis flood.
RESPONSE
LThe general arrangement of the intake ' structure is provided
.in Fig ures 1. 2-4 0 a nd 1. 2-41. Section AA of Figure 1. 2-41 is reproduced- here as Figure 410.69-1 which identifies the water-tight areas and the walls and slabs designed to accommodate flood loads. As de scrib ed in Sections 2.4.2 and 2.4.5, the south' and west exterior walls of the intake structure are sub-ject to a maximum wave run-up elevation of 134.4 feet due to the probable maximum hurricane (PMH). Such waves could overtop the roof of the western portion of the structure at elevation 128 feet. However , a rigorous analysis has been per formed to deteomine the depth of water in the low area (elevation 122.0 feet) af ter wave impact and to confirm that water does not enter the building through the air intake control dampers (bottom elevation 128.~5 feet) . There fore, flood water will not*
enter into the dry area of - the intake structure. On the north side of the intake structure, the maximum water level will be only slightly higher than the still water elevation (113.8 feet) during the PMH. According to Table 2.4.6, the maximum Lwave elevation for the north side of the intake structure is
- 26.3 feet MSL ( elevation 115.3 feet ) d ue to a postulated mul-tiple dam break. Therefore, flood protection of the north exterior wall to elevation 121.0 feet is adequate.
On the east side of the intake structure, the maximum wave ,
run-up elevation due to the PMH equals 122.3 fee t. This ele-vation is due to a 14 wave traveling in the direction of Fetch i
"A". Fetch A, which is rotated about 15 degrees from Fetch 1 i- ( as shown in Figures 410. 69-2 and 410. 69-3 ) , is chosen to maxi-mize the wave run-up elevation. Elevation 122.3 fee t exceeds the elevation of the bottom of the RVAC exhaust openings at
- elevation 122.0 feet by 0.3 fee t. Curbs will be added at the L bottom of -these openings to prevent water from entering into L .the building.
In addition the following assessments have been made to confirm
.the adequacy of the structure and interior components 6or the overtopping wave
- a. The exterior walls are designed to withstand -the flood i
loads including the dynamic wave action ef fects.
- b. The roof hatches at both elevations 122.0 and 128.0 feet have been sealed (caulking , gas kets , etc.) to prevent any intrusion of water. The hatch covers are keyed into 410.69-1
~ . - - , . .. - , - . - - , - - . - - - - - - - . . - . . . - - - . . - _ . _ - - . . - - -
SEP - S N O 2 7 C fi26 RESPONSE - cont' d the openings to prevent any adverse slippage due to wave induced loadings.
- c. All Seismic Category I ccuponents except for the travel-ing. water screens are 'located within the dry areas of the structure.
- d. The traveling water-screens, located in the " wet" area between column lines B and C have electric motors which are "ully protected against the flood water level,
- e. A condition was postulated where suspended moisture enters the dry areas of the structure through the air intake control dampers. It has been assessed that all of the Seismic Category. I components subjected to this environment will continue to function as required.
Section 3.4.1 and Table 3.4-1 have been revised for clarifica-tion.
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SERVICE WATER INTAKE
$ STRUCTURE - FLOOD 8 PROTECTION 3 .
Pl0URS 410.001
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- Fig. k ? Fetches for Serv 5.ce Water Intake Structure, Hope Crhek Generating
- Station s
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700,6
- ..' / / August 6, 1984
. d:p3 Crsek Gen 2rsting Statien Analysis cf Overtopping cf Service Water Intake Structure
.I. Wave Calculations- ,
o Wave heights and periods as well as still-water levels and runup elevations are as given in Table 2.4-10a of FSAR (Amendment 5, April 1984).
II. Overtopping Calculations o Overtopping rates were calculated for west face and south face where top of wall elevations are 128.5 and 122.0, respectively.
o Equations from Wiegel (1976) were used for the overtopping.
calculations.
o = (e o? Hs) exP #$ lose { ~ff 9,,,(ft)*(N)%h(@)
N x
s 72
. si. n 2
. El. I28E q)
, 160' o where E was taken as 1/21T in order to maximize the value of Qo*
(see Figure 6 of Wiegel's paper) o c( was taken as 0.06 in order to maximize Q (see Equation 4 of l Wiegel's paper).
i "
o Conservative assumptions in calculating overtopping rates were:
It was assumed that waves attacked normal to the wall of the structure.
It was assumed that the train of waves was made up of all 1%
waves.
It was assumed that wave height was constant along the crest.
o Calculated overtopping rate was increased to allow for wind speed .
using Equation (7-11) of the 1977 edition of the U. S. Army Corps
.of Engineers Shore Protection Manual.
W = l.0 + Wf- + 0.l) Sin 8 1
- . . _ .- _ . _ _ _ _ . . . _ . ~ . _ ~ . _ _ . - . , _ . . _ _ . _ , _ _ _ _ . . _ _ - _
\
' SEF -5 '84 0 2 7 0 6 :' 8 In making the wind adjustment the fccter Wf was catum2d to be 2.0 for onshore winds greater than 60 mph. The angle 0 was 90*.
o After adjustment for wind the overtopping rates were adjusted for angle of attack by multiplying the overtopping rate by the sin of
-the angle between the fetch vector and the wall.
III. Mai== water surface elevations were calculated by backwater calculation starting from the north end of the roof.
i o N separate overtopping rates were added and the total was assumed to flow off the top of the structure at the north end.
o critical depth was assumed to occur at the downstream and of the channel and was calculated as:
mr//GY fc = ^ 32.2 where. Q is the rate of flow from the west side in cfs/f t.
h backwater calculation assumes a gradually varied steady flow.
~
o t
2,bQ
- b X * &n 2 y, * ^ + fx .-
/6*32.2 *$x .
o Calculations were performed moving upstream starting with the depth at the north end.
o h calculations showed that fetch 3 was the critical case. N total flow rate for fetch 3 was 0.5 cfs/f t from the west and 14.7 cfs/ft from the-south end.
o - N maximum water surface elevation reached was 126.9 for the fetch 3 condition which is well below the critical 128.5 elevation at which flow could enter the air intakes.
IV. A separate calculation was made considering a surge generated by flow coming over the south end of the building. N depth of flow and velocity of flow ahead of the surge resulting from the previous surge had to be assumed. Velocity ahead of the surge was assumed to be zero, since that condition maximizes the surge height. Depth ahead of the surge was assumed to be 1.0' and does not have a really significant '
affect on the height of the following surge. N resulting elevation l-i, of the crest of the generated surge was 126.9 which is below the 128.5 l elevation at which water can flow into the air intake.
l V. A check was made to see if flow could surge into the air intakes as a i result of plunging from the roof at elevation 128.5.
l l
l
SEP -5 '84 0 2 7 0 6 0 6
, o Lora coefficiants cf 0.5 at the entranco to the air inecke cyaning and 0.5 at the bend (see attached sketch were assumed). ,
I o Velocity at the edge of the 128.5 elevation roof section was l calculated assuming critical depth there and was increased by 50%
for reasons of conservancy. i l
o N velocity approaching the entrance to the air intake chamber was calculated using the energy equation and neglecting losses.
1~
o Losses incurred by turbulence and impact of the jet enterits water ponded on top of elevation 122.0 were neglected.
o Headloss through the screens was neglected.
o ~ The maximum elevation schieved was calculated to be 126.3 or well below the 128.5 elevation at which water could flow into the building. V o A separate analysis was made using a one-dimensional momentum .
approach. N presence cf the louver on top of the outer wall was neglected. A velocity of 26 feet per second was assumed to occur over the top of the lower outer wall whose top elevation is at 124.0. This velocity was calculated assuming that the total potential energy in a wave runup to 134.4 would be ccaverted to {
kinetic energy at elevation 124 without energy loss. N one-dimensional energy analysis, assumius a flow rate of 5.75 .-
i cfs/ foot indicates that the water surface within the intake could '
, rise to elevation 127.0 which is below the 128.5 elevation at which water could flow into the service-water intake structure. The assumption of a flow rate of 5.75 cfs/ foot is very conservative since that is the total overtopping rate from the west side of the i structure for the critical fetch conditions assuming the wave strikes normal to the structure wall.
o The total pressure of the air intake fans equals
, 4.5 inches of water. The-maximum elevati'ons of 126.3 feet and 127.0 feet given above result in
. margins of 2.2 and 1.5 feet respectively with respect to the 128.5. feet elevation at which water could flow into the building. Therefore, there is sufficient margin to accommodate a rise in water level due to fan suction pressure.
i 3
. SEP -5 '84 0 2 7 0 6 2 6 9 128 5 4 4 4 6 i L gQwes+) 4 I w v t 2
- OsouTM s i YC J/gg); covery
' l i e i e i i i i i ii i , i i , 9 1220 f ..
Flow along center of building
- ~Yi~
Con 4rd 10dit!1 h
o 3.5 r
/
y 12 8 1 } q, sza h v l28' V IN Tr Fix
< AirInkle
.- 6
. , : .- ) 1"
[
/
9 21 12 HMU f/M4 // F4 Sketch of flow conditions at intrance i:o air intakes r .
4
~
. . , < < SEP -5 'e.4 0 2 7 0 6 0 0 References
- 1. Wiegel, R. L., " Wave Overtopping Equation" Proceedings of the 1976 Coastal Engineering Conference.
6
- 2. Jackowski, R. A. (Editor) Shore Protection Manual, U. S. Army Corps of Engineers, Coastal Engineering Research Center, 1977.
P t
i O
e e
5
HCGS .
El 30 '84 0 2 6 8 6 4 0
- D8ER Open item No.'140_(DGER Section 9.1.2) -
SPENT FUEL STORAGE Since the applicant's applidation for an operating license.was docketed in 1983, which is af ter the Novembgr 17,.1977 date specified in the SRP, the. applicant must pr6 vide the results of-an analysis which shows that a failure of the liner plate as a result of an SSE will not cause any of the following (1) significant. releases of radioactivity due to mechanical damage to the fuelt (2) significant loss-of-water f rom the pool which could uncover the fusi and lead to release of radioactivity due to heat ups (3) loss of the ability to cool the fuel due to flow blockage caused by a portion of one or more complete section of the. liner plate falling on the top of the fuel rackst (4) damage to safety-related equipment as a result of the pool leakager and (5) uncontrolled release of significant quantities on radioactive fluids to the environs in accordance
~
- to the Standard Review Plan. These buildings are also designed against flooding and tornado missiles (refer to Section 3.4.1 and'3.5.2 of this SER).. We cannot conclude that the requirements of General Design Criterion 2, " Design Bases f or Protection Against Natural Phenomena," and the guidelines of Regulatory Guides 1.13, " Spent Fuel Storage Facility Design Basis,"
Position C.3, 1.29, " Seismic Design Classification," Positions c.1 and C.2, have been met. ,
The applicant has not provided the design details of the spent fuel storage racks, the results of an analysis of impacts onto the racks, the bundle to bundle spacing, the design maximum i enrichment (weight percent of U235), a description of i calculational methods used for criticality analysis (along with the results), a tabulation of the nominal value of K egg of the
[
t racks along with the various uncertainties and biases considered in the analysis, and a tabulation of the reactivity effect of each of the abnormal accident situations considered for our review. Since credit is taken for gadolinia in the fuel, the applicant must provide a commitment that every fuel bundle will have a specified minimum amount of gadolinia distributed over a specified number of specific fuel pins, for the entire length i
of the fuel. As an alternative, the applicant can provide the results of the criticality analysis without taking credit for the gadolinia.
Thus, we cannot conclude that the requirements of General Design Criteria 61, " Fuel Storage and Handling and Radioactivity
( control," and 62, " Prevention of Criticality in Fuel Storage l and Handling," and the guidelines of Regulatory Guide 1.13, l Positions C.1 and C.4, concerning fuel storage facility design are satisfied.
l.
s 140-1 l
- a. g - w v , w -
~*
, JJL 30 '841) 2 6 8 6 4 0 .
DSER Open Itek No. 140 (Cont?d) ?
Wecannotdokcludethatthespentf0elstoragefacilityisin conformance with'thalrequirements of General Design Criteria 2, 61, and 62 as .they relate to protection of the spent fuel against na'turaliphenomena, . radiation protection,-. and prevention of criticality; and the gui'delines of Regula$ory Guides 1.-13, Positions C.1, C.3, and C.4 and 1.29, Posit fens C.1 and C.2, relating.to the facility's. design basis and seismic x
. classification.The' spent" fuel storage facility does not meet
~
the acceptance.criteriu.of SRP'Section 9.1.2. We will report
, resolution'ofythisitemin'asuppfementtothisSER.
Additionally, therinformation. provided through Amendment 3 was not sufficientEfor~the staff to complete the evaluation of the
~
compatibility land-chemical stability of materials wetted by spent fuel pool water. To complete 'the review, the following information is re, quested:: e' ,
(1) Identify and listiall materials in the spent fuel storage pool' including the neutron poison material, rack leveling
- feet, sand rack frame.
-(2) Provide, test or'operatinhdatashowingthat the neutron poison _m'aterial will not degrade during the lifetime of the spent f u' 'le ' storage . pool.
(3) Provide a' description of any materials monitoring program 3 for the pool.Y,In particular, provide information-on the frequency oft inspection'and type of samples used in the monitoringVprogram.- ..
(4) Providfidn't' alls of ihe spent fuel ~ racks to show.that no
~
buildup df gases will occur in'the cavities containing~the s
poison materials. s
- w RESPONSE -
Thi spent fuel pool liner plate Mas not' designed to seismic Category-I, requirements becabse SRP 9.1.2, Revision 2
.(March 1979), which first 'inv'oked the seismic Category I requirement, was not issued'until af ter the' design and procure-ment'ofsthe liner plate was complete and fabrication had begun (November 1978) . Howeverf the? liner plate was designed to act as a form for the concrete'in the spent fuel pool walls. To-perform this function a' system of channels, wide flanges and angle stiffeners was welded _to the back surfaces of the liner and connected t'osthe outside ' formork w with. form ties. Thus,
. during the' c'oncrete placing o'peration the welds between the stiffenirs and the liner were subject to the lateral pressure effects'of'the weteconcrete. This may be considered a ' test' load in tiiat af ter' the ' concrete sets, the anchoring capability
- g s
"% 0 _ ,- 140-2.
s- ,
'..,- - e-sn, .--,nn. ,,--n. . - . ,, . _ . . . . _ - - - _ . -
i RESPONSE (Cont'd) .
"\ of the stiffener system in holding the liner plate against seismic loads is at least equal to the form pressure load. The estimated test load during construction (approximately 300 lb/ft2) was lower than the design value of 690 lb/ft2 This construction load induced a correspondingly lower stress in the stif fener-to-liner welds.
An analysis, performed to evaluate the effect of SSE loads on the liner, shows that the resultant stresses would be insignifi-cant.(approximately 1% of the stresses due to concrete placement) whan added to the residual concrete load. SSE induced loads imposed on the floor liner by the spent fuel racks would also be insignificant, and will not cause a liner failure.
Based on the cons'iderable design margin for form pressure load and the acceptable performance of the wall liner plate when sub-jected to this ' test' load, it is concluded that the liner plate .
is capable of withstanding SSE loads without any loss of functiqn. [
Thus, the design of the liner plate satisfies General Design i Criteria 2, 61, and 62, Regulatory Guide 1.29, Positions C.1 and C.2, and Regulatory Guide 1.13, Positions C.1 and C.4. Refer to l Section 9.1.2.5 for additional justification of the non-seismic ,
Category I liner design. For additional information on the -
design and analysis of the liner plate, refer to Appendix 3F. .
For a discussion of the liner leakage collection system, which
' permits expedient liner leak detection and measurement, and prevents uncontrolled loss of contaminated pool water, refer to Section 9.1.2.2.2.1.
The spent fuel storage-facility design meets the intent of Regulatory Guide 1.13 Position C.3, as described in Section 9.1.4.6 and 9.1.5.6. pw.If 4,o,3 y The spent fuel storage rack design details have been providedThe in the response to Questions 81.2, 281.13, 410. 3 9 and 410. 4 2.
- l j
' : - n" = e
- i s c.; 2 2 0.10 o..d 41G.30 -ill te-p o nded by Septext:r, 1^C0. This information webL supportsthe se85miC and criticality review 3 and demonstrates that the design satisfies General Design Criteria 61 and 62, and Regulatory Guide 1.13 positions C.1 and C.4.
l The materials used in the spent fuel storage racks were included in the response to Question 281.13.
140-3 l
i
,. , . _ . , _. . _ . _ _ . . . . , _ - . _ . _ . . . . .. _y.- r -
?
^
-RESPONSE'(Cont'd).
Similar. rack designs, with' vented Boral poison in stainless steel
, racks, have been_ licensed and have proven successful. HCGS's maximum anticipated radiation exposure for the Boral'is 1.0 x 10llerads.
LThisiradiation exposure assumes-freshly discharged fuel assemblies c
are stored in .each cell for. a :20. year period and then replaced with ifreshly discharged fuel for a second 20 year period. Brooks and
.Perkins Product Performance Report No.~624 documents Boral's
" capability to withstand exposure of 1.0 x 1011 rads gamma'and 5.3'x:1019 neutrons per sq. cm. in demineralized water without<
' detectable outgassing attributable to Boral,' decrease in neutron
.."^
attenuation,<nor any discernable physical changes. This~ testing was
'; performedLat the Phoenix Memorial Laboratory of the University of
. Michigan'using the. Ford Nuclear Reactor. Ongoing tests have exposed Boral to accumulated radiation doses up to 7 x 1011 rads. These specimens were also found to be structurally sound and neutron attenuation capabilities were~not degraded by irradiation.
,' LI n order to-continually assure'the adequacy of the poison material,
, test 5 coupons are provided for a Boral surveillance program. Forty-five coupons are installed in high radiation areas of the spent '
fuel pool. . However, .because- vented stainless steel spent- fuel racksLwith Boral poison material are already in use in other BWR fuel' pools, such as.Monticello and Browns Ferry, a Boral surveillance
. program is. not : planned at HCGS.
PSE&G will' develop a program.to monitor the Boral surveillance-program' of either Fermi, Monticello 'or Brown's Ferry by March,
- 1985. ~The response to' Question 281.14 has.been revised to reflect .
Lthis response.
- The spent fuel rack poison cavities are vented to. prevent any buildup ofLgases.. Response to Question 281.13.provides further
-inf ormation on venting .
l*
^
No-4
-l -
HCGS_FSAR 6/84 .
3.8.4.8.3 Spent Fuel Rack Design l Acceptance Criterion II.4.f requires that the spent fuel racks be designed in compliance with Appendix D of SkP'a.8.4, which requires that construction materials should conform to Section III,-Subsection NF of the ASME Code. !
= /A/ SERT C#
The spent fuel racks are constructed of ASTM A-240 and ASTM A-564 '
stainless' steel. The A-240 and A-564 material specifications are identical-to the ASME SA-240 and SA-564 material specifications.
All' rack steel is supplied with certified material test reports. !
The rack materials are procured under a Q.A. Program that is intended to comply with:
10CFR50, Appendix B, " Quality Assurance Criteria for a.
Nuclear Power Plants and Fuel Reprocessing Plants".
- b. ANSI /ASME N45.2, " Quality Assurance Program "
Requirements for Nuclear Facilities", and .
I (-
- c. ANSI /ASME NOA-1, " Quality Assurance Program Requirements for Nuclear Power Plants".
3.8.5 FOUNDATIONS Foundations for all Seismic Category I structures and the turbine building and the administration facility, which are non-Seismic Category I structures, are described in this section.
l 3.8.5.1 Description of the Foundations j
l The configuration of the foundation mats for the various structures is shown on Figure 3.8-37.
L i
j Reinforced concrete matJfoundations are provided for all
- structures. 'Except.for the station service water system (SSWS) intake structure, the mats rest either on the Vincentown Formation or'on. engineered structural backfill placed on the Vincentown Formation. ~The mat and the Ican' concrete leveling 3.3 43b Amendment 6 DSER OPDI ITD( /$ /d.
L l _ _ _ . . - - . . _ . _ . . _ - _ . _ . , , . _ _ _ _ _ . _ . . _ _ _ . _ _ _ _ . . _ . . . _ . _ . _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
1 1'
,i
'. . 1 . .
1 -
. . . . _ . . . . . In J"d r1. ._....C. .__._ _ . . . __._... . -
. _ . . . . . . . - - . - - . . . ,. . . - . . - _ h. .. .Sfenf {uel- = ..
_7h e _clesi4n., autysa sud._fa bnes hon...sf .fle,ra.c k.s -
con &rm wi+h Re appkcah/e punivns sl.. Su bsee.h%.
. . . NF. See Appendrx 18. hr a desenybon oI &e de.1i , .
au tysa and cons huc hn o f. . M. e ..ra ck s.. .. . . . .. . -.. . . . . .
DSER OPEN ITEM /h - -
.- 't HCGS FSAR 12/83 CHAPTER 9 TABLES
\-
Table No. Title 9.1-1. Fuel Pool Cooling and Cleanup System and Torus Water Cleanup System Design Parameters 9.1 Fuel Pool Cooling and Cleanup System He'at Removal Capacity and Makeup Requirements 9.1-3 Fuel Pool Cooling and Cleanup System and Torus Water Cleanup System Failure Modes and Effects Analysis 9.1-4 Tools and Servicing Equipment 9.1-5 , Fuel Servicing. Equipment 9.1 Reactor Vessel Servicing Equiptient .
9.1-7 In-Vessel Servicing Equipment ,
9.1.8 Refueling and Storage Equipment 9.1-9 Under Reactor Vessel Servicing Equipment and Tools 9.1 Overhead Heavy Load Handling System Data Summary I 9.1-11 Reactor Building Polar Crane Data 9.1-12 OHLHS Loads Over Safety-Related Equipment 9.1-13 Reactor Building Polar Crane Design Comparison With NUREG 0554, Single Failure Proof Cranes for Nuclear Power Plants 9.1-14 , Hope Creek Polar Crane Special Lifting Devices as,d Slings 9.1-15 Refueling Floor Heavy Load Height Restriction i
- g. 9.1-16 Not Used l
l 9.1-17 Spent Fuel Pool Liner Drain Lines l' g Decay Heat and Evaporation Rates for Loss of Spent
[g 9.1-18 Fuel Pool Coc, ling Spe5 Fuel Rak uiWeality htalysis Enpt %efe,'s.
}il 4.l -l9 f' l-20 C, rih *cadt fy A nafy . tis 9-vil Retitlff Amendment 3
.%I-2.l Poiwssl fS ed Fuel KutinpaF Pawelv't Special For Cf(HcaWonNy hta(p*rt
, n
_ ~
CHAPTER 9 .
FIGURES Ficure No. Title
's, 9.1-1 New Fuel Rack Arrangement 9.1-2 General Arrangement of Spent Fuel Storage Pool 9.1-3 A Typical Spent Fuel Rack cM 9.1-4 Spent Fuel Rack Arrangement in Fuel Pool l yy0 9.1 ' Fuel Pool,Coolipg and Torus Water Cleanup,,P&ID 9.1-6 Fuel Pool Filter Demineralizer, P&ID 9.1-7 Fuel Preparation Machine Shown Installed in Fuel Pool -
9.1-8 New Fuel Inspection Stand .
9.1-9 Channel Bolt Wrench 9.1-10 Channel Handling Tool 9.1-11 Fuel Pool Sipper 9.1-12 Channel Gauging Fixture 9.1-13 Fuel Grapple I
9.1-14 General Purpose Grapple l
~
9.1-15 Fuel Inspection Fixture 9.1 Refueling Outage Flow Diagram j 9.1-17 Plan View of Refueling Floor During Refueling i 9.1-18 Simplified Section of New Fuel Handling Facilities
! (Section X-X, Figure 9.1-17) 4, l'I'{ R Specia[ Sped Fuel Racic
(
9, V LD SPEWT FUEL RAct Clf171CAUTY (rENIETRY DSER OPEN ITD4 / hk) 9 311 Amendment 7 i s
.- i HCGS TSAR
- g/34
- 1. Normal storage conditions exist.when the fuel storage racks are located in the pool and are covered with about 25 feet of water for radiation shielding, and with the maximum numbe'r of fuel assemblies or bundles in their 5esign storage
. position.
- 2. An abnormal storage condition may result fro buddl6 accidental dropping of a( empt( fuel se , or from damage caused by the horizontal movement of fuel handling equipment without first disengaging the fuel from the hoisting equipment.
b.. It is assumed that the storage array is infinife in all c directions. Since no credit is taken for leakage, the values reported as effective neutron multiplication factors are in reality infinite neutron multiplication factors. The biases between the calculated results and experimental results and the uncertainty involved in the calculations, as well as other uncertainties, are taken into account as part of the calculational procedure to ensure that the specified K,gg limits are met.
The racks are designed to protect the fuel assemblies t
c.
from physical damage caused by impact from fuel assemblies. The rack design would prevent the release of radioactive materials in excess of 10 CFR 20 and 10 CFR 100 allowances.under normal and abnormal storage conditions. .
- d. The racks'are constructed in accordance with the OA requirements of 10 CFR 50, Appendix B.
e.- The spent fuel storage racks are constructed in The accordance with Seismic Category I requirements.
applicable code for the design of racks is ASME Section III, Subsection NF.
Spent fuel storage space is provided in the fuel i f.
l storage pool to accommodate 5.3 core loads of fuel assemblies.
' DSER OPIN ITEM / N 9,j_7 Aggndggnt 7 0
- --v r.e,- , e-, n ,, .- v,-.e..-n ----~--+,,,,,,--,,-,-,--,,-,,.~.-,-.---na..,-n-,-...,n,,,- _ _,__ ~
.,..-,-----m,,..,-~.---,-
. i.
HCGS FSAR 10/83 9.1.2'.2.2.2 High Density Spent Fuel Storage Racks High density spent fuel storage rack,s in th'e fuel pool store spent fuel transferred from the reactor vess' elk These are -
- top-entry racks.
The spent fuel storage racks are of freestanding design and are
- not attached to either the fuel pool wall or the fuel pool liner plate. The racks are constructed of stainless steel, and the details of rack construction will be provided prior to fuel load.
See A ppenclix 9A for a clescriphnn of %e .desfg n, a naly sis and con s truc %n of the k3 h &.u ej agent h el sforage ra c ks .
. s i
f 9.1-10b Amendment 2 DSER OPEN ITEM //y/g f
--- ---w-r-~.,#.,,_m. , , . _ _ , . _ _ . _ , , . , , _ . _ _ _ , , . _ . . _ , , _ . _ _ , _ _, . _ _ _ __
HCGS _FSAR 3/84
- b 4, . .
The maximum stress in the fully loaded rack in a 1.
faulted condition will be provided prior to fuel load. - -
s,
- j. .The spent fuel storage racks also have the capability
' of storing control. rod guide tubes, control rods, and defective fuel containers. When the spent fuel is stored in the spaces provided for storing the above the K,ff does.not exceed 0.95.
- k. Several design features reduce the possibility of heavy objects dropping into the fuel pool. The main and
..* auxiliary hoists of the reactor building polar, crane are single-failure proof. In addition, the main hoist is physically prevehted from traveling in the truncated segment shown on Figure 9.1-31 by mechanical stops on the girders of the pol'ar crane. The crane design is discussed in Section 9.1.5. The removable guardrail and the four-inch curb around the refueling cavities J
further limit the possibility of heavy objects dropping
.i into the fuel pool.
- 1. The fuel storage pool has water shielding for the stored spent-fuel. Liquid level sensors are installed L
to detect a low pool water level. Makeup water is available to ensure that the fuel will not be uncovered should a leak occur.
- t
- m. Since the fuel racks are made of noncombustible material and are stored underwater, there is no potential fire hazard. The large water volume also protects the spent fuel storage racks from potential pipe breaks and associated jet impingement loads.
- 9. I . 2.h3 Spent DVS MO 9.-l.2.4 Fuel Rack Inservice Inspection I -
Ad -$niseritTe'5hsppetTopyro7am 9 in eff)ct th ug,hoGT t)Ht711 l
Af $he' racks-to 'ensute that'.the qu it'y c(f th6 isong's! 'radks Tnsed B 9.1-14 Amendment 7 DSER oPEN ITD( /v0 t-L
. - . . . . - - . . - - _ . _ . - . - ~ . . . . - _ _ _ . . - , _ _ . . _ . . . _ _ , - , . - _ _ - - . , . _ _ _ . . .._ - - - _,,. .-_ ..-.-
on i
,s 9.1~2.4.1
. Test Coupon Description and Installation Details of test coupon description and installation will be provided prior to fuel load.
9.1.2.5 SRP Rule Review In SRP Section 9.1.2, Acceptance Criterica II.1 requires conformance to ANS 57.2, Paragraph 5.1.1, which states that the spent fuel storage facility, including its equipment'and safety-related structures, shall be designed to Seismic Category I requirements. ,
The spent fuel liner plates are non-Seismic Category I and are not considered safety-related. These liner plates are welded to Seismic Category I embeds in the pool walls. Their primary
- g
- functions are to minimize pool leakage and facilitate _
- - decontamination of the pool walls. Since they are essentially 1 nonload-bearing, they will not adversely affect the structural integrity of the fuel pool and the spent fuel storage racks, and therefore do not have to comply with Seismic Category I requirements. Any pool wall attachments will always be affixed to the wall embeds.
Acceptance Criterion II.6, ANS 57.2, Paragraph 5.4.1 states that at least one radiation monitor with audible alarm should be t
installed on the fuel handling machine.
1 l :
At HCGS, permanent radiation monitors scanning the entire refueling floor are mounted on the reactor building walls. These t-monitors indicate and actuate audible alarms locally and in the
- l. control room. In addition, portable health physics
! instrumentation will be installed on the fuel handling platform whenever the refueling machine is used over the spent fuel pool and the reactor core. The radiation monitoring system, including the portable platform mounted health physics instrumentation, is considered to be adequate for protection of personnel in the reactor building during all phases of station operation.
1 osER oPEN ITEM /t[M 9.1-15 ,
i I
~~
INSERT A paf6 l#f li HCGS FSAR 9.1.2.3.3 CRITICALITY ANALYSIS AND RESULTS The criticality analysis was performed us.ing the input parameters contained in Table 9.1-19. Figure 9.1-20 ghows the reference
. nt cr
- 'f* *d F G4 an* 1*y geometry!"g j'_ *ffp b!pticali,ty le
'N f *criticality The
- flat U-235 enrichment of 3.4 w/o. No credit is in.the analysis is based on new fuel with a nominal, taken for the
-simu Min.If f bel' [t burnable poison fuel rods which may be present fuel
. assemblies.2 The analysis uses Utility Associates International's 4;aJ' (UAI's) dif f usion theory model, CHEETAH-B/CORC-BLADE /PDQ7 as the main working model. The analysis includes the various criticality
+
4 safety % elated aspects of the The rack design, Monte Carlo including transport various model, sensitivity calculations.
AMPX/ KENO -IV, is used as the verification model to verify the reactivity of the nominal rack design.
UAI performed similar criticality analyses for Limerick and c"L susquehanna.aTheanabisincludesallthenormal, abnormal,and N -
accident conditions described in Section 9.1.2.3.1.
Table 9.1-20 summarizes the nominal value of K ef fectiveThe of the racks under normal, abnormal, and accident conditions.
various uncertainties and biases considered in the analysis are also included. - .. ,--- - - - . .. .. _.. _ .
f i
.4 DSER OPEW ITEM /y() .
l .
gy f h [481#
IlCGS FSAR CALCULATIONAL MODELS This section presents a description of the calculational models and the basic assumptions used in this ' criticality so,alysis.
i
- i The Working Model I
'l l
The criticality analysis for the Hope Creek BWR spent fuel racks
- employs the CHEETAH-8/CORC-BLADE /P0Q-7 model as the basic engineering tool.- CHEETAH-BN is UAI's BWR lattice code based on the original LEOPARD code and uses a modified ENDF/B-II cross section library. CORC-BLADE enerates equivalent diffusion theory cross sections for the control blade. The PDQ-7 rogram is the well-known, few-group spatial diffusion theory code widely used by the industry. The CHEETAH-8/CORC-BLADE /PDQ-7 model, which is also a part of the LEAHS (Lifetime Evaluation and Analysis of Heterogeneous Systems) nuclear analysis series of Control Data Corporation, has been extensively tested ,
through benchmarking calculations of measured criticals as well as through core physics calculations for several operating power reactors.
A zero current boundary condition was applied to the four sides of the unit reference storage rack cavity ,
"to pmduce an infinite array effect. The two-dimensional, ?DQ-7 calculations were made for four neutron energy groups, two mesh intervals per fuel pin, a flat U-235 enrichment description and a zero axial buckling to simulate infinite fuel length.
~~
The Verification Model The verification calculation employs the KENO-! PX del.
The basic r.eutron cross section data comes from the master libraries of AMPX - e 123 group GAM-THERMOS neutron library prepared from ENDF/B version II data. The NITAWL module of the AMPX program is osan orrw ITex / Vd -
= . . . . . .
. Lrtsen H pp s e (6 ,"
FSNR GCGS used to perfonn a Nordheim integral treatment of the U-238 resonances accounting for the self-shielding effect. The working. library produced by the NITAWL/AQX module retains the 123 group energy structure and is used directly by KENO-IV.
In the KENO-IV calculation, the spent fuel rack geometry including each fuel and water rod cell is represented discretely.
To sisiulats the arrangement of a large number of storage rack units, and for a non-leakage condition in the axial directions, a specular reflective condition is applied to all six sides of the reference case storage rack cavity '";- - M
-.3JL Basic Assumptions To ensure that the analysis follows a conservative approach and conforms to the general guidel,ines of criticality safety analysis in Refennce W the calculations are perfonned with the following ;
assumptions:1.1-10
- 1. A flat 3.4 w/o dis.tribution in an 8x8 bundle, with U-234 neglected
- 2. Fresh fuel, no burnable poison
- 3. Minor structural members replaced by water, i .e.. spacer grids
- 4. Fresh water
- 5. Fuel is channeled.
DsER oPEN ITEM /kd I
yLSO Y $ f df8 N k (b HCESPMR p....r.:
REFERENCE CASE CALCULATIONS Physical Parameters and the Basic Storage, Rack Cavity Geometry The reference storage rack cavity 'J:p.. das a pitch of 6.308" + 0.030". The stainless steel canister has a nominal inside clearance of 6.080 to accommodate 8x8 fuel assen61y channeled in 0.080" thick Zircaloy-4. Plates of the neutmn absorter mater f al Boral, consisting of B C 4 in an aluminum matrix core and clad with an aluminum sheath, are fastened to the outside of the canister.
The Boral plate has a nominal total thickness of 95 mil's and a minimum 2
S-10 density of 0.028 g/cm . Table #contains the values of the input parameters used in the anclyIts.
The rack must accomodate both channeled and unchanneled fuel. .
Studies reveal that the channeled fuel in the rack is more reactive than the unchanneled fuel. Taking the conservative approach, the study here involves channeled fuel (except in the accident condition where the dropped fuel is unchanneled in order to pemit the closest contact between the dropped fuel assembly and the rack).
Two small, but non-conservative changes were made to the reference case in order to facilitate modeling. First, the boral width was set at
' 4.48" instead of 4.465" . Second, the stainless steel flanges used in welding the outer wrapper to the inner can were deleted. An adjustment was made using P0Q to account for these differences.
- Results of the Reference Case Calculations 4,I-I1 ' T./ -
Using the input data fmm Table fand Figure (except as noted above),theK,ff values of the reference case at 68'F were cal-culated for the calculational model described in Section 0,0."
The results are:
y ,-
os . .. - no
~~~
Inserf N f a9sF oF I& _
dc(rs =sAR PDO-7 KENO-IV 0.9229 O.9306 + 0.0042 k,ff, reference calculation ,
955 confidence intarval 0.9222-- 0.9390 G
e e e
p l
I-l l
i !
t ll
!I D5ER OPEN ITEM /yd i
- -,,r-,,,,,,- -
---,------,mw,vww,,,+w,e--e- e-,<---w---
Q$gtt N (V
~
. . McG5 ~
PMR
~
- s. i. 2.3.3. 2. .
andW8 SENSITIVITY AND TOLERANCE REACTIVITY CALCULATIONS 4 Temperature Effect * -
y ,
Using the reference storage rack cavity geometry, the temperature of the fuel and pool water was varied. In addition to the nominal 68'F, 40',F and 212*F were studied and the results of the CHEETAH-B/
9.l-20
~
. CORC-BLADE /PDQ-7 runs are given on . Table As shown, reactivity decreases continuously as temperature increases g Void Effect' The effect of boiling (assuming equal voids inside and outs;de of the rack) was studied by varying the voids from 0% to 20% at a temperature of 212*F with the reference geometry. The{H,E,QAH-B/
CORC-BLADE /PDQ-7 results are shown in SeguemmemusE Tabley., As*
indicated, k,ff decreases con'tinuously as the void fraction increases.
sof5' Pitch Sensitivity The rack design permits the storage cavity pitch to differ from
[ the 6.308" nominal value by +_0.030". The pitch sensitivity calculations of this analysis show the reactivity effect of these tolerance components as well as the reacEiif ty pitch sensitivity by expanding the calculational range from -0.060" to
+.030" at .030" intervals. The results, which are 36seguesuper
. ct. - w 1 L
-.m tabulated in Table E, Fndicate 'that in the neighbor-hoo'd of the nominal pitch, the pitch reactivity coefficient is about .15%Ak per .030" pitch change.
W osta onw Inn / #C
'~ "'
Insed R l' + '
HC&5 FS/?R Effect of Baron, The Boral Plates which separate two' adjacent fuel assen611es have a nominal thickness of .095" (consisting of an 73 mil core .
and 11 mil aluminum sheaths) a nominal width of 4.465" and an overall length of 11 feet 3 inches. The minimum B-10 loading
~ ~ ' - - ~ ' " '
is 0.028 g/cm2 . .,
(a) Baron Width Tolerance .
The effect of reducirg the Boral width was examined.
The PDQ-7 calculation for the neference case con-figuration with the Boral width reduced by 0.0625" yielded k, = 0.92641. Hence, the reactivity increases due to the -0.0625" tolerance on Boral width is Ak = +0.0029... m .
(b) Baron Density 2
The boron density was maintained at .028 g/cm for all calculations. This areal d'nsity e is the minimum
~
density allowed by manufacturing design spe'cificatiopt. _ _...
(c) Boral Core Thickness Variation The sensitivity to the Boral core thickness was de-tennined by calculations in which the thickness varied from 61 mils to 80 mils (the aluminum sheaths were
~
. _ . . varied within..tileran.ce to obtairj [Grst cas.e i:5nt_ thickness)."
The results, tabulated in Table show a continuous increase in reactivity as the core thickness increases. This is due to the fact that the areal
~ ~ '
density i,s_, held constant, so an increase in thickness reduces volumetric density and. 6i a smafl~ degree, the boral effectivenes:
EsER oPEN ITEM ///O M
Enre<+ k fqe 6 of I4
~
((c-GS FSBR d Dimensional and Positional Tolerances The total Ak bias for diinensional'anaspositional tolerances are calculated from five separate contributions:
(i) Pitch Reduction .
(ii) Boral Width Reduction f (iii) Inter-Cavity Spacing Reduction *
(iv) Off-center Loading (v) Boral Thickness Increase *
(1) Pitch Reduction.. The effect of redu:Ing the canter-to-center spacing of the rack cavities is obtaired from time.
Tal/e T.l-Iv; ' r ' t ; ' E _ . : .i . . " " - - - _ " " ' ' ' _ t -- , _,
.ar1cl l'.$ Akj = 0.0015.
(ii) Boral Width Reduction.'. The Ak bias due to reducing the Boral width by its tolerance. 0.0625" is obtained from GamW85m4M Tale 9.l 20 and is ak2 = 0.0029.
(iii) Inter-Cavity Spacing Reduction. Any seismic effect that may reduce the separation distance between adjacent cavities can be detennined from the pitch sensitivity studged elegdendprS.' Bringing two adjacent cavities closer by 0.048" results in the canisters togng'and a reactivity i increase ak j = 0.0023 (from Table : _ ~ , 5). Since I
this reduction is the maximum reduction of pitch possible in this design, this effect will not be added to item (1),
L but will replace it.
i (iv) Off-center Loading. The free space existing between a properly center fuel assembly and the top casting allows an a:sembly to be loaded off-center in a cavity. It was DSER oPEN ITEM gd 1
l V
,T o 3 , 4 & r af e T ,A //
f/ CGS FSAll i
shown that this condition causes no adverse reactivity effect since ,the resulting kaff for off , centered loading is less than that for properly centered assemblies.
N (v) Boral Thickness Increase. The worst' case boral core thickness reactivity effect calculated due to manufacturing tolerance
..,-indisakj=.0001
^ '
~
stackup (.080") is 'obtained from f [.I
~
The above positive ak contributions are statistically conbined to give the total Ak bias for mechanical and seismic uncertainties.
Ak = /(Ak)2j +- (Ak2) +(Ak)5 = 0.0037
- e OsER oPEN ITEM /YO
InwtnyeWoV16 f c(,S. FShR .
9.t.u.a .
y SPECIAL CASES
- Grappler Drop Accident .
s.
The accident considered is the inadvertent drop of the assedly grappler used in lifting assemblies within the spent fuel pool. In this accident, the grappler is dropped in such a way that assemblies in adjacent rack cavities are displaced such that'they are resting in an off-center loading arrangement.
The reactivity effect for this off-center arrangement'was dis-cussed'in ~Secti' a n,@ ([' _
- , .ta. u. 2'0]v)
(al, Assembly Drop Accident (a) Single Assembly Dropped on Top of Rack. No adverse reactivity effect is expected from dropping a fuel '
assembly on top of a fully loaded storage rack during fuel' handling because of the large water thickness
(-14 inches) existing between the top of the assemblies already inside the cavities and the dropped assembly resting on top of the rack. Moreover the P0Q-7 model assumes an infinite fuel length in the axial direction.
(b) Single Assembly Next to Rack. The dropping of an assembly outside the rack is a possible event because of the un-obstructed water area existing between the periphery of the storage racks and the side walls of the pool.
A conservative analysis to evaluate this situation is illustrated in Figure P. An assedly, presumed to be 9.I-20 W .
DsER oPEN ITEM jgp
~~
Q ge;- y e.II o Ylb ~ ' FSAR HCGS dropped during handling, lodges paralled to an assembly in the outer cavity with no Boral slab separating the two assemblies. The dropped assegly is unchanneled to pennit the closest contact with the racgh The dimensions used are those of the reference case. This arrangement of the dropped fuel assembly with a 31/2 x 3 finite fuel racg is reflected on three sides as indicated in Figure Tit $'e'fourtli side is a zero flux boundary. The k,ff result for this ca!e was 0.9120. The result for the same geometry without the dropped . fuel was 0.9064 giving an increase of rdactivity of l 1 Ak = 0.0063 for the above dropped assembly configuration.
>s s> ? -_ _ n v w __pt.e -
N 7 Assembly Moving Between Two Storage Racks The rack structural design does not allow sufficient room tc fit a fuel assembly between any two of the high density spent fuel racks. Therefore, the movement of assemblies
. between racks.1s, precluded.
l i-l l l
. . - . . = - - -. - . - .. .
Jyggf A y 9 6 52.oY '9 ' s
. 14 CGS FSR((
( l,1,3,3,N New Fuel Storage in the Spent Fuel Racks The feasibility of storage of fresh fuel in the high density spent fuel racks was analyzed. Storage of new fuel in the mist, partly flooded, and dry conditions are addressed below. osan oPEN ITEM = /YO Y
'-*-'-'r'- v -me,-
._.y,y __
L nierr rryye w ** to tic 6 5 FSMt 1
- 25% Mist Condition The storage of new fuel of uniform 3.4'w/o U-235 enrichment in the high density spent fuel rack in a 25%
aqueous' mist environment was analyzed with the KENO model ' . '- #-
"';r: OkThe resulting k, and 95%
confidence interval,are shown below:
. 95% Confidence Interval 25% Mist .63751 0054 .6267 .6483 Dry Condition UAI experience in the analysis of poisoned rack criticality
. indicates that the fully flooded rack configuration is the most reactive with reacitivy decreasing with a decrease in moderator density. The 25% mist condition analysis confirms this as shown below. For this reason a dry condition analysis was not performed since it too will be less reactive than the flooded condition.
K Moderator Density - 95% Confidence Interval 3 Reference Case: 1.00 g/cm .93061 0042 .9223 .9389 3
- 25% Hist Condition:0.25 g/cm .63751 0054 .6267 .6483 i
W Partly Flooded Condition The totally flooded condition as analyzed in the reference l
- case is,more reactive than that of the partly flooded i condi tion .
C[' l. 2. 3 3
- f l
M Special Spent Fuel Rack Storage
',' 5x6 non-horated special rack is to be installed in thel'ged spent fuel pool . Storage of control rods, control rod guide tubes and defective fuel is provided for by this special rack. This rack was analyzed for storage of ruptured fuel as shown in Figure CSpecial rack input parameters are summarized in Tabl%
DSER oPEN ITEM /9d
. 4e5
- . - - , , - _ , - - - - , , , , - - , . - - . - - , . - - - - - ,,,-,.-m,-- -
,--,,a,- ,,-w --,,,---w,..,-.,,,w---,n,----
, , --.- _ , - --,e.,,n-
Inseef & f c)* IfrVl6 ! [fC 6S FSA R . 1 The storage of ruptured fuel is a more reactive evaluation than that of control rods or control rod guide tubes. . W Storage of Ruptured Fuel in the Fully Flooded Special Rack The s,torage of ruptured fuel assemblies within defective fuel ' storage containers inserted into the special ' - ! x i:r:tdack was analyzed using the CHEETAH-B/PDQ-7 diffusion theory model. The case was analyzed a an infinite ai' ray in ~ order to simulate storage of .. :: :duptured fuel assemblies in the special rack. The resulting K,ff for this case was .6589. Considering that this K,ff accounts for no radial or axial leakage, the reactivity for the storage of fuel in the special rack is well beiow the design limit K;ff o f .95'. Storage of undamaged fuel within the special rack is less reactive than storage of damaged fuel.
- This is due to the fact that in the ruptured fuel case, the defective fuel storage container displaces water. For this reason, the storage of undamaged fuel w'as not analyzed.
l I { l l l l osta opzw ITra / 9 0 __ . . _ _ . . . _ _ _ _ __ ._.~.__.______.___.
.~ - . . - _ . _ . , , . _ - . . - _ . . _ _ . _ . . . _ , . - -_.
Ensed M p ase i5 c,1-u
. k 005 YS$E
- q. I. L. B . 6 g
SUMMARY
AND CONCLUSION The final result as calculated by both the working model (CHEETAH-8/CORCBLADE/ PDQ-7) and the verification model (AMPX/ KENO-IV) is sumarized in this section and compared to the NRC regulation k,77 limit of 0.950. The
" Reference Case" r this report uses the nominal dimensions given.inFigureh,gerredtoy#and. Table h wtthout the dimension tolerances included. .
g Results of the Transport Monte Carlo (AMPX/ KENO-IV) Verification
, Calculations and the Calculational Bias
- 9. l.2 3.3. (
k,ff. Reference Case WWWMuneddis 0.9306 t 0.0042 ~ Benchmark bias. Ak -0.001
.9296 t 0.0042 95% Confidence Intenal k,ff .
0.9212 4 0.9380 The bias of the KENO-IV vs. measurement is based on criticality experiments perfonned with fixed neutron poisons U. These experi-ments were chosen because they approach the fuel storage rack configur-ation in that they used fixed poison plates between fuel rod clusters, The result of the benchmark calculations was that the KENO-IV results were 0.001Ak above the mea,sured value. This demonstrates a negative bias of 0.001Ak. ! g Summary of Results k,ff, adjusted (KEN 06) 0.9296 t 0.0042 Dimensional and Positional Tolerance. Ak (P0Q, N ) 0.0037 PDQ correction for non-conservative assumptions in the reference case, 0.0006 ak (PDQ M ) Dropped Assembly ok (PDQ,.A Etuum C ) 0.0063 , DSER oPEN ITEM / 70 2 % :. . . _ ._-. , _ _ _ __. . __ .._. ..__._______. __. _ _ _ _.. _ ..,_,_.__ ~ __.__.-_._.___._
~
w.
] fLS&W
/ ,
6 l - g N, . f i 4 l In order to. continually assure the adequacy of the poison material,
, test coupons are provided for a Boral surveillance program.
Forty-five coupons are installed in high radiation areas of the (, -
. spent fuel pool. However, because stainless steel spent fuel t .
' racks with Boral poison material are already in use in other BWR lg fuel pools, a Boral surveillance program is not planned at HCGS.
. If information from these lead plants indicates any problem
- _vith the'Boral, a surveillance program can then be initiated.
I p. 9 I
\
A s e WW
- i f
p S-s4
.i e=
l_ l r . DSER OPTM'f. TEM //o ,
- _ ,- . . _ . - - _ ..c __.,_..,.._.___.,.,_.-...~._,_.,_m._.,__,._. . - - _ _ - . _ . . _ , , , - _ _ _ _ . . _ , _ . .
HStN A f *f e. (( e f* (6 l{CGy ff}k Final K,f ,
'.9402 0 1 0.0042 95% Confidence Interva3 ,
0.9318 - 0.9486 Design Limit, k,ff - . - ; .0 " " :.0 0^' '
,% 0.950 The final k,ff value (0.9486) includes all the design specification tolerances, the postulation of a dropped fuel asse21y, the model
- bias, and the 95% confidence interval from the KENO calculations. However, the negative reactivity effect (- 0.5% Ak) due to the presence of U-234 and the parasitic structure materials (i.e., spacer 2 rids) in each assenbly was not included. e o 8 9 8 osan open Itsa / q'O
4 JO s . _
*~ "
HCGS PSAR
.. . c
+ ^T ,
gates,~Ind"anyother'noncoutineheavyloadsthatmustbecarried over,the spent fuel pool. .
' \ -
, l 9.
1.6 REFERENCES
9.1-1 - C. L. Martin, Lattice Physics Method, NEDO-20913, General Electric, June 1975. < s l 9.1-2 AISC Manual of Steel Construction 9.1-3; AGMA Gear Classification Manual
- Aluminum Construction Manual, Aluminum Association
~
, 9.1-4 9.1-5 AWS D1.1, Structural Welding 9.1 NEMA MG-1, Motor and General Standards .
. s 9.1-7 '
National Electric Code V 9.1 OSHA 1910.179 OhHA,.Vol37, No. 202,,Part 191 ON
~
9'I'9 l
' n 9,,[p objec fJe>
l kb6 T ll/ zW - 87Q -ue, Spent Reac/w w L.yf Wder '#' F#### Stora e FaciNkeJ a5 W Raision 0, Samry RepidoV R',l-lI ()AL B4-4L Nuclea,~ Cri%fl} Af & For tile SgFd l Racks & Hope a<eek sewaGy stdion. DEER OPEN ITEM ///d
"* f- , . ;- . . ,,.m._._ , _ ,,. m , . . . , . . . . . . , . , , , , - ., _.;._,_., , _ . , _ , . . . _ _ . .
MCCr3 PSA R
' TABLE 4.l-14 pqe sopz
(. N CitlTIC4LITT REFERENCE CASE SPENT FUEL RACK INPUT PARAMETERS 1 FUELASSEMBLY.(8x8) -
. FUEL GATA O.410" Pellet 00 .
Clad OD 0.483" Clad Thickness O.032" Clad Material Zr-2 Fuel Rod Pitch . 0.'640" , Active Fuel Length 150." U-235 Enrichment 3.4 w/o Effective (Stacked) Density 96.51 Theoretical WATER ROD D'ATA (2 per Assembly) , Water Rod OD 0.591" Water Rod Thickness 0.030" Water Rod Material Zr-2 CHANNEL DATA Channel Inside Dimension 5.27" Channel Thickness . .080" !- Channel Material Zr 4 l .. (; - DSER oPEN ITEM j y/)
- , E.,-m-... - . - - - . . . . , . . . . . . , - - _ - _ . - _ _ . . . . , _ , - , . . . . - . - - -
MCGS.FSAR (,, , TABLE
@~O (continued) helef2 BORAL PLATE DATA ,
Total Thickness 0.095" + .005"
- .01 0 l
Width 4'.465" t .0625" Length , 135"1 0.25"
- Sheath Thickness 0.011" 1 001" Sheath Material Aiuminum (1100 series)
Core Thickness - 0.073" (nominal;s ' range:.061" to .080
~
Core' Material Boral
- 2 B-10 Density 0.028 g/cm CAVITY DATA ,
( Can Inside Dimension 6.080" (nominal) Can Thickness - 0.090" (nominal) Outer Wrapper Inside Width Dimension 4.562" 1 020" Outer Wrapper OutsideWidth Dimension 5.36" (nominal) Outer Wrapper Inside Th,1ckness Dimensf6n- 0.101" (nominal) l- , Outer Wrapper Haterial Thickness 0.024" (nominal) Stainless Steel ! Cavity Material
' Rack Cavity Pitch
- 6.308" 1 0.030" l
l [ 1
/
DsER OPEN ITEM /fb e e a ~o - ,. --- e e -e-m- ,W e - v-. - .w,~, e, .m, ,e.e >w,,,--w-, . -,w--g.,,,, wee,ww,+ - ,.m-.mm.-,--,.-
i I hC&$ f$$ (- ' mup- 7,/-20 rvaoM j CRITICAUTV . M REFERENCE CASE - SIMMRY OF' 4-GROUP PDQ-7 RESULTSl k,77 AS A FUNCTION OF TEMPERATURE AND VOIDS REFERENCE GEOMETRY - FIGURE p y,20 [ E 1 Voids eff Temperature 'F 0 0.9265 ' 40 . 0 0.9235 68 0 0.9028 212 10 0.8845 212 20 0.8630 212 ( l l l l (
/ O DSER OPEN ITEM
M ( ' [4782 Ob / TABLE
- 9. [I-ro N
k,g A5 A FUNCTION OF PITCH VARIATION PITCH E (INCHES) - eff (INCHES) 6.338 .9220
+ .030 ,
- 6.308 .9235 Base 6.278 .9250
.030 6.248 .9264
.060 .
6 ( DSER OPEN ITEM
/%
-e ,= , . ,w-w -y..., ,.#< , ..,,--,w-~,-wwwy...~,w-%,,,-.--s~,w-,,*v--.w,e. ,-ry. --,e--..w,,.wr-e, ,----y-- ,vw as c ,-y%. --gg,,m., 3--,-
'.( .
TABLE M @ # T. ( ~Z O i l I l i
~
k,ff AS A FUNCTION OF BORAL CORE THICKNESS - AREAL DENSITY CONSTAN Thickness of l Boral Core g Inches eff
- i .
j e 0.061 0.9233 0.9235 (base) 0.073 0.080 0.9236 l
- l. .
l-1 l-r ( . I ~DSER OPEN ITEM /YO l I-l _._ i __ ,
WCGS PSAR
~
( . TA8LE V T. [- 1 [ h5PECIAL NON-POISONED SPENT FUEL RACK IllPUT PARAHETERS FOR TifE CFITICAlITY AMAtysts .
~
~~ ~ ~ ~~""'
FUELESSEMBLY(8x8) .
-p i .
FUEL DATA-Pellet 00 . 0.410" Clad 00 0.483" Clad Thickness 0.032" , Clad Material Zr-2
- Fuel Rod Pitch -
0.640" Active Fuel Length 150." U-235 Enrichment 3.4 w/o . Effective (Stacked) Density 96.5% Theoretical WATER R00 D'ATA (2 per Assembly) - {. Water Rod OD 0.591" Water Rod Thickness 0.030" Water Rod Material Zr-2 CHANNEL DATA Channel Inside Dimension 5.27" Channel Thickness .080" ! Channel Material Zr-4 CAVITY DATA i can Inside Dimeniton - ' 11.50 + .00"
.06" l
! Can Thickness 0.165 (nominal) Cavity liatarial Stainless Steel Rack Cavity Pitch 11.66.7 + .020"
.000"
(
- DSER OPEN ITEM [%
l l * - 1 .._. _.
+:
I e \
... .. To
.+T TOP VIE W Ve a . -v' :
~
TYr
- Borad y Centeral .,
** ;~YPICBL CRW(STER EQar tut TH BOML AWD WR A PPERS o s j' ;
l i
- 1 l \ Canister .
I 1
/ .
l 4
+
- -& , W Oater Vhayei
~
I j Ay Vent rre o s v l
, a f(
N ?[ e ! 3
~.
; B*al .,t N_;~ A
\
\ 'h C
ml 1 il l 1 ( [ %.L ' n N I l l ~- % :1: > i ; o i vedtre i . v s s '- r
,~
e
%. ' ' HOPE CREEK g I GENERATING STATION s
~ FINALSAFETY ANALYSIS REPORT 3( l
}
"g
~~
O ~ ~- r A TYPICAL SPENT FUEL RACK o o , FIGURE 9.13 $f ge /. y ,fif.. __
l I le 8 fOEF ECTIVE FUEL CONTROL RO) GU10E TUS E. , G.E.e10tO5419 . y/ STORAGEG.E. e 741E720 CONTAINER e * - CONTRO L ROO 11.865 G .E. e 814E749 11.50 - l . l TYP. INSIDE OPENING
" 1 I 25
! ! ELEv.180.25 max.
l 7g 37s.25 uiN. J l -
/
' l .
' \. t i e i 1 I e !
-W
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!- 71.16 SECTION 0"O O BOTTOM VIEW SCALE: NONE N SCALE: NONE HOPE CREEK GENERATING STATION 5
p NOTES: FIN A.; SAFETY ANALYSIS REPORT s ~
- 1. REGULAR FUEL OUNOLES COULO ALSO BE STORED
! IN THIS RACK ON A TEMPORARY 8ASl5 IF ff[C/[(
0* NECESSARY' Ai WiCAfSPENT FUEL RACK [ o at 2. 5 m 6 $PECIAl. RACK FOR STOR AGE OF CONTROL
$hggf l0f1, l
l W ROD G.E. *014E749 AND OEFECTtvE FUEL STOR AGE CONTAINER G.E. e 761E720 AND CONTROL G RE -[gg ROD GuiOE TUB E G.E. e '0505619. l .
.- , - , , - - . . _ - . - - . . - . .-. . _ - . . . - - . . _ _ . - . - _ . _ . . - . ~ - . . _ _ _ . . . . - . . .
~ M SPECIAL M .70 fr fy RACK ,
( ,
+.020 c 11.565- .00039
- A ll // // / / llN//.
/ -7.716 10 8.03 CD MAX
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or ( . l DSER OPEN ITEM /t/d
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$ 9are 9.(- M slieef 2<
l
HOPE CREEK $ PENT FUEL RACK AEFERENCE CASE GEOMETRY Sc% *\ *s%* p. 0600 O O 'O 'O 2.n < 0 0 .0 0 l 0000 . 00 00 l 3,. . 0000 ; O000; O O O Oz.e
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pgfPEResteten-MMt?T SPENT FUEL STORAGE (DROPPEDASSEMBLYACCIDENT Mck,
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l _ _ . _ _ _ _ _ _ _ _ _ . . . __ A 1.717 N D i x 98__. . . _._ _ _ ... ..._ . _ _ _ . DES I 6Al I A^l A LN.5 ( 5_. AN.D ' 9qN 5 Tg uc.Tt art ...._
= . .._ . . _ o F imin . L GM.J' 5 PEN T EH EL _ s To rc A AR.4c ._ E k.5 . _ . _
__. _ . . _ _ . . _ _ . . _ . _=.. . . . _ . . . _ . - . _ _ . _
. 90 I . S t o P E ._ _ . _ _ _ _ _ . _ _ _ . . _ _ _ _ . . _.
Th6 a.7pcn,d,i3 _ de sc ri.69 ,,,& e,. de3(3n, aaalysa amc{ . co ns hmc.byn o f Of. 5 y e d .... b I _.rs.c kJ. . _. . _.... . ...._. . _ _.. 1 #f $ . 2. , , 9 6S( R.L P Ti dAl oF S P EAJT Pt4 a. foo t AND RACKS .. . Sec Ron 9.1.2.a. c o n+ncns a de3crcp w n of ec s7< t b I S h ra. e facili ih c lad (n de hc h c(e.tsi s7eai Gel s brage racks. The syeat fue ( racks are of free s kading desy n aad are n o t a ffac hed to e r&er +h e fuel pool wall or +ls e fuel pool liner pla te . Figures I. 2 - l o a nd I2-52 show [ lh e spen t h<el pool in retahon lo o + lie r jo/a 4 s tru c + a re.s . 9.l.5 ad 9. l. 4 sh ow dekils of de spen f fuel Fc3 ares [ ra.eks.... . . . . . . The. spenk fuel ruks are des cgned +o withs ts.ul & c. poski4.hd drop of a fue i bunJ(e. Sec fron 9. /. 5 con k(ns DSER OPEN ITEM /(/O 9A-I o . ._ . _ - - - _ _ - .
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os d haud/s>tf ... .
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(
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A II par f3 o f A e sp,nf. .fxeI rack 3, eycegF Ae ' g u36ty ,
. . _ .s.c re ss . tn . h e. fte f _o f .each modale and Ae. p o rs o n .
l ___ exte r cal , are ma.de bm A STM A 2 4 0, Typ e 304L
. 5_laioIe55 5!ceI The_,oidjas htt$ scr.em s._4re usad e Q.m. . . . . .
. ASTM A%4, TypeJ 30 s facn less she I. Bor a.l cs. Oc
. _2 ; f o sHM MA h e PC .. _, ,. ,... ._ . _ . , .. . . . . _ . .
[ Desr3n, b.b rcc o. ht n and installa htn a f th e sgen t fue I rs.c ks are ge r 5ormed hasec4 upon Sv.bs ec hmn NP i ceou..re mea b ef Re ference 9a-I fror Cla s5 3 co,wg on e .t i-
. su.h por ts. .
n
._ g. 4 .
.S E t $ in t C AND i tn PA C r L O A D S . _ . _ . . . . . . _ . _ . . . .
Th e _s e rs m cc cn gu.f._ fo r-4e spea t fuel. racks costs 44 .
. . . a.f 8t= r ...re s pons.A _sge.c fra. . h.r_.fh e.. syca.t. fue l pool sI46.
..f_lo o r r o po4.c _ s r ec.+v. _4 re ..de v e.I a p ed Mm 3'rou nel....
.._ _ _. $h .~. 'L.. . . . . . . . _ _ _ . . . . . . . . . ..
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- y. with d e . reg ut,est en fs
.of . Regulabry.. Guides 1,4.0 a>ld. 6AJ . . Asceleva h.vru _
.__ -. .h.% < h a hiries ..are.. de.velop.ed.. Joe _ L,a._ h ortson laf .. .. _ ..
..d ove c h v.r.tS . aad .pne...rer_h'c.af_. 41e c h_vn_.. fro.m . fit c .. -{loo r._ .__. .
- Tit e,s.s. Mr.ee._ km e_ha.& rces.. a re_
rtSfarts e. . spec.fra ., . i _.... impos ed . simu flueously, Th e. pea k ..respenses . (ner ca c h . dire ch_'on. p.cg_,.com bin ed.. ky . Jy ga re roc / d.. flt e
._.-- .-.S am
- f. +l, e . s fu a ees. sh.. .a eco rdaat ce with Regu la lv y . . ...
-_..... .Ga ede. . I 9 2 . _ . . .
In sy a.e F (oads_.due._k.Jue I. ra 1+lmy..are ca icwla Hd . . m e 14t ods deserib ed .in Seehun 98
- G. /mpaet ust'ny loads a re consedered kr lac a./ a > we ll as overa y e ffee h on the rack desyn.
98.5 LCADS A N D i.o A D com a /N A riodS j L oads and toad com bena +uvn.s are in. a.;re emed with l t
. Ta ble I af Reference 98-2. Therial e ffeefs are I rncluded . by astny decreased ma ferts./ properh'e> s / +he applec a ble tempera ture le ve I . Sin ce the ra c le.3 a re free shn diny , +be re are no thermal s tresses. . .
om om a j.g - - - - - - - - - - i
'f8 - 5 .
l 98 6 D ES i Gh1 AND AAIA M $($ HLO C E D U R.E S_,.. .. . . . _ . . . T
. e uSiny he ANSYS c'oylet grain.
Each fue l.. rec.k.J .fdeakt.ed as _.g. , 3D bck .elemeat _ modc([ .Ff$ u.ye ..'!8 - I _.s .4 ows _st.._ Ji'ye cao rs le ie.. g e r fron . o F a. ro.e h . Jh.e .. c a. rte.> k es_.esd.. .ho &m. g rectgla le
. a re .._ m o del ed .wi%. . y la.k . ..e l.c m m b.,__T.h e ..ye n m e k rf b a r. , .
u keek s e c a r e s . %. e . e a. n cs.k es_a +... A c 4e p , aa} . A e . .
.sbTfenin3 b4.rs ...fo r N y ted__gI4.k are ud peel *(&
. b eo.m e le m en ts. T h e. th.@ _ga.cn.le ss steel' d.m pper c a n += in cy . _+h e. .n e u.b on. _a h s or_ber. ae.cl A e. . s. %% le s s s Fe e
_. yaneIs a s e d . lo.. c'lo s e.. . F F +k e._ a He ena. te . ._ca , ckes . a ve . . . . __ . k F m.odeled _M 44seir- w s se.s <.re. iuc iad ed . The hei L a s s ea tre s a re mod el ed o s bea.m e le m enfs . Fcga.r e 'l6-2 5 hows a dou.ble_ rm.ck mede ( cn sche,u.ke. form. 3D inf e r (m.c c elemc th etre used fo re f res e a t f h e. fuel- b- c a ncs h r. . clearance as well as de. ra.ck. b_ rack ga.p. These son icn ea r e le me,1(s re p roduc e krecs clae Av. Gu e ( ra.+tleng _ad y a ss i.ble. . r a.c G- k~ <ack. l .. . in +e e a c h v o __ .. . .._32_g_e.g ._s (e ese net 3 _. w ith m4.k real.. _ p roper fees _ b4 sed on +ke_. faler fa c.e &cchvn .coeffeece~ts are used do s tunt a.k_3Le_co r ne e _s yyar b'y le et. .
.. .. .ukeek uq ...s kcle_.. or. . l(4 t ._d f__ Ae. y co L4 loo.cs. ho osa om zm /VO g..g 4.._. . . _ . . . . _ _ . . . . .
~
I . . _ . _ . . _ - . . _ . . _ , _ - .__,... . ._,, ____ ,...___.. ,_, _ , _ , _ , , _ _ _ , _ _ _ _ _ _ , , , _ , _ . . , _ , _ _ _ _ _ _ _ , _ _ _ _ , _ . _ _ _ , , , _ _ _ _ ,
1 1 _.beaad,y .yafue.s o f h, c.hvn .co e./kccea t-_ ( 0. 2._a_d
.D.8)...Ar_C._u3ed.I'h order. h A en.ff. he..nos.f__c.r.Lhts)
.c o.ndtbbn3. hr sIfdth. and_hr_matmuon_rca e h.w. a f _..
fu f. ._..y h.. s<yyort . ... - - - _ _ _ . . __ 3}ru c lura. l..da mp01)..coe fkei.en h .of. 2 .per c.ea L.for oaG . . .
.. . a.nd. 4 pec i.enf.. br. p_5E. 4 re.. a s e.d, _ xcey f _ Vita i i'mpa e t -... .
da mprag of.._lo_per_cen t_ a f_.crt.' hts l U3 sed b.r fA s .y af .__. . eIem en b S!.hce . impa.cf .drsstpa fes. su hs /an Hu / .aonours h a f . . .
. .en er3y. . lain . 20. .fe el. _af su. bmerg en ce, . s losIwtf e ffe c h . . .
are nef leyibl.e and. 8eeefare are aef leehd. Flued damping . e f F<e h aec. a Iso ..nejteckd. To sanufale. Mt e anaersion e f fech, a lI h in terna./ wa /er en tespped utMen +4 e ra ch en velope a added k &e 11oritoit tal mns . Th e ex +eaa f waler be lweest adjacent racks n wde/ed usi tbc bydrv dy n.smec cou,oby element .rliswn in Ryure M 'l. A ps.au.e 6-ee iu am aan h, o f . skudy , ta keck conscders vs. '
.. b I ik . A. . n3 e _r=.ck I y J.h_ co 4c.fecl_ .+o. dclee m h c._ w h e.k
. _e( 4Le. b llo w coadekcons s kou.id ._h e ..conwdwed .. n. o,de r do e v.A i w .4kc. seesma. reSgonse .of & e caekS. .. .
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.F &c A.e l en ene scJe of .h_ Fack _.3 cox3rc{e e .. _ _ .
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... . . _ . . . . - . . . . . . . . ._ _._.._.._..i....-........
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_.. . fte G re u e 9A .l. _.S M ss . .e-( Is.6,. plo.h..ete - A - . eq un m.le.s. . Ar .. ple 4e J....s it e ll .+y e . sy g ov % . s mcc c4ress keld3 N dese coyo,1e h cm W a ,us( .
%r Ae (m! d>-og con.drhwn ,.. local perusaed de bruhm psu6() ,-equcri r<.gw rs ge~n s, b(e poacteel +ka.+
ue ca. l\ s M sse3 Ac ac( exce ed . va. lues ye r uc+ted bre l.ese \ 'D s e rvce.e h%.c h aJ de res44, de 6 rua.% l
. .. does wi' perud. +Le _fs.e. l . .c o..n_frey.ra hvo ._K e pp. fo .e.xceed 0 9 '5 .,__._.._ __. _.__ .. . . . .
- - . .. ._ . . . _ . . . . . . _ . _ . . . . . . . . _._ .. _ . . . . _ _ _ . _ . . -_. - = _ _ _ .
t osta orru zrza /90 _ . _ . __ l ._. .. . . . .". . A . .b. ._ L
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. . 9 A ,. 8.___..._ M.4 r e s.ia o , a u4 wry CON Tsct. MO JPECIAf . _ . . _ - .
co,u s r exertou 'T ec H'MI Q M E5.s RI. Af.e r_ra.l s_.4 c.c._d e s e r .b ed_.in S e dron_SA . 3 ...._.6 a It+y ... _ controI .. proce du res lor....ma te.rta.Is,_.hb ricakan an d.
. design. con 4vo I ..and veriHcoMon_comply . with. A^lS I _. . ._N A 9._'2. . . .C o n v en ka n aA_. sons kc.kon me % o ds ' a re.-
LL S tb o . A1, de sc rdied .cn... Sec hvn _.1. I. 2. 2. 2. e , . appro xansa kyt __ __ . . 2 5...f e r cen + . o f. +kek kl. spea f _fu e / s kraf e capa cWy. ..
. .svilI be prodded by rscks .i>tsk fled prior k int & .
The ra eIcs a,itI be jilu t opera hun. remain org inifa llen la ler. Th e inth a try ins ht ifec/ rs c les a r e y e n'erally l o ca led a t +b s norFh end a f he spea f 4 eI yco I. Ther e fo re, +ke addehunal racles es n he assla fled la ler wifhou.f be(+tf frsitsjsorfed orer r<cks exis fot} y u h cc h co n la in spent fael. 1 _ .9 8 9 .... R. EF G R GMc E'S . . . . . - ._. - l- __ Ab-l
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152 A. U DSER OPEN ITEM /t(C f l k r
l l l HCGS FSAR e l QUESTION 22M 15 (SECTION 3.5.4) Provide sketches of the mathematical models used in the design of spent fuel racks. Describe in detail, the methods of analysis by which seismic and other loads are applied tosthe racks and the pool.
. pzsronss n.: ::w;:::: :::: ,__m.__ = ' :: :::: i:__2=,_:il:ti: :::, ;;:
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' r Mformabbn. .
8 6 e l i l l i 98 D i G DSER OPEN ITEM //d N 220.15-1 r
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.3 HCGS FSAR 4/34 OUESTION 281.13 (SECTION 9.1.2)
Identify the materials, including the neutron absorbing material
'(poison), used in the fabrication of the high density spent fuel storage racks and all other structural components wetted by the pool water. Indicate how the poison-contai< ling cavities are
- vented.
RESPONSE
All parts of the spent fuel racks, except the adjusting screws in the feet of each module and the poison material, are made from ASTM A240, Type 304L, stainless steel. The adjusting screws are made from ASTM A564, Type 630 stainless steeb. Boral is the poison material. fwAth H llCO hed ttedbHeIt. l{dd treatmenf $ Cafe )b
. . . tenttwe4. .
Thin (0.024 inch thick) outer canister sheets hold the'Boral The tighltly ag& inst the 0.090 inch thick inner canister walls. outer canisters are spot welded to the inner canisters along the bottom and both vertical sides of the outer canister. The top edge of each outer canister is seam welded to the inner canister. N g;p b e tuc e.. ti.e sp;t u; M : pre"idFj;he poison 'vente ce
%wrt. an F@w'e 1.I-3.
I l e DSER OPEN ITEM /pO 281.13-1 Amendment 5
,, _ ,_,-.-,,~.~_._,.-c , , _ , _ . - - - - , , , . _ - _ . - ~ -
c j l I HCGS FSAR l
' QUESTION 281.14 (SECTION 9.1.2)
Provide details of the materials monitoring . program for the spent fuel pool, including type of samples used and frequency of inspection.
RESPONSE
There are no plans' to provide a materials monitoring program at
' HCGS, as-vented stainless steel spent fuel racks with Boral poison material are already in use at other BWR spent fuel storage pools, such.as Monticello and Browns Ferry. .PSEG will develop a program to monitor the Boral- surveillance program of either Fermi, Monticello or Brown's Ferry by' March 1985. HowcVer, forty-five test coupons have been installed in high radiation areas of the spent fuel pool in case'a need for a HCGS materials monitoring program is indicated by the Boral surveillance program.
DSER-OPEN ITEM 140 281.14-1 Amendment
, , JLt 30 '84 0 2 6 8 6 4 0
...t o HCGS FSAR 8/84 l.
QUESTION 410.38 (SECTION 9.1.2) Insufficient information is provided for review of the criticality gf the spent fuel pool. Ihe design bases are acceptable with respect to criticality. Ihe (nformation required for the review is promised for later. Such information should include the following:
.a_ .
Sufficient structural detail to permit an independent calculation of the criticality of the racks.
- b. A description of the calculational methods used along with the results of'the. verification of the methods.
Ihis may be by reference to documents previously submitted by the I organizations doing the analysis. l c. A tabulation of the nominal value of k effective of 'the racks along w_ith the various uncertainties and biases considered in the analysis. . d. A tabulation of the reactivity effect of each of the abnormal 1 accident) situations considered. . l RESPONSE , Owff kieni th= criticality of the spen * . Iuel veel, including r.ha c liuied ;bo te gill be avaWbl-e-by --)---- SeptemM'er 1964, ana will ce addeTTtrSec4 i nn 41 S O __ - Seche 1 Cl. ( . 2 3.3 has I:een rev'Ne/ tc iachek ll: e in he a h.bi 1 e eguer fa f a ksw. i l l l
- IN / Yd 410.38-1 Amendment 7 l
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TL '~ BROOKS &Perxins,; ADV80Ceo stroCrures An AAR Company [ . [~ / AAR BROOKS & PERKINS CORP.
-}2633 Inkster Road
-:Livonia, Michigan 48150 E.
E
/
E . Report 624
.!, Boral
' Neutron Absorbing / Shielding Material Product Performance
* , Prepared By: Report
/
// -L. Mollon ,
Nuclear Prpgrams Manager
- 7( July 20,1982 -
[ .
Rspsrt 624 (. BROOKS &Perxins, f.~ ADVanCBD STPUCTur8S '1 An AAR Company TABLE OF CONTENTS Item - Page [. General 1 Boral Material Characteristics 2
- Aluminum 2
- Boron Carbide 4
- Material Compatibility 6 Boral Physical Characteristics 7 Dispersion Uniformity 9 Corrosion Resistance 10
- General Corrosion 11
.p - Galvanic Corrosion 15 Pitting Corrosion 15 h' -
- Crevice Corrosion 16 ,
.p - Intergranular Corrosion 16
( - Stress Corrosion Corrosion Monitoring System 16 J7 Radiation Resistance 17 Neutron Shielding Performance 21 Boron and Halogen Leachability 21 Residual Activity 23 Installations Using Boral 24 References 25 IE . . g . . lli r !N l - l l-11
. Report 624 . It
[- BROOKS &PerxinS.% ADV80CBDAnStructures AAR Company v
. BORAL
,NEUTRO*' ABSORBING / SHIELDING MATERIAL
. Product Performance Report
[L GENERAL
- Boral'is a thermal neutron poison material composed of boron carbide and the 1100 alloy aluminum. Boron carbide is a compound having a high boron content in'a physically stable and chemically inert form. The 1100 alloy E aluminum is a light-weight metal with high tensile strength which is protected from corrosion by a highly resistant oxide film. The two materials, boron carbide and aluminum, are' chemically compatible and ideally suited together -
for long-term use in the radiation environment of a nuclear reactor or in r r i m-spent fuel containment. Boral is an ideal neutron absorbing / shielding material because of the following reasons:
- 1. The content and placement of boron carbide provides a i
very high ' removal cross section for thermal neutrons.
- 2. Boron carbide, in the form of fine particles, is homogenously
.' dispersed throughout the central layer of the Boral panels. - l
- 3.
- khe boron carbide and aluminum materials in Boral are totr.11y unaffected by long-term exposure to gamma radiation.
i
- 4. The neutron absorbing central layer of'Boral is protected by permanently attached surfaces of aluminum.
Report 624 r BROOKS &PerxinS 4 t: ADV80CBD Structures %# u m c ,,
.p L 5. , Boral is stable, strong, durable, and corrosion resistant.
Boral is manufactured under the control and surveillance of a computer-aided Quality Assurance / Quality Control Program that conforms to the requirements of 10CFR50 Appendix B entitled, " Quality Assurance Criteria l for Nuclear Power Plants". For further discussion on Quality Control see Brooks & Perkins Bulletin No.102. { Boral has been licensed by the USNRC for use in BWR and PWR spent fuel storage racks, shipping and storage containers and for many other shielding i . l uses including control blades. For spr cific applications see later in this report. Boral panels can be used in the flat panel form or fabricated into a variety of geometrical shapes by standard metal working methods and techniques. .E The shielding capability of Boral is assured by wet chemical analysis or neutron attenuation testing and is specified as a minimum of grams of B 10 per square centimeter of surface area. Boral can be provided at any B loading up to 0.06 gm/sq cm as required. I BORAL MATERIAL CHARACTERISTICS ,
' Aluminum.* Aluminum is a silvery-white, ductile metallic element o
that is the most abundant in the earth's crust. The 1100 alloy aluminum r- {i is used extensively in cooking utensils, heat exchang'ers, pressure and i - storage tanks, chemical equipment, reflectors and sheet metal work.
Report 624 BPOOKS &Perxins .4 {' ADVanCeo An sTruCrures AAR Company v
. It has high. resistance to corrosion in rural, industrial and marine atmospheres.
Aluminum 'has atomic number of 13, atomic weight of 26.98, specific gravity { of 2.69 and valence of 3. The physical and mechanical properties of the 1100 E alloy aluminum are listed in Table 1. [ Table 1 - 1100 Alloy Aluminum Density 0.098 lb /cu. in. 2.713 gm/cc Melting Range. 1190-1215 deg.F 643-657 deg.C Thermal Conductivity 128 BTU /hr/sq ft/deg.F/ft (77 deg.F) 0.53 cal /sec/sq cm/deg.C/cm
/deg. F Coef. of Thermal Expansion 13.I x 10 (68-212 deg.F) 23. 6 x 10 /deg.C Specific Heat 0.22 B TU/lb /deg. F (212 deg.F) 0.23 cal /gm/deg. C 10 x 10 psi Modulus of Elasticity Tensile Strength 13,000 psi annealed 2 (75 deg.F) 18,000 psi as rolled Yield Strength 5,000 psi annealed (75 deg.F)
~ 17,000 psi as rolled Elongation 35-45% annealed ,
, . (75 deg.F) 9-20% as rolled Hardness 23 annealed . f, Li (Brinell) 32 as rolled Annealing Temperature 650 deg.F 343 deg.C
Report 624 BROOKS &Perxins,s j {: ADVANCED STPUCTur85 %8 An AAR Company Chemical Composition - Aluminum (1100 Alloy)
. 99.00% min. - Aluminum 1.00% max. - Silicon and Iron
. 05 .20% max. - Copper
. .05% max. - Manganese
.10% max. - Zinc
.15% max. - others each The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general ,
corrosion or pitting corrosion and the film remains stable between a pH range
- of 4. 5 to 8. 5. More detailed corrosion data.is provided 1ater in this report E.
and in Brooks & Perkins Bulletin No.101. Boron Carbide. The boron carbide contained in Boralis a fine granulated powder that conforms to ASTM C-750-80 nuclear grade Type III. The particles range in size between 60 and 200 mesh and the material conforms to the chemical composition listed in Table 2. g . 0 E
. - 1,. .. .. - - . - .
Report 624
- BROOKS &PerxinS a
, ADVanCeoAnstruaruresv AAA Company r ; Table 2 - Boron Carbide Chemical Composition, Weight %
Total boron r 70. O min. . B isotopic content in natural boron 18. 0 min. F L- Boric oxide 3. 0 max. i Iron 2. 0 max. Total boron plus total carbon 94.0 min. E The general physical properties of the boron carbide powder are listed in E
= Table 3.
{. n Table 3 - Boron Carbide Physical Properties T Chemical formula BC 4 h Boron content (weight) 78.28 % Carbon content (weight) 21. 72 % Crystal ' structure rombohedral Density 2. 51 gm/cc-0. 0907 lb/cu. in. Melting point 2450 C-4442 F 1 3500 C-6332 F
-{ Boiling point Microscopic capture cross section 600 barn iE is
Report 624 BROOKS &PerxinS:- !
-b- ADVanCeoAnsTruaruresJ ,
AAR Company
- Materials Compatibility. The materials contained in Boral are compatible with all pa'rts of a spent fuel storage system in either a boiling-wati r (BWR)
{ or pressurized-water react'or including the fuel assemblies, the cooling system, the cleanup system, the pool liner and the structures of the storage racks. This compatibility is evidenced by more than seventeen years of continuour service in both types of pool water (1) (3). None of the following materials are contained [ in Boral nor do they come in contact with Boral during its manufacture and therefore Boral can not cause these materials to come in contact with the fuel ass emblies: E a. Any material that contains halogens in amounts exceeding 50 ppm, including chlorinated cleaning compounds.
- b. Lead
- c. Mercury
- d. ' Sulfur i[
l i e. f. Pho spho rus Zine
- g. Copper and Copper alloys
- h. Cadmium
- 1. Tin
- j. Antimony
). k. Bismuth l- 1. Mischmetal ! m. Carbon steel, e.g. , wire brushes
- i. '. n. Magnesium oxide, e.g. , insulation
- o. Neoprene or other similar gasket materials made of halogen-containing eb.stomers.
- p. Viton
- q. Saran ,
- r. Silastic Ls-53 E s.* Rubber-bonded asbestos
- t. TFE (Teflon) containing more than 0.075% total chlorine (glass-filled) and TFE films containing more than 0.05%
((i total chlorine. , 1 E j . _ _ - . _. . . . . -_ - - . .___ - __ .-. -.
l Report 624 BPOOKS &Perxins 4 ; {; - ADVanCBD Structures- . um conm l l T. l ' u. Nylon containing more than 0.07% tutal chlorine. !
- v. - Polyethylene film (colored) with pigments over 50 ppm ;
fluorine, measurable amounts of mercury or halogens, or l [~ more than O.05% lead.
-w. Grinding wheels that have been used on other than stainless :
I steel or Inconel material. [' x. Water containing more than 25 ppm halogens during any cleaning operation. Any material that forms alloys or deposits on the fuel [ y. as s embly. E BORAL PHYSICAL CHAR ACTERISTICS Boral is a clad composite of aluminum and boron carbide. The Boral panel consists of three distinct layers. The outer protective layers are solid 1100 alloy aluminum. The central layer contains a uniform aggregate of fine boron carbide particles tightly held within an aluminum alloy matrix. The boron carbide particle in the central layer averages 85 micons in diameter. The average spacial separation is 1.25 to 1.50 particle diameters. The overall 10 thickness of the three layers will vary depending on the B content in accordance with Figure 1. The physical characteriestics of a Boral panel will vary of course, according to clad thickness, overall thickness and B content. A typical Boral panel for spent fuel storage can be described as having 0.020 grams of B per sq. cm with an overall thickness of .075 + .004 inches including a nominal clad
' ~
l of .0095 inches on each side. The physical characteristics for that typical panel is as shown in Table 4. 'L
- 1 .
w
Report 624 1 1 BROOKS &Perxins ,s :
- h. ADV80Ceo AnsTruarures AAR Company v
[ 20 Figure 1 Boral Thickness Versus B10 Content j .j. l
.j i-
-l
, ;- . r. . . . .
9 p. l . l. l; .:. r - l q, - ,1 .. I,-. , .l. ,
. [. .
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. 15 ,
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=M.y E
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.5 s'
; Total! Panel ; .;. . ; / ..uu [
[ -
-E *
~
; . [ ~. . . .
_: : I.
~
.;. / ~ .j_ -
,y Q . b .:-~.'li ..1
[ o W Cladding .(2) - 2 > d e E-* i t .. .
.. u:: .
.s .05 . / i E - . / t --
- . ._ . .__ _Co r_e. $ :I :9I-
' /._
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i !
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-l
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.... / _____.l... . . .. . . _ _.! ,,~ _ ' ;, , , [-
'L_
0 / ;~" i P :d j-r , f t - -i -T" { 0 .01 .02 BIO Content - gm/sq cm
.03 .04 .05 .06 BIO Equiv. Total Thickne s s Including Cladding
_ Content Boron Inches 1 Tol. L mm 1 Tol.
.005 .028 .075 .004 1.91 .10
.010 .056 .075 .004 1.91 .10
.015 .083 .075 .004 1.91 . 10
.020 .111 .075 .004 1.91 . 10
.025 .139 .083 .004 2.11 .30
.030 .167 .096 .005 2.41 .13 r
.035 .194 .108 .006 2. 74 .15
.040 .222 . 121 .006 3.07 .15
.045 .250 .133 .006 3.38 .15 -
. .050 2 78 .146 .007 3.71 .18
.055 .306 .158 .007 4.01 .18
.060 .333 .172 .009 4.37 .23 This tabulation is for Boral with thin cladding for use in high density spent fuel racks.
Boral with thicker cladding is also available for other applications. 7 i E .e.
Report 624 y- BPOOKS &PerKinS : ADVanCeo sTruerures 1 An AAR Compan/ Table 4 '- Boral Panel B content 0.020 gm/sq cm Boron content .111 gm/sq cm Thickne s s-overall .075 .004 inches
.190 . 010 cm Thickne s s-clad .0095 inches (nominal) .024 cm Neutron attenuation 935 (at 0.06 eV)
- Total weight .42 gm/sq cm
. 86 lb /s q ft.
7 C Dispersion Uniformity. The aluminum and boron carbide ingredients , f. E in the central core of the Boral panel are combined in powder form. The methods used to weigh and blend the powders as well as the design and construction of the ingots necessary to produce acceptable Boral panels are patented and proprietary processes of Brooks & Perkins. The manufacturing methods used include a sintering process and hot rolling. The final outcome of the entire manufacturing cycle is Boral panels having l l boron carbide uniformly dispersed throughout the central core. The amount of boron carbide per unit area is directly related to the panel thickness. { 10 The minimum B content per unit area and the uniformity of dispersion - g . within a panelis verified by wet chemical analysis or neutron attenuation te s ting. For details of the verification methods see Brooks & Perkins Quality Assurance Procedures BP-11002-QAP and BP-11004-QAP. l L, . 1 9
Report 624 {' BROOKS &Perxins ,
'{i ADVanCOO STPUCTur8S 1 An AAR Company f The acceptance standards in these procedures are controlled by statistical data to ass'ure the minimum requirements are achieved with 95/95 confidence
~
level. The maximum variation in the manufacturing processes (statistical tolerance interval) over a significantly large sample size has been determined and is utilized in the establishment of acceptance criteria. CORROSION RESISTANCE The useful service life of Boral will exceed 40 years when in contact with the E..- storage pool water of either a boiling-water or pressurized-water reactor.
- This fact is evident through laboratory testing and is further supported by the longest continuous, in-pool, service by Boral over any other thermal neutron -
L shielding material. This excellent corrosion resistance is provided by the protective nature of the aluminum cladding that in an integral facing on the Boral panels. The corrosion of aluminum is negligible in fuel storage pools of either type reactor when the water quality and temperatures are maintained within the normal operating limits as listed in Table 5. The boron content in the Boral will not be reduced below the specified limit during the forty or more years of exposure under those operating conditions.
- E In order to understand the total corrosion resistance of aluminum within the
. normal operat'ing conditions of the storage pools a discussion of that resistance l p must consider all forms of corrosion. A detailed discussion follows for general, i
l- { galvanic, pitting, crevice, intergranular, and stress' forms of corrosion.
- P l{
l ._ _
Report 624 BROOKS &Perxins * [~ ADV80 CODAnStructures vr AAR Company Table 5 - Chemistry of Pool Waters Reactor type PWR BWR Cooling medium
- D-M water D-M water
[' Boron content, ppm 0 to 2000 0 pH range 4. 5 to 6. 0 6. 0 to 7. 5 Temp range, F 80 to 140 80 to 125 C 26 co 60 26 to 52 Conductivity @ 25 C 1 to 30 1 micro mho/cm Chloride ions, ppm, max. 0.15 0.20 E Fluoride ions, ppm, max. 0.10 ---- g . Total solids, ppm, max. 1.00 0.50 {. Heavy metals, ppm, max. ---- O. 10 0.15 t Halogens, ppm, max. ----
- demineralized water General Corrosion. General corrosion is a uniform attack of the metal
- over the entire surfaces exposed to the corrosive media. General corrosion
' L, is measured by weight loss or decrease in thickness and is generally expressed I in mils per year (mpy). The severity of general corrosion of aluminum depends upon the chemJeal nature and temperature of the electrolyte and can range from superficial etching s,nd staining to dissolution of the metal. I L
Report 624 I. t BROOKS ADV80 COO &Perxins Structures,3,< s An AAR Company 1 i Figure 2 shows a potential - pH diagram for aluminum in high purity water
~'
at 25 C (77 F). The potential for aluminum coupled with stainless steel and the limits of pH for BWR and PWR pools are shown on the diagram to be well E within the passivation domain. The passivated surface of aluminum (hydrated . oxide of aluminum) affords protection against corrosion in the domain shown A because the coating is insoluble, non-porous and adherent to the surface of the 7
' aluminum. The protective surface formed on the aluminum (gibbsite and bayerite) 14 known to be stable up to 135 C (275 F) ( and in a pH range of
- 4. 5 to 8. 5 (6) ,
I L 7 Figure 3 is also a potential-pH diagrams for the aluminum-water system but . at 60 C (140 F) which also shows the potential for the aluminum / stainless
. steel couple and the BWR and PWR limits for pH at this upper limit of temper- ,- ature.
i a 7 The ability of aluminum to resist corrosion from the boron ions la evident a from the wide useage of aluminum in the handling of borax and in the manu-facture of boric acid. (7) Aluminum racks with Boral plates in contact with ! ,; the 800 ppm max. boron water showed only small amount of pitting but maintained good structural integrity after seventeen years in the pool ., g .
- E l[
l w
l P- l Rsport 624 1 i BROOKS &PerxinS:s
.: ADVanc8D STfUCTureS An AAR Company se l
7 4 Figure 2 l
- Potential Versus pH Diagram
"- For Aluminum-Water System
, . At 25 C (77 F) (10)
L 0 7 14 2
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... . . _ .... .4, . .. . . . g t._: .-- .t. : -
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-2 0 7 14 pH of Water .
O O ^ 9 4 a 8
RGport 624 rL BROOKS &Perxins a ADVanCeo AnsTruarures AAR Company v Figure 3
. i. .
Potential Versus pH Diagram For Aluminum-Water System At 60 C (140 F) (5) [ 2 0 7 14 2
,I _. ' '
y'___ ' . 2_'. 2. I '
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. : . .:. === : . ._. . .--- ..t= -=t = : ._ ; _.. v -t: =
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. . . . = - Passivity L=_ =_
[ r- - : = .= i -
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=Do_ma:m=._ r .. _.
.. .. .. .. . Domam _ __ ..
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-2 0 7 14 E- pH of Water g
f
pr Rcpart 624 [ BPOOKS &PerxinS: [- ADVanC80 Structures' An AAR Company F- Galvanic Corrosion. Galvanic corrosion is associated with the current of a galvanic ' cell consisting of two dissimilar conductors in an electrolyte. The two dissimilar conductors of most interest in this discussion are aluminum and [ stainless steel while in an electrolyte similiar to the pool water from either a BWR or PWR. There is less galvanic current flow between the aluminum-stainless steel couple than the potential difference would indicate because of the greater than normal resistance at the metal-liquid interface on stainless steel f which is known as polarization. It is because of this polarization character-istic that stainless steel is compatible with aluminum in all but severe marine, or high chloride, environmental conditions. Test data for aluminum coupled with 304 stainless steel in 5.0 pH water at 100 C (212 F) with flow rates ranging from 0. 5 fpm to 81 fps show weight losses of 0.1 to 0.2 mpy and rando.nly spread pits that were not of major consequence. This performance indicates a
. projected service life much greater than forty years.
Pitting Corrosion. Pitting corrosion is the forming of small sharp cavities in a metal surface. The first step in the development of corrosion pits is a local destruction of the protective oxide film. Pitting will not occur on commer-cially pure aluminum when the water is kept sufficiently pure, even when the aluminum is in electrical contact with stainless steel. l Pitting of aluminum has been observed when in contact with stainless steel where the electrolyte can stagnate and the conductivity of the electroylyte i {,'
- L inc rea s es .
Report 624 BPOOKS &Perxins,s {! ADVanCeoAnstrucruresv AAR Company 2 This pitting has not been significant in spent fuel environments and it is not Iikely that pitting of the aluminum would have any influence on the neutron { shielding performance of the Boral. (4) [ Crevice corrosion is the corrosion of a metal that Crevice Corrosion. is caused by the concentration of dissolved salts, metal ions, oxygen or other gases in crevices or pockets remote from the principal fluid stream, with a resultant build up of differential galvanic cells that ultimately cause pitting. E Testing has confirmed that after 2000 hours, under a centro 11ed environment, the Boral and 304 stainless steel combination exhibited little or no corrosion of the aluminum cladding of the Boral. In a separate 2000 hour test at 90 to 180 C the maximum pit depth of corrosion of the Boral surface was reported at less than five mils giving a projected life much greater than forty years. Intergranular Corrosion. Intergranular corrosion is corrosion occurring e preferentially at grain boundaries or closely adjacent regions without appre-ciable attack of the grains or crystals of the metal themselves. Intergranular corrosion does not occur with the commercially pure aluminum (alloy 1100) and other common work hardening alloys.
~
Stress CorrosJon. Stress corrosion cracking is failure of the metal by cracking under the combined action of corrosion and high stresses approaching the yield stress cf the metal. The 1100 alloy used in, Boral is not susceptable to stress corrosion and Boralis seldom if ever subjected to high stresses when used as a neutron shield in a spent fuel rack.
r Report 624 r f :POOKS &Perxins a
; ADVanCeo sTruCruces An AAR Company v
Corrosion Monitoring System. A corrosion monitoring system is a
\
program whereby a series of surveillance samples are placed in the spent fuel radiation and pool water environment and are periodically examined L for physical and chemical changes. It is important the physical configuration of the samples be carefully selected so they are representative to the con-struction and design of the spent fuel racks and are positioned in the pool to be exposed to representative pool conditions and radiation environment. l The physical and chemical characteristics of the samples must be precisely established before insertion into the pool so precise quantative comparisons can be made after each exposure period. The procedure for the manufacture and testing of surveillance samples recommended by Brooks & Perkins is contained in Procedure No. BPS-454. For further discussion on corrosion see Brooks & Perkins Bulletin No. 101. RADIATION RESISTANCE Boral has the ability to shield thermal neutrons from nuclear fuel assemblies without physical change or degradation of any sort from the accompanying exposure to heat and gamma radiation. This ability is attributable to the fact that Boral is a thermal neutron shield that contains no organic nor polymeric type binders which undergo extensive crosslinking and oxidative scission type ,E degradations from both heat and radiation exposure. Boral utilizes an all metallic binder which is stable and unchanged under long-term gamma and neutron irradiation and heat up 540 C (1000 F). L aw ..
Report 624 r BROOKS &PerxinS a 11 ADVanCeo-struCruresv co, Boral, in addition to having the longest history of use in spent fuel storage { application's (since 1965), has been subjected to accelerate irradiation tests which fully support the stability of Boral under these environments. Boral 11 test specimens have been exposed to cumulative doses of 10 rads gamma 9 and 5. 3 x 10 neutrons per sq cm in demineralized and borated water with-out detectable out-gassing attributable to Boral or any decernible physical [ changes. [ The testing referred to was performed at the Phoenix Memorial Laboratory of the University of Michigan using the Ford Nuclear Reactor. The. purpo s e of the test was to determine changes to physical and chemical properties . E of Boral as a result of irradiation under conditions similiar to those en-countered in PWR and BWR spent fuel storage pools. The data recorded during this testing effort is available upon request and includes the following: Total radiation exposure and residual radioactivity Dimensions Weight Specific gravity Hardness Mechanical strength Neutron attenuation Solution boron content, pH, conductivity, and leachable halogens p E-During irradiation, gas evolution rate, total volume of gas evolved, and gas compositloh were determined. The Boral samples were irradiated in air, l demineralized water, and 2000 ppm borated water which simulate both the vented and sealed enclosure of Boralin both PWR an'd BWR spent fuel storage ' (,'
- k. environments.
f_
1 1 Report 624
.E F
O BROOKS &PerxinS a ADVANCED Structures y An AAR Company 1 The test results show conclusively there is no out-gassing from Boral when irradiated in dry air. The same was also true for boron carbide powder in a dry aluminum sample container. This clearly show that Boral is unaffected [' by radiation exposure making Boral a neutron absorber that can be safely exposed while being contained in a sealed enclosure. This characteristic of Boral, no out-gassing from irradiation, also clearly shows that the sour ce of the evolved gases when water was added to the sample containers with Boral has to be from the water itself. There are two mechanism by which water will evolve gases under these circumstances and only one of which requires a radiation environment. The one mechanism requiring a radiation field is the hydrolysis of the water. The disassociation of water into its hydrogen and oxygen elements also requires the presence of free radical scavengers which could well be the E boron carbide powder, impurties within the powder, impurtles in the water, or surface irregularities on the Boral sample. Gaees evolved by hydrolysis i would be a hydrogen-oxygen gas mixture in a 2:1 ratio. . L,. The other mechanism by which water will evolve gases is from the chemical reactions between aluminum and water. The sample containers were made of aluminum with an internal surface area of approximately 9.5 sq. in. The
- m3 surface area of the aluminum cladding on the Boral samples were approxt
- -
[ mately 3. 5 sq. in. The gas released from the water-aluminum reaction is fi hydrogen as shown in the following reaction:
- b
Repart 624 BROOKS &Perxins
- {'- Advanced STRUCTURES An AAR Company
[ A1 +HO2
) A1 (oH) +H 2 A1 + 6 H O + 6 electrons
~
) A1 2 +
3* 2
+
[. 2H + 2 electrons )H 2 (5) The water-aluminum reactions are self-limiting because the surface of the
. aluminum becomes passive by the formation of a protective and impervious coating making further reaction impossible until that coating is removed by 5
mechanical or chemical means. The volumes and types of gases collected from the Boral in demineralized and borated water strongly indicate the gases resulted from one or both of - the two described mechanisms and did not result from cross linking or oxidative scission of any of the Boral materials. In summary Boral does not out-gas or change physically or chemically as a result of exposure to gamma radiation. Water in contact with aluminum will release hydrogen chemically until the aluminum surface is passivated f N and water will disassociate through hydrolysis from gamma radiation. It is therefore necessary to provide a means for venting the hydrogen and oxygen gases if water is' allowed to come in contact with Boral in spent fugl storage applications. l li L
Rsport 624
?-
BPOOKS &Perxins .* ADVanC80 STRUCTURES An AAR Company v I L NEUTRON SHIELDING PERFORMANCE
"* 0 The thermal neutron shielding performance of Boral is obtained from the B Isotopes contained within the boron carbide particles in its core. This F
L: performance is directly related to the amount of boron carbide provided and the spacial relationship between the particles of boron carbide. Figure 4 shows the actual performance of Boral as compared to an ideal (unobtainable) layer of B IO isotope s. The shielding performance is measured as a neutron attenuation factor and is plotted against the surface density of B isotopes in grams per square centimeter. For further discussion on the shielding s properties of Boral see Brooks & Perkins Bulletin No.100. The neutron chielding performance of Boral was unaffected after exposure to 1.03 x 10" 19 rads gamma 'and 5. 3 x 10 thermal neutrons per sq cm. Boron and Halogen Leachability. The boron leachability and the halogen
** 1eachability was evaluated for Boral during irradiation testing conducted at the University of Michigan. The test solutions were analized for boron and
.{' halogen contents before and after radiation exposure when sufficient solution
.E was remaining after the test. The volume of solution was reduced to zero in
- some cases by the radiation. The analysis of the test solutions showed no increase in boron or halogen that cannot be accounted for by the decrease
- E in test solution volume or pickup of the soluble boron on the external edges of the Boral. The boron carbide is allowed to contain, by the ASTM Speci-fication C750-80, up to a maximum of three percent (3.0%) soluble boron In the form of boric oxide (B2 03).
0 -21 De s - o
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- topug uo})tnua))y meag pagenua))y Rcport 624 r BROOKS &Perxins L ADVanCeo sTruCruces s,e An AAR Cm The amount of boron carbide that can come in contact with water is limited to that which is confined to the outer edges of the Boral panel. This wettable amount of boron carbide is of course influenced by the geometrical size and shape of the panel but is less than one percent (1.0%) of the total boron carbide contained therein. In any regard, the total boron content of the panel will remain above the specified minimum content in the event the total soluble boron content were somehow lost through dissolution.
Residual Activity. The re'sidual radioactivity of the Boral was measured following the irradiation testing conducted at the University of Michigan. The activation was limited to trace amounts of impurities contained in the boron , carbide and aluminum materials from which Boral is produced. The specific [ results are avaliable upon request. l- . E E E E E E -23 r, we
Rsport 624 {- BROOKS &Perxins%If ADVanCOO Structures An AAR Company E . Installations Using Boral I. Spent Fuel Storage Racks A. . Pressurized Water Reactors R eactor Utility Water Contact Service Date
- 1. Yankee Rowe Yankee Atomic Electric Co. Yes 1964
- 2. Maine Yankee Maine Yankee Atomic Power Co. No 1977
- 3. Cook 1&2 Indiana & Michigan Electric Co. No 1979
- 4. Sequoyah IL2 Tenn Valley Authority No 1979
- 5. Zion 1&2 Commonwealth Edison Co. Yes 1980
( 6. 7. Salem 1&2 Bellefonte 1&2 Public Service Electric & Gas Co. Tenn Valley Authority No No 1980 1981
- 8. Yellowereek 1&2 Tenn Valley Authority No Indef.
B. Boiling Water Reactors R eactor Utility Water Contact Service Date y 1. LaCros se Dalryland Power Coop. Yes 1976 6 2. Pilgrim 1 Boston Edison Co. No 1978
- 3. Monticello Northern States Power Co. Yes 1978 g 4. Vermont Yankee Vermont, Yankee Nuclear Power No 1978 E 5. Peach Bottom 2&3 Philadelphia Electric Co. No 1978
- 6. Fitzpatrick Power Authority of State NY No 1978
[ 7. Cooper Nebraska Public Power District Yes 1979 h 8. .Duane Arnold 1 Iowa Electric Light & Power Co. No 1979
- 9. Susquehanna 1&2 Pennsylvania Power & Light Co. No 1979
- 10. Perry 1&2 Cleveland Electric Illuminating Co. No 1979
[' 11. Limerick. Philadelphia Electric Co. No 1980
- 12. B rowns Ferry 1,2, &3 Tenn Valley Authority Yes 1980
- 13. Dresden 1,2, &3 Commonwealth Edison Co. Yes 1981
{ Yes k 14. Hatch 1&2 Georgia Power Co. 1981
- 15. Brunswick 1&2 Carolina Power & Light Co. Yes . 1981
- 16. Clinton . . Illinois Power Co. Yes 1981
- 17. Hartsville 1&2 Tenn Valley Authority Yes Indef.
- 18. Phipps Bend 1&2 Tenn Valley Authority Yes Indef.
[ t . _
Report 624 BROOKS &Perxins . [* ADVanceo stroCruces; An AAR Company References
. 2
- 1. Yankee Rowe, Rowe, Mass., Boral Spent Fuel Storage Rack in 800 ppm boron max. water, installed Aug.1964, removed in 1981, small amount of pitting,
[i good structural integrity (2).
- 2. F.M. Kustas, S.O. Bates, B. E. Opitz, A.B. Johnson Jr. , J.M. Perez Jr. ,
R.K. Farnsworth, " Investigation of the Condition of Spent Fuel Pool Components", Battelle-Pacific Northwest Laboratory, PNL-3513/UC-85 Sept.1981 pg 5.
- 3. Brookhaven Medical Research Reactor, Boral in fuel storage area since Jan.1959, ir. demineralized water, no loss of boron carbide s.fter more than 19 years (4).
- 4. C. Czajkowski, J.R. Weeks, and S.R. Protter, " Corrosion of Structural and p Poison Material in Spent Fuel Storage Pools", Paper No.163 presented at
.O Corrosion 81, Apr.1981, Torento, Canada, p 5. D.D. Mac Donald and P. Butler, "The Thermo-dynamics of the Aluminum -Water ' L' System at Elevated Temperatures" Corrosion Science 1973 Vol.13 pgs.264,265 &
266. ,
- 6. K.R. Van Horn, " Aluminum", American Society for Metals, 1967 Vol. I pgs.211, 220 & 221.
- 7. T. Lyman, " Metals Handbook" 8th Edition,1961 Vol.1 pg.930.
.p 8. J. L. English and J.C. Griess, " Dynamic Corrosion Studies for High Flux Isotope L Reactor" ORNL-TM-1030 Sept.1966 pgs.1,2,3,4,23,26,27 & 31. . p' 9. K.Videm, " Pitting Corrosion of Aluminum in Contact with Stainless Steel" L Institute for Atomenergi, Kjeller Research Establishment, Lillestrom, Norway.
- 10. E. Deltombe, C. -Vanleugenhaghe and M. Pourbaix, " Aluminum", pg.172 E
11 . E ! -25 n .
i DSER Open Item No. 143 (DSER Section 9.1.5) OVERHEAD-HEAVY LOAD HANDLING We cannot conclude that the overhead heavy load handling sys-tems are in . compliance with the Phase I and Phase II criteria contained -in -NUREG-0612 until the applicant provides an accept- ,
- able response : to the guidelines. The overhead -heavy load l handling systema do not meet the acceptance criteria of SRP ;
Se ct ion 9.1. 5. We will report -resolution of this item in a l supplement to.this SER. Guideline 1 - Safe Load Paths [NUREG-0612, Article 5.1.1(1)] ( Re ference: DSE R, Ap pe n dix . A , Se ction 2.3.1) Recommendation a. Submit drawings to show the locations of the cranes in relation to the safe load paths. t
' RESPONSE
- Table 9.1-10' provides the figure' number of the plant equipment .
- location drawing and the area on that drawing, defined by column lines, that shows the floor location below each HCGS heavy load handling . crane ,or hoist. For example, the equipment location
. drawing for the personnel airlock hoist (Table 9.1-10, item 2)
- is . Fig ure l . 2-28, and the floor location that envelopes the areaR, below the airlock hoist monorail is bounded by column lines P, 20R, and 23R. To show the relative elevations between the loads and the floor below the loads, Section 9.1.5.2 has been revised to provide the . elevation of the rail (s) and of the book in its raised position for each non-exempt crane listed in Table 2.1 of HCGS DSER, Appendix A, except the Reactor Water Cleanup Filter /
Domineralizer Holst, RCIC. Pump and Turbine Hoist, Turbine Building Bridge ' Crane , Turbine Generator Auxiliary Crane, and the Dominer-alizer Removal Hoist which are discussed below. .The location of a-non-exempt crane in relation to its safe load path (s) can be determined by referring to the appropriate equipment location drawing and floor area as provided in Table 9.1-10 together with the rail and hook elevations provided in revised Section 9.1.5.2. LThe Reactor Water Cleanup F/D Hoist has been reclassified from non-exempt to exempt status because there is no- safe shutdown or decay l heat removal equipment beneath the load path. - This interpretation is based on-the guidance provided in Item 2.1.1
= of Enclosure ' 3 to the NRC letter to PSE&G of December 22, 1980 (Reference?3 of HCGS DSER, Appendix A. ) The design uof this heavy 143-1
j 5' y-. ~ - - - - - - . - - ~ - - - - - - - --
/}
.j~,
n , w 'rf. f
, ur
- g. ' DSER Open Item No.-143.(Cont'd)
/
.J - w'
- !1oad handling ' system has been- changed -to employ - two hoists b
'insteed of ' 6ne, a shorter monorail, new load paths, and a
-- removable stop on the monorail. ,' These changes provide the
' basis for the. reclassification. TheThe two new hoists, lAH220 shortened monorail
-and 1BH220r replace hoist 10H213.
' ' extends 8;ft-10 in, beyond the? south wall of the F/D cells (orfil'in.
north of column 15R). Figure 1.2-31 has been
. revised! to show'this. 'The new load paths do not extend beyond P the i south wall of .the F/D Cells because the revised plan is to : stackithe four shield. blocks for a given cell onto the roof of the adjacent' cell at elevation 178.f t-6 in. , instead
-w of lowering : them ;to t.he 1 elevation 162 f t level as originally planned. : The orientation and size of the shield blocks . s i
such that they cannot freely pass through the available Tho open- , ing -from elevation -178 f t-6 in. to elevation 162 ft. ' available opening is a rectangle with a north-south dimension of:6:ft-10 in. (constrained by the length of the shortened
-monorail)land a east-west dimension of 7 ft-9:in. (constrained j !by the' dryer-separatoripool wall on the west and The a floor two uppee fram-
- ing . beam a t elevation 178_ f t6 in. on the east) .
blocks must t>e lif ted simultaneouslynorth-south by the - two (compared hoists. Togethe with r, they = form a ^ rectangle with a 11 f t. 8 ft-10 in. available opening) and a 10 f t-9 in. east-west The (compa' red with 7 f t-9 in, available opening)Each dimension. is a rectangle
,two : lower blocks are lif ted . individually.
with a 4 ft-3 in. north-south (compared with 8 f t-10 in.) and a j a f t.' 9.in. east-West
- (compared with 7 'f t-9 in.) dimension.
The removable stop will prevent a shield block from being carried
'over the opening. - The load handling procedure for tho' blockc ~
.." .will require the operator to verify that the stop is bolted in t i
place before beginning' the lif t. The FSAR has been revised E ' -(9.1.5.2.2.c, 9.1;5.3.3.c, Table 9.1-10,~ Table 9.1-12, and g 7 'Figuret 9.1-38) , to describe ,the new design. I ' ' Of -The RCIC Pump'snd Turbine Hoist has been reclassified from 5 y' non-exempt to exempt status because' it does not handle heavy
!1oads during maintenance. The RCIC' turbine case weight of' l '
i. 2.3 5 '. tons, originally -given in Table 9.1-12, was the total weight ~of"the RCIC to turbine, including baseplate and stop valve. > The maximum weight be lif ted is the casing, which we ighs - l 785_ pounds. Af ter tho' upper half of the turbine ' case (785 pounds), .the next ' heaviest RCIC turbine ~ maintenance loads (400 pounds) and the rotor (325 pounds). [ ' - areithe stop valve i' Se ction s ' 9.1. 5. 2. 2. 3 fa nd 9.1. 5. 3. 3. e , Tab le s 9.1-10 a nd 9.1-12, Y .and Figure 9.1-34 have been revised to describe the exempt ', . status of this hoist.
/
r' f: , ( I' , , [ 143-2 1
- e.-., ru ew h , >. .. raw-ve ww,- ,+.m e,mnarv a-m .. .. . . .y wmeye----rmwem-sw w-wr.e
DSER Open It^n No. 143 (CCnt'd) Die Turbine Building Bridge Crane, Turbine Generator Auxi-liary Crane, and Domineralizer Removal ' Hoist have also been reclassified from non-exempt to exempt status because- there is no. safe shutdown or decay heat removal equipment b'eneath z, and their load paths as originally stated in 9.1.5.3.3.k, aa respectively. Tables 9.1-10 and 9.1-12, and Figure 9.1-39 have been revised to be consistent with the FSAR text for these hoists. Note also that the outboard MSIV Hoist, 10H214, has been in-corporated into the design of the new Main Steam Tunnel (MST) underhung crane. The FSAR has been revised (9.1.5.2.2.f, ' to 9.1. 5.3.3. f , Tab le 9.1-10, Tablo 9.1-12 a nd Fig ure 9.1 .1 describe the new design. . Rail elevations, but not raised hook elevations, have been added to Section 9.1.5.2.2 for the CRD Service Hoist (9.1.5. 2.2.h), SACS Pumps Hoist (9.1.5.2.2.11), and SACS Heat Ex-changer Hoist (9.1.5.2.2.mm) because the hook elevations are not known. The raised hook elevations are not known because instead of dedicated hoists for each service, the hoists will be transferred from other locations as required. The heck heights are expected to be approximately two (CRD) , six (SACS Pumps), and four (SACS heat exchangers) feet below the corresponding rail elevations. Recommendation b. Check the loads carried by the rigging beam hoists for servicing CRD, SACS pump, and SACS heat exchanger to determine the exempt or non-exempt status of these load handling
' systems, and list.the non-exempt systems in Table 2.2.
RESPONSE
The loads for the CRD service hoist are identified in Section 9.1.5.2.2.h as a control red (450 pounds), neutron monitoring cask ( <1150 pounds), and unspecified CRD maintenance equipment (up to 2000 pounds). Control rods and the neutron monitoring cask are not heavy loads for HCGS. A maximum weight of 2000 pounds is chosen for the unspecified CRD maintenance equipment because that will be the capacity of the hoist that will be used on the CRD service monorail. Because a dedicated hoist will be transferred was not purchased for this service, a hoist None of the three from another location when one is needed. d 143-3 ' 8 W 4Pd
DSER Open Item No. 143 (cont'd) exclusion criteria of Table 9.1-10 apply to this hoist. There-fore,-it would tm ciassified as non-exempt if it handled a maintenance load equal to or greater than 1200 pounds. This hoist has been added to FSAR Tables 9.1-10 and 9.1-12, instead of to Table 2.1 (not Table 2.2 as typed on page 22) of DSER Appendix A, because the Table 2.1 format is NRC controlled, and is a summary of information first provided in Table 9.1-10. The loads for the SACS pumps hoists are identified in Sec-
~
tion 9.1.5.3.3.11 as the SACS pump motors (1155 pounds). The revised motor we ight is 6160 pounds. 'Section 9.1.5.3.3.1 has been revised to incorporate the new weight. A dedicated hoist was not purchased for this service because the heaviest anticipated SACS pump maintenance _ load is the upper half of the pump ' casing' ( 82 5 pounds) . There fore , this hoist does not i routinely handle heavy loads. Instead, a hoist will be l borrowed. from another location when needed. None of the three exclusion criteria of Table 9.1-10 apply to this hoist. There-fore , it would be c'lassified as non-exempt if it lif ts a heavy load. This hoist has been added to Tables 9.1-10 and 9.1-12., . The load for the SACS heat exchanger hoists is identified in
.9.1.5.3.3.mm as a SACS heat exchanger return.end cover. The i
return end cover weigns 18,400 pounds. Dedicated hoists were not
. purchased for this service. Instead, hoists will be borrowed from another location when needed. None of the three exclusion Therefore, .' criteria of Table 9.1-10 apply to these hoists.
j they.are classified as non-exempt. These hoists have been
- added to" Tables 9.1-10'and 9.1-12.
Recommendation c. Provide the missing information concerning the safe load paths for the non-exempt cranes.
RESPONSE
The missing information has been added to Section 9.1.5. The load paths for the CRD service hoist, the SACS pumps hoist, and'the SACS heat exchanger hoists -have been added to Figure 9.1-35. Because the turbine building bridge crane and the turbine- !' generator auxiliary crane have been reclassified as exempt, their safe load
-(see response to Recommendation b above)
Teth drawings have not been added to Section 9.1.5 as re-Because l quested in Item B.2 on Page -9 of DSER, Appendix A. ~ the RCIC Pump and , Turbine ' Hoist has been reclassified as exempt, it's load path has been deleted from Figure 9.1-34. Because the Reactor Water Cleanup Filter-Demineralizer Hoist and the Domineralizer Removal' Hoist have been reclassified as exempt, their -safe load ' path drawings (Figures 9.1-38 l-l and 9.1-39) have'been J*leted. I ' 143-4 l l
- -. _ _ - - . _ _ _ . . . ~ . _ _ _ . . _ ~ ~ , . _ , . . - . . . . . _ _ _ . - . _ _ , - . _ . , _ _ _ _ . _ .
~
DSER Open Item No.-143 (Cont'd)
.The load paths for the seven loads listed in Itam B.3.a of '
DSER, Appendix A have been added to Figure 9.1-32. The 4-ft by _4-f t. hatch and the lO-f t by lO-f t hatch paths were added to Sheet 1 of Figure ' 9.1-32; the spent fuel gates path and the flux monitor shipping crate paths were added to Sheet 2 7 Sheet. 6.was added for the refueling bellows guard ring path; ,
' Sheet 7 was. added for the channel handling boom crane path; ,
and Sheet 8 was added for the RPV head stud rack path. < The load path _ for the new f uel vault covers has been deleted from Fig ure . 9.1-3 2, Sheet 1, because the covers are not heavy loads. Ea ch of the three sections is made of 0.25 in thick steel plate and weighs' less than 900 pounds.
- Safe load paths for three additional polar crane heavy loads that are listed in Table 9.1-12, but not mentioned in Item B.3.a of DSER, Appendix . A, have been added to Figure 9.1-32. The main hoist load block, auxillary hoist load block, and spent fuel cask yoke are shown on new ' Sheets 9, 10, and 11, respe ctively.
The dryer-separator sling has been added to Table 9.1-12, and .
.the; load path is shown on new Sheet 12. The spent fuel rack -
modules and the fuel rack lifting fixture have been added to
' Table 9.1-12, and the load paths are shown on revised Sheet 4 of Fig ure 9.1-3 2. The. reactor well shield plug sling and the dryer-separa tor pool plug grapple have been added to Table 9.1-12,
- and the load. paths are shown on.new Sheet 9 of Figure 9.1-32.
The justification in Section~ 9.1.5.6 for the HCGS deviation ' from Acceptance criterion 2 of SRP Section 9.1.5, has been revised to clarify the HCGS position that load paths will not be ~ painted on the floor. Instead, the alternative method of using a signalman and temporary load paths as suggested in
. Item C.4 on Page 11 of DSER, Appendix A, will be used.
i l b l [. f-i. r s - 143-5 ; u . -u.._ i.,._.._____._.._.._.__._____._..___.._...___.._.
DSER Open Item No. 143 (Cont'd) Guideline 2 - Load Handling Procedures (NUREG-0612, Article 5.1. .. ( 2 ) ] (
Reference:
DSER, Appendix A, Se ction 2. 3. 2) Recomme nda tion: Provide evidence to show that the load handling procedures have been developed.
RESPONSE
As agreed in the May 23, 1984 conference call between the appli-cant and the NRC, the load handling procedures described in FSAR Design Basis Section 9.1.5.1.1 will be developed before fuelto load. HCGS conside rs Se ction 9.1. 5.1.1 to be a commitment comply with Article 5.1.l(2) of NUREG-0612. e d 143-6
DSER Open Iton No. 143 (Cent'd) Guideline 3 - Crane Operator Training [NUREG-0612, Article 5.1. l( 3 ) J (
Reference:
DSER, Appendix A, Se ction 2.3.3) Re comme nda tion : Provide information on the status or plan for crane operator training in accordance with Chapter 2-3 of ANSI B30.2-1976.
RESPONSE
As discussed in the May 23, 1984 conference call between the applicant and the NRC, crane operators will be trained, quali-fled, and conduct themselves in accordance with Chapter 2-3 of ANSI B30.2-1976, as stated in FSAR Design Basis Section 9.1.5.1.j. HCGS considers Section 9.1.5.1.j to be a commitment to comply with Article 5.1.l(3) of NUREG-0612. e e o e l 143-7 l
DSER Open Item No. 143 (Cont'd) RESPONSE i Guideline 4 - Special Lifting Devices (NUREG-0612, Article 5.1.1 (4)? (
Reference:
DSER, Appendix A, Section 2.3.4) Rec omme n da tion a . Recalculate the stress design factors based on the combined maximum dynamic and static load. These factors must be equal or greater than those specified in ANSI N14.6-1976.
RESPONSE
The recalculated stress design factors based on the combined maximum dynamic and static loads are provided in revised Table 9.1-14. The title of Table 9.1-14 has bgen changed from " Hope Creek Polar Crane Special Lif ting Devices and Slings" to " Hope Creek - Special Lif ting Device Factors of Safety". The fuel pool slot plug ' sling (Item 7) has been deleted from the table to reflect design evolution. The slot plugs will be lifted with conven-pional slings instead of with a dedicated lif ting device. Two new special lif ting devices (SLD), the personnel air lock strongback (Item 9) and the fuel rack lifting fixture (Item 10) Item 9 was added when the scope have been added to the table. of Table 9.1-14 was expanded to incl'.de all HCGS SLD instead of just the polar crane SLD. Item 10 is supplied with the s pent fuel racks. The HCGS plan is to install partial spent fuel storage capacity during construction and additional racks ! as needed af ter plant operation.begins. Item 10 would be used to install these spent fuel rack modules. Se ctions 9.1.5.3.2 and 9.1.5.3.3.a have been revised to be L consistent with the recalculated factors of safety provided in revised Table 9.1-14. The reference factors of safe ty are obtained from ANSI N14.6-1978, instead of N14.6-1976 as stated in Item 4.c on Page 25 of DSER, Appendix A, because L Paragraph -5.1.l( 4) of NUREG-0612 invoke s the - 19 78 ve rsion. l l l l 143-8
--gis' .e-i.. en v y-m-g- ---.y9 y-.ec wa* --T * -r -ygi . --g- -wpg--<,n ---M.-e- -
i.,e y y g-ing-- ce e-e--r'ey i.y-*m
DSER Open Item No. 143 (Cont'd) I
~ The missing safety evaluation information for the RPV stud.
tensioner sling , noted in Item B on Page 14 of DSER, Appendix A, is provided in revised Section 9.1.5.3.2. As stated in Section 9.1.5.3.2, a single-failure proof Be con-ventional sling is used- to lift the fuel pool gates. cause it is a conventional sling and not a special lifting device, it does not appear in Table 9.1-14. Recommenda tion b. Address the requirements of ANSI N14.6-1976 in addition to the requirements for stress de sign factors.
RESPONSE
As shown in revised Table 9.1-14, the stress design factors for all but two of the Hope Creek special lif ting devices meet or exceed the values of 3 versus yield strength and 5 versus ultimate strength required by ANSI N14.6-1978. The
- design' of the two Hope Creek lifting devices (dryer-separator.
sling and RPV service plat form Sling) that do not meet the , safety factor criteria is the same as the design of the corresponding Washington Nuclear Plant No. 2 and Limerick Generating Station special lifting devices. Therefore, as discussed in the August 24, 1984, telephone call be tween the
' applicant;and the NRC, no additional information regarding compliance with ANSI N14.6-1978 is provided.
l i- [ L 143-9 i
Y DSER Open Item No.-143 (Cont'd) l Guideline 5 - Lifting Devices (not specially designed) [NUREG-0612, Article 5.1.l(5)? ~ (
Reference:
DSER, Appendix A, Se ction 2.3.5) Recomme nda tion : Provide information concerning the installation and use of the lifting devices as required. 9
RESPONSE
~ The heavy load handling system Design Bases have been revised to provide the requested information in FSAR, Se ction 9.1. 5.1. n.
HCGS considers Section 9.1.5.1.n to be a commitment to comply with the provisions of Article 5.1.l(5) of N UR EG-0612. As discussed in the September 6, 1984 telecon between the applicant and the NRC, a dynamic load will not be added to the static load when a sling is selected for use with a hoist *that has a maximum hoisting speed equal to or less than 30 ft./ min. , e e l 143-10
+- e* p , .- +-e-- g---y- -y,- ---m.-- p - g - -- y--wwe' Tw--et-y- - -T w--'W-F-Tw m-' w' y*m7
4 DSER Open'Ites No. 143 (Cont'd) Guideline'6 - Crane Inspection , Testing and Maintenance
' [NUREG-0612, Article 5.1.1(6)]
( Re ference: DSER, Appendix A, Se ction 2.3.6) Recommendation: Provide information for inspecting, testing, and maintaining the cranes including those that are not listed in FSAR, Table 9.1-10, but may carry heavy loads over safety related equipment.
RESPONSE
All cranes and hoists at HCGS that may carry heavy loads over sa fety related equipment, including the CRD service hoist, SACS pumps hoist and SACS heat exchanger hoists which were not The originally listed, are now listed in revised Table 9.1-10.
- procedure for inspecting , testin,g , and maintaining each of these non-exempt cranes (those'not identified by exclusion criteria A, 8, or C) will comply with Chapter 2-2 of ANSI B30.2-1976.
co ns ide rs FS AR Se ction 9.1. 5.1. k. , together with revised HCGS Sections 9.1.5.4.1.3 and 9.1.5.4.2.3,. to be a commitment to c omply with. Article 5.1.l( 6) of N UR EG-0 612. l t l l 143-11 E
DSER Open Itam No. 143 (Cont'd) Guideline 7 - Crane Design [NUREG-0612, Article 5.1.l( 7) i (
Reference:
DSER, Appendix A, .Section 2.3.7) Recomme n datio n: . Provide information explicitly directed at c om- f plying with the guidelines of ANSI B30.2-1976, ; Chapte r 2-1 and of CMAA-70.
RESPONSE
As agreed in the May 23, 1984 conference call between the appli-
-cant and the NRC, the requested information is already provided in Tab le 9.1-10. The Design . St3ndard column identifies the standards tha t were applied for the design of each crane and hoist in revised Table 9.1-10.
- The HMI 100 design standard was added for 7tems 3,18,19,31 and L3'7, and Note 4 (use of ANSI B30.17) was added for Items 31 and 32 to correct their inadvertent omission from the original sub-mittal. The HM! 100 standard was deleted for Item 4 (reactor '
water cleanup F/D hoist) becausa the motorized noists were replaced with manual hoists. The ANSI ~ B30.2 and B30.17 design ' standards were added to Item 7 because the outboard MSIV hoist was incorporated in the expanded main steam tunnel underhung
. crane design.
I l-i I 1. 143-12 l l
-y -e --p--- ,- --g ---y- -m-g -,--y a. 7yp, yg g y w ---ew3 w. .,. ,wwp v-- wi--w-y-y-e,,-yee-op -we w-g--- -ww'q-=wve9 yy--w
DSER Op7n Itan No. 143 (Cont'd) Section 3'.3 ' Interim Protection [NUREG-0612, Article 5.3]
-(
Reference:
DSER, Appendix A,-Sections 2.4.2 and 2.4.3) Rec ommendatio n: The Interim Protection Measures 2 through 6 should be initiated to provide safe operation of the cranes before implementation of the guidelines of NUREG-0612,~ Article 5.1 is i c om ple t ed . l
RESPONSE
Interim Protection Measures 2 through 6 are not applicable to HCGS because the plant is not operational . HCGS's intent is
-that implementation of the guidelines of NUREG-0612, Article 5.1 will be completed before the plant is operational.
HCGS's detailed position with respect to Measures 2 through 6 is clarified below. Interim Protection Measure 2 (load paths) . Safe -load paths have been defined per the guidelines of Sec-tion 5.1.l(1) of NUREG-0612, as described above in the response to DSER, Appendix A, Section 2.3.1, and in the associated re-Therefore, this measure will
. vision of FSAR Section 9.1.5.6.
not be required when the plant becomes operational. l Interim Protection Measure 3 (procedures) Procedures will be developed and implemented per the guidelines i of Section 5.1.l(2) of NUREG-0612 be fore fuel load, as described ab ove in the response to DSER, Appendix A, Se ction 2.3.2. There- [ ! fore, this measure will not be required when the plant becomes ope ra tion al . Interim Protection Measure 4 (operators) Operators will be trained, qualified, and will conduct them-l L selves per the guidelines of Section 5.1.l(3) of NUREG-0612, as described above >in the response to DSER, Appendix A, L Section'2.3.3. Therefore, this measure will not be required when the plant becomes operational. Interim Protection Measure 5 (crane maintenance) Cranes will be ins pe cted , tested, and maintained in accordance with the guidelines of Section 5.1.1(6) of NUREG-0612, as des-cribed above in the response to DSER, Ap pendix A, Section 2.3.6, and in the associated revisions to FSAR Sections 9.1.5.4.1.3 and 9.1.5.4.2.3. - Therefore, this measura will not- be required when the plant becomes operational. 143-13 u
._. _. ._._ __. . _ . . . _ . _ . _ - _. , ~ _ _ . _ _ _ _ . . _ . _ . _ . . . . . _ . , . _ _ . . . . . . _ . . . -
DSER Open Item No. 143 (Cont'd)
~
Interim Protection Measure 6 (heavy loads over core) As described above in the response to DSER, Appendix A, Sec-tion 2.3.2, specific load handling procedures for each load ' will be developed before fuel load. During procedure develop-ment, ' rigging or lif ting device installation and load _ movement rege'raments will be reviewed to assure that sufficient detail and clear, concise instructions are provided. As stated in-revised FSAR Section 9.1.5.6, the polar crane signalman will
- review the specific load handling procedure before each lift.
This will supplement the original procedure development review for detail, clarity, and conciseness. As stated above in the response to DSER, , Appendix A, Sec-
-tion -2.3.6, cranes will be inspected in accordance with Chapter 2-2.of. ANSI B30.2-1976. Chapter _2-2 requires periodic As visual stated inspection of the crane load bearing components.
above in the response to DSER, Appendix A, Section 2.3.4, Re-commendation b , the s pecial lif ting devices _ comply with AN SI N14.6-1978 as described in Table 9.1-19. That compliance in-cludes periodic visual inspection of the load bearing welds add '
-critical areas in accordance with Paragraphs 5.3.l(2), 5.3.6, As stated above in the response and.5.3.'7 of ANSI N14.6-1978. Section 2.3.5, slings will be used in
-to DSER, Appendix A, accordance with the guidelines of ANSI B30.9-1971. Such use-
. includes periodic visual inspection of slings in accordance with Sections 9.1-8 (steel chain), 9. 2- 8 (wire rope ) , 9.3-7 (metal mesh) 9.4-6 (fiber rope), and 9.5-6 (synthetic webbing)
-to identify flaws or deficiencies that could lead to failure.
As stated above in the response to DSER, Appendix A, Sec-
' tion 2.3.6, cranes will be maintained in accordance with Chapter 2-2 of ANSI B30.2-1976. Chapter 2-2. As requires repair stated above and replacement of defective components. Section 2.3.4, Recom-in the response to DSER, Appendix A, j
mendation b, the special lif ting devices comply with ANSI i That compliance N14.6-1978 as described in Table _9.1-19. includes repair and replacement of defective components in accordance with Paragraphs 5.4.1 and 5.4.2 of ANSI' N14.6-1978. As stated above in1the' response to DSER, Ap pe ndix A, Se c-l- tion 2.3.5, slings will be used in accordance with ANSI B3 0.9 -19 71. - Such -use includes repair and replacement of defective components in accordance with Section 9.1-6 l (steel chain), 9.2-8(wire rope ) , .9. 3-7 (metal mesh), 9.4-6 l
-(fiber rope), 9.5-6 and 9.5-7 (synthetic webbing).
143-14 i _ _ _ . . . _ . . . - _ - . - - _ . _ _ - _ - . . _ _ . . _ . _ , _ . _ ~ _ _.-- - -_ _ _.._ ._ ,.-.-- _ ,. _ b
)('
4 DSER Open Item No. 143 (Cont'd) As stated above in the response to DSER, Appendix A, Se ction 2.3.3, crane operators will be trained, qualified, and will conduct ~ themselves in accordance with Chapter 2-3 of ANSI B30.2-1976. The - examination required of each operator for compliance with Chapter 2-3 will include a requirement to demonstrate familiarity with the hand signals specified in Fig'ure ~ 5 of B30.2. The specific load handling procedure for each heavy load handled over the core will require tnat the . crane ' operator certify his f amiliarity . with the content of that procedure by signing and dating the procedure before the lift. Therefore, Interim Protection Measure 6 will not be required when the plant becomes operational. p I-l l' I i l l ! F71(5) 143-15
- 1. O PsAR E M M"
- 3. NO.
- d. DISCIPLINE hikt) cd
- 4. ORIGINATOR D, Q, Qi, 5.DATETf7,e,fgg .
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- 6. REFERENCED SECTIONS OF SAR q.l.5 Nhoad ibeg Lcod hndig b y%m) i OE NhClM GTfS bC. dnh.l.b QS hd2 3 neu_ssexp a c.ddah sw tk. wwnse. 4,at s.c. cemmeA in pseg. Appenex A. Aho, irmepck&. se9ttra eAtod c.%es.
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fj 1-'.I'3 3% on!5 +39 I I' 2O f 8. REFERENCED SPECIFICATIONS OR DRAWINGS l %nt I l l
- 9. JUSTIFICATION %BELOw O ArrACHED .
($Xtd Of C.y I.E.,. 4h CN]tS TESQolti k DSER. Ib l@.at -
- 10. DISTRIBUTION / INTERFACING DISCIPLINE REVIEW ORIGINATING GROUP SUPERVISOR INTERFACING DISCIPLINES:
(POAE FOR CHAPT.17) .
- 15. CONCURRENCE DATE SAR COORDINATOR S.3 Ibh (NSSS)
PROJECT ENGINEER dLS CHIEF NUCLEAR ENGINEER (INFO ONLY)
- 13. APPROVED SY DATE 14. CONCURRENCE DATE h 11. APPROVED BY DATE 12. AH60VED SY DATE NA (PROJECT ENG.) (CLIENT)
(GROUP SUPV) SAR g, 7,
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.-__. _ , _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ _ , . . _ . _ - _ . . _ _ . __,__ _-~.
i, I i l HCGS FSAR l
~
9.1.5 OVERHEAD HEAVY LOAD HANDLING SYSTEMS 9.1.5.1 Desian Bases
- a. The overhead heavy load handling systems (OHLHS)'are designed to move heavy loads from one location to another within the various plant structures.
- b. The OHLHS are designed to safely handle all plant heavy loads that range in weight from a maximum of 150 tons in the reactor building and 220 tons in the turbine building to a minimum of 1200 pounds.
- c. The reactor building polar crane main hoist and auxiliary hoist are designed to be single-failure proof
- in conformance with NUREG-0554 and NUREG-0612. .
- d. The OHLHS in the reactor building are designed so that releases of radioactive material that could result from damage to spent fuel, due to a postulated heavy load drop, will produce doses that are within 10 CFR 100 limits.
l-The OHLHS in the reactor building are designed so that e.- damage to fuel and fuel storage racks due to a postulated heavy load drop will not result in a fuel configuration that causes K ,fg to exceed 0.95.
- f. The OHLHS in the reactor building are designed so that damage to the reactor vessel or spent fuel pool, resulting from a postulated heavy load drop, will not cause water loss that could uncover spent fuel.
- g. The OHLHS are designed so that damage to equipment resulting from a heavy load drop will not prevent safe
~ '
. shutdown of the reactor.
- h. The OHLHS are designed to minimize the potential for heavy load drops on spent fuel or safe shutdown equipment by carrying their loads over safe load paths to the extent practical. They are defined in written
.9.1-69 L-~-,-. - . _ . - . - - - . - . - - . - . . . . . - . _ . - _ . -
HCGS FSAR'- 1 load handling procedures, and are shown on safe load i path drawings.
- i. .The reactor building polar crane and other OHLHS that handle loads over safe shutdown equipment are operated in compliance with written procedures that include identification of the required equipment, inspections and acceptance criteria required before load movement, sequence of steps to be followed for load movement, definition of safe load path, and any special precautions, for each known load.
< j. The OHLHS' cranes are operated by operators who are trained, qualified, and conduct themselves in compliance with Chapter 2-3 of ANSI B30.2-1976.
- k. The OHLHS cranes are inspected, tested, and maintained .
-in compliance with Chapter 2-2 of ANSI B30.2-1976. ',
l ! 1. The OHLHS crane designs include electrical interlocks and/or mechanical stops to restrict crane travel to , those areas that are necessary.
- m. The OHLHS-cranes are designed to meet the applicable criteria of CHAA-70, and Chapter 2-1 of
. ANSI B30.2-1976.
f, l$3a 'A* k\ n. fElng devices are designed to meet the applicable
\Mff riteria of ANSI B30.9-1971.
- o. Special lifting devices are designed to meet the applicable criteria of ANSI N14.6-1978.
l' 9.1.5.2 System Description The cranes and lifting devices that comprise the OHLHS are
' described in the following sections. Table 9.1-10 includes a summary of the design data, seismic category, and code or standard used for design and manufacture of each OHLHS crane. It also includes monorails ;f liftin; 50 -- for which no dedicated hoists exist but which are used occasionally for equipment I
9.1-70 e +- --w-r---, , , - -w.,- .---..-.,-.~,.,-%.*we-w,,w--,mw.w.--ew.,--.e,+w-ww---n.--,,---.. .-.---*wwe--rww.~4,.--- ~
l (Ref. 9.1.5.1.n) INSERT 1 l 1 Lifting devices that are not specially designed are installed and used in accordance with the guidelines of ANSI B30.9-1971. For all hoists that have a maximum hoisting speed equal to or less than 30 ft./ min., the load used to select the proper sling is the static load. For hoists that have a maximum hoisting speed greater than 30 ft./ min., the load used is the
- sum of the static and maximum dynamic loads. The maximum
- dynamic load is determined by multiplying the maximum hoisting speed by of 1% of the static load.
- The rating identified on the sling is for the static load
- that corresponds to the maximum total static plus dynamic load. If a-sling is restricted to use with only certain hoists, the identification of the acceptable hoists is clearly marked on the sling.
6
* - w , , ,e ~ - , - - , - - - - , - ,, - - - - - , , - - - - ,,,
HCGS FSAR s 9.1.5.2.1.'1 Structural Components I All the structural components and, machinery of the reactor building polar crane are designed for a full capacity of 150 tons, with a minimum safety factor of 10 against ultimate l failure.for the load-carrying parts, including hoist ropes, and the machinery. The calculated stresses of all load-carrying parts are in accordance with the requirements of Crane. Manufacturer's Association of America (CMAA) Specification 70. l Structural design of the crane complies with the following seismic loading combinations and criteria: 4 8
- a. Dead load plus live load plus operating basis earthquake (OBE) resultant stresses are less than the normal AISC code allowable stresses. -
b.- Dead load plus live load plus safe shutdown earthquake i (SSE). resultant stresses are less than 1.5 times the normal AISC code allowable stresses, less than 0.9 yield in bending, 0.85 in axial tension or compression,
- and 0.5 yield in shear. )
l i e l l The earthquake motion considered consists of two horizontal and one vertical component. The total structural response is predicted by combining the applicable maximum codirectional responses, calculated from the three.(two horizontal and one vertical) alyses, using the square root of the sum of the squares-( S) method. x i l The structural members.of the reactor building polar crane are ! . designed for a fatigue loading of.100,000 to 500,000 cycles, with each completed lift representing one cycle. The rotating machinery is designed for a fatigue life expectancy of 2,000,000 cycles, with each rotating component cycle represented by one revolution. Any load below 50% of the crane rated capacity does not reduce the_ life expectancy of the crane. 9.1.5.2.1.2- Mechanical Components The. crane is of a double-trolley, indoor, electric overhead,' bridge crane design. The main trolley layout is shown on L i 9.1-72
-. _- - .. = .
HCGS FSAR maintenance. Table 9.1-11 includes a more detailed listing of the design parameters for just the reactor building polar crane. Cranes are included in the OHLHS if their capacity is greater than 1200 pounds. This is the designated weight of a heavy load for Hope Creek Generating Station (HCGS). It is defined as the weight of one spent fuel assembly and its handling tool. For HCGS, the 1200-pound value consists of assumed weights for a fuel assembly (650 pounds), a fuel assembly channel (100 pounds),.and the refueling platform grapple (450 pounds). Section 9.1.4 includes a description of those aspects of new fuel receipt,and storage, reactor refueling operations, and spent fuel shipment that involve the reactor building polar crane. 9.1.5.2.1 Reactor Building Polar Crane , dc,W b 'A@ bY. ' The reactor uilding polar crane is a bridge crane mounted on a ' circular rail that is supported by the reactor building superstructure. The bridge consists of two welded box girders with full depth diaphragms. The bridge girders are held together by structural end tie girders. Two dual-wheeled trucks that travel on top of the runway rail support each of the two end tie girders and drive the bridge. The crane is shown on ' Figure 9.1-21. i' Two electric-motor-driven trolleys, one for the main hoist and one for-the auxiliary hoist, provide the structural frame support for the polar crane hoisting machinery. The trolleys travel on a single set of rails secured to the tops of the two bridge girders. The main hoist design capacity is 150 tons, and the auxiliary hoist, design capacity is 10 tons. Both hoists are single-failure proof. The electric-motor-driven hoists raise and lower their loads using wire rope that is dual-reeved through upper and lower sheaves. The lower sheaves are an integral part of the load block. Each hoist includes a hook that is attached to the load block. The. heoka re, d .kwkum a.w Ge.it wh% .thyte EvuvYsSc4 The design parameters for the reactor building polar crane are listed in Table 9.1-11. l The reactor building polar crane design includes the features described below. 9.1-71
HCGS FSAR End of travel limit switches on the auxiliary trolley, and at both ends of the bridge, permit it to travel nearly the fullat each Bumpers length of the bridge, as-shown on Figure 9.1-31. end of the auxiliary trolley travel path, designed for 100% unloaded impact, back up the auxiliary trolley limit switches. The bridge and. trolley limit switches cut the power to their respective drive motors and set the corresponding brakes when they trip. 9.1.5.2.1.6 Thermal Overload Protection 1 Thermal = overload protection is provided for motors on the craile In addition, to prevent continuation of motor-stalling torque. ' thermal overload warning lights in the operator's cab indicate bridge, trolley, or hoist motor high temperature. , 4 9.1.5.2.2 Other OHLHS Cranes All plant OHLHS cranes, except14-6-' the reactor i "- building
---- polar--crane,
--d-- '
t..q ' are described below. 10."- The equipment tag nu=hers are shown i An Th-u = = Tdle ?.1 6.,,m u l.. hh4 h,p.c pw 61h .d T Parentheses. e\c h % Geet 4Its s k
- a. Personnel air lock hoist (10H217 chfGn and O c M M iS ai in, Seed 3.'2s incks vAe.n h Cay wwd. - >
This 30-ton capacity monorail hoist is located above 1 elevation 102 feet in the reactor building.#Tt is used t to remove nine shield blocks and the drywellThe personnel upper air lock (30 tons) during plant shutdown. shield _ block includes retractable wheels that permit this hoist to tow it forward along the monorail and position it to be lowered by two adjacent 15-ton hoists (item 11.). The 30-ton hoist lifts each of the eight
' lower shield blocks, moves it a short distance, and lowers out it onto a cart. The cart carries the nine blocks of the reactor building.f The personnel air ock is moved along the monorail and set down on a predetermined spot. A portion of the primary containment suppression pool is located below the load path of this hoist on the next lower elevation.
h eHwr S h 4b. So- b k 15 0585 ! The. pers.anel p n 4 k air lork. b mo w ahA. %locA eshopa,d & d npact A mhn 43 9.1-77 m = 3:-K . O.
forcAk kh c4 4%w, ygg h f ~g L20C+ 16. and +t. heeK is -
. d alt % ' m Q , w hen it is HCGS FS Q 4, m
- b. Reactor recirculation pump motor hoist (1AH201, 1BH201)
Each hem These 24-ton capacity monorail hoists are cated above elevation 102 feet'inside the drywell. lif ts the m. I recirculation pump motor (24 tons) out of the pump housing for inspection and repair. To remove the motor, the hoist raises it (in place).about 3 feet to l clear the pump. The removal cart is then moved beneath l the motor. When.the cart is in place, the motor is .
. lowered to rest on the cart and tilted to a horizontal
. position. It la tilted by simultaneously moving the chain-operated trolley and lowering the lift point. l With the motor secured, the cart is pulled out of the containment.
- c. Reactor water cleanup filter-demineralizer hoists
( G;;;ul l Aguo, gg ano ) g ' qw, . e . ThJsa.10-ton capacity maaorail hoistr.45 located above elevation 178 feet in the reactor building. Lt-4e4used for removal of the four concrete shield blocks above each RWCS filter-demineralizer. cell, and the filter g
/
tube bundle. The two heavi'est blocks weigh 8 tons each. The tube bundle weighs less than 500 pounds and - therefore is not a heavy load. .
- d. HPCI pump and turbine hoists (1AH211, IBH211) ,
forcar_% hoM 4w.ke,' ec We ve\ k et e}eva% 11Q4. A.in g o.nd +he_ h a A h d e. b m W W . Gt.i o W . A W w Q o us v w , These 4-ton. capacity monorail hoists are loc ed above elevation 54 feet in the reactor building. They are used during maintenance-of the HPCI pump and turbine. There is no' lower floor elevation. l~
. e. RCIC pump and turbine hoist (10H212) f ,
' This 3-ton capacity monorail hoist is located above elevation.54 feet in the reactor building. It is_used L during maincenance of the RCIC pump and turbine. There is no lower floor elevation, g hgf,g. g g g l w hals 4 +k We. ms') =% 1s pounk. h6 du bit am net knk h=g i.4. i ) l 9.1-78 .
. .- _ . - . . - . - . - _ _ - . . _ . _ - _ - . - - - - - . _ - _ . - ~ _ . _ _ . .
HCGS FSAR Main GemmTunne.) dnderkong CWne.
- f. ^"' M--e : in et; = i;;l;t.;n ::1r: ("f!") 5:ict (10H214K lettt13)
(7s a z This,Y Y'$ t:n ;;p ;it/ Mi-* is located abov elevation 102 feet in the. reactor building. It is used to lift the'oper tor off of the outboard d'WW b MSIVi r > maintenance. weda deem shotwwas pas *a-bao morar ogman M she dwac. eves
- g. -Inboard.MSIV hoist (10H219) psetT 3 o This 2-ton capacity hoist is located above elevation 102 feet in the reactor building. It is used to lift the operator off of the inboard MSIV for i
maintenance. l .
- h. CRD service ri;;in; bre- hoist (' t ;;! ,
monovad This ri;;ing .~e.T. is located above elevation 102 feet c in the CRD maintenance area of the reactor building. TOP 8 Oc gQakM0" pit is designed to accommodate a hoist for lifting N O M^- control rods (450 pounds), CRD maintenance equipment (up to 2000 pounds), and the neutron monitoring cask (less than 1150 pounds). S a no k a. descated c.t.D ss. E ' N we not puuhased 3 one. mM k % crowed b ano%sy tecdho'vs when neded.
- i. Vacuum breaker valve removal hoist (10H207)
L This 2-ton capacity circular. monorail hoist is located l above elevation 54 feet in the reactor building. The monorail is located inside the suppression pool chamber for maintenance and removal of the vacuum breaker
. valves. The valves weigh 912 pounds each and thus do l_
not constitute a heavy load.
- j. Main steam line relief valve removal hoist (10H202)
This 1-ton capacity circular monorail hoist is located above elevation 121 feet inside the drywell. The monorail is actually at elevation 135 feet. It is used
.to remove the main steam line relief va lves as requ ired for maintenance. The main steam line relief valves weigh 1100 pounds each and do not constitute a heavy
~
load. 9.1-79
HCGS FSAR
/ )
- k. Turbine building bridge crane (10H102)
This crane consists of a 220-ton capacity main hoist and a 45-ton auxiliary hoist. It is located above elevation 137 feet in the turbine building. It is used to lift the parts of the turbine-generator. A second crane, identical to 10H102, and originally intended for i use with the Unit 2 turbine, travels along the same rails as 10H102. A stator lift beam, supplied with the , cranes and designed to be simultaneously supported by l the main hoist of each crane, is used to lift the ! 366-ton stator of the turbine-generator unit.
- 1. Feedwater heater removal hoist (1AH103, IBH103)
These portable 24-ton capacity, manually (chain)
- operated hoists are designed to operate in tandem on e one of the nine I-beam monorails located above .
- elevation 120 feet in the turbine building. The beams serve the nine condenser-mounted feedwater heaters.
The hoists are used during feedwater heater tube removal.
- m. Heating and ventilating equipmsnt removal hoist '
(10H104) l This 15-ton capacity monorail hoist is located above elevation 171 feet in the turbine building. It is used for moving heating and ventilation equipment through the equipment removal hatch at elevation 137 feet.
- n. Motor-generator set hoist (OAH105, OBH105)
These 15-ton capacity monorail hoists are located above elevation 137 feet in the turbine building. They service and replace components of the two reactor j recirculation pump motor-generator sets.
- o. Secondary condensate pump hoist (10H106) l This 15-ton capacity monorail hoist is located above ,
elevation 54 feet in the turbine building. It services t ' ).' i l 9.1-80 I i
~ - - - -, - - , - . , _ . _ . , _ _ -
(Ref. 9.1.5.2.2.f) INSERT 2
~
Two parallel manually driven bridge beams connected by end trucks travel on two fixed girders located in the
- main steam tunnel. The top of the bridge beams is at l elevation 140 fe e t. A manually operated 2.5 ton capacity E trolley a nd hois t ( 10 H214 ) i s mo un t ed on one b r id ge be am ,
and. a manually operated 3 ton capacity trolley and hoist (10H223) is mounted on the other. .The book is at eleva-tion 137 f t.-4 in, for 10H214, and 137 f t.-3.5 in. for 10H223 -when f ully raised. (Ref. 9.1.5.2.2.g) INSERT 3 e
-It is moved between any of the five monorail beams as n e eded . The top of the rails are at elevation 119 ft.-
7.5 in., and the hoist hook is at elevation 117 ft.-
. 5.5 in, whe n it is fully raised.
L K54/14-2
. . -._- . ... ~_ - - - . .-.
HCGS FSAR the three secondary condensate pumps and their electric in motor drivers from one common-i %;;no;robc::. d
-p. Reactor feed pump hoist (1AH107, 1BH107, 1CH107)
These 15-ton capacity chain-operated monorail hoists are located above elevation 137 feet in the turbine building. They service the reactor feed pumps and their turbine drivers.
- q. Water box removal hoist (10H109, 10H110)
These 12-ton capacity monorail hoists are located above elevation 77 feet in the turbine building. They are used for removal of the condenser water boxes that have
- inlet and outlet nozzles. .
- r. Steam packing exhauster hoist (10H115)
This 10-ton capacity chain-operated monorail hoist is located above elevation 77 feet in the turbine building. It is used during removal of the tube bundle from the steam packing exhauster condenser.
- s. Steam jet air ejector hoist (1AH117, 1BH117, 1CH117, IDH117)
These 8-ton capacity chain-operated monorail hoists are l located above elevation 77 feet in the turbine building.- They are used during removal of tube bundles from the steam air ejector interim and aftercondenser. l t. Water box removal hoist (10H111, 10H112) These 8-ton capacity chain-operated monorail hoists are
- located above elevation 77 feet in the turbine o
[ building. They are used for removal of the condenser water boxes that do not have inlet and outlet nozzles. l l i 9.1-91 .
,- y - - , ,~,,w-w --v-,,me--,,,.een----,,.,e.,,--,.,--w,,.,-.. ,,..-,,,.wam,,..,-an,,n,-,n,,,,,-,.w-,.,,-,,,,--.-..,,,,,,,,,,--w --~,-.-,--v, ,an,---,,-,a..
HCGS FSAR , l s 4
- u. -Chiller tube removal hoist (10H118) ,
This 5-ton capacity chain-operated monorail hoist is located above elevation 171 feet in the turbine It is used for removal of chiller tube 4 building. bundles. v._ Emergency air compressor hoist (10H114) This 4-ton-capacity chain-operated monorail hoist is located above elevation 123 feet in the turbine
. building. It is used to service the emergency instrument air compressor.
- w. Main air compressor hoist (00H113, 10H113) r These 3-ton capacity, chain-operated monorail hoists are located above elevation 123 feet in the turbine building. They are used during replacement of the station air compressors and their motor drivers.
I i x. Vacuum pump water cooler hoist (10H116) This 2-ton capacity, chain-operated monorail hoist is located above elevation 77 feet in the turbine building. They are used for removal of tube bundles ! from the mechanical vacuum pump seal water coolers.
- y. Heating and cooling coil removal hoist (1AH119, 1BH119)
These 1.5-ton capacity monorail hoist are located above elevation 171 feet in the turbine building. They are ! used for removal of the cooling and heating coils that are located inside the air supply plenum. ,
- z. Turbine-generator auxiliary crane (00H100) l This 10-ton capacity bridge crane serves the turbine-generator. A set of rails is provided over the turbine-generator above elevation 137 feet. The
)
9.1-82
. . , . ~
^ ,,.,_y,,-,,,.__, .. _ .._-.
a l l i HCGS FSAR 3 radwaste area. They lift the WER pump (3300 pounds) and motor (3835 pounds), separately, for maintenance. ff. Waste evaporator hoist (00H312, 00H313) These 1-ton capacity, hand-operated monorail hoists are located above elevation 87 feet of the service and radwaste ares. They lift the tops of the evaporators and miscellaneous parts. ,
.gg. Diesel generator underhung crane (1AH400, 1BH400,
%c b 4 A bgc 'n d && M k%% g 1CH400, 1DH400) 4k ho'd hook e.k.vdLim is yt 2. 44. wbn i+ is . folk) kdwd--
I These 2-ton capacity-underhung bridge cranes are located above elevation 102 feet of the control anda . diesel generator area. + They are used to lif t and move miscellaneous _ diesel generat'or parts and equipment. .
.The four cranes share a single interchangeable hoist.
.4 + b.c. %ob bk.eW O.
hh. Intake structure 1N4f.A g.ws down on Byegantry at.1-vi,crane (00H500)
+he. ettwah ).' .%% g\ g ad of We is kn hoot is 1too &t. whm Sony re%Aq 'gu # This gantry crane is located above elevationjl23 feet of the intake structure.F It has a 30-ton capacity main hoist and a 15-ton capacity auxiliary hoist. The crane's heaviest lifts are parts of the traveling screens (19 tons).
ii. Reactor building. personnel lock shield removal hoist (1AH218, IBH21.8) These 15-ton capacity monorail hoists are located above elevation 102 feet in the reactor building. One is L locsted on each side of the personnel air lock hoist (item a.). After the personnel air-lock hoist tows it I l into position, they work in tandem to lower the upper shield block onto a cart that carries it out of the building.
}-
L l' 9.1-84 l~
- / ,
HCGS FSAR
,/ auxiliary crane is moved by the turbine building bridge
~
crane (item k.), as required.
.aa. ~ Domineralizer removal hoist (00H302)
.J This 10-ton capacity monorail hoist is. located above s - , elevation 102 feet in the service and radwaste area.
It lifts the fuel pool filter domineralizer and liquid radwaste filter elements out of their vessels for maintenance and replacement. The heaviest load j (1800 pounds) is a liquid radwaste filter bundle.
/
- I bb. Decontamination evaporator hoist (00H305)
This 7-1/2 ton capacity monorail hoist is located above elevation 54. feet of the service and radwaste area. . It . 4 is used during maintenance of the decontamination ' solution evaporator. The hoist lifts the top (1.'2 tons) and middle (2.9 tons) sections separately and sets then down on predetermined spots. Equipment' decontamination room hoist (00H314)
'cc.
This 5-ton capacity bridge-type crane is located above elevation 102 feet of the service and radwaste area. It is used for lifting and moving miscellaneous equipment.
-dd. Machine shop underhung crane (0AH301, OBH301, OCH301, l
ODH301) These 5-ton capacity acnorail cranes are located above , l - elevation 102 feet of the service and radwaste area. They are used for lifting and moving miscellaneous' plant ~ equipment and parts. ee. Waste evaporator recirculation (WER) pump hoist (00H309, 00H310) These 2-ton capacity, hand-operated monorail hoists are i located above elevation 54 feet of the service and 9.1-83 l
HCGS FSAR .
.jj.. Solid redwaste monorail hoist (00H316)
This 1-1/2-ton hoist is. located above elevation 102 feet in the auxiliary building. It transfers filled 55-gallon radwaste drums from the two i extruder / evaporator turntables to the capper / scanner infeed conveyor and replaces them with empty drums. kk. Solid radweste bridge crane (00H317) This 7-1/2-ton double girder crane is located above elevation 102 feet in the auxiliary building. It also 1 serves the radweste drum storage area loft at elevation 126.5 feet. It moves filled 55-gallon drums within the storage area, unloads the outfeed conveyor, assists in removing the shipping cask lid, and in truck loading.
- 4 ,
- 11. SACS pumps ri;;ir; L;;; hoist 'f_t :)
monoenils Fifteen-ton capacity ri;;in; i:r r above the SACS pumps
- are designed to ac ate hoists for removal of the pump motors. One serves pumps A and C in SACS gf d(Ayy The loop d the other serves pumps B and D in loop'B.
are located above elevation 102 feet in the eko reactor building.3 Beamoc, deAndad sac.s gum hists were noY dAens wh8" 4 ,y)twP. @ 4 p . " c. d 4Wg w'e M. bottouxd h>m okt 3 l ' ! mm. SACS heat' exchanger ri;;in; brr- hoist 'f;turr) monotEM Two parallel 2-ton capacity ri;;in; i: r at one end of each SACS heat exchanger are designed to accommodate hoists for of the heat exchanger end covers. One set of ves both SACS loop A exchangers, and the other set serves the loop B exchangers. The wenowd h beams are located above e}evation 102 feet in the reactor building. The. 40p of ear.h wM is al c.bab 12M l.5 k - , he*0W dedkokd SACS kat cachan hohh w not pare.bewd' Ng M M. borrows.4 from seh I II,hr.n10H315 of.edg nn. Recombiner system hoists These 1-1/2 (00H318) and 2-1/2 (10H318) ton capacity , chain-operated monorail hoists are located above elevation 67 feet 3 inches in the service and radweste area of the auxiliary building. Each hoist removes the i 9.1-85
HCGS FSAR j 'T valve 2 operator from one of-the control valves in the feed lines to.th= offgas recombiners, carries it to the hatch in the valve cell, and lowers it to a maintenance cart in the the access corridor at elevation 54 feet. EachLvalve operator weighs 943 pounds. 9.1.5.3 Safety Evaluation l All of the OHLHS cranes are evaluated in Table 9.1-10 with respect to'whether they carry heavy loads over safety-related equipment. located under the load path or on the next lower elevation. Table 9.1-10 excludes from further evaluation those OHLHS cranes that have no safety-related equipment below their load paths or only handle loads lighter than 1200 pounds although their design capacity is greater. Those OHLHS cranes not excluded in Table 9.1-10 are listed in . Table 9.1-12 along with the loads they carry, the lifting device, ' - if any, for each load, and the safety-related equipment beneath the load path. m Hazard elimination criteria are applied to each load handling situation identified in. Table 9.1-12 to determine
' if it-can be excluded from further evaluation. All equipment hatch load handling situations are dealt with in compliance with the guidelines of NUREG-0612. .
f Application of the-NUREG-06'12 guidelines, the exclusion criteria in Table 9.1.-10, and the hazard elimination criteria in Table 9.1-12. snow that there are run remaining OHLHS for which heavy load dro. might prevent safe' shutdown or decay heat l removal, cause anacceptable radioactivity release, or expose o spent fuel. The safe load paths for the OHLHS load situations in y Table'9.1-12 are presented on Figures 9.1-32 through 9.1-357 9.1.5.3.1 Reactor Builling dolar Crane Figure 9.1-32.shows tre w1 4 .aths for this crane. The reactor p- building polar crane in che s. hay one of the OHLHS cranes, that is physically capable of carrying heavy loads over irradiated fuel. Both the main and' auxiliary hoists are single-failure proof. Trolley and~ bridge travel' limit switches, plus a set of bridge stops on the railp and main trolley stops near the middle of the bridge, together ensure tha the main. hoist cannot travel over the fuel pool. Figure 9.1-31 shows the main hook exclusion area. The cask loading pit is outside the exclusioi: area and separate from the spent fuel pool. The spent fuel cask, therefore, can 9.1-86 E
.. .m- ....i.".-_m __i___._..,-...J_, ......--.__,__.,-._,_____,,.-__.-_-.__,.,_m.-..~.-..-_,,.. .
HCGS FSAR not accidentally drop into the spent fuel pool. The cask is moved directly between the~ hatch, the cesk washdown area, and the cask loading pit on the refueling floot as shown on the load path drawing, Figure 9.1-32. Some safety features of the polar crane design are discussed in Section 9.1.5.2.1. In addition, the crane is designed to Seismic I Category I criteria so that either hoist will retain its load I during and after a SSE. Manually engaged anti-derail devices on 1 both trolleys secure the trolleys when not in use and prevent rolling during an earthquake. Flat plate earthquake restraints welded onto the bottom of the girder end ties transfer the seismic loads to the reactor -building wall through the crane rail. The single-failure proof aspects of the polar crane design include complete redundancy for the sheaves, ropes, reeving,
- reducing gears, holding brakes, and other load path components.of both the main and auxiliary hoist *s. -
Figure 9.1-30 illustrates the single-failure proof auxiliary hoist design. The load is supported by the hook and two shackles, one on either side of the hook. The two separate load paths fr9m the hook and shackles extend through the four side plates up to two separate sheave pins. Each of the two plates on either side of the load block is designed to support the design load. The' trunnion applies the hook load to all 4 plates. Each shackle applies the hook load to the two side plates on its side. The side plates transmit the load to the two sheave pins. Each pin holds a sheave that is reeved independently. The block l housing includes two through-bars that are designed to catch the l- wire ropes and/or' sheaves if a sheave or sheave pin fails. Each sheave is. independently reeved to the hoist drum, where the ropes are dead-ended to the drum. l L Table 9.1-13 presents a point-by-point comparison of the reactor ' building polar crane design with the criteria of NUREG-0554, Single-Failure Proof Cranes for Nuclear Power Plants.
-9.1.5.3.2 ' Reactor Building Polar Crane Lifting Devices Lifting. devices used by'the polar crane are listed in Table 9.1-12. The special lifting devices, as-defined by
! NUREG-0612, are listed in Table 9.1-14 along with the Of
-6uz;1irn:0 rith f. :: ::;.'-;;70 : d the design safety facto tzt:hf l
E 9.1-87 ; I
HCGS FSAR
., 'T A single-failure proof spent fuel shipping cask lifting device and cask lift point design in accordance with the requirements of NUREG-0612 will be selected for HCGS.
A single-failure proof conventional sling selected in accordance with NUREG-0612, Section 5.1.6(1) is used to lift the fuel pool gates. The fuel pool gates are the only heavy loads which must vo$ndy be carried over the fuel pool. There are two lift. points on each fuel pool gate. They are designed with a minimum static factor of safety of 20 with respect to material ultimate strength. This l f}hhhisatisfiestheNUREG-0612,Section5.1.6requirementforasafety * ' 4 q
! factor of 5.
omitvdi5mol *dk l The fuel pool slot clug sling is a single-failure prooff<C;eri: 1
# dtd liftir; d:ti:: deci;nedito meet the requirements of NUREG-0612, Section 5.1.6. Each fuel' pool slot plug has a single lifting point designed with a minimum static factor of safety of 20 with .
respect to material ultimate strength. This satisfies the NUREG-061,2, Section 5.1.6 requiremenf for a safety factor of 10. .
.Although the special lifting device for the dryer-separator pool The plugs is single-failure proof, the lift points are not.
dryer'-separator pool plugs each have four lift points designed ' with a minimum s.tatic factor of safety of 10 with respect to material yield strength. Although not in strict compliance with NUREG-0612, Paragraph 5.1.6(3)(a), which requires redundant ~ points, each having a design safety factor with respect to ultimate strength of five times the maximum combined concurrent-static and dynamic load, the design is conservative and satisfies the intent of NUREG-0612. L The'special I?fting device for the reactor well shield plugs is i single-faibace proof in accordance with NUREG-0612, Each shield plug has ' Section 5.1.6, but tne lift points are not. fore lift points to prevent uncontrolledEach lowering of thehas load, ! as:,cing a single lift point failure. lift point a staci; design safety factor of 5 with respect to yield strength. Although not in strict compliance with NUREG-0612, Paragraph 5.1.6(3)(a), thedesignisconservativeands[atisfies the intent of NUREG-0612. The dryer-separator pool plugs and reactor well shield plugs ' discussed above are not carried over the fuel pool, but are carried over the. reactor vessel. They are only carried over the reactor vessel when both the drywell head and the RPV head are in place. A shield plug drop will not damage fuel or cause ) 9.1-88
INSERT 4 - 1 l
)
i'
~
The fuel rack lif ting fixture will be used for several c non-routine ' heavy load lif ts over the fuel pool. It is used for installing the ' spent fuel rack modules. As described ~ in Section 9.1.2.2.2.2, a base capacity of 1078 ispent ' f uel cells . plus 30 multipurpose cavities will be
~
"ipstalled for : initial plant operation. The remaining cap (- r.
- gity of 17 rack modules, providing an additional 2976 cells, will be installed during plant operation. The lif ting fixture design factors of safety versus yield and ultimate strengths are provided 'in Table 9.1-14. These factors meet the criteria of _ paragraph 5.1.6(1)(a) of N UREG-0 612 for a single-f ailure-proof single load path special lif ting device.
The - lif ting eye of the fixture is connected to the crane hook by a sling arrangement.- The slings are selected to meet the single-f ailure-proof criteria of Section 5.1.6(1)(b). of.NUREG-0612. .The four legs of the- fixture each have a - , J-shaped plate at the . bot tom . .The fixture legs are lowered ' through four of the empty cells of' the rack module being lif ted , moved horizontally a short distance, and raised to hook to the module base. The . four J-shaped plates contact the . underside of the module base when it -is being
. lifted. This design eliminates the need for lif ting eyes on the module. The weight of the module, together with the shape' of the .lif ting fixture pla te s , provides ass urance -
that the _ fixture is securely attached to the module during lifting. - -Thus, because there are no lif t points on the modules, and bothlthe crane,and lif ting fixture are single-failure-proof, the modules will be installed with a single-f ailure-proof handling system. The modules will be lif ted with the main hoist of the polar crane. Limit switches . and travel s tops , de scribed i n Section 19.1. 5. 2.1. 5, will be - removed as 'necessary to permit the main hook to travel into the main book exclusion area stown on Figure 9.1-31 when the modules are installed.
;K54/14-3
, , . - . .- , .-.-._.....-- -.. -.- . - - ,.......-.., - .. - .-..-. - ~ -. -
c HCGS FSAR unacceptable water leakage from the reactor. This conclusion is based on the assumption-that a plug drop could damage the drywell head and seal plate, but would have a less severe impact than a drywell or RPV head drop. In the highly unlikely event of a plug
-drop, the consequences would satisfy the four evaluation criteria of NUREG-0612, Section 5.1.
The.drywell head is lifted by the RPV head strongback. It is carried over the reactor vessel while the RPV head is in place. A drywell head drop will not damage fuel or cause unacceptable water leakage from the reactor. This conclusion is based on the assumption that a drywell head drop would be less severe than a RPV head drop. Depending on orientation, a drywell heed drop could damage the insulation support structure, rupture the RPV vent and head spray piping, damage the seal plate, and hit the RPV itself. But because the drywell head weighs about 2/3 as much as the RPV head, and because'some of its kinetic energy would be: absorbed by the insulation support structure and head - piping before it strikes the RPV head, which is still in place, a drywell head" drop would not cause fuel damage or unacceptable , water leakage. In the highly unlikely event of a drywell head drop, the consequences would satisfy the four evaluation criteria of NUREG-0612, Section 5.1. The RPV head strongback lifts the RPV head. The strongback design satisfies the guidelines of ANSI N14.6-1978, Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kilograms) or More for Nuclear Materials, in general. 4 However, it does not explicitly comply as recommended by NUREG-0612, Section 5.1.1(4). Further, the design satisfies the minimum design safety factor of 5 with respect to the material ultimate strength requirement of Section 5.1.1(4), but not the single-failure proof criterion of Section 5.1.6(1)(a) for a design safety factor of 10. Because the strongback is not single-failure proof, an RPV head 1. drop onto the open reactor vessel has been analyzed. Results show that vessel and core integrity would be maintained within
.the guidelines criteria of NUREG-0612, Section 5.1. The effects j
would be less severe than'those due to the fuel handling accident analyzed in Chapter 15. Damage to the vessel would not be severe enough to cause water leakage that uncovers the fuel. 2 The dryer-separator sling lifts the steam dryer and the moisture f' separator. The sling design satisfies the guidelines of ANSI N14.6-1978 in general, but does not explicitly comply as recommended by NUREG-0612, Section 5.1.1(4). The design eteo Cactors,4 s4ch em S4 arul o$mdt %M mt. emvM in Ta d 9.f-lQ . TQ et. le.n 9.1-8 9 +hn +be 4x4 eG S 9exsus pu ad s .qtu obdit giwag Sech ELAN.
HCGS FSAR
)4 reti:fie. G., ;;f:ty f::t;; ;f 5 ::;uir....uc or 5-siler. 5.;.;(0) ,
En* aat th- =iacie-fril :: p ::f ::quirc;::.t of 5.;.'(1)(e) f; :
- =='-t; E::ter of ;0.
Because the sling is not single-failure proof, both a dryer drop and a separator drop have been analyzed. Results show that vessel and core integrity would be maintained within the guideline criteria of NUREG-0612, Section 5.1. Damage to the reactor vessel would not be severe enough to cau'se water leakage that uncovers the fuel. The service platform sling lifts the RPV service platform. The sling design satisfies the guidelines of ANSI N14.6-1978 in . ggjI general, but does not explicitly comply as recommended by MUREG-0612, Section 5.1.1(4).4 Al;; ". d;;i;n ::ti:fic: tr. 0"' I ' f;ty f::ts; : ::;;ir== ant af e etie- 5.'.'(4), h"t net th - ein;1 -friler preef ::;;ircrent of Caetina 5 .6(')( ) f: e -
- f;ty f::t; Of 10. -
- I l Because the service platform sling is not single-fa'ilure proof, a service platform drop has been analyzed. Results show that vessel and core integrity would be maintained within the g guideline criteria of NUREG-0612, Section 5.1. I, The fuel pool jib cranes are carried over the reactor vessel when I
I the R9V head is off, but only when the RPV service platform is in l place on the RPV flange. A jib crane drop could damage fuel if it managed to cause structural failure of-the service platform. A conventional sling, selected in accordance with NUREG-0612, Paragraph 5.1.6(1)(b)(ii), is used to lift the jib crane. The
- load used to select the sling is two times the sum of the maximum static plus dynamic load. The dynamic load is assumed to be 0.25W, where W equals the weight of the jib crane. The load used is, therefore, 2(W+0.25W). The jib crane design has a single lift point with a design safety factor of 10 times the maximum combined concurrent static and dynamic load with respect to material ultimate strength as required by NUREG-0612, Paragraph 5.1.6(3)(b). The jib crane handling system, therefore, meets the single-failure proof criteria of NUREG-0612, ,
- Section 5.1.6.
l t No other heavy loads will be carried over the open reactor vessel. i 9.1-90
-. --.-..-_l_._.--_---.-..-.. ._ -
(Ref. ~p . 9 .1- 9 0 )' '. INSERT 5 The design factors of safety versus yield and ultimate strengths are provided in Tab le 9.1-14. The factor versus yield is greater. than- the value of 3, and the factor versus ultimate. is less than the value of 5 required by Section 5.1.1( 4 ) of NUREG-0612. e f 4
~ K54/14-4
+ 2 HCGS FSAR
+bt EEVb6dd bybEI The RPV head insulation and its support structure is carried over It is lifted by ...-,_ _....... ..
the RPV when the head is on.
- t'the rin;l; f ilure preef criteri Of .JSrc ? 00:2,'~
4eeti n 5.1.5'134 The support structure is lifted in two pieces. The lift points on.each piece are designed to meet the l single-failure proof criteria of NUREG-0612, Section 5.1.6(3)(a). i The other heavy loads carried over the RPV while the head is on are the RPV stud tensioner and the RPV head stud rack. They will
~
not cause fuel damage or unacceptable leakage because the drop would be less severe than a drywell or RPV head drop. . All heavy loads that need not be carried over theAdministrative reactor well are restricted from this area during refueling. ' _p rocedures help to control safe movement of all heavy loads. In summary, a. load drop into the reactor well could not affect . ' safe shutdown capability since the well is only open when the , ' reactor is_ shut down. Decay heat removal capability could be threatened only by a load large enough to damage the seal plate. Failure of the seal plate would not allow the large, heavy loads 4 to fall into the drywell because their size is greater than the space between the RPV and the drywell. The reactor well and the drywell are lined with steel' plate which will retain It anyisconcrete which is-fragmented by swinging or falling loads. doubtful l that other debris large enough to damage shutdown cooling piping could fall through the' labyrinth of' intervening piping and structural steel, including the massive primary containment radial box beams. The RHR shutdown cooling subsystem described in Section 5.4.7 includes a single suction line from reactor reciculation loop B. Therefore,,a load drop into the reactor well could disable the shutdewn : cooling function of the RHR system. The design basis for this event is that any debris that. managed to fall and disable RHR shutdown cooling would not have enough residual energy when it reached the components of this subsystem to do sufficient damage to prevent manual restoration of the cooling function. Damage such as a severed or crimped pipe, or complete loss of function of aShutdown suction cooling line valve would be operator is not considered credible. If manual manually restored as described in Section 5.4.7.1.5. restoration cannot be achieved, an alternate flow path as described in Section 15.2.9 could be used. Similarly, if debris from the load drop were able to cause leakage from exposed . reactor vessel piping, makeup water could be supplied by any of a number of RHR and core spray injection lines until the leak could be repaired. Therefore, the drop of a heavy load into the reactor well would not affect decay heat removal capability. 9.1-91
*m-e-vw -w e e +we-++--wee-ee en , %%--,-meweer-see--- --,w-e---. -,--w-r -w - - - -~=---o-=--v--me.--oe-+er=*----= ===--u----~ -we -
HCGS FSAR 4
. s The fluz donitor shipping crate is carried over the refueling floor by slings selected to meet the single-failure proof guidelines of NUREG-0612, Paragraph 5.1.6(1)(b).
Heavy loads carried over the refueling floor that employ lifting devices or lift points that are not single-failure proof weigh up to 107.5 tons. These loads include the items listed below and are also tabulated, with their weights, in Table 9.1-12. r
- a. RPV head
- b. Drywell head
- c. Reactor well plugs-curve,,
d 4 e
- d. Reactor well plugs-straight, 2
- e. Dryer separator pool plug-curved
- f. Dryer separator pool plugs-straight, 3 l
- g. RPV service platform
- h. RPV stud tensioner
- i. RPV head stud rack The RPV and drywell heads each have four lift points. The drywell head lift points meet the single-failure proof guidelines of NUREG-0612, Section 5.1.6. The heads are handled as close to the refueling floor as is practical. Both heads are lifted by the RPV head strongback. As described above for loads handled over the reactor, the head strongback is not single-failure proof. However, the design is conservative and the potential for a load drop is very smal'1.
9.1-92
l HCGS FSAR l The reactor well and dryer separator pool plugs are handled as ! close to the refueling floor as is practical. As described above i for loads handled over the reactor, the four lift points of each
' plug are not single-failure proof. However, the design is conservative and the potential for a load drop is very small.
1 The RPV service platform has three lift points. The platform is handled as close to the refueling floor as is practical. It is lifted by the service platform sling. As described above for loads handled over the reactor, the sling is not single-failure proof.- However, the design is conservative and the potential for a load drop is very small. The RPV stud tensioner has four lift points. The tensioner is handled as close to the refueling floor as is practical. The (66pf)l stud tensioner lifting device consists of four slings supplied i
- p) ) with the tensioner., 7be-deei;; i: : rerestiv : d th: p:t:nti:1-
< . f;; 1 erd dr:; is very crril. - The RPV head stud rack has a single lifting point. The stud rack' is handled as close to the refueling floor as is practical. The stud rack is lifted by a sling selected to meet the single-failure proof criteria of NUREG-0612, Section 5.1.6(1). Because the polar crane main hoist is prevented from traveling over the fuel pool, as described in Section 9.1.5.3.1, a load drop would not damage the fuel pool, spent fuel racks, or spent fuel. The RPV service platform, stud tensioner, and head stud rack are light enough to be handled by the polar crane auxiliary ! hoist. The loads paths are administrative 1y controlled to keep these loads out of the main hoist exclusion area, i.e., from over the fuel pool. a-In summary, a load drop on the refueling , floor of any of the ! loads normally carried over the floor by.nonsingle-failure proof , overhead handling system would satisfy the four evaluation criteria of NUREG-0612, Section 1.1. Table 9.1-15 presents a failure modes and effects analysis for the reactor building polar crane. 9.1-93
,- w e. -+ ...,---,ev.nern--n.n.n~,,,en,,,-,_,__,,.a.nn.. ___,n.,.,___ _-
HCGS FSAR
}
9.'1.5.3.3 Other OHLHS Cranes All plant OHLHS cranes, except the reactor building polar crane,
.are evaluated below. They are li=*=d 4s th: :: : : d:: :: th== ^ -
2;;: r in ?:51: 0.1-tSS--The equipment tag numbers are shown in parentheses. Each motorized hoist includes one 125%-capacity mechanical and-one 125%-capacity electrical brake that is automatically applied on loss of power. Each bridge drive includes a 125%-capacity brake that automatically sets upon loss of power. Each trolley includes one 100%-capacity electrical br'ake that automatically sets upon loss of power. The cranes and hoists shown as seismically secured in Table 9.1-10 have positive restraints that prevent crane derailment or crane parts from falling during an earthquake. , These cranes are designed so that their parts will remain in * ' place under a seismic acceleration of 7g vertical and 7g horizontal. The design also includes locking devices for use when the cranes are parked.
- a. Personnel air lock hoist (10H217) ,
This crane's load path is shown on Figure 9.1-33.
,___ ,,\ There is no safe shutdown equipment directly below the
}p6FE3 } load path. A portion of the primary containment
- .. suppression pool is 1 ted below the load path on the 1
next lower elevation.1 he personnel air lock is part h'of the primary containment pressure boundary. It is only moved when'the reactor is shut down. The air lock lift height above the floor is administrative 1y limited to less than 2 feet 6 inches. This A is the calculated load drop would not makimum allowable lift height. penetrate the floor if dropped from less than 2 feet 6 inches above it. Movement of the nine shield blocks in front of the personnel air lock is administrative 1y limited to reactor shutdown. The calculated maximum l allowable lift height for the shield blocks is 1 foot. I When the upper seven shield blocks are moved, they are higher than this. Administrative procedures require that the removal cart be in position below these seven
' blocks before the blocks are moved. The cart would L
4 absorb some of the energy of a load drop. A major i portion of the remaining energy would be absorbed as 4 the load punched through the floor. The low velocity J 9.1-94
.._..._____. E __.___.____ _ ____ _ _ _ _ _ _ _ _ _ _ _ __.
s
- (Ret._p.9.1-93)
INSERT 6 The tensioner , sling design factors of safety versus yield and . ultimate strengths are provided in Table 9.1-14. The f actors calculated for the maximum combined static and
. dynamic load , assuming the entire load is carried by only two of the four wire ropes, are greater than the-values of 6 versus yield and 10 versus ultimate required by paragraph 5.l'.6(1)(a) of NUREG-0612 for a single-f ailure-proof single load path special 'lif ting device.
e o K54/14-5
(Ref-p.9.1-94) INSERT 7 The air lock strongback design factors of safety versus yield and ultimate strengths are provided in Tab le 9.1-14. They meet the safety factor requirement of paragraph 5.1.6(1) (a) o f .NUREG-0 612 for a single-f ailure-proof
~ single ' load path special lif ting device.
h 9 e e 4 'e O i i K54/14-6 L
a HCGS FSAR impact on the suppression pool shell below would probably deform but not punch through it. If the dropped block managed to penetrate the upper suppression pool shell, the residual energy would almost.certainly be dissipated by the internal hardware 4 (piping and catwalk) and the water itself before the block ruptured the lower portion of the shell and caused any water loss. Because the reactor would be already shutdown at the time of a shield block drop,
-the suppression pool would not have to be available for t decay heat removal. The residual heat removal (RHR) system, operating in-the' decay heat removal mode, would take suction from a reciculation loop, pump through a RHR heat exchanger and back to the reactor. Therefore, a load drop that caused suppression pool water loss would not prevent decay heat removal.
- b. Reactor recirculation pump motor hoist (IAH201, 1BH201)
Figure 9.1-33 shows this hoist's load path. The hoist-is only operated during reactor shutdown. Dropping a motor during the short time it is raised and free hanging is unlikely. The load is positively attached to the hoist hook by the hook safety latch. No intermediate lifti.ng device is required. The hook directly engages the shackle pin on the top of the motor. The motor cannot be raised more than 5 feet because of the space limitation. It is normally raised tno more than 3 feet. h Ifthemotorweredropped,itwouldhitthepumpf'and 3 prob.bly.):::;: it: ecupling, seals, shaft, and bearings. The motor mount and the pump casing and its supports would absorb most of the energy and thereby protect the pump suction line between the pump and its upstream isolation valve from severe damage. The shutdown cooling line required for decay heat removal originates from the recirculation loop B suction line
- only. A motor drop could not prevent decay heat removal because the line branches from the recirculation loop piping about 15 feet above and more j
than 20 feet to the side of the potential motor irpact point. ' s 9.1-95 e
- ,. _. . ~ , , - - - - - - . . - - - , - . , -_w-ye-m.._._
\
l HCGS FSAR 3
- c. Reactor water cleanup filter-demineralizer hoist f t9Ht+9tIAn2.20, IB Auo) . ,
7v.tv. 4 no s&. shstamon or dq. hen.4 re.vnoalGetPmed . l b!nean un foot, patta of %cse. hoth on W wt Euerclardhon. : Fi;;;; s. ;-se _teu; Ois ::ci:t': 1::e p:th. c Llw tray? that ::rry ~~=" ==d incts;;;ateti:n erhies
- smiet J ith cin;1e **Ve ch====1 - e Ic:: "
adj:::nt to the 10 d path :t its eeuth end. d peration of this hoist is unlikely because the pressure precoat type reactor water cleanup filter domineralizer vessels ' are designed to operate for the life of the plant without undergoing maintenance. In th: .likely event of -- ;e te a cable trey thet caus; less of a FRis unit, the rMund:nt ""VS units -esid eilli be avai44ble, The nnly-safety-selsted ;quive A un the-nort lower elevation (= = ventil tien dr:t. The duct isprtsfthw uuntainment ycepsrs: el :n;p eyotwe. -I t orriy operates prior-to- occupancy of th: d:,a;i1 er terue. T* dn== =ct oper t: during er=al pl:nt eparatica er d ing hutdews. It i: net re";rie:d for , sefe shstde n es dec:y heat rrrevel, ,
- d. HPCI pump and turbine hoists (IAH211, 1BH211)
Figure 9.1-34 shows this hoist's load path. The only
)'
safe shutdown or decay heat removal equipment located in the load path is associated with the HPCI system. There is no lower floor elevation. A load drop during plant operation that disables the HPCI system would not prevent safe shutdown because HPCI does not function i during normal shutdown. It may not be necessary to shutdown the plant, provided the applicable requirements of the plant Technical Specifications are 1 : t. Th:t 10, u?CI c a Le inopstabl: for 14 days if other ECCS divisions are available. Otherwise, hot shutdown must be achieved within 12 hours, or cold shutdown within 24 hours.
- e. RCIC pump and turbine hoist (10H212)
Tk heis+ does noi handte. heaug tods, Theonlysafel [ [FTgure 9.1-34 shows the safe loaa patn. s tut $n o decay at remo 1 equi m nt located in the I l oad!phth asso i ted with he RCIC Wystem. hergnis low (r flo e e)evations A oad ystein rop {duri g p a Id n attorf th tfisables/the R IC IC dos w event safe (nutdoQecause ot function during norma.1 shutdown. It may not be necessary to t 9.1-96
. . . _ . . ~.._.,__.,_..,c_.._ _ . , . . . , , , , . . . _ , , . , , , , . _ , , , . , . , , . _ , . _ , . . . , , , , , , , . , . , , _ _ . - _ . , _ , _ . _ _ , , , , _ , . _ _ _ _ , , , _ _ , _ , . . _ . , _ . _ _ .
HCGS FSAR butdowntheplant 3rovi the a p licable nsar]e r0gui en of th p lant echnic Speci at I mat. ha is , RC C c an irope able f I ays ons avai le. erwise, hot o th r C S divi s i td must e achieved within 12 hours, or cold ( utdown within 24 hours. f C f. g;4n
. ts::deamTonnd e r.::: :iztOndeAog'mneionzts)
(sour:4j Figure 9.1-35 shows the safe load path. TheThere reactor is no l will be shut down when this hoist is used. decay heat removal equipment located in the load path.
' All of the equipment below the valves is associated with either the main steam or feedwater systems. If aws 4he-operator were dropped, it would hit one or more of the following items before it could hit the steam tunnel floors its valve body; the pipe on either side of the valve body; one of the other three main steam '
pipes; one of'the feedwater lines; restraint steel; structural steels and miscellaneous small ;;;;;;: pipe andth: valves of the main steam drains system.
;;;; ter ;;ig;.; 0000 ; nd;, ::: ;cep hei;ht reul ;;
1;;; tt;n 10 feet, nd th interv: in; t;;1 -vold itT ci;;iL acet Of tt; e...cg,, it _i: d;;;;d in;;;dib!: th:t TM7 # : d::;;:f re!r: :;;;;ter ;;alc* punch through the- seeem-4enne4 floor. If a dropped operator managed to cause 4 spalling after striking the floor, the concrete could hit one or more of the pipes in this area, or the torus itself. The pipes are associated with nuclear boiler
-instrumentation, liquid radwaste, RCIC, reactor water cleanup, core spray, fire protection, HPCI and primary containment instrument gas. None of the equipment e
l below the load path is required to remove reactor decay l heat. Therefore, decay heat removal' ability would not-l be affected by a-load drop from this hoist. A
- g. Inboard MSIV hoist (10H299)
L \ Figure 9.1-35 shows the safe load path. The reactor is There is no decay shut down when this hoist is used. heat removal equipment located in the load path. The tops of the drywell radial structural steel and drywell floor framing cross beams are located at elevation 100 feet, just below the main steam lines. All of the equipment above this structural steel network As associated with either the main steam, primary containment instrument gas, or breathing air systems. 9.1-97 ef..-__,_.~.,.,._-- -,_,_..w. _ _ . - , , , , . _
.. _ . - - - -- -=_- - .
1 HCGS FSAR
) ,
None.is required for decay heat removal. If the operator were dropped, it would hit its own valve body or_ steam line, or one of the three other main steam pi p before it could contact the structur.a1 steel below, unless it were dropped in the removal space - between main steam lines A and D. It would hit the steel directly if it were dropped in the removal space. The steel would stop a dropped valve operator. It would not fall to the-lower elevation (drywell floor). There is no decay heat removal equipment on this lower , elevation.
, h. CRD service s:. h hoist ;t;;:1 Figure 9.1-35 shows the safe load path. There is no safe shutdown or decay heat removal equipment in the load path. The torus is below the load path on the
- next lower elevation. It is-doub.tful whether a dropped.
load could punch through the elevation 102 feet floor. . Most loads actually weigh less than the 1200-pound l heavy load limit. All . loads are carried as close to the floor as is practical. ' 1 The following piping is located above the torus on the next lower elevation under the load path: 1 18-inch RHR pump A discharge L 2. 20-inch RHR shutdown cooling suction
- 3. inch HPCI pump discharge
- 4. 12-inch HPCI turbine steam supply.
4 Three 1-inch channel A reactor vessel level, pressure i and differential pressure instrument lines are also located in this area. If a dropped load during plant opration managed to penetrate the elevation 102 feet floor, or cause concrete spalling, and disable the ! shutdown cooling line, cold shutdown could still be i achieved. As discussed in Section 15.2.9 for this l
' situation, an alternate method to achieve and maintain l
cold shutdown that involves'the safety / relief valves, , 9.1-98
i
^l (Ref. 9.1.5.3.3.f) INSERT 8 )
l Because of- the congested piping and massive restraint steel beneath the load path it is nearly impossible for a dropped valve operator to reach the steam tunnel floor.
. Together the congestion and energy absorbing capability make it certain that a dropped operator will not e
e ) f 4 K54/14-7
l HCGS FSAR l that the impact could cause water loss. However, water loss would not prevent decay heat removal. l jf. Solid radwaste monorail (00H316) : The hoist is remotely controlled with the aid of closed-circuit television from the drum-handling control panel located in the radwaste control room. If the hoist becomes inoperable, a mechanical retrieval device permits removal and/or repair as necessary, while keeping operator exposure as low as reasonably achievable. There is no safe shutdown or decay heat re'moval equipment in the load path or on the next lower floor elevation. The drop of a drum could require - implementation of isolation and decontamination . procedures, but could not affect safe shutdown of tho ' - plant.- L kk. Solid radwaste bridge crane (00H317) The hoist is remotely controlled with the aid of
' closed-circuit television.from the drum-handling control panel located in the radwaste control room.
. Independent motors control low and high speed crane movement. Eyelets on the bridge provide attachment points for'a winch-type' retrieval hoist in the event of a loss of crane electrical power.
i There is no safe shutdown or decay heat removal ( equipment in the load path or on the next lower floor elevation. The drop of a drum could require I. l implementation of isolation and decontamination i procedures, but could not affect safe shutdown of the plant.
- 11. SACS pumps -!;;ir; i::: hoist ut :]
l monoted '-- serves the two pumps associated with ! One -^- '- safety auxiliaries cool'ing system (SACS) loop A, and the other serves the two pumps associated with loop B. A pump motor is only removed when the SACS coooling (6140 po0ndd 9.1-105 l _._,,_,-.,,m..,______,-__..._._ _-,._ _ _ _ _
A haswd an%dd -nimisown. loma w h op gg ppd i (sas A') Jeh w oh = Mag w=A. HCGS FSAR d sr.pavah b A elig loop associated with that pump is shutdown and completely isolated from the other (redundant) loop. This is not a normal maintenance lift. It would be done inf requently, if at all.T The '- '- "-- monowdd restricts the load path so that a load drop could only disable a pump or other equipment associated with the down-loop. A dropped motor would not- punch through the elevation 102 feet floor.because the deformation of the motor shroud, the intermediate pipe restraint steel, and the floor strength would absorb the kinetic energy of the dropped load. A OA00 p.;p ;;ts: ;;ig.; ;;; ;;;;d;; SACS heat exchanger ri;;ing b:r hoist ' ;. .; mm. monom'd
- Two hoists, one mounted on each ri;;ing i;;;, work in e tandem to remove a SACS he'at exchanger return end .
cover. The configuration includes a separate sling and lifting point for each hoist. Each of the two hoist, sling, and lift point combinations is capable of independently supporting the cover. The OHLHS is thus 4 ~ single-failure proof in the sense that a single failure )
- would not cause uncontrolled lowering of the load.
nn. Recombiner system hoists (00H318, 10H318) This hoist does not handle heavy loads. 9.1.5.4 Inspection and Testina 9.1.5.4.1 Reactor Building Polar Crane Final assembly and initial power operation of the bridge, both ' trolleys, and both hoists is done on site rather than in Paceco's shop. All crane parts subject to hoisting or seismic loads are i nondestructively examined as described in Section 9.1.5.4.1.1. l i 9.1-106
--w,. ,- + - , --,,..,.,mm.-_.-,_,_e_,_,_,,..,,,__g.w . n me , e.,rw.,%,w,, ,
HCGS FSAR The following steps are used to determine which items must be repaired ~or replaced after construction operation:
- a. A review of maintenance ~1ogs to be aware of any crane operation difficulties and any special or unusual lifts that were accomplished during the construction program
- b. A thorough visual inspection of all load bearing members
- c. Crane is operated to clock speeds and motion smoothness
- d. Maintenance personnel remove safety guards and access covers and clean the gears. Gears are then examined, relubricated, and replaced as necessary *
- e. Motor-coupling-reducer is checked for proper operation' p
- f. Limit switches are checked for proper operation
- g. Crane electrical control system is checked for proper ,
sequencing and operation. Preoperational tests of the polar crane include all of the specific heavy load handling operations that are performed during a normal refueling outage. 9.1.5.4.1.3 Operational Tests gg) In compliance with NUREG-0612, Section 5.1.1, the crane is inspected, tested, and maintained in accordance with Chapter 2-2 t i of ANSI B30.2-1976, Overhead and Gantry Cranes. :: ;pt rhr ::::: -
"== f* T - r; i. 1 - - th ; ;t,; ;;;;ini 2 tre :- in;;;;;;;;-
> 1eaguaney the test ar 4- ;:: tire i f::: prier t: ;;;ae .... ^- l i 9.1-109
._.-_.-._L___. _ _ _ . . _ _ _ . _ _ _ -._
HCGS FSAR [ 9.1.5.4.2 Other OHLHS Cranes Shop, preoperational, and operational tests on OHLHS cranes other than the polar crane are discussed in this subsection. I 4 9.1.5.4.2.1 Shop Tests , All of the OHLHS cranes listed in Table 9.1-10, except items 1 (reactor building polar crane), 11 (turbine building bridge , crane), 27 (solid radweste monorail), 28 (solid radwaste bridge I crane), and 38 (intake structure gantry crane) are functionally ' tested without load and at 150% of, rated capacity. Each hoist , brake is tested to confirm ability to brake the load from rated speed and hold it without slipping. , t Shop testing of the reactor building, polar crane is discussed-in e *; i Section 9.1.5.4.1.1. i . The turbine building crane is shop-assembled, except for the r, ope and blocks, to check fit. The trolley is powered along the
- bridge to check tracking. The hoist, trolley, and bridge drives are operated in the shop for 15 minutes. ,
The intake structure gantry crane is shop-tested at rated load. Each hoist brake is tested to confirm ability to brake the load ~ from rated speed and hold it without slipping. The solid radwaste monorail and solid radwaste bridge cranes are
- shop-tested at 125% of rated load.
9.1.5.4.2.2 Preoperational Tests Each of the OHLHS crades listed in Table 9.1-10 is given an : i operational performance test, a rated load test, and preoperational inspection in accordance with ANSI B30.2-1976, Chapter 2-2. 1 Preoperational testing of the reactor building polar crane is
, discussed further in Section 9.1.5.4.1.2.
, ) 9.1-110
HCGS FSAR I After preoperational performance and rated load testing, per ANSI B30.2-1976, the turbine building bridge crane is operational 1 and rated load tested in accordance with Paragraph 1 .179(K) of OSHA. Each hoist brake is tested to. >< confirm abi y to brake the load from rated speed and hold it I without slipping. 9.1.5.4.2.3 Operational Tests l i \ All the OHLHS cranes listed in Table 9.1-10 that carry heavy Ioads over safety-related equipment (those not identified by < exclusion criteria A, B, or C) are inspected, tested, and (' maintained in accordance with ANSI B30.2-1976.- '
..: ::::;ti:n i; "-
"'-a th- crrn ;;; f cey.ency is 1;;; ths= the ; :i'ied teet er "-
ta pect,ie- rrncy, in rhich :::: the teet er in ;;;ti;n i; - __. .fr 2___ u_ . . . . . . . . - i . 9.1.5.5 Instrumentation
- Instrumentation and controls for the reactor building polar crane are described in Sections 9.1.5.2.1 and 9.1.5.3.1 and Table 9.1-13. Supplemental information is presented below in Section 9.1.5.5.1.
'9.1.5.5.1 Reactor Building Polar Crane Bridge and trolley controls are the variable speed, reversing, magnetic, five-step type. Cab control handles are deadman-type with spring return. Hoist controls are A.C. static stepless-type in accordance with NEMA Industrial Control Standard ICS-3-442 Class III and OSHA. Release of a hoist controller stops the motion and sets the' brakes.
i The hoist control system limits lowering speed to 120% of full load hoist speed. Each hoist-holding brake system includes and i overspeed switch that stops the motor and applies the brakes at 120% of maximum no load hoist speed. The hois3s limit hook l
~
movement when starting from a standstill to 1/3.' inch for the main hook and 5/16 inch for the auxiliary hook in either the hoist or lower direction. Simultaneous motion of the bridge, trolleys, and hoists is possible whether-control is from the cab or the pendant. Cab 9.1-111 i l .- - . - _ _..--- - - - . . _
r r HCGS FSAR 3 control includes a maintained contact, master on-off switch. Cab
- . control is not possibl e unl ess the pendant is stored in its full
- up position. All pendant controls are momentary contact return to off pushbuttons. A deadman foot switch must be held down during crane operation fecm the cab.
For both the main and auxiliary hoists, a rotary limit switch coupled to'the drum trips at the normal up and extreme low hook position. A block-operated overhoist limit switch backs up the normal "up" limit switch by stopping the drive and setting the brakes. Hoist-overload switches shut off hoist. power and set the brakes if the design loads (150 or 10 tons) are exceeded. End of. travel limit switches stop the main and auxiliary trolleys and the bridge at their normal stop positions. The bridge, trolley, and hoist motors include overtemperature - protection. . i' 9.1.5.5.2 Other OHLHS Cranes ! ,All cranes include a drum overspeed system to automatically set the load brake when hoist drum speed exceeds motor synchronous speed. A phase-loss protection system automatically stops the hoist and sets the holding brakes when hoist power is lost. The turbine building bridge crane (item 11 in Table 9.1-10) control system includes redundant 125%-capacity hoist holding brakes that are automatically applied upon loss of power. The 125% trolley and bridge brakes are also automatically applied ( upon loss of power. The design includes hoist raising or lowering overtravel limit' switches. Bridge and trolley travel limit switches cut power at the bridge or trolley travel limits. All crane action control switches and pushbuttons are momentary-contact-return-to-off type. The' intake structure gantry crane (item 36 in Table 9.1-10) design includes automatic application of the mechanical hoist load brake and electrical hoist holding brake upon loss of power. Trolley end of travel limit switches cut motor power when the travel limits are reached.
- I
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9.1-112
HCGS FSAR 9.1.5.6 SRP Rule Review i In SRP Section 9.1.5, Acceptance Criterion 2 refers to Regulatory
. Guide 1.13,. Position C.3, which requires that' interlocks be 1 provided to prevent cranes from passing over stored fuel when !
J fuel handling is.not in progress. i At HCGS, only the main hoistlof the polar crane is physically prevented from traveling over the spent fuel pool. The auxiliary hoist has no travel restriction.- Preventing its travel over the fuel-pool is not an auxiliary hoist design basis. Instead, the alternative basis of a single-f ailure proof hoist described 'in l Section 9.1.5.3.1 is used. No loads are required to be routinely handled over the fuel pool when fue? handling is not in progress. J ' The fuel pool gates are the only heavy loads routinely handled ~
~over the pool when fuel handling is in progress. A single-failure proof handling system lifts the gates, and any other 4
noncoutine heavy loads that must be carried over the spent fue) . pool. . Acceptance Criterion 2 also refers to NUREG-0612, which, in Paragraph 5.1.1(1), states that load paths should be clearly marked on the floor in the areas where heavy loads are to be handled. , At HCGS, load paths are not painted on the floor. They are i omitted to avoid possible operator confusion in areas such as the refueling floor where'aultiple paths would cross. The paths are defined in the specific load handling procedures and shown on equipment layout load path drawings that are incorporated in the
, procedures. Deviations from defined load paths require written l 3
- alternative procedures approved by the plant safety review l
committee. l
- Acceptance Criterion 2 also refers to ANS 57.1, which in Paragraph 15.2.1.1(a) requires that the auxiliary fuel-handling l crane be provided with an underload interlock that is actuated upon a reduction in load while lowering, to prevent any further i
downward travel. At HCGS, the polar crane auxiliary hoist functions as the auxiliary fuel-handling crane. It does not have an underload interlock since it was purchased before ANS 57.1 was issued. The fuel pool gates are the only heavy loads normally handled over the fuel pool. A single-failure proof handling system lifts the 9.1-113
---.-..,-,,,n,...n-,,,,,-n-,.,-_ .,n.. . ~ _ .n _ n _.
HCGS FSAR gates, and any other noncoutine heavy loads that must be carried over the spent fuel pool. l 9.
1.6 REFERENCES
9.1-1 C. L. Martin, Lattice Physics Method, NEDO-20913, General Electric, June 1975. 9.1-2 AISC Manual of Steel Construction 9.1-3 AGMA Gear Classification Manual 9.1-4 Aluminum Construction Manual, Aluminum Association , 9.1-5 AWS DI.1, Structural Welding 9.1-6 , NEMA MG-1, Motor and General Standards 5I 9.1-7 National Electric Code . 9.1-8 OSHA 1910.179 9.1-9 OSHA, Vol 37, No. 202, Part 191 ON
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9.1-114 ,
l (Ref. 9.1.5.6) INSERT 9 Because opaque plastic sheets may be taped to the floor where the potential for radioactive contamination exists, polar crane load paths painted on the ref ueling floor (elevation 201 feet) may not be visible. The alternative method that is used at HCGS for the polar crane is to make a.-person other than the ; crane operator ( i.e. , a signalman) responsible .for assuring that the load path - is followed. The signalman inspects the load path before the lif t : to ensure that it is clear , reviews the specific load -_ handling procedure before the lif t, and provides direction to the crane operator to ensure that the pres-cribed path is followed. The specific load handling pro-cedures clearly define the duties- and responsibilities of the operator, the signalman, and any other members of the load handling party. The appropriate polar crane load path is temporarily marked with rope or pylons to provide a visual reference for the operator. If it is not possible to temporarily ' mark the load path, permanent or temporary match marks
- are useo to assist in positioning the bridge and/or
~
l trolley fer the lif t. The method of marking the load path is defined in each specific load handling procedure. bb t enEn deam N
,.The reactior building ' polar crane is the only non-exem t udtMqcMnt cab-operated crane at HCGS. Other non-exempt craneg are simple hoists on monorails where the load path cannot vary.
Most lif ts are short' lif ts where movement is limited to one ' coordinate axis "in addition to the vertical.3 For these no n-exemp t , non-cab-operated hoists the specific load handling procedures define whether a signalman is used [ and whecher. the load path will be marked. 6m m.my- a am
+ % 4mnaaum - sis ma sh*.
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h' #. . j ( . C OVEN Equipment Item Tag Floor Elev toc Fig Crano or Hoast Nunner Buildtno (fti NumDer Nunner Reactor building polar crane 10H200 Reactor 201 1.2-32 1 Personnel air lock hotst 10H217 Reactor 102 1.2-28 2 s Recirculation pump motor 1AH201 Reactor 102 1.2-28 3 { hoist 1BH204 (Drywell) A to 4 menetor water clean-up tilter / 1FH2 M - Reactor 178-6 1.2-31
- demaneralizer hoist IM N EPCI p o p and turbine hoist IAH2ft Reactor 54 1.2-26 5
g 18H211 6 RC2C pump and turnine hoist 10H212 Reactor 54 1.2-26 7 O .- M MN heactor 102 1.2-28 tou1L13 C1 3 8 innoard MSIV hoist 108219 Reactor 102 1.2 28 (Drywell) -
;s. , ,. .
9 Vacuum breater valve removal 105207 Deactor 54 1.2-27 hoist . . , . . (Torus) L main steam line relief valve 108202 Reactor 135-6 1.2-29 ( 11 removal hoist (Drywell)
=
11 Turbine building bridge crane 10st02 Turatne 137 1.2-16 12 reedwater heater remoeal naast IAN103 Turoine 102 1.2-14 13H103 13 56V equipment removal hotet 10R104 Turonne 171 1.2-17 14 Motor-generator set hoist OAH102 TurDine 137 8.2.=16 OBH105 , d .. T1002771? t, o C -
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T YAB12 9.1-10 ,- i Page 1 of 3 RAD NEAVY IAAD flANDLING SYSTEM 3 DAYA S1259ARY l Is load Over j Nas Vert is load Over Safety-Related M Coloma Capacity Litt Setsste Destgn Safety-Related M Equtpment on Esclusion j Cat i Standardf2) Eautoment? Next tower Elev Criter son t i) i l Area trons) fft int Yes a, b Yes Yes None u-V:14R-232 150 matn 129-0 [ 10 aum f NOI33 c, d No Yes None
,P-R 20R-23R 30 16=3 t
I Yes None i Sa-Q19R-203 24 s12-0 NoIII gd ~ Yes
- SbY318.9-17R ts.-o E. 63 0 sao .
lA-031p{R-17R10 h No ,g[d e goe= etene D No III Yes MA Mone 48-v310R-212 4 9-10 c, d Yes MA 1eumee C, IMFitTR-ISR 3 9-0 teoI3I c, d
- 6 Q, d Yes None
.P-931?R-20R 2-1/2 4 -0 (Iset n ) NOI3I ( Yes M LwG8#)
NoI3I 7es Yes None Q-R17R-20m 2 15-5 .d ' d Yes leo ,s _C , p-v 1!R-22R 2 7-0 sooI31 % ,
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.A NoI3) e, d I'* T -C <5 0-Y:17R-20R 1 32-2 i I *"
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- 12-29 '220 72-3 peo a, b *I- 4 esp N -r' =IBe* M .eeene. S ,,
main mean 45 122-0 % ., aus aus
! 3 No d 2 No No
- E-13:10-2; 24 12-6 c, d iso No a 15 37-0 No g R-Ft12-13 No leo B
- 15 16-5 teo e, d !'
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1/05/83 APERTURE ! [i dr[" CARD >
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- T1 Equipment Tag rioor r.iev 1ac rig i Item fft1 Number Nuncer crane of Hoist Nuncer Buildino Secondary condensate pump hoist 10H106 Turbine 54 1.2-12 15 Reactor feed pump hoist IAH107 Turbine 137 1.2-16 16 s
1BH107
' ICH107 Water bon removal hoist 10H109 Turbine 81 1.2-13 17 10H110 Steam pacting exhauster hoist 10H115 Turbine 77 1.2-13 ,
*13 Turbine-generator auxiliary crane 005100 Turbine 137 1.2-16 19 Steam jet air ejector hoist 1AH117 Turbine 77 1.2-13 20 IBH117 ,
ICHt17 108117 Water boa removal hoist 10H111 Turbine 81 1.2-13 21 10H112 C.g 171 1.2-17 22 Chiller tiube remoeal hoist 10H118 Turbine amorgency air compeessor hoist 108114 Turbine . 123 1.2-15 23 24 Main air compressor hoist 00H113 Turbine 123 1.2-15 105113 4 e 25 Vacuma pump water cooler hoist . 10H116 Turbine 77 1.2-13 26 meeting and cooling coil remoeal 1AH119 Turbane 171 * .2-17 hoist 15H119
' 27 Solid radweste monorail 008316 servtce and 102 1.2-20 radweste 28 Solid radweste bridge crane 005317 servtce and 126-6 1.2-21 radwaste 29 Domineraliser remoeal hoist 008302 service and 102 1.2-2f radwaste 30 Decontamination evaporator hoist 008305 service and 54 1.2-18 radwaste .
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ALE 3,1-10 (conti ( l
'I Page 2 og 3 Is toad Deer Mas Vert !s load Deer Safety-Related I hiumi Capacity Lift seismic Design safety-nelated Drutpoent on Exclusion see ftonsi (ft int Cat i standardf21 Eautoment? Nest inwer Elev Criterion l c, d No No B l-Fal2-13 15 12-9 leo '-11:17-22 15- 19-9 No c, d No . No 3 Ftgsl7-23 12 17- 4 NoI3I d, e too No B W :16-17 10 14-11 NoIII Rj d, e too No S e sgal3-26 10 26-1/2 NoI3I 3 dI 'I who hh h6 His 17-26 8 13-4 No III < de e leo No S 17-4 NoI3I d, e Ito No B P Etst7-23 0 k-al:17-21 5 14-4 NoI3I d, e les No S I i i . 1 4 7-0 sooI3I d, e i too No S frdste-16
- l. No B-P-rds11 3 7-1 teoI3I d, e No ,
l i k 6-9 leoI3I d, e leo too B 5-l'314 2 1 e, d les Iso S Hisl3-10 1-1/2 13-0 Iso 4 c, e ' Iso No 3 Re-Ms 36.9-4 2.6 1-1/2 ton 13-0 No a, b No B be-Me 44.2-45.4' 7.5 ton 39-6 No 10 0 18 tom 21-8 4eo e,' d - No 9er k teuer S ,3.9- 34.6 . i e, d too NA B f-E30.3-33.1 7-1/2 21-3 leo i' e l Bee A Neo Available On
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; APERTURE i CARD.
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P e- - T I 8 . TAE Equtpoent Tag Floor Eley toc Fig Ca Item , Ar Crane or Holst Nuncer Building (ft) Nuncer Nuncer Equipment decontamination room 00H314 Service and 102 1.2-20 54 31 notst radweste CAH301 Service and 102 1,2-20 N4 32 nacatne snop undernung crane radwaste 1: OBH301 OCH301 00H301 Waste evaporator recirculation COH309 Service and 54 1.2-18 Hs 33
. pump notat 00H310 radwaste waste evaporator hoist 00H312 service and 87 1.2-19 Nd
*34 ~
008313 caowaste , 35 Diesel generator undernung crane 1AH400 Control and 102 1.2-35 5-IBH400 diesel gene-1CH400 rator 10H400 36 Intake structure gantry crane ODH500 Intate struc- 122, 1.2-41 A, ture 126
. 't
/ 37 Personnel lock shield removal 1AH218 Aeactor 102 1.2-28 P v' hoist 1BH218 38 asconniner system hoist 10H318 Control and 67-3 1.2-24 &
. 00H318 dtesel gene-i.
y , CRO Mroute - IS w .. Os) rator Reme.W 127. 1.T. 'll *I SML poetas MLSg . - (gI a40 . 11'8'S
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# \0'5- M-W 04 II3 Exclusion criteria I.2. 2-S %
A. This crane is located in a cuildina or structure tnat contains no saf ety-related or saf e snutdown equipe
- 5. This crane's load pata does not pass over any saf ety-relatea or saf e snutoown equtpoent on tne floor bei C. Although this Crane's capacity is greater tnan 1200 pounos, its oedtcated load as lionter enan 1200 poun
~
III esign D standards: Tof" MMtvt3D h ar.g
- a. ANSI 330.2.0 Overhead and Cantry Cranes kultiple Carder)
- h. CMAA 70 Electrte overnead Traveling Cranes
' c. Ist! 100 Electrac Ware appe Hotsts
- d. AHS1 830.16 Overnead Hotsts (dMM - ~
- e. ANSI B30.11 Nonorail Systems and Undernuno Cranee
< . -- e (33S etssically secured (designed so that au parts romaan in place under 7g nnetrontal and vertical seismic accel restratats and locking devices). ,
Tee b44ng j (4)The destgn also uses ANSI 330.17 (overhead and Gaatry Cranes Esingte c.ar,se- as a guide. s s
$ tM1l h ) [ [ k p v y o( M b lelte 9felg W,lafid ir N4d c,*W,qidtt.d B e o.) % hit a w. w .a h w w. W ww naa.t p -
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kume . - Capacity _ , Lift, Setssic Design Safety-Related Equipment on Exclusion
~' toutament? Critetton' na (tonst fft snt Cat T Standardf2) Next Lower Elev 328.8-33.1 10-2 NO III C d. eh No too 3 W: $ 11-0 No d e[,.th No No 3
' 6-25.9 2 9-3 NOI33 de No NA 3 hE15.8-19.9 ft$.8-19.9 1 9-4 NoI33 d. e No No 3 324.3-34.6 .2 19-4 NOI3I c, de e Yes Yes None I M35-9 30 65'-0* No b, d e Yes Yes None main main is 8. -n-
.. -s c saon-22m 15 23-0 NoI33 Ed too Yes None ,
,d a iso c . s ,34.6-19.1 . 2.st10 21s> ts-a No mio 1.st00ssis ,,. ' $lm-na i a_ .,M ,M . .M
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2= .2 m.
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or on the next lower etevation. , l.. i
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HCGS FSAR T TABLE 9.1-12 page1ot/ , OHIJtS LOADS OVER SAFETY-REIATED EQUIPMENT First Elevation , secone Elevatton satety-Relates, satwty-petatee, Sate Snuteoesu, e, Sate Sate Shutdown, Load or Decay Hasare or Decay Itasere Litting Path Heat Renovat Elimination Heat Demovat Einstmatton Load Equipment critertonsea rg goutoment Guten2e _ lleavy lose Weight Device Fig . Feet crane /noista Reactor Buildine polar crane (Item 1. Table 9.1-101 testas a, t 162 testas a, t Reactor ese11 shield 107-1/2 Shield plug 201 a. pluga tons eeneenghaen dq, 9.1-32 eaeeaa a, e, t 162 ee ea ae a, e, t Dryssell itead 65 torts RPV head 9.1-32 201 b. strongback tantas a , e, t 162 test se a, e, t Reactor vessel 97 tone RPV head 9.1-32 201 c. head strongback 201 eesea: a, e, t 162 eaeaaa a, e, e
- 4. Ieoisture 13-1/4 Dryer / 9.1-32 separator tons separator sling .
castas a, e, t 162 tastas a, e, t
- e. Steam 45 tons Dryer / 9.1-32 201 dryer separator
- sting testas 162 tastas a, t
- f. Dryer / separator 90 tons Pool plug 9.1-32 201 as t.
pool plugs -* : _. 7 - r grapple, 201 eaaeas a, d 162 eeaaaa a, e
- g. Spent fuel 110 tons Fuel cast 9.1-32 shipping cask yoke ea 3a aa 4 162 eeseae a
- h. Auxiliary hoist 1 ton (None 9.1-32 201 load block required) 201 eeaeaa a, d 162 eeaaaa a, a 14 stain hoist 10 tons (teone 9.1-32 load block requireal d 162 488488 e
- j. Spent fuel pool 9 tons Single- 9.1-32 201 te st a s slot plugs failure proot .
sling 4
- s< s a d 162 ee se ae e
- k. Spent fuel /I tona Single- 9.1-32 201 pool gates 3,4 failure proot sting
'HCGS FSAR .
TABLE 9.1-12 (cont) Page2ot[ ' 'First Elevatton S - f Elevattom Satety-Related, satety-metatee, Sate Shutdown, Sate Snetdoun, Sate Easere-Load or Decay Kasard 'or Decay. Heat Renovat Elimination Igoat Demovai Einmanetton l Imad Litting ' Patta gestonent Ct&tgrh Equ tostent Critettonsas g gg teeteht Device A F Feet saaeaa e, i 162 saaeaa e, g
- 1. Rrw service A tone -Service 9.1-32 201
_ platform p platforan j sting ,* i caa:aa 162 seasaa e, t
- m. Road stud 1.5. tons Single- 9.1-32' 20g e, t rack failure proof sling
[. p-T-beend E.*.'1-3 2 saae8e e, t 162 easeea e, g 9 201
- n. Vessel head 5 tons f
i insulation fee _. novs ! Aruk Qvosne -' ' r i seasaa 162 s*e<88 e 245 tons single- 9.1-32 201 e .. !' o. Fles monitor failure proof shipping I crate . slings T.3 sassea e, t 162 easeaa e, t 4-4M- prv stud 9.1-32 201
- p. Stud tensioner frams tensioner sting a, e, t 162 s a ss se a, e, t 201
- q. Road strongback f tone (Isone required l 9.1-32 s stac a s l
seaeae a, 4 162 seeaas a, e 201 j r. spent fuel 6 tone (teone 9.1-32 cask yotte required) j seaaaa 162 s a as aa t
- s. Batch cover 2' 4 Single- 9.1-32 201 i 4' x 48 tons tailure proof stings f
sessas 162 sestas t
- t. Eatch cover 7.5 Single- 9.1-32 203 t
{ 10' x 10' tons failure proof slings lI 10 Single- 9.1-32 201 tescas e, t 162 <*8t*8 e, t j u. Refueling bellows guard tous failure proof j j ring sting . i 1.b b 3946-4b Single- 9.1-32 201 cassas d 162 s e st as a i
- v. Jib crane tailure proof i
sling
- w. Channel handling (teone 9.1-32 20g :aseaa e, [h 162 ea as aa e, boom crane requirad) l pe.pr _.seeavab 2 h hagehA 7.t-3r. zol Cuaeal cl xot G>nerr) d x.
lole + ty*;L LM2 '" f*" 3 J u t Gma j i_ .-__-
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e.% 0.n.1 ntK
e ne oS
/ n e
rae
- t. gt 3 aat
't st s as 3 M&s 1
n E WC o t d , P t eus a tes t v ac a e le vt s <- l et ov 0 L E Seyso
- it aea e yScpc n t e t s
0 t u o eeptu , c tt e aareS s Ssoe so g e u t e e F n8 o8 i8 dtn . rao - ant 9 sar ame Htt ti Er n, c o d , t en . t tw t 4ao a vt e2l detyon \ t tRuas e n g n E hcem ySeRo o c t c D t i ( s ee tu r tt raq / 2 i aeoeE - h 1 FSS H 1 t 9 e e 7 E F L C 8 k T edh fat q aoai @ sLPF { s g ne M e g y
. ic u ti tv ie LD w t
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4 HLMS FSAR nl 7 Page / Of TABLE 9.1-12 (cont) First Etewation Second Elevetton Satety-Related, satety-metatee,
' Sate Shutdown, Sate Shutecem, Safe Nasare or Decay Hazard or Decay Load Litting Path Heat Femoval E11manation Heat Demowai 511minatt3a Imad Equ1roent Criteraont*8 Peet Equapaent Mrapg .
1 ' Iposvv Los4 Pe.,Asht. Dev1ce _ Fto Feet 4 Crane / Moist: Personas 1 Air ' Tack Hoist (Item 2. Table 9.1-10L 102 None b FF Torus, and core 0,, c 30 tons Air lock 9.1-33
- a. Air lock strongback spray Mrct, ane sav eksemarge Paping 1
4 b #7 Torus ane core e, e None required 9.1-33 102 Mone
- b. Upper shield block 21 tone spray ,MPCE, and av das-charge paping L
b 77 Torus ane core e, c None required 9.1-33 102 Hone l c. Immer shield 17 tone geray, HPCI and i blocks (0) JfSRv etscharge Papang i Crane /Moista Recirculation Pump Motor Holst (Item 3. Table 9.1-101 102 Recircu- b, c sf Mone Recirculation 24 tone. None required 9.1-33 pump motor (incide lation (botton erywell (1 AP201, et ery-1BP201) weil) ,i and asso-ciated piping and-i conduit
~
l l Crane /Beist: R= tor Water etwooup Filter /Demin.' it' M item 4 Tahie T 1-101 ( V BMCS filter / V V %/Cdhventional 9.1-38 ~-
'1'7cable M'trays M . b y ,162 s
-Vent 11atton euct c I dominergleegr 8 tons f #sh /
j 11 er \ 4 blocks f 1 Crane /Notett RFCI P E and Turbine Hoist (Item L Table 9.1-101 i ' HPCI pump and 3.'75 tons conventional 9.1-34 $4 Pumps b,_c (No lower etewetton)
- turbine parts slings (10P20s, 10P217)
] (turbine case) turbine (10S211) 6 HPCI*paping
- l i
, ~ . ~ <.
I i HCGS FSAR
. T TABLE 9.1-12 (conti Page 4 og First Elevation Second Elevation Safety-Related, Satety-Relatee, Safe Sate Shutdown, sete senteown, Load or Decay Hazard or Decay Masare Load Lifting Path Heat Removat Elaminatton Heat Removat E11mination We&qht Device Faq Feet __ Equipment Crstertona s a Leg Equ1pme;st Crgtet30*g_
lleavV Load 6~ C=---# M ett RCIC Puse ame Tur62ne Hotst (Item 6. Table 9.1-101 5 tone Co entio al 9 -34 4 op D, c (no elevat - IC pump ngs (10P20 )
. . ame t Ebi pa t
( (turb case) turbin (10S12) RCIC piping l l i Crame/Boiet h%'n Mbana\ MMYM
- -'___i"*""__.^. fItom 7. Tabie 9.1-10) 44st. *Mmm2bes*. c de Torus plus core D, c plain steen N Conventional 9.1-35 102 MSIVs (HV F0 2W A-Dl , spray, contain-
- isolation valve g,g g slinge main steam, ment anstrument
- opsentoso gas, ac1C ane and feedwater i M s%eam M dakit. O.8t h s piping HPCI papang, l and nuclear
' MA N & c M ydst,0i h5 Dotter sjstem anstrumentatton 4 Crane /Noist: Th_r4 ISIV Boist (Item 8. Table 9.1 80) 9.1-35 1u2 Ms1Vs c et namn steam 4 c, e plain steam deat.46- Conventional J isolation valve gg slings (HV F022 (bot- containment A-D) and tom Enstrument gas operators or ans areatning
- main steam paptng ery- air papang we111
! Crane /Noiett 'furbine Building Bridge Crane . (Item 11. Table 9.1-101 202 tons Conventional La r 137 RPS (rea b 102 RPS condent o i Turbine-generator rotor slings tor pro ec-tion system) condugt j Cran Noist! 'Itsrbine Gener'ator Aux ary Crane & tem Ta le 9.1-10
/ 102 RPS conduat o ter shield 6.75 tons venti 1 Later 13 P S conduit b .
e ings Crane /Noi W i=or iter Demoval Moi Ites 29. Table 9.1-101 Liquid dwas e 1000 lb Conventional 9.1-39 102 toon, - ul Res coneust a f11ter slings
HCGS FSAR l 6 Y TABLE 9.1-12 (cont) Page /1 ot/ First Elevatton Second Elevataca , Satety-Belated, Satety-petates, Sate Shutdoesn, Sete Seet40em, Sate or Decay Sasare l Load or Decay Hazard Neat Demove! E11manatace 1 Imad Laiting Path Heat Penovat Elaminataan Equionent Craterton s *8 [331 _ Eeuansent Cr4tersee i Ileavy Lead M Devace h feet s Crame/noiett Diesel C r ~mter Unde-hune Crane (Iten 35. Table 9.1-101 Diesel b, c #7 Assocaates De C 3549 ab Conventional 9.1-36 102 coolang papang paesea generator generators elings parte, e.g., (IAG400-combustion air 1DG400) cooling test.or and assoc heat eschanger cooling l tube bundle piping i Crame/meiett Intake Stahure Srtry Crane (Iten E Table 9.1-10) . 93 stratnere e, c 19 tone Conventional 9.1-37 123 screens n, c Traveling screen, (S$01) (FSO91 S.W. pump, and slings E heaters maec equipennt (vES01) s S.W. pumps + (PSO2) , (Item 3 7. Taple 9.1-101 j Crane /Boiett Reactor sui 14 Lao Personnet Lock shield memovat noist e, c Hone - 34 Torus one - T-shaped shield 21 tone Hone required 9.1-33 102 core spray, i neCI, ans saw block
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$- 7 TABLE 9.1-12 (cont) Page 4 ot y sea gasard elimination criteria:
.' a. Crane travel for this area / load combination is prohibited by electrical interlocks or mechanical stops. s l b. . System redundancy and separation precludes the loss et the capaD111ty ot the system to perform its safety-related function to11owing this Rosa drop , ir this area.
- c. Site-specific considerations, such as maintenance sequencing, e11minate the need to consider this load / equipment combination.
- 4. The likelihood of a handling system taiture for this load as extremely j oma11; 1.e., Section 5.1.6 of NUREG-0612 is satistied, the OHS is single-failure-proof.
; e. Analysis demonstrates that crane failure and load drop wait not prevent safe shutdown or decay heat removat, or cause unacceptaole radiation t release.
4 f. The likelihood of a handling system tailure is small. The system design meets the intent of MUREG-0612 (not dropping the load) . ta: Irradiated imet. Ea3 mesctor vessel. . 1 a i I i I . 0 w
HCGS FSAR f TABLE 9.1-13 Page 1 of 5 REACTOR BUILDING POLAR CRANE DESIGN COMPARISON WITH NUREG 0554, SINGLE FAILURE PROOF CRANES FOR NUCLEAR POWER PLANTS
-(MAY 1979)
Does Not NUREG Section Complies Comply Notes
- 1. INTRODUCTION X
- 2. SPECIFICATION AND DESIGN CRITERIA 2.1 Construction and Operating Periods X (1)
Maximum Critical Load (2), < 2.2 X 2.3 Operating Environment X (3) 2.4 Material Properties X (4) 2 . 5< Seismic-Design X (5) 2.6 Lamellar Tearing X (6) 2.7 Structural Fatigue X (7) 2.8 WeldingProcefbres X
- 3. SAFETY FEATURES 3.1 General X 3.2 Auxiliary Systems X 3.3 - Electric Control System X (8) 3.4 Emergency Repairs X i
- 4. HOISTING 4.1 Reeving System X (9) g-4.2 Drum Support X (10)
! 4.3 Head and Load Blocks X 4.4 Hoisting Speed X 4.5 , Design Against Two-Blocking X (11) i
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- 9. OPERATING MANUAL
- 10. QUALITY ASSURANCE X (17)
YW5 (1) Section 2.1 - The load lifts during construction were not greater than those for plant operation; therefore, no separate specifications were prepared.
-(2) Section 2.2 - The reactor building polar crane main hoist id '
designed to' handle a maximum critical load (MCL) of 130 tons. The MCL rating will be clearly marked on the main hoist. The design rated load (DRL) of ISO-tons provides an overall increase of 15% in the crane's load handling ability ,
, above its MCL capacity to compensate for wear and exposure. l The reactor building polar crane auxiliary hoist is designed to handle a MCL of 8.7 tons. The MCL rating will be clearly
- marked on the auxiliary hoist. The design rated load (DRL) of 10. tons provides an overall increase of 15% in the crane's load. handling ability above its MCL' capacity to compensate for wear and exposure. <
(3)- Section 2.3 - All identified parameters, except maximum rate !- of pressure increase and emergency corrosive conditions, were ! specified. A maximum rate of pressure increase was not ! specified because it was judged not significant to safe L design of the crane. Because it is in the reactor building, l .outside the drywell, the crane would not be subjected to the high accident pressure (62 psig) possible inside the drywell. The maximum pressure increase specified for crane design is
.25 in.'wg minimum to +7 in. wg maximum. Emergency corrosive conditions were not specified because none we're identified that would prevent safe crane operation.
(4) Section 2.4 - The minimum specified operating temperature is . 600F. Materials for structural members essential to structural integrity are impact-tested unless exempted by the provisions of Paragraph AM-218 of the ASME Code, Section VIII, Division 2. All structural members, except the l main hoist-drums, are exempt under Paragraph AM-218.2, which i withdraws the impact test requirement if stress intensity is
',less than 6000 psi. The main hoist drums are Charpy-tested I
f-
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l HCGS FSAR s .. - [ TABLE'9.1-13 (cont) Page 5 of 5 stainless steel auxiliary hoist wire ropes, with independent wire rope center, are 1 inch in diameter with an ultimate 4 breaking strength of 77,200 pounds each.
- (10) Section 4.2 - The main hoist and auxiliary hoist drum assemblies, each.with its shafts and bearings, are designed
.at factors of safety not less than 10. Safety lugs are !
.provided inside each trolley truck to sustain the: drum :
assembly hubs in the event of drum shaft failure at.either end. Upper sheave shafts and block swivel assemblies are provided with safety retainers and block housings capable of sustaining the load in case of shaft or swivel failure. Drum movement in this event is mechanically limited so that the gears and holding brakes remain engaged. (11)'Section 4.5 - Dual upper limit switches of diverse design in series, and an overload cutoff switch on each hoist stop the, hoist motor and set the brakes. Motor overtemperature ' switches activate warning lights in the cab and on the pendant. Each limit switch allows the hoist motor to be operated in reverse after it has opened. (12). Section 6.1 An. emergency breaker switch located at the refueling floor level cuts power to the crane independently of the crane controls. (13) Section 6.2 - The crane Lo rd 'oes not lift spent fuel assemblies. :::.;Llies s_ h (14) Section 6.4 - Jogging and plug'ging are considered in the crane controls design. Drift point is not provided for bridge or trolley movement. (15)-Section 6.6 - Manual controls for hoisting and trolley movement are not provided on the trolley. Manual controls for the bridge are not located on the oridge. (16) Section 8.3 - The crane design does not include an energy controlling device between the load and head blocks. Therefore, the two-block test is not done. Instead, the > two-block test consists of verification that the two uptravel limit switches on each hoist function as designed.
- (17) The crane is procured under a QA program that complies with the applicable provisions of ANSI N45.2-1971. Field installation, testing, operator qualification, and crane operation comply with ANSI B30.2.
l
. . ~ , , , , . . , . - . - , . . . - . . - - . . . . - - - - - . - - - - , . . - - , _ . - . . - . - ~ . - . , - ~ - - - - - - - - _ . - - - - - _ - _ . - - - .
a ,,, , a A.e a .ssa, n - m HCGS FSAR i TABLE 9.1-13 (cont) Page 4 of 5 per ASTM A 370. The crane was not subjected to coldproof testing because low alloy steel, such as ASTM A 514, is not used. Cast iron is not used for any crane parts. (5) Section 2.5 - The SSE design vertical acceleration is less than Ig. Therefore the bridge and trolley wheels will not jump up off their tracks during a seismic event. The bridge and trolley designs include horizontal seismic restraints that would prevent the wheels from leaving the tracks. . (6)- Section 2.6 - Nondestructive examination (NDE) was done on all welds whose failure could cause a drop of a critical load. Section 9.1.5.4.1.1 describes the NDE in more detail. Lamellar tearing of these welds is not expected to occur. ( 7 )' _Section 2.7 - A structural fatigue analysis was not part of 4 the design requirements for the reactor building polar , crane. The crane is classified as a low-use crane.according , to the guidelines of CMAA Specification 70. Structural fatigue s.not considered necessary in view of the low number.o load cycles expected. Wdps (8P Section 3.3 - Cab controls are deadman-type with spring A deadman foot switch in the cab must be held down return. during crcne operation. Release of the switch will stop the crane and' set the brakes. Overspeed switches on the hoist drives stop the motors and set the brakes at 120% of no load L speed.- Pendant controls.are momentary contact pushbuttons '- that return to off when released. Pendant control includes an emergency stop pushbutton that stops power to all drivers.
' (9) Section 4.1 - The maximum fleet angle from drum to lead sheave in the load block or between individual sheaves does not exceed 3-1/2 degrees at any one point during hoisting.
Reverse bends are not used in the reeving system. Each main ; hoist rope is reeved through block and upper sheave' assemblies so that its eight parts provide two parts iri each quadrant of the load block about the vertical axis of the - hook. With both ropes effective, the load is' supported by sixteen parts at an effective static factor of safety of 10. If one rope loses its effectiveness, the load is supported by the eight parts of the remaining rope at a static factor The extra improved plov steel main hoist j g __ of~ wiresafety ropes, of,$4f. with independent wire rope center are 1-1/2 inches in diameter with an ultimate breaking strength . of 228,000 pounds each. With both auxiliary hoist ropes
- effective, the load is supported by four parts at an effective static factor of safety of 15. If one rope loses its effectiveness, the load is supported by two parts of the
, , remaining rope at a static factor.of safety of 5. The t
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(w 2) DSER Open Item No. 166 (Section 12.3) AIRBORNE RADIOACTIVITY MONITOR POSITIONING The applicant should clarify how he intends to use the ventilation monitors to accurately monitor plant iodine levels when the air being monitored by these monitors has been filtered through the plant HEPA and charcoal filter banks.
RESPONSE
FSAR Section 12.3.4.2.2 has been revised to address how HCGS intends to accurately monitor particulates and iodine from
' any compartment which has a possibility of'containing airborne radioactivity and which normally may be occupied by personnel, taking into account dilution in the ventilation system. .
MP84 95/17 1 h d-
HCGS FSAR 8/83 taps are located in the ducts next to the detectors so'that grab samples can be taken. Additional mobile samplers with monitoring detectors that are ' displayed, controlled, and recorded by the CRP are provided for use if needed. , More details about airborne radioactive material sampling and monitoring are included in Section 11.5. , The above described airborne radioactive material monitoring i equipment and procedures are used to meet the applicable parts of
! Regulatory Guides 1.21, 1.97, 8.2, 8.8, 8.12, and ANSI N13.1-1969.
Acceptance Criteria II.B.17 of standard review plan 12.3 - 12.4 provides criteria for the establishment of The locations for fixed specific document continuous area gamma radiation monitors. , ref erenced is ANSI /ANS-HPSSC-6.8.1-1981. The locations and numbers of monitors used at HCGS are not in full compliance with this standard. The location of these monitors are in the vicinity of personnel access areas only. These locations are
' based on the dose assessment and operating experiences from other nuclear power plants. In addition, these locations were finalized prior to the issuance of this standard and provide pn acceptable method of monitoring area radiation levels.
-Inse_ri- s,
'I;.pter.caC;;i...vu ;;...w.; 6-yuise. 1-uilletien evniiw6. Le be HCGS design places the p upstream of the HEPA filters.
venti on monitors downstream of the HEPA filter in orde is assess th lant's effluents. This is achieved best a location as: I a. It is more eff ent to have a gle monitoring point
"*r '"*" ""2*i' '"'
dekk 7__ .
- b. The instrument i ufficienti nsitive to ensure compliance w technical specifi on release limits.
entilation effluent monitors referred t we and
- c. T e HVAC in line monitors (see P& ids in Section .
are scintillation detectors. These monitors are us te d;tect ;;es; ectivity ead ee esch will indieete j i 12.3-43 Amendment 1 DSER OM IM j(g(,(ggg, .
l HCGS FSAR ,8/83 , dele +t l iner Maintenan s in airborn radioactivity concentrations. of iodine concentration within 10 - ours will be assur the use of several met including these monitors, i ant surveys, an etable particulate and iodine ling tors. Grab samples may be obtained from the ystems or the room air by using the portable ers. ese samples are then analyzed in the la tory by mult nel analyzer (MCA). (See ion.12.5 for further ation about MCA). T ore, particulate and iodine sa mon are not provided upstream of the HEPA rs. 12.
3.5 REFERENCES
12.3-1 J.J. Martin and P.H. Blichert-Tof t, " Radioactive Atoms, Auger Electrons, and X-Ray Data," Nuclear Data Tables, Academic Press, October 1970. 12.3-2 J.J. Martin, Radioactive Atoms Supplement 1, ORNL 4923, Oak Ridge National Laboratory, August 1973. 12.3-3 W.W. Bowman and K.W. MacMurdo, " Radioactive Decays Ordered by Energy and Nuclide," Atomic Data and Nuclear Data Tables, Academic Press, February 1970. 12.3-4 M.E. Meek and R.S. Gilbert, Summary of X-Ray and Gamma- Ray Energy and Intensity Data, NEDO-12037, General Electric, January 1970. t 13.3-5 C.M. Lederer, et al, Table of Isotopes, 6th edition, John Wiley, New Yore, 1967 (1st corrected printing March.1968). , 12.3-6 D.S. Duncan and A.B. Spear, " Grace 1 - An IBM 704-709 Program Design for Computing Gamma Ray Attenuation and Heating in Reactor Shields," Atomics International, NAA-SR-3719, June 1959. l 12.3-7 D.S. Duncan and A.B. Spear, " Grace 2 - An IBM 709-Program for Computing Gamma Ray Attenuation and 12.3-44 Amendment 1 l nsER oFEN ITEM l(a[o(fed.!) , l
~
4
Insert Acceptance Criterion II.4.b.3 requires ventilation monitors
.to be placed upstream of HEPA filters. The HCGS design places scintillation detectors in ducts that are tributary to the release vent in order to provide warning of increased releases within the plant. These instruments detect increases in the gross noble gas concentrations of the effluent. Hence, placement of the detectors relative to l HEPA and/or charcoal filters does not significantly affect 1 their response. Since releases of iodines and particulates will be accompanied by much larger releases of noble gases, , - the changes in ventilation monitor readings provide !
indication of a change in airborne activity concentration in ' one or more of the plant's areas. If an increase is i detected, its source and magnitude will be determined using j portable samplers. Normally occupied non-radiation areas in the plant do not have potential for significant airborne concentrations of particulates and iodine during plant operation because:-
- a. The ventilation systems are designed to prevent the' :
spread of airborne radioactivity into normally occupied areas.
- b. Highy radioactive piping / components are not located in normally occupied areas.
Certain activities, such as refueling, solid waste. handling, or _ turbine teardown may . increase the possibility of encoun-
,tering significant airborne activities in some normally occupied areas. Continuous local airborne monitoring will be provided during these activities, as needed.
I Exposure of personnel to high concentrations of airborne l activity in radiation areas will be prevented through - [ in-plant surveys and these portable particulate and iodine h sampling monitors prior to personnel entrance. Continuous l (- monitoring will be provided as required by area conditions and the nature of the entry.f Administrative control will l , 'l prevent inadvertent entry of personnel into normally ___ q unoccupied areas (Zone III and above). The provisions ;IpsEIT Aj discussed above ensure that personnel will not be inadvertently exposed to significant concentrations of airborne activity. Therefore, continuous ventilation radioactivity monitors capable of detecting 10 MPC-hrs of particulate and iodine o from any normally occupied compartments are not provided as permanently. installed equipment. DSER OPEN ITEM l (p (a W-t .
>~==---v~rmew-,,m.-,_,._,, , . _ , _ _ _ _ _ _ , _ , , _ _ _ _ , _ _ _ _ _ _
f .. INSERT A The location of. portable monitors which will be positioned within the station to provide supplemental inplant monitor-ing of particulates-and iodine levels will be provided by
~ July 1, 1985. The positioning of supplemental' continuous air. monitors is part of the Radiation Protection Program and a July 1, 1985 date is consistent with finalizing other details of-the program (i.e'., instrument and equipment cali-bration). The location, quantity, and monitor type will be provided at that time.
RSC:srd I- 9/6/84 M P84 93/10 1 s 5
HCGS FSAR 4/84
- OUESTION-421.10 (SECTION 7.1 & 7.2).
I The staff. believes that the physical separation provided in the
- design of the RPS cabinets may not satisfy the requirements of Regulatory Guid 1.75 or the plant separation criteria and is, therefore, unacceptable. As an example, it has been noted on similar plants that the cabinet lighting and power circuits (which are not treated as associated circuits) becomes associated with Class IE circuits.inside the RPS cabinets. Section 8.1.4.14 includes a brief discussion on the physical separation'provided within. panels, instrument racks and control boards for the instrumentation and control circuits of different divisions.
Review the design of all Class 1E cabinets for separation between non-Class IE and Class IE circuits. Provide the staff with a
- listing of the cabinets which-were reviewed and' describe in detail how physical separation is maintained within the panels, racks and boards for those cases where a 6 inch air space cannot ;
be maintained.. Provide a summary of the analysis and testing ; performed to support this lesser separation. Include.in the discussion the separation provided for associated circuits, internal wiring identification and the use of common terminations.
RESPONSE
The HCGS RPS cabinets (10C609, 10C611, 10C622 and 10C623) meet the requirements of'IEEE Standard 384 as modified and endorsed by Regulatory Guide 1.75, as stated in Section 1.8.1.75. Cabinet lighting and receptacle power circuits are physically separated l from RPS circuits by being routed in metallic conduit or by F structural steel barriers.
- Physical-separation between non-Class IE and Class IE instrumentation and control circuits is provided in panels, instrument racks'and control boards in accordance with IEEE Standard 384, as modified and endorsed by Regulatory Guide 1.75 as stated in Section 1.8.1.75. The following is a listing of Class IE panels, instrument racks and control boards reviewed for the separation requirements of-IEEE Standard 384: ,
Panels-1AC200 H,/0, Analyzer A Panel ' - 1BC200 H,/O, Analyzer B Panel ICC200 H,/O, Analyzer Heat Trace Panel IDC200 H,/0, Analyzer Heat Trace Panel 1AC201 SACS Control Panel A IBC201 SACS Control Panel B 1CC201 SACS' Control Panel C 1DC201 SACS Control Panel D 10C202- RACS Heat Exchanger and Pumps Control Panel. 421.10-1 Amendment 5
- - - , - ,-v,- ,, , , . . , , . . - , - , , -,c -.,,r.+, . , . . , - - - ,--,n,a-,n-w ,w ,..- n,----wm ,-,-,.m.v., ,,,.--,w,n-- ,.w,,-.,,.m.m--,v--
~
HCGS FSAR 4/84 IAC213 Instrument Gas Compressor A Control Panel IBC213 Instrument Gas Compressor B Control Panel 1AC215 H, Recombiner A Power Distribution Panel IBC215 H, Recombiner B Power' Distribution Panel 1AC281- Reactor Building Unit Cooler Control Panel 1BC281 Reactor Building Unit Cooler Control Panel ICC281 Reactor Building Unit Cooler Control Panel-IDC281 Reactor Building Unit Cooler Control Panel
- 1AC285 Reactor Building FRVS Control Panel 1BC285 Reactor Building FRVS Control Panel ICC285: Reactor Building FRVS Control Panel IDC285 Reactor Building FRVS Control Panel 10C286 Reactor Building Equipment Lock Ventilation 10C399 Remote Shutdown Panel 10C401 Diesel Generator Area Battery Room Panel
- 10C402 Diesel Generator Area Battery Room Panel 1AC420 Diesel Generator A Exciter Panel IBC420 Diesel Generator B Exciter Panel ICC420 Diesel-Generator C Exciter Panel IDC420 Diesel Generator D Exciter Panel 1AC421 Diesel Generator A Local Engine Control Panel IBC421 Diesel Generator B Local Engine Control Panel ICC421 Diesel Generator C Local Engine Control Panel 1DC421 Diesel Generator D Local Engine Control Panel 1AC422 Diesel Generator A Remote Control Generator Panel 1BC422 Diesel Generator B Remote Control Generator Panel 1CC422 Diesel Generator C Remote. Control Generator Panel 1DC422 Diesel Generator D Remote Control Generator Panel 1AC423 , Diesel Generator A Remote Engine Control Panel 1BC423 Diesel Generator B Remote Engine Control Panel ICC423 Diesel Generator C Remote Engine Control Panel
- 1DC423 Diesel Generator D Remote Engine Control Panel 1 4 1AC428 Diesel Generator A Load Sequencer Panel IBC428 Diese) Generator B Load Sequencer Panel-ICC428 Diesel Generator C Load Sequencer Panel IDC428 Diesel Generator D Load Sequencer Panel 1AC482 Electric Heater Control Panel 1AVH403 i IBC482 Electric Heater Control Panel IBVH403
! 1AC483 Diesel Area HVAC Control Panel. 1BC483 Diesel Area HVAC Control Panel , l
- 1CC483 Diesel Area HVAC Control Panel l l 1DC483 Diesel Area HVAC Control Panel l' 1AC485 Control Area-HVAC Control Panel l l IBC485 Control Area HVAC Control Panel 1 I
1AC486 Diesel Area Panel Room Supply System
~ 1BC486 Diesel-Area. Panel Room Supply System 1AC487 Water Chiller Panel 1BC487 Water Chiller Panel j IAC488- Chiller AK403 Power Panel 1BC488 Chiller BK403 Power Panel 1AC489 Electric Heater Control Panel 1AVH407 IBC489 Electric Heater Control Panel 1BVH407 421.10-2 Amendment 5
, HCGS FSAR. 4/84 1AC490 Water Chiller A Control Panel 1BC490 Water Chiller B Control. Panel 1AC491 Water Chiller A Power Panel IBC491 Water Chiller B Power Panel 1AC492 Electric Heater Control-Panel 1BC492 Electric Heater Control Panel IAC493 Control Panel - Auxiliary Building Diesel 1AC494 Control Panel - Auxiliary Building Diesel 1AC495 Control Panel - Auxiliary Building Diesel IBC495 Control. Panel - Auxiliary Building Diesel ICC495 Control Panel - Auxiliary Building Diesel
- 1DC495 Control Panel - Auxiliary Building Diesel 1AC515 Traveling Screen Control Panel-1BC515 Traveling Screen Control Panel ICC515 Traveling Screen Control Panel
' 1DC515 Traveling Screen Control Panel IAC516 Service Water Pump Panel 1BC516 Service Water Pump Panel ICC516 Service Water Pump Panel IDC516 Service Water Pump Panel 1AC581 Intake Structure HVAC Control Panel 1BC581 Intake Structure HVAC Control Panel 1CC581 Intake Structure HVAC Control Panel IDC581 Intake Structure HVAC Control Panel 10C601 RRCS Division 1 Panel 10C602 RRCS Division 2 Panel.
10C604 Class 1E Radiation Monitoring Instrumentation Cabinet 10C617 Division 1 RHR and Core Spray Relay Vertical Board
~
10C610 Division 2 RHR and Core Spray Relay Vertical Board
- 10C620 HPCI-Relay Vertical Board 10C621 RCIC Relay Vertical Board 10C622' Inboard Isolation Valve Relay Vertical Board 10C623 Outboard Isolation Valve Relay Vertical Board 10C628 ADS Division 2 Relay Vertical Board 10C631 ADS Division 4 Relay Vertical Board 1AC633 Post LOCA H, Recombiner A Control Cabinet 1BC633 Post LOCA H, Recombiner B Control Cabinet 10C640 Division 4 RHR and Core Spray Relay Vertical Board 10C641 Division 3 RHR and Core Spray' Relay Vertical Board 10C650 Main Control' Room Vertical Board 10C651 Unit Operators Console 1AC652 1E Solid State Logic Cabinet Channel A IBC652 1E Solid State Logic Cabinet Channel B 1CC652 IE Solid State Logic Cabinet Channel C IDC652- IE Solid State Logic Cabinet Channel D 1AC655 IE Analog Logic Cabinet Channel A IBC655- IE Analog Logic Cabinet Channel B
' ICC655 IE Analog Logic Cabinet Channel C IDC655 IE Analog Logic Cabinet Channel D 1AC657 IE' Digital Termination Cabinet Channel A IBC657 IE Digital Termination Cabinet Channel B ICC657 1E Digital Termination Cabinet Channel C 4
421.10-3 Amendment 5
HCGS FSAR 4/84 1 IDC657 IE Digital Termination Cabinet Channel D 1 1AC680 IE Electrical Auxiliary Cabinet Channel A 1BC680 IE Electrical Auxiliary Cabinet Channel B ICC680 1E Electrical Auxiliary Cabinet Channel C IDC680 1E Electrical Auxiliary Cabinet Channel D Instrument Racks 10C002 Reactor Water Clean-up Rack 10C004 Reactor Vessel Level and Pressure A Rack 10C005 Reactor Vessel Level and Pressure C Rack 10C009 Jet Pump Rack A 10C014 HPCI A/HPCI Leak Detection A Rack 10C015 Main Steam C/D and Recirc A Flow Rack 10C018 RHR A and ADS Rack 10C02) RHR B and ADS Rack 10C025 Main Steam C/D and Recirc A Flee Rack 10C026 Reactor Vessel Level and Pressure D Rack 10CO27 Reactor Vessel Level and Pressure B Rack 10C037 RCIC D/RCIC Leak Detection D Rack 10C041 Main Steam A/B and Recirc B Flow Rack 10C042 Main Steam A/B and Recirc B Flow Rack 10C069 RHR D and ADS Rack 10C208A RCIC/ Reactor Cooling 10C211 RCIC Pump 10C212 RCIC Pump Instrument racks are separated into channels. No two redundant piped or tubed safety-related instruments are located on the same rack. Where a 6-inch air space cannot be maintained between instrumentation and control circuits of different channels (both Class 1E to Class 1E and Class 1E to non-Class IE), barriers are provided in accordance with IEEE Standard 384. These barriers are metallic conduit, structural steel barriers, or non-metallic wrap (Havey Industries Siltemp Sleeving Type S or Siltemp Woven Tape Type WT65). The metallic conduit and structural steel barriers are noncombustible materials. The nonmetallic wrap (Siltemp) was successfully tested for use as an isolation barrier j (reference Wyle Laboratories Test Report Number 56669). - For ce ain types of isolati devices, barriers f the type s noted a ve are not feasible. or these cases, r quirements of Section 7 .2.1 of IEEE Standar 384 are met, as llows:
)
"The se ration of the wiring t the input and autput terminal of the isolation devi e may be less t,an 6 inches ( .15 m) as required in .6.2 provided hat it is not less han the distance bet en input and o tput terminals. g Add Tnseel A 421.10-4 Amendment 5
Y4 to
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- /w scar ,4 -
A+ Hess , 16ae. is ola Non Anus 9" "S'l a AicA fla 6 i"'A 4r 50*a do eof sa+1s}Asiy r:eguireme,{ ad, &y , dom ers o-f A fye idex/;Ged shove. are Aof S<asI4/e - %
& ihc2 air space regained is inair-hi>rac/ -Se wiri assoc;< hd w;K Lsa Jewcas c f +q0a dovice e <cyo+
i hetf sala ,.c & sya ,sfbn is niaiafaind [ efis faace 6doveen no+&/<rs hyaf&a aed.t-t,e p/wl a ai ,w,gs 00 +la isolaNem dov/ce . 71au drwas are : A) T 6C- a.tralo isoo ory yt^oda ' / 5L prot /ido s c. e-(c(s s lG .to now- class i E isolc+i% %r loco let/t ' awalog im+3 fo & g(cetco9 eg t
'o') S+v sef%<s'% m Qye 2.I9 relay provicbs
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HCGS FSAR ( , 4/84
~
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Minimum separa ion requirements do ot apply for iring and components ithin the isolation device; howeve , eparation shall be provided whereve practicable." Testin , in accordance ith IEEE Standard 72 (Surge Witn tand Capabil ty) will be per ormed to ensure tha the Class IE nputs to the i olation devices are not affected by short-circuit
! failures, ground faults voltage surges on he output side of the isola ion devices. -
s ( _ So ngI c. Teu k re we y e.
-The enly analysis th:t "'I be performed to support air spaces less than.6 inches, sir.cc the requireeents of--IEEE-St-andard 384 '--
ec 5etisfied, lecfor the Neutron Monitoring System Panel (LOC 608) and the Process Radiation Monitoring System Panels (10C635 and 10C636). % s repor / wa s s u s., su/ . < oce e r 2, cy<e r< < / c c-ove r ( g .t- pt .' H I tco 11 Sc hwersc e r cAoAecl S ejd e m be n 7,19W) No associated circuits have been identified in the non-NSSS panels, instrument racks, or control boards. Internal wiring identification is done using color coded insulation or insulation marked with color coded tape. For panel sections of one channel only, internal wiring identification may not be one. Where common terminations are used, the requirements f IEEE Standard 384 are satisfied as stated above. Electrical equipment and wiring for the reactor protection system (RPS), the nuclear steam supply shutoff systems (NSSSS) and the engineered safeguards subsystems (ESS) are segregated into separate divisions designated I and II, etc., such that no single credible event is capable of disabling sufficient equipment to prevent reactor shutdown, removal of decay heat from the core, or closure of the NSSSS valves in the event of a design basis accident. No single control panel section (or local panel section or instrument rack) includes wiring essential to the protective function of two systems that are backups for each other (Division I and Division II) except as allowed below:
- a. If two panels containing circuits of different separation divisions are less than 3 feet apart, there shall be a steel barrier between the two panels. Panel ends closed by steel end plates are considered to be acceptable barriers provided that terminal boards and wireways are spaced a minimum of one inch from the end plate.
- b. Floor-to-panel fire proof barriers must be provided between adjacent panels having closed ends,
- c. Penetration of separation barriers within a sub' divided panel is permitted, provided that such penetrations are sealed or otherwise treated so that an electrical fire could not 421.10-5 Amendment 5
4/n HCGS FSAR 4/84 reasonably propagate from one section to the other and destroy the protective function.
- d. Where, for operational reasons, locating manual control switches on separate panels is considered to be prohibitively (or unduly) restrictive to normal functioning of equipment, then the switches may be located on the same '
~
panel provided no single event in the panel can defeat the automatic operation of the equipment. With the exception of panels 10C608, 10C635 and 10C636, internal wiring of the NSSS panels and racks has color-coded insulation. Associated circuits are treated within a panel or rack in the same manner as the essential circuits. Where common terminations are used, the requirements of IEEE Standard 384 are satisfied. E w k -l gedwbx a ss e m6 v,s La Le m okbd 6e+ west % A)es ( wcuorr) avv/If5 per twomap' /zo v ac po u L vrs, pome -9:e.ufers a s d20scr; bd A rev is.<d S cceto w 7.6.l. +.z--, J
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I l 421.10-6 Amendment 5 N
. HCGS FSAR 4/84 CHAPTER 7 FIGURES (Cont) Figure No. Title 7.6-2 NMS IED 7.6-3 Detector Drive System 7.6-4 Functional Block Diagram - IRM Channel 7.'6-5 APRM Circuit Arrangement - Reactor Protection System Input 7.6-6 Power Range Monitor Detector Assembly Location 7.6-7 NMS FCD 7.6-8 Redundant Reactivity Control System Initiation Logic 7.6-9 HCGS Redundant Reactivity Control System ARI Valves .
, 7.6-1 De1eted [g 9y;c,,g p7,4c ko^ b5cMblie5 (6'PM l 7.7-1 CRD CD 0*"
p N'Mn NG0b56g E ph 7.7-2 RMCS Block Diagram 7.7-3 Reactor Ma.:ual Control System Operation 7.7-4 Reactor Manual Control Self-Test Provisions 7.7-5 Eleven-Wire Position Probe 7.7-6 Recirculation Flow Control 7.7-7 Feedwater Control System 7.7-8 Simpli'ied Diagram Turbine Pressure & Speed Load Control Requirements 7.7-9 Deleted l s' s 7-x Amendment 5 i
HCGS FSAR I' coaxial cable. The amplifier is a linear current amplifier whose voltage output is proportional to the current input and therefore propo.rtional to the magnitude of the neutron flux. Low level
- output signals are provided-that are suitable as an input to the computer, recorders, etc. The output of each LPRM amplifier is 3
isolated to prevent interference of the signal by inadvertent grounding or application of stray voltage i t the signal terminal point. rom. (sa. FNun f+)'3 Std.3) - Power for the LPRM is supplied two non-Class 1E )
- uninterruptible power sources ApproIlmately half of the LPRMs are supplied from each bus. Each LPRM amplifier-has a separate power supply in the main control room, which furnishes the detector polarizing potential. The LPRM amplifier cards are mounted into pages in the NMS cabinet, and each page is suppl,ied operating voltages from a separate low voltage power supply.
- M/SE:&- A -
The trip circuits for the LPRM provide signals to actuate lights
-and annunciators. Table 7.6-3 lists the LPRM trips.
l
' Each.LPRM may be individually bypassed via a switch on the LPRM amplifier card. Placing an LPRM in " bypass" sends a signal to the assigned APRM, electronically causing it to adjust its averaging amplifier's gain to allow for one less LPRM input. In this way, each APRM can continue to produce an' accurate signal
. representing average core power even if some of the assigned LPRMs fail during operation. If the number of functional assigned LPRMs drops to 50% of the normal number, the APRM automatically goes inoperative and a half scram (one trip logic channel deenergized), rod block, and appropriate annunciation are generated. Administrative controls ensure that a minimum number of LPRMs at each level (A, B, C, and D) in the core.are maintained or the APRM is declared inoperative and manually placed in the tripped state.
~
In addition to the signals supplied to the APRMs, the LPRMs also send-flux signals to the rod block monitor (RBM). When a central control rod is selected for movement, the output signals from the amplifiers associated with the nearest 16 LPRM detectors are displayed on the main control room vertical board meters and sent to the RBM. The four LPRM detector signals from each of the four detector assemblies are displayed on 16 separate meters. The ,. operator can readily obtain readings from all'the LPRM detectors by selecting the control rods in order. These' signals from the 7.6-10 4
,--w-- t .,y.+, ,,w,, , - , , - -,,,,w, v.-ww,-,...-,,,,,-% y, - , , . , , , - . - - ,
4 INSERT A Electrical protection assemblies (EPAs) identical to those used in the reactor protection system (RPS) . (described in Section 8.3.1.5.4) are installed between the power range NMS and the two 120V AC
~
feeders from the UPS power sources (see Figure 7.6-11). The EPAs ensure that the power range NMS never operates under degraded bus voltage or frequency conditions (undervoltage, overvoltage, underfrequency). The power range NMS panel (10C608) was analyzed with this power suprly configuration to ensure that no single f ailure of the power range NMS could inhibit the proper operation
~
of the reactor protection system or any other safety system required for the safe operation of the plant. The interf aces between the power range NMS and the RPS have adequate provisions for separation. The RPS cabling external to the NMS panel conforms to the separation p idelines-of Regulatory Guide 1.75, which the RPS must satisfy. fithin the panelj ydhere the cable and wiring runs to the different RPS divisions do not conform to the Regulatory Guide 1.75 separation criteria, fire-resistant "Sil-Temp" tape is wrapped around the cables and wires. This eliminates the possibility of fault propagation between the RPS divisions. In accordance with paragraph 5.6.'2 of IEEE Standard'384, this tape has been demonstrated to be acceptable. , 4 -1
l i HCGS FSAR l four LPRM strings (16 detectors) surrounding the selected rod are used in the RBM to provide protection against local fuel n overpower conditions. Average Power Range Monitor Subsystem 7.6.1.4.3 . The APRM subsystem-monitors neutron flux from approximately 1% to above 100%-power. There are six APRM channels, each receiving core flux level signals from 21 or 22 LPRM detectors. Each APRM : channel averages the 21 or 22 separate neutron flur signals.from the LPRMs assigned to it,'and generates a signal representing core average power.
- This signal is used to drive a local meter and a remote recorder located on the main control room vertical board. It is also applied to a trip unit to provide APRM downscale, inoperative and upscale alarms, and upscale reactor trip signals for use in the RPS or RMCS.
, Refer to Section 7.2.1.1 for a description of the APRM inputs to the RPS, and Figure 7.6-5 for the RPS trip circuit input arrangement. APRM trips are summarized in Table 7.6-2. The APRM scram units are set for a reactor scram at 15% core power in " refuel" and "startup" modes. When the mode switch is L in "run," the APRM trip reference signal is provided by a signal that varies with recirculation flow. This provides.a power
- following reactor scram setpoint. As power increases, the L reactor scram setpoint also increases up to a fixed setpoint
- above 100%. Reactor power is always bounded with a reactor scram, yet the change in power required to generate the reactor scram does not vary greatly with the operating power level.
Provision is made for manually bypassing one'APRM channel at a time. Calibration or maintenance can be performed without tripping the RPS. Removal of an APRM channel from service without bypassing it, by unplugging a card, by taking the APRM function switch out of " operate," or by having too few assigned LPRM signals to the APRM, will result in an APRM " inoperative" condition which causes a -half scram, a rod block, and annunciation-de 54%ut [ non-Class 1E uninterruptible The APRM channels receive power from, drip unit is supplied from power sourceg. Power for each APRM Y {v.d s ufply Oe.DRN5(Sed fy 7 7* b I + 7-) -
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