ML20079P589

Forwards Responses to NRC Request for Addl Info Re Generic Ltr 82-14 Concerning Structural Engineering Audit

ML20079P589
Person / Time
Site:	Satsop
Issue date:	01/17/1984
From:	Sorensen G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Knighton G Office of Nuclear Reactor Regulation
References
GL-82-14, GO3-84-30, NUDOCS 8401310276
Download: ML20079P589 (155)

Text

{{#Wiki_filter:Washington Public Power Supply System Box 1;223 Elms. Washington 98541 1223 (2061482-4428 Docket No. 50-508 January 17, 1984 G03-84-30 Director of Nuclear Reactor Regulation ATTN: Mr. G. W. Knighton, Chief I Licensing Branch No. 3 U. S. Nuclear Regulatory Comission l Washington, D. C. 20555 L

Subject:

NUCLEAR PROJECT 3 RESPONSES TO NRC QUESTIONS i

Reference:

a) Letter #G03-83-0889, G. C. Sorensen to G. W. Knighton, dated November 18, 1983. [ In accordance with the guidar.ce of Generic Letter 82-14, the Supply System hereby submits 40 copies of responses to the NRC's requests for Additional Information as shown. In preparing this submittal it was necessary to include several full Since as a practical matter it is quite difficult to size drawings. include a copy of each for each copy of this letter, the NRC Licensing Project Manager will receive three copies of each for Dittribution. l In reference a) the Supply System indicated that a response to one item deriving from the Structural Engineering Branch audit, held in September 1983 at the offices of Ebasco Services Inc. in New York Due to the City, would be forthcoming in this group of submittals. interre it is more appropriate to submit the responses to these items as a Consequently the response to Audit Finding 19 will be delayed pending a complete response for all of the Structural Engineering group. Branch Audit Findings. This situation has been discussed with the NRC Licensing Project Manager for WNP-3. g\\ 8401310276 840117 (h0go PDR ADOCK 05000500 a roe i

Mr. G. W. Knighton Page 2 NUCLEAR PROJECT 3 RESPONSES TO NRC QUESTIONS If you require additional infonnation or clarification, the Supply System Point of Contact for this matter is Mr. D. W. Coleman, Licensing Project Manager (206/482-4428 ext. 5436). Sincerely,

L ' =)

' G. C. Sorensen, Manager Regulatory Programs AJM/kh Attachments:

1. item 10 2.

NRC Question No. 210.03 3. NRC Question No. 220.18 4. NRC Question No. 220.25 5. NRC Question No. 220.26 6. NRC Question No. 220.30 7. NRC Question No. 220.37 8. NRC Question No. 241.20 9. NRC Question No. 270.01

10. NRC Question No. 271.01 11.

NRC Question No. 281.06

12. NRC Question No. 311.03 13.

NRC Question No. 410.40

14. NRC Question No. 430.23 15.

NRC Question No. 450.01

16. NRC Question No. 471.01 17.

NRC Question No. 471.12

18. NRC Question No. 471.23 19.

NRC Question No. 471.25

20. NRC Question No. 480.12

21. NRC Question No. 480.19 l

22. NRC Question No. 480.24 23.

NE Question No. 480.26

24. NRC Question No. 640.16 cc:

P Christofakis - Ebasco NY0 N. S. Reyncids - 0 & L J. A. Adams - NESCO D. Smithpeter - BPA A. Vietti - NRC A. A. Tuzes - CE Ebasco - Elma WNP-3 Files

REQUEST FOR ADDITIONAL INFORMAT10N - ENCLOSURE 4 10) EFFECTS OF CONTAINMENT C0ATINGS AND SUMP DEBRIS ON ECCS AND CONTA SPRAf OPERATION A copy of the staff concerns on this issue, including a request for additional information which has been sent to a number of OL applicants, is provided as 0 (Attachment I).

Response

The Supply System responded to this question in letter #G03-82-1228 dated November 30, 1982. That response contained one open item pertaining to the amount, types and Manufacturers of the thermal insulation to be used at WNP-3. Subsequently, the Supply System indicated (by letter #G03-83-889, dated November 18,1983) that a response to this remaining open item muld be forthcoming by December 1983. The Supply System has not awarded a contract for the purchase of thermal insulation as yet. Since this will not occur until a resumption of full construction activity at WNP-3, any response at this time would be premature. The Supply System will provide the requested information when it becomes available from the vendors. l ~ l i -.__,_v_. ,_._..~.,.-,,_._-.o

n. I Question No. 210.3 Either supply the information idertified as later in Appendix (3.9.38) 3.9.38 or provide a schedfle for submittal of this infomation.

Response

In letter G03-82-1085, from G. D. Bouchey to J. D. Kerrigan dated October 22, 1982, a commitment was made to provide the information identified as "later" within Appendix 3.9.38 by September 1983. Attached are tipdated pages of Appendix 3.9,38 providing some of the missing information. The rest of the information will be provided upon receipt from the valve manufacturer. l ? -..e ,-e,,. ,e- - - -- -, -, r-,-- -n-, v w- --- +- - - - - -, n,-

Qggd6yb. WNP-3 FSAR j m,s* DESIGN / SEISMIC QUALIFICATI0d REPORTS TCR A/E SUFFLIED ACTIVE VALVES Manufacturer's Valve Tas No. Manufacturer Resort No. 2AF-VD035SA Target Rock Corp T 28 l 2AF-VD036SB Target Rock Corp TRP-2%1 Rev B l 2AF-VD0395A Target Rock Corp TRP-2341 Rev 3 2AF-vD040SB Target Rock Corp TRP-2341 Rev 3 l 3AF-VD098SB Target Rock Corp TRP-234' Rev B I 3AF-VD0995A Target Rock Corp TRP-2341 Rev B \\ 3AF-VD100SB Target Rock Corp TRP 2341 Rav B l 3AF-VD101SA Target Rock Corp TRP-234J Rev B 2BD-VE037SA Borg-Warner Corp v-2BD-VE038SA Borg-Warner Corp 2BD-VE039SB Borg-Warner Corp 2BD-VE040SB Borg-Warner Corp 2BD-VR033SA Borg-Warner Corp 2BD-VR034SA Borg-Warnec Corp 2BD-VR035SB Borg-Warner Cor; 2BD-VR036SB Borg-Warner Corp 2BD-VR077SA Borg-Warner Corp 2BD-VR078SA Borg-Warner Corp l l 2BD-VR07953 Borg-Warner Corp 2BD-VR080SB Borg-Warner Corp 2BD-VR089SA Borg-Warner Corp f ~ 2BD-VR090SA 3org-Warner Corp = i.,

• Later 3.9.33-1

h:A.Lv h n We. WNP-3 ,, '. T' ~ FSAR DESIGN / SEISMIC QUALIFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Valve Tan No. Manufacturer Report No. 2BD-VR091SB Borg-Warner Corp 2BD-VR092SB Borg-Warner Corp 3CC-5006SA Litton Contramatics No. 14225-5A Rev 1 3CC-B005SB Litton Contromatics . No. 14225-5A Rev 1 3CC-B066SN Litton Contromatics No. 14225-5A Rev 1 3CC-B0675N Litton Contromatics No. 14225-5A Rev 1 3CC-3068SN Litton Contromatics No. 14225-5A Rev 1 3CC-B0695N Litten Contromatics No. 14225-5A Rev 1 3CC-B507SA Litton Contromacies No. 14225-2A Rev 1 3CC-B508SA Litton Contromatics No.14225-2A Rev 1 3CC-3509SB Litton Contromatics No. 14225-2A Rev 1 3CC-3510SB Litton Contromatics No.14225-2A Rev 1 '# # ~ 3CC-3511SA Litton Contromatics No ' ##~ 3CC-B512SB Litten Contromatics 3CC-B513SA Litton Contromatics No.14225-2A Rev 1 3CC-BS14SA Litten Contromatics No.14225-2A Rev 1 3CC-B515SB Litton Contromatics No. 14225-2A Rev i 3CC-B516SB Litton Contromatics No.14225-2A Rev i 2CC-B521SB Litton contrematics No.14225-2A Rev 1 2CC-BS22SB Litten Controcatics No.14225-2A Rev 1 2CC-3523SA Liccon Contro=acies No. la225-2A Rev 1 2CC-3524SA - Litton Contromacies No. 14225-2A Rev 1

• Later l

3.9.33-2 l

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2. s C. 3 DESIQt/ SEISMIC QUALIFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Valve Tan No.

Manufacturer Reoort No. 2CC-B525SB Litton Contromatics No.14225-2A Rev 1 2CC-B526SB Litton Contromatics No.14225-2A Rev 1 SCC-B531SA Litton Contromatics Ko. 14225-2A Rev 1 3CC-B532SB Litton Contromatics No. 14225-2A Rev 1 3CC-55395A Litton Contramatics No. 14225-5A Rev 2 1 3CC-5540SB Litton Contromatics No.14225* 5A Rev 2 3CC-VE255SB Target Rock Corp TR26pRevB 3CC-VE535SA Target Rock Corp TR 2684 Rev B TR 2684 Rev B 3CC-VE536SB Target Rock Corp TR 2684 Rev B 3CC-VE638SA Target Rock Corp 2CE-7P011 SBR Borg-Warner Corp NSR-307FCB6-1 Rev B 2CH-VP029SAR Borg-Warner Corp NSR-107DCB6-1 Rev B 2 G-VP030 SBR Borg-Warner Corp NSR-107DCB6-1 Rev B 2CH-VSO41SA Borg-Warner Corp NSR-304ECB6-1 Rev A 2G-VSO42 SBR Borg-Warner Corp NSR-306XCB6-1 1 2G-VSO44SA Borg-Warner Corp NSR-306xCB6-2 i 2CE-VSO47 SBR Borg-Warner Corp NSR-304CCB6-1 l NSR401HDB4-001 2G-VWO37SAR Borg-Warner Corp i:5-$$ i a Cup ' MM dC' 2G-VWO40 SBR 2G-VW401 SBR Atwood & Morrill No. 62 Rev 0 2CR-VW404SAR Atwood & Morrill No. 62 Rev 0 U EW M*"O*"

• 3.9.3B-3

( Q,, w.o b % b _. WNP-3 z,p,3 FSAR DESIGN /SRISMIC QTIALTFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Report No. Manufacturer Valve Tar No., SRV-1502 Rev 1 2CS-8001SA McNally Pittsburg Mfg Corp SEV-1502 Rev 1 2CS-8002SB McNally Pittsburg Mfg Corp of TRP-2 Rev B 2CS-M005SAR Target Rock Corp TRP-2 Rev B 2CS-M OO6 SBR Target Rock Corp No. 34 Rev B 2CS-VS015SAR Atwood & Morrill 2CS-VS016 SBR Atwood & Morrill No. 34 Rev B No. 31 Rev C 2CS-VS0175AR Atwood & Morrill 2CS-VS018 SBR Atwood & Morrill No. 31 Rev C No. 32 Rev A 2CS-VS021SAR Atwood & Morrill No. 32 Rev A 2CS-VS022 SBR Atwood & Morrill No. 34 Rev B 2CS-VSO23SAR Atwood & Morrill l No. 34 Rev B 2CS-VSO24 SBR Atwood & Morrill No. 31 Rev C 2CS-VS079SAR Atwood & Morrill No. 31 Rev C 2CS-VS080 SBR Atwood & Morrill N6. 33 Rev B 2CS-VSO91SA Atwood & Morrill No. 33 Rev 3 2CS-VSO92SB Atwood & Morrill NSR-402HDB3-001 2CS-VUO55SAR Borg-Warner Cory 1 NSR-402HDB3-001 2CS-VUO56 SBR Borg-Warner Corp No. 14225-3A Rav 1 3EC-3001SA Litton Contromatics No. 14225-3A Rev 1 3EC-30033A Litton Contromatics No. 14225-3A Rev 1 3EC-B004SA Litton Centromatics No. 14225-3A Rev 1 3EC-3007SB Litton Contrematics Amendment No. 1, (10/82) 3.9.3B-4

h M *.i. /! - ? WEP-3 FSAR g g,3 ~ DESIGN / SEISMIC QUALIFICATICN REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Reoort No. Valve Tar No. Manufacturer 2PV-3025SA McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-30295A Litton Contromatics No. 14225-1A Rev 1 3PV-3033SA McNally Pittsburg Mfg Corp SRV-1502 Rev 1 SRV-1502 Rev 1 3PV-B034SA McNally Pittsburg Mfg Corp 3PV-B035SA McNally Pittsburg Mfg Corp SRV-1502 Rev 1 3PV-B036SA McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-B0375A Litton Contramatics No. 14225-3A Rev 1 2PV-B038SA Litton Contromatics No. 14225-3A Rev 1 ^ r-2PV-B039sA Litton contramatics No. 14225-2A Rev 1 V 2PV-B041SA Litton Contramatics fv'o M M S- /*d C f*i 2PV-3042SA Litton Contromatics l 2PV-B043SA Litton Contromatics 2PV-3044SA Litton Contromatics 1 2PV-3045SA Litton Contrematics 2PV-8046SA Litton Contramatics 2PV-B0475A Litton Contromatics J 2PV-B048SA Litten Contromatics 2PV-3052SA Litton Contramatics No. 14225-3A Rev 1 227-B054SA Litton Contromatics No. 14225-% Rev 1 i 2PV-B055SA Litton Contromatics No. 14225-3A Rev 1 37V-5060SA Litton Contromatics No. 14225-1A Rev 1 3PV-BC61SA Litten Contromatics No. 14225-1A Rav 1 (.. 6 ater L v 3.9.3B-9

b O nr-3 PSAR k 0 .,0 ,i DESIGN /SRISMICQUALIFICATICNREPOR2S FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's j Valve Tee No. Manufacturer / Report No. 3PV-B062SA Litton Contrematics No. 14225-1A Rev 1 3N -5063SA Litton Contrematics No. 14225-1A Rev 1 2?v-5064SA Litton Contramatics No. 14225-4A Rev 1 2PV-B101SB McNally Pittsburg g Corp SRV-1502 Rev 1 2PV-B103SB McNally Pittsbu g Mfg Corp SRV-1502 Rev 1 2PV-B104SB McNally Pitt urg Mfg Corp SRV-1502 Rev 1 2PV-B105SB McNally Pi tsburg Mfg Cory SRV-1502 Rev 1 2PV-B1075B Litton C tromatics No. 14225-2A Rev 1 2PV-B1095A Litt Contromatics No. 14225-3A Rev 1 2PV-B110SB Lit Contromatics No. 14225-3A Rev 1 2PV-B111SB ally Pittsburg Mfg Cory SRV-1502 Rev 1 2PV-B112SA Tally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-B113SB McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-B114SB McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-B123SB Litton Contromatics No. 14225-4A Rev 1 2PV-B124SA Litton Contromatics No. 14225-4A Rev 1 2PV-B125SB McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-3129SB Litton Contramacies No. 14225-IA Rev 1 l SRV-1502 Rev 1 3PV-B1335 McNally Pittsburg Mfg Corp 37V-B1348B McNally Pittsburg Mfg Corp SRV-1502 Rev 1 3PV-31 SB McNally Pittsburg Mfg Corp SRV-1502 Rev 1 / 3PV-31/36SB McNally Pittsburg Mfg Corp SRV-1502 Rev 1

• Later 3.9.3B-10

~- 0kBS k%s b_. WNP-3 290.3 pggy DESIGN / SEISMIC QUALIFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVE 3 Manufacturer's Valve Tar No. Manufacturer Resort No. 2PV-B137SB Litton Contromatics No. 14223-3A Rev 1 2PV-B138SB Litton Contromatics No. 14225-3A Rev 1 2PV-B140SB Litton contramatics h. 14225-2A Rev 1 / 2##'# # 2PV-B141SB Litton Contramatics No

• I 2PV-B142SB Litton Contramatics 2PV-B143SB Litton Contromatics 2rV-B144SB Litton Contramatics 2PV-B145SB Litton Contromatics I

2PV-B146SB Litton Contromatics %) 2PV-B147SB Litton Contromatics 'PV-B148SB Litton Contramatics 2 i 2PV-B152SB Litton Contromatics No. 14225-3A Rev 1 l l No. 14225-3A Rev 1 2PV-B154SB Litton Contromatics No. 14225-3A Rev i 2PV-B155SB Litton Contromatics No. 14225-1A Rev i i 3PV-B160SB Litton Contromatics No. 14225-2A Rev i 3PV-B161SB Litton Contromatics No. 14225-IA Rev 1 3PV-B162SB Litton Contromatics No. 14225-1A Rev 1 3PV-B163SB Litton Contromatics No. 14225-4A Rev 1 2PV-3164SB Litton Contromatics ISI-VP091SAR Atwood & Morrill 361-14215-00 ISI-VP092SAR Atwood & Morrill 361-14215-00 1 ISI-VP097 SBR Atwood & Morrill 361-14215-00 m C* Later, 3.9.33-11 Amendment No.1, (10/82) i l

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DESIQt/SEI2fIC QUALIFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Valve Tan No. Manufacturer Recort No. ISI-VP098 SBR Atwood & Morrill 3 u s - t4 ;. i S~- c c 1SI-VP104SAR Atwood & Morrill 3 9 o - i <# 2 e 5~ - C O I 1SI-VP107SAR Atwood & Morrill ISI-VP110 SBR Atwood & Morrill ISI-VP113 SBR Atwood & Morrill f ISI-VP185 SBR Borg-Warner Corp ISI-VP186 SBR Borg-Warner Corp 2SI-VQO11SAR Borg-Warnar Corp 2SI-VQO27 SBR Borg-Warner Corp ^ 2SI VQO33SA Borg-Warner Corp ,esA-ya9xc.33- > ( -a l 2SI-VQO35SB Borg-Warner Corp .r 2SI-VQ046SA Borg-Warner Corp l 2SI-vQ049SB Borg-Warner Corp 2SI-VQ087SA Target Rock Corp ne n c.y ze oa 6 NovB TRP-23 2SI-vQ088SB Target Rock Corp 2SI-VQ115SA Borg-Warner Corp ~ 2SI-VQ170SA Target Rock Corp T-g p 23 eq 2.,v 6 l fI y 2SI-VQ171SB Target Rock Corp T32-23biev 3 2SI-VQ172SAR Target Rock Corp II l TRP-23(J+ ' Rev 3 2SI-VQ173 SBR Target Rcck Corp TR 2684 Rev 3 2SI-VQ201SAR Target Rock Corp TR 2684 Rev B 2SI-VQ202 SBR Target Rock Corp TR 2684 Rev B 2SI-VQ2f"SA Target Rock Corp

• Later 3.9.3B-12

M@k.*//b. WNP-3 '5^^ z.,c.3 DESIGN / SEISMIC QUALIFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVES Mtnufacturer's Velve Tag No. Manufacturer Report No. 2SI-VQ204SB Target Rock Corp TR 2684 Rev 3 TR 2684 Rev 3 2SI-VQ2075BR Target Rock Corp TR 2684 Rev B 2SI-VQ208SAR Target Rock Corp TR 2684 Rev 3 2SI-vQ209SB Target Rock Corp TR 2684 Rev B 2SI-VQ210SA Target Rock Corp 2SI-VS054SAR Nissho-Iwai American Corp W5A ott Re/o 2SI-VS060 SBR Nissho-Iwai American Corp TR 2684 Rev 3 2SI-VS131SA Target Rock Corp TR 2684 Rev B 2SI-VS134SA Target Rock Corp TR 2684 Rev B 2SI-VS137SA Target Rock Corp TR 2684 Rev 3 2SI-VS140SA Target Rock Corp TR 2684 Rev 3 2SI-VS193SB Target Rock Corp TR 2684 Rev B 2SI-VS194SB Target Rock Corp TR 2684 Rev 3 2SI-VS195SB Target Rock Corp TR 2684 Rev 3 2SI-VS196SB Target Rock Corp No. 39 Rev A 2SI-VUOO1SAR Atwood & Morrill No. 39 Rev A 2SI-VUO21 SBR Atwood & Morrill No. 38 Rev 3 2SI-VUO55SAR Atwood & Morrill No. 38 Rev B 2SI-VUO61 SBR Atwood & Morrill i 1 2SL-VP102SABR Borg-Warner Cor? NSR-107XCB6-1 v

• Later 3.9.33-13 Amendment No. 1. (10/82)

b'@ e /e - f WNP-3

2. O.3 i

FSAR i J DESIGI/ SEISMIC QUALIFICATION REPORTS FCR A/E SUPFLIED ACTIVE VALVES Manufacturer's Report No. Panufacturer Valve. Tar No. 2SL-VP103SABR Borg-Warner Cor? NSR-107XCB6-1 2SL-VP104SABR 3org-Warner Corp NSR-107XCB6-1 1 2SL-VP105SABR Borg-Warner Corp NSR-107XCB6-1 I l l J L Anendment No. 1, (10/82) 3.9.3B-14

P Question No. I The rationale for disregarding rotational and out-of-plane 220.18 (SRP 3.7.2, translational degrees of freedom, in the justification of use of II, 3, d two-dimensional seismic models, is considered to be incomplete. FSAR Provide the numerical bases to support your position that these 3.7.2.3) degrees-of-freedom are believed not significant.

Response

By letter dated November 2,1983, frcm G. W. Knighton to 1 D. W. Mazur, the NRC transmitted the results of an audit conducted September 26 through 30, 1983 by staff reviewers at the offices of Ebasco Services Inc. in New York City. The Supply System considers NRC Questions Number 220.18 to be sub-stantially similar to the staff's Finding Number 17. Accordingly, the Supply System response to Question Number 220.18 will be com-bined with the response to Audit Finding 17 which has been sched-uled for March, 1984.

Question No. . 220.25 Provide additional detailed sketches or production drawings of (SRP 3.8.2, the following: II, 1-(Descrip-1. Area where the steel containment is embedded in the tion of the basemat and particularly where the steel shell meets Containment) the concrete. FSAR 3.8.2.1) 2. Details of the personnel hatch sad equipment hatch including attachment and welding to the containment shell. 3. Details of the crane girder including its attachment to the steel containment including welding.~ 4. Details and material descriptions (including radiation resis-tance and aging characteristics) of all seals.

Response

1. The Chicago Bridge and Iron (CBI) production drawings of the Containment Vessel showing where the steel containment is embedded in the basemat listed below are attached in response to above request: 74-3431-1 General Plan 74-3431-2 Basic Curve 74-3431-3 Bottom Head Plan 74-3431-4 Bottom Head Section 2. The CBI production drawings of the Containment Vessel listed below showing details of the personnel and equipment hatches are attached in response to above request: 74-3431-100 General Arrangement for Personnel Lock 74-3431-135 Sub-Final Assembly for Personnel Lock 74-3431-61 Insert Assembly for Personnel Lock i 74-3431-69 14' Dia Maintenance and Equipment Hatch 74-3431-70 14' Dia Equipment and Maintenance Hatch The CBI production drawings of the Containment Vessel listed 3. below showing details and of the crane girder are attached in response to above request: 74-3431-13 Polar Crane Rail and Girder 74-3431-14 Polar Crane Rail and Girder Details 74-3431-15 Crane Girder Assembly 74-3431-18 Crane Girder Assembly v.u., ,.-- -,, ~, -. - - <a-.,,,,,m,,,-y.-,-,., ,,,,,,,a, ,,,-,,,-,, - - - -,-, + ,e-_,_-,

Question No. 220.25 Response (Cont'd) 4. a) Details of the containment seals are prcvided in the drawings listed below which are attached for your use: EMDRAC Drawing No. 3240-34321 SkV Medium Voltage Penetration EMDRAC Drawing No. 3240-33737 15kV Medium Voltage Penetration EMDRAC Drawing No. 3240-34320 Low Voltage Power Control and Instrumentation Penetration CBI Drawing No. 74-3431-69 14' Dia Maintenance & Equipment Hatch CBI Drawing No. 74-3431-100 General Arrangement - Personnel Lock CBI Drawing No. 74-3431-135 Sub-Final Assembly - Personnel Lock CBI Drawing No. 74-3431-136 Sub-Final Details & Section For Personnel Lock CBI Drawing No. 74-3431-137 Sub-Final Misc. Sections & Details For Personnel Lock CBI Drawing No. 74-3431-147 Shaft Seal Assembly For Personnel Lock CBI Drawing No. 74-3431-148 Valve Assembly For Personnel Lock CBI Drawing No. 74-3431-200 General Arrangement - Escape Lock CBI Drawing No. 74-3431-202 Lock Sub-Final Assembly - Escape Lock CBI Drawing No. 74-3431-203 Misc. Details - Sub-Final Assembly - Escape Lock CBI Drawing No. 74-3431-230 Sub-Final Misc. Sections & i Details For Escape Lock CBI Drawing No. 74-3431-246 Escape Lock Shaft Seal i Assembly b) Material Descriptions of Seals i - Electrical Penetration Seals I Electrical penetrations for low and medium voltage L cables are sealed by Silicone and Ethylene Propylene "0" rings as shown on the attached EMDRAC (Westing-house) drawings. The entire module, including the r {

1 i Question No. 220.25 Response (Cont'd) seals, are tested and qualified by the Venaar (Westinghouse) in accordance with IEEE Standard 323-1974 for 40 years of installed life while oper-ating at 70*C (158'F) plus one year DBA conditions. The seals show no degradation after exposure to 2.1 x 108 Rad of gamma radiation. Details of the elec-trical penetration seal qualification will be submit-ted as part of the WNP-3 Equipment Qualification Program. ii - Maii.tenance and Equipment Hatch, Personnel Lock and Escape Lock Seals The material used for the permanent seals of the 14 ft diameter Maintenance and Equipment Hatch and the doors for the Personnel and Escape Locks of the Con-tainment Vessel is silicone rubber, General Electric compound SE-555 (Methyl Phenyl Vinyl Sil1 cone). The initial properties and effects of gamma radiation on GE compound SE-555 is shown in Table 1 obtained from a technical service report by General Electric on silicones, entitled " Nuclear Radiation Resistance of Silicone Rubber" dated November 30, 1960. The material used for the ball seal in the relief valve on the locks is DuPont Nordel, an ethylene pro-I The initial properties and effects of pylene rubber. gamma radiation on DuPont Nordel (EPDM) is shown in Table 2 obtained from a paper titled " Insulations and Jackets for Control and Power Cables in Thermal Reac-tor Nuclear Generating Stations", authored by R. B. Blodgett and R. G. Fisher and presented before the IEEE Summer Power Meeting, Chicago, 1968. The material used for the flange seal of the shafts for the locks is Tefzel, a DuPont plastic (Fluoro-polymer). The effect of radiation dose on room temperature tensile strength and elongation retention is shown in Figure 1 obtained from the Design Hand-book on Tefzel by DuPont. l I

Question No. 220.25 Response (Cont'd) TABLE 1 INITIAL PROPERTIES AND PERCENT CHANGE Dose Hardness Elongation Tensile (r v in6) shore A; A% %; a % psi ; A% 0 60 590 1590 5 6.7 -24.1 -7.1 10 16.7 -36.4 -8.6 50 25.0 -86.4 -41.3 100 45.0 -91.5 -51.6 l l l l

Question No. 220.25 Response (Cont'd) TABLE 2 INITIAL PROPERTIES AND PERCENT RETENTION AFTER IRRADIATION Dose 200% Modulus Tensile Strength Elongation psi (rads) psi _ 0 1033 1433 470 5 x 105 100 104 111 5 x 106 94 97 102 5 x 107 120 93 47 1 x 108 (Less than 200%) 79 32 l . ~.

G n c.2 7 i 1 Figure T./TEFZEle 200 - EFFECT OF RADIATION DOSE ON h00M TEMPERATURE TENSli.E STRENGTH & . ELONGATION RETENTION (By ASTM D 638. Note: No ) measurable :liftere.we if the ambiest atmosphere is air or nitroges. M e i 1 ~ -i i I i l I l l I m I I g t t l - i I. i 1 "e w e e ."a e i

Y a no. Ai l9 Figure 1/TEFZELe 200- EFFECT OF RADIATION DOSE ON N00M TEMPERATURE TENSILE STRENGTH & . ELONGATION RETENTION (By ASTM D 838). Note: No asseursbie differesca if the ambient atmosphere is air er nitrogen. I w - 1 3 i ..I f l -1 .t i l l i i l l t - 8 i i i l-l i l 8 e l e I i i i l i i i L l l \\

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Question No. 220.26 In the description of computer programs used for the con-(SRP 3.8.2, tainment stress analysis no validation information is pre-II, 4, c sented. Provide validation information for each program (Refers to as outlined in the staff's position in the referenced section 3.8.1, II, of NUREG-0800 (SRP). 4,e) FSAR 3.8.2.4.5)

Response

The following computer programs are used by CBI in the stress analysis of the containment vessel: a) Program 405 b Program 7 81 4 c Program 1017 d Program 1027 e) Program 1036M f) Program 1392 Validation information for each computer program is as follows:

1) CBI proprietary Program 781 is a version of a recognized pro-gram in the public domain and has had sufficient history of use to justify its application and validity without further demonstration. This proprietary program / methodology was also used for St. Lucie II (Florida Power & Light Co.) and its approval by the staff was benchmarked via St. Lucie II's Safety Evaluation Report (NUREG-0843).

2) CBI proprietary Programs 405, 1017, 1027, 1036M and 1392 have been demonstrated to obtain solutions substantially identical i

to solutions obtained from classical problems, accepted exper-imental tests, or analytical results published in technical literature. The test problems are hand solutions and/or com-parisons to solutions developed by other computer programs, such as 781. The use of the programs are within the range of applicability of the classical problem analyzed to justify acceptance of the program. These programs / methodologies were used for St. Lucie II and/or Waterford 3 and approval by the staff was benchmarked via their Safety Evaluation Reports (WUREG-0843 and NUREG-0787). e

l Question No. It was noted on Figure 3.8.3-1 of the FSAR that the applicant E20.30 l (SSP 3.8.3, is installing a foam material behind the upper structural II, 1 support for the reactor vessel. The following information is 3.8.3-1 ) requested for such materials. 1. A complete list of places where such plastic foam type materials is to be used as fillers or separators in Cate-gory I structural applications where material is left in place. Description of material characteristics (including aging, 2. potential thermal deterioration effects, potential radia-tion deterioration). 3. Purpose of these materials for each use. Item 1 & 3) - A listing of places where plastic foam type mate-

Response

rial has beest used along with their purpose is described below: Location of Filler Material Type of Filler Material Purpose a) Behind Reactor Vessel SE - F0AM, Filler material used to Horizontal Supports Manufacturered by BISCO create a gap behind (See FSAR Fig. 3.8.3-1) (Brand Industrial support to minimize Services Inc) transfer of compression loads to concrete wall. Filler is compressible but occupies space during concrete placement to create a gap. b) Between Reactor Vessel ETHAF0AM - 220 Filler material used to and Reactor Bldg Polyethylene manufactured maintain a gap between top Internal concrete (See by Sentinel Foam Products edge of RB internal base concrete and steel con-Dwg. 3240-G-2520 S1-Det tainment liner as shown 'D' in referenced dwg. This gap will prevent transfer of loads due to expansion of containment liner to the top edge of the base Concrete. ~

Question No. 220.30 Response _(Cont'd) Location of Filler Material Type of Filler Material Purpose c) In RB Radioactive Pipe SE - F0AM, Filler material used to Chase Area to Separate Manufactured by BISCO create a gap between steel Steel Columns, Contain-columns, containment liner & secondary shield wall ment Liner & Secondary Shield Wall from Pipe and pipe chase walls & slabs to prevent inter Chase Walls & Slabs action of these elements (See Dwgs 3240-G-2585, due to deflections caused 2588) by thermal, seismic or other loads. Filler material used to d) Between Fuel Transfer SE - FOAM, Tube Shield Walls and (Impregnated with Lead) maintain a gap between Concrete Shield Bldg Manufactured by BISCO fuel transfer tube con. crete shield bidg wall Wall (See Dwg 3240-to prevent interaction of G-2530) these structures due to displacements under seis-mic or thermal loading. The filler material is also impregnated with lead to act as radiation shield to reduce " shine" through gap when fuel transfer tube is in use. e) Wrapped around Circum-ETHAF0AM 220 Filler material used to ference of Penetration Manufactured by Sentinel maintain a gap between Penetrations 23, 24 and Nozzle and Reinforcing 44 (which are embedded in I' late of Penetrations RB base concrete) to pre-No. 23, 24 and 44 (See load transfer to penetra-Sketch from DCN-AS-tion from concrete. 48SAttached) .. -. -. -. ~.. -. - - -. -. - - -.

t Question No. 220.30 Response (Cont'd) Location of Filler Material Type of Filler Material Purpose Item 2) Materials used as fillers are installed prior to concrete place-ment. Deterioration of the fillers, if any, after the concrete hardens will not be detrimer.tal to the structure. Properties of the material described in item d) above, which is used as shielding, are included below: SE-Foam a) Fire Resistance - ASTM E-84 flame spread test indicates flame spread of zero, fire resistance as measured per ASTM E-119 is rated 3 hours. b) Radiation Resistance - Maintains integrity at gamma radiation levels in excess of 1 x 1@ rads. c) Water - Long term innersion in water shows no deterioration. For the other WNP-3 Category I structural separation joint applications, i.e.; 1) between the Shield Building and Fuel Handling Building 2) below the cask loading pool floor in the Fuel Handling Building and, 3) between the Dry Cooling Tower Foundations and the Diesel Fuel Oil Storage Tank Enclosure foundations. Premolded joint fillers, manufactured by either W. R. Grace Co. (Fiber Expansion Joint Filler) or ED0C0 (cane fiber preformed expansion joint) were I used. - These products are bituminous impregnated fibrous materials. l l l

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Question No. Provide a description and sketches of arrangement and details 220.37 (SRP of the spent fuel rack seismic restraints. Provide results of 3.8.4, II key calculations for the structural design of the spent fuel FSAR 9.1.2) racks and seismic restraints for the racks.

Response

The FSAR will be modified as shown. i e ,-----a,-,,

1 %..4 i 0 s N dc_._ WNP-3 A.'10.3 7 -FSAR g, ' The absorber is fabricated of 0.21 inches (nominal) thick x 8.25 inches MEs wide x 30.44 inches long plates., Five such plates are inserted in each ' mir cavity on the outer surfaces of each storage location. The Bt.C plate is inert in water and under long-term irradiation, and is thermally stable under all temperatures expected in the pool. In the unlikely event of a postulated seismic event, the neutron absorber plates are restrained from shif ting by the adapters and cover plates. In the extreme case of the worst anchanical tolerances assumed to be present with the seismic event, there will be no settling of the neutron absorber asterial below the tcp of the active fual. Also, all seal welds of the absorber cavity will be dye penetrant inspected to assure j integrity of the seal. c) Rack Assembly The assembled storage cans cre formed into rack assemblies by attachasnt to stainless steel bars whic'n run horizontally near the top and bottom of the rack to form a unitized grid arrangement. The storage cans are oriented such that alternate storage cans are positioned with the 3 C plates in the north-south and east-west 4 directions, respectively. This results in neutron absorber plates between all adjacent storkse locations. Each storage can is continuously welded on all four sides to the lattice at the top and bottaa to form a single rigid rack structure. The center-to-center spacing between storage locations is maintained at 11.12 inches by the grid structure. Each rack is supported by four legs. Each of the legs has the capability of being adjusted to compensate for any unevenness i in the pool floor. The les supports are located under storage ~" locations,replacingthefuelsupportplate,andarereinfprcedby guasets. The leg support, dici i: 21:: ^ 5 i;;i : ;iid, contains five holes to provide a flow of cooling water to the fuel assembly. i Twenty storage racks will be located in the spent fuel pool as shown on i Figure 9.1.2-2a. i m iksub 4 i i s GJ JH( Structural ! n.: :7 Dg3l7 Cf.k r.*A The structural integrity of the fuel racks has been analyzed to assure that the required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of the AISC (includes Supplement 3, June 1974). The analysis considered the following load combinations and section strength limits: LDAD COMBINATION STRENGTH LIMIT a) D+L 1.05 b) D+E 1.0S 1.5S c) D+To+E l 1.6S' d) D+To+E 1.6S (Note exception described below) 5: a) D+To + LyAp . ac. o 1_ st s~ A~n e w 1 (1 n /e)

r - l 1528W-4 i GW'S*u de 'WNP-3 ro,3 7 FSAR 1.65 (Note exception described below) f) D + T, + LFAS

===:

Dead loads which include dead weight of rack and fuel l1 D = assemblies. Live loads due to lif ting the empty rack during iastallation. L = Thermal loads due to the unifora thermal er.pansion of racks. T = o An average pool temperature change from 40F to 150F at a 20F thermal gradient.between adjacent storage locations. Loads generated by the OBE (severe envirotunental load) E = El Loads generated by the SSE (extreme environmental load) = LFAD = Accidental spent fuel assembly' drop. Postulated fuel assembly stuck causing 5000 lb. upward force. l1 LyA3 = Required section strength based on the elastic design methods S = and allowable stresses defined in Part 1 of the AISC " Specification for the Design Fabrication and Erection of Structural Steel for Buildings.* Nste that local strasses may be allowed to exceed limits in load combinations o and f. S: ::i zi: 1 :f f:: Oh !x1 ::;i n: 2;;;;;' :f f:

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:d;. Damping factors of two percent for the OBE and four percoat for the SSE vere used in confermance with Regulatory Guide 1.61.

Closely spaced modes were combined directly and then combined with other significant modes using the SRSS method. The structural responses to each of the three etaponents of earthquaka action were combined by taking the SRSS of the =m'rfata representative values of the co-directional responses in conformance ~ uith Regulatory. Guide 1.92. ( [M u r+ 2.k Impact loads caused by fuel movement within the can during the seismic event were considered in the analysis. Since a gap on the order of 1/4 in. exists i between the sides of a fuel assembly and the can, the fuel will actually move within the can during a seismic event and cause impact loads to be transmitted 'to the fuel rack restraints. The effects of this fuel-can interaction were cualyzed using a energy balance technique. An analysis of a single can and fuel assembly performed to determine the shear force and bending acaent which may occur at critical sections of the can as a result of the fuel assembly impacting the can. These forces were.added directly to those cbtained from the modal ana g ig.g g impact velocity is taken to be a T fraction of the peak velocity 3 representative-ed e. can within :u4 rack,as 3&d 1 determined from the modal analysis of the rack. This factor vss used to cecount for hydrodynsaic coupling, as treated by Fritz (Raference 4) between the fuel and can.a M ysg.Q Ce can has stiffness characteristics representative of a can within a rack. Th The can is restrained at the upper grid elevation by a spring representative of_the average grid and horizontal rack restraint stiffness. $wt: El 'inw E61,

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'r- -* 6.ee c b 9.1.2.3 Safety Evaluation 42.- The spent fuel pool storage rack design and location, discussed in Subsection rg 9.1.2.2 provides assurance that the design bases of Subsection 9.1.2.1 are met as discussed below: 9.1-9 Amendment No. 1, (10/82)

GuPsNe><il). 1484W-5 .ws7 3 z,w,37 FSAR / E, CALPLOTF -CALPLOTF II WHAMHOC II b) The following computer program is employed by the A/E for genera:; ion of the seismic Response Spectra Values used in the structural dynamic stress analysis of seismic Category I piping systems; for further details on the program refer to Subsection 3.7.2.1. It was also used I ( for stress analysis of SIRS forgings. 7 NASTRAN l*LWrh-G \\ \\ 3.9.1.3 Experimental Stress Analysis 3.9.1.3.1 Experimental Stress Analysis Within the NSSS Vendor Scope of Supply Refer to CESSAR-F Subsection 3.9'.1.3. 3.9.1.3.2 Experimental Stress Analysis Within the A/E's Scope of Supply Requirements for experimental stress analysis methods to be used in lieu of analytical methods have not been imposed on any seismic Category I mechanical systems and components. 3.9.1.4 considerations for the Evaluation of the Faulted Condition 3.9.1.4.1 Seismic Category I Within the NSSS Vendor Scope of Supply Analyses of the Reactor Coolant System components (reactor vessel, steam generator, reactor coolant pump, pressurizer, and reactor coolant piping) and their supports have been performed in accordance with the methods described in CFSSAR-F Subsection 3.9.1.4.1. For each component and support member, the esiculated loads, in combination with the seismic loads, are below the loads .specified for design and the stresses (pipe rupture in combination with SSE) are below those listed in CESSAR-F Table 3.9.3-2. No components or supports of the Reactor Coolant System main loop for WNP-3 were designed using the inelastic methods defined in Section III of the ASME 4 Code as plastic instability or limit analysis methods. The reactor vessel lower key horizontal supports include load limiting devices in accordance with Subsection 5.4.14.2(e) of CESSAR-F. These load limiters are designed to remain elastic for all normal, upset and the SSE loadings, and elastic system analyses are used to establish or confirm the loads specified For loads for design of the components and supports for these conditions. resulting from postuisted pipe breaks, the load limiter devices are designed to deflect plastica 11, and nonlinear system analyses are used accordingly for 7 proper calculation of the distribution of the loads among the system of supports. ... = I 3.9-5 Amendment No. 4, (12/83)

s N3 Ous bb Il.o. 1484W-6 110.37 s s .i Table 3.9.3-2 for ASle'. Code Class 1 Piping and Piping Supports

=

and ' Table 3.9.3-5 f or ASME Code Class 2, 3 Piping and Piping Supports. Inelastic methods of analysis are used in cases where it is deemed necessary to permit significant -(local) inelastic response. In these cases, the system or subsystem analysis perf ormed to establish the loads are modified to include the inelastic strain compatibility in the regions at which significant (local) inelastic response'is paraitt'ed. Piping, nd p.iping supports are always a maintained,within the_ ' elastic region. Inelastic response is permitted f or both piping and pipe whip restraints in the piping system where the, break is 1 postulated. Calculated inelastic strains, produced by tha f aulted cond,ition, are limited to 50 percent of the ultimate strain (strain at ultimate stress of the material).- For design parameters on pipe whip restraints ref er to Subsec. tion 3.6.2. [%wkFh 4 \\ [5 ~\\' x i. o. A a,,,andment No. 1. (10/82)

Question No. 220.37 INSERT A

Response

(cont'd) d) Seismic Restraints lateral seismic restraints are provided between the spent fuel racks and pool walls, as indicated by figure 9.1.2-2a. These restraints are located at the top and bot-tom of the racks (as indicated on figure 9.1.2-1) and con-sist of structural members cantilevered off the fuel racks with a pad mounted on the end of each restraint. The pad is designed to bear on the pool wall during seismic events with a small gap provided between the wall and pad for thermal expansion during normal operation. INSERT B f) Seismic Analysis The seismic loadings on the fuel racks are determined from a response spectra modal dynamic analysis using the STARDYNE computer program. Storage racks are modeled in detail using beam and plate finite elements. A detailed model of a corner rack is used, with the adjoining 3 and 4 rack arrays also modeled but in a less detailed fashion, using beam and lumped elements. 4 The detailed corner rack is modeled with individual storage cans, top and bottom grids and support feet. The adjoining racks are modeled to provide the proper mass and stiffness characteristics considering interactions that occur during seismic events. Also the stiffness of the fuel assemblies is neglected. However, the fuel assembly mass and effec-tive mass of the water are considered te be uniformly dis-tributed along the storage cans. INSERT C g) Structural Analysis Results l Stress analyses were performed for critical rack sections using the loading conditions and the methods previously outlined. In general, stresses were calculated either automatically via STARDYNE or through the use of standard engineering handbook formulae using load developed by STARDYNE. The analyses demonstrate that the structural design of the fuel racks complies with the desi9.1.2.2.2 3)gn acceptance criteria of this FSAR. discussed in Section . =. - -

Question No. INSERT D 220.37

Response

(cont'd) The key results of the seismic analysis are as follows: East / West Horizontal Direction The first two modes contribute over 80% of the modal effec-tive weight. The corresponding natural frequencies are 13.14 and 13.99 Hertz respectively. North / South Horizontal Direction The first and fourth modes contribute over 80% of-the modal effective weight with the corresponding natural frequencier being 13.4 and 15.4 Hertz respectively. Vertical Direction The fundamental frequency of 39.3 Hertz contributes over 80% of the modal affective weight. INSERT E The ANSYS code was used to perform the nonlinear impact analysis. INSERT G 3.9.1.2.3 Computer Programs used in the dynamic and static analysis of the Seismic Category I Spent Fuel Racks. 3.9.1.2.3.1 STARDYNE The STARDYNE Analysis System (reference 1) con-sists of a series of compatible digital computer programs designed to analyze linear elastic structural models. Dynamic analysis is accom-plished via the normal mode technique. 3.9.1.2.3.2 ANSYS ANSYS is a general purpose nonlinear finite ele-ment program with capabilities for structural and heat transfer analyses, as described in reference 2. l INSERT F References for Section 3.9.1.2 1. MRI/STARDYNE 3 (Version 3) developed by Mechanics Research Inc., Los Angeles, California. 2. ANSYS Computer Code, Update 180 Revision 2 ceveloped by Swanson Analysis Systems, Inc. Elizabeth, PA. l _... _ _. _ _ _ _... _. -, _. _ _ _. ~ _ _

Question No. 241.20 In interpreting the dynamic laboratory test results, you have: -(SRP 2.5.4) (i) divided the measured axial strains by a correction factor of 1.5 for the 13 strain-controlled cyclic triaxial tests on residual soils. (ii) multiplied the measured peripheral shear strains by a cor-rection factor of 0.7 for the Resonant Column test results on Residual Soils, and (iii) divided the measured axial strains by a factor of 4 for strain controlled tests on fresh sandstones. Justify these correction factors. In addition, submit any labora-tory test data used for developing strain-dependent modulus and damping curves for weathered sandstone. Provide and justify your basis for assuming the same dynamic properties for weathered sand-stone as for fresh sandstone.

Response

The correction factors for the strain-controlled cyclic triaxial test results were selected based on published data and data in Woodward-Cycle Consultants' files. Using an electro-optical tracker device, Moore and others (1969) indicate that the axial strains at a mid-section of a triaxial specimen measured by the tracker were smaller than the overall strains computed from the vertical displacements between the top and bottom loading plat-tens. The differences between the two measured strains increase with a decrease in strain level and with an increase in stiffness Primarily based on these data, the correction of the specimen. factors of 1.5 and 4 were selected for the strain-controlled tri-axial test results on residual soils and fresh sandstone, respec-tively, based on the appropriate strain levels and stiffnesses. For the resonant column tests, because of boundary conditions, the shear strains induced in a specimen are not uniform and vary from zero along the axis of rotation to the maximum strain at the peri-meter of the specimen. The average strain induced in a specimen is equal to two-thirds of the peripheral strain (for a ready re-ference see Drnevich et al.,1978). Thus, the factor of two-thirds (round up to 0.7) was applied to obtain the average shear strains of the specinm. from the measured peripheral shear strains. i I 1

Question No. 241.20 Response (Cont'd) [ t i After applying these correction factors to the s" rain-controlled triaxial test results and the resonant column test results, the dynamic properties of the residual soils from bo"h types of the test were found to fall in a consistent trend for defining the variations with shear strain levels. Similarly, the shear strain-dependent dynamic properties for the fresh sandstone were found to be consistent with those developed for rock by Schnabel et al. (1971). Therefore, the selected numerical values for the cor-rection factors Are judged to be reasonable and appropriate. The attached pages E-3 and E-4 of Ebasco Design Specification 3240.052 504 - Reactor Auxiliary Building, Appendix E, present typical and design values for both index and engineering pro-perties. It can be seen from these pages that there is a marked similarity between both static and dynamic properties of each material considering the normal range of variation of soil and rock parameters. Some examples are the dry unit weight of 103 to 112 pcf for weathered sandstone and 103 to 114 pcf for fresh sandstone and geophysical measurements of 2300 and 4100 ft/sec for weathered sandstone and 3000 to 4000 ft/sec. for fresh sandstone. Considering the overall agreement of both the index and engineer-ing properties for both fresh and weathered sandst'one (as exhibited by pages E3 ana E4), it was felt that assuming the same dynamic properties for both fresh and weathered sandstone is justified. REFERENCES Drnevich, V. P., Hardin, B. O., and Shippy, D. J. (1978), " Modulus and Damping of Soils by the Resonant-Column Method," Dynamic Geotechnical Testing, ASTM STP 654, American Society for Testing and Materials, pp. 91-125. Moore, W. M., Swift G., and Milberger, L. J. (1969), " Deformation Measuring System for Repetitively Loaded, Large Diameter Specimens of Granular Material," Research Report No. 99-4, Stress Distribution in Granular Masses, Research Study Number 2-8-66-99, Texas Transporation Institute, Texas A & M University., College Station, Texas. l

Schnabel, P., Seed H. B., and Lysmer, J. (1971),

" Modification of Seismograph Records for Effects of l Local Soil Conditions," Earthquake Engineering l l Research Center, Report No. EERC 71-8, College of l Engineeing, University of California, Berkeley, Calif., December. ='w' ~ -

$ 14 {. AD Ebasco Design Specification Reactor Auxiliary Building Appendix E (Cont'd) Soil and Rock Properties WEATHERED SANDSTONE INDEK AND ENGINEERING PROPERTIES Parameter Typical values Design Values Index Properties Dry Unit weight Y ' IDII" 103 - 112 d 3 Saturated Unit Weight Y, Ib/ft 126 - 133 129 g 18 - 21 20 Water Content w % Apparent Porosity n 0.32 - 0.39 .35 Specific Gravity Gs 2.71 - 2.74 2.72 Core Recovery, % 85 - 100 Rock Quality Designation RQD, % 90 - 100 Static Properties Ultimate Strength e, Ib/in. 350 - 800 350 g 3 Tangent Modulus at 50%,,; g, Ib/in, x10 100 - 400 110 0.25 - 0.30 .27 Poisson's Ratio at 50% e,, g Dynamic Properties - Use These Geophysical Measurements Compressional Wave Velocity (V )f, ft/s 5000 - 6000 p Shear-Wave Velocity (V )g, ft/s 2300 - 4100 3500 l g 3 Tangent Modulus E, lb/in. x10 650 - 750 700 f '4 Poissun's Radtio pf

2. 7-El ShearModulus-FSARFig({}

Damping -MFig ( L)2.-T-I2.% fik{L. Laboratory Sonic Velocity Measurements Compressional Wave Velocity (V )g 8000 - 12,000 8400 p 3000 - 4,000 3100 Shear-Wave Velocity (V ); 3 Tangent Modulus E lb/in. x10 700 - 1,500 750 g 0.35 - 0.45 .42 Poisson's Radio pg E-3

b Ebasco Design Specificction Reactor Auxiliary Building Appendix E (Coat'd) Soil and Rock Properties i FRESH SANDSTGIE INDEX AND ENGINEERING PROPERTIES Parameter Typical Values Design Values - Index Properties 3 Dry Unit Weight Y, Ib/ft 103 - 114 d 3 Saturated Unic Weight Y, Ib/ft 125 - 133 128 g Water content w, % 18 - 23 20 Apparent Porosity n 0.32 - 0.39 .35 2.75 - 2.78 2.75 Specific Gravity Ga 85 - 100 Core Recovery, % Rock quality Designation RQD, % 90 - 100 Static Properties Ultimate Strength e, Ib/in. 300 - 850 300 d 3 100 - 250 110 Tangent Modulus at 50% a,a E, Ib/in. x10 g 0.35 - 0.50 .44 Poisson's Ratio at 50 % a, p T c 1 Dynamic Properties Geophysical Measurements Compressional-Wave Velocity (V ) f, ft/s 7000 - 8000 p ^ 3000 - 4000 3500 Shear-Wave Velocity (V )f, ft/S 3 3 Tangent Modulus E, Ib/in. x 10 850 - 950 900 f 0.3 - 0.4 .35 Poisson's Radio pg Shear Modulus - M ig ( ) F*dA. Damping -% Fig ( b2.T-72.t ) FSM Laboratory Sonic Velocity Measurements 7000 - 9500 Compressional Wave Velocity (V ); p 2500 - 4000 2800 Shear-Wave Velocity (V )g 3 Tangent Modulus E, Ib/in.2 x 10 400 - 1000 600 g .35 .45 .43 Poisson's Radio pg E-4

Question No. 270.1 Tables 3.11-1 and 3.11-2 are not complete. Provide the missing (3.11.1) information or a schedule for providing it.

Response

Tables 3.11-1 and 3.11-2 will be revised as committed to in Letter G03-82-1085 dated October 22, 1982 from G. D. Bouchey to J. D. Kerrig+.n to incorporate information as it becomes avail-able. The first submittal was issued on December 1982 with an overall completion date scheduled for June 1984. The Supply System considers this information adequate response to enclosure 4, item 1 of the August 20, 1982 letter from D. G. Eisenhut to R. L. Ferguson. For previous submittals refer to the following Supply System letters;

1) G03-83-0046, dated 01/17/83

2) G03-83-0105, dated 02/03/83

3) G03-83-0333, dated 04/20/83 The Supply System has no information ready for submittal in this quarter.

l l

i Question No. 271.1 Table 3.10-1 is not complete. Provide the missing information or l (3.10) a schedule for providing it. {l

Response

Table 3.10-1 will be revised as committed to in Letter dated October 22, 1982 from G. D. Bouchey to J. D. G03-82-1085 Kerrigan to incorporate information as it becomes available. The Supply System considers this information adequate response to enclosure 4, item 11 of the August 20, 1982 letter from D. G. Eisenhut to R. L. Ferguson. For previous submittals refer to the following Supply System letters;

1) G03-83-0046, dated 01/17/83

2) G03-83-0333, dated 04/20/83 Attached is the information for December 1983.

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Question No. 281.6 Regarding the Spent Fuel Pool Cleanup System, provide the (SRP.9.1.3) following information: Describe the samples and instrumentation and their frequency of measurement that will be performed to monitor the spent fuel pool water purity and need for ion exchanger resin and filter replacement. State the chemical and radiochemical limits to be used in monitoring the spent fuel pool water and for initiating corrective action. Provide the basis for establishing these limits. Your response should consider variables such as: gross gamma and iodine activity, demineralizer ar.d/or filter differen-tial pressure, demineralizer decontamination factor, and pH and crud level.

Response

Periodic grab samples are used to monitor the pool water in order to confirm that the water chemistry is within the speci-fied limits. These limits are based upon providing a proper environment for fuel assembly materials, maintaining radiation levels within acceptable limits and maintaining clarity levels sufficient to support refueling operations (primarily fuel handling). Each sample is analyzed in the plant chemistry laboratory on the applicable instruments. Ion exchanger accept-ability will be determined by the plant chemist using as a basis The filter the decontamination factor across the ion exchanger. differential pressure limits used to determine acceptability are found in FSAR Table 9.1.3-3. Spent Fuel Pool Water chemistry is given in Table 9.1.3-2. Expected nuclide concentration is given in CESSAR-F Table 11.1.7-l&2. FSAR Tables 9.1.3-2, 3 and 11.5-2 will be amended to reflect the response to this question. l l I l

kb e tt 1616W-4 WNP-3 M I * (# FSAR. ~ / A low pressure switch on charging pump suction will then stop the ~~, proccure.De fuel pool cooling suction line is located two and one half feet pumps. The volume of balow the siphon breaker holes to ensure fuel pool cooling. 1 boratsd. water available to the CVCS from the fuel pool is consistent with CE intarf ace requirements as described in CESSAR-F Subsection 9.3.4.6.F.2. During normal operation, losses due to evaporation are made up by unborated ankaup water which is supplied from the nonseismic deionized water system via th) deionized water transfer pump and valve _3FS-1002N. De average water l1 makeup requirements based on fuel pool temperature of 13W is 1.5 gpa. In order to prevent possible boron dilution during the refueling operation, the fuel pool water is borated to refueling concentration (4000-4400 ppa borsa) from the seismic Category I INST via the Boric Acid Makeup hap (BAMP). The makeup capacity from the BAMP is 165 gpa. l1 Energency makeup to the fuel pool is provided from the EWST via the M*=lcal The askeup capacity of the cad Volume Control System (CVCS) charging pumps. Energency makeup to the fus1 pool following a 1 charging pumps is 44-88 gym. Design Basis Accident (DBA) is provided by the seismic Category I condensate scarcge tank via condensate makeup pump (see Subsection 9.2.6). In the event of a LOCA, initially when the component cooling water is not available for Spent hel Pool Cooling use, it will take a miMann of 11 hours 1 b fore pool water starts boiling (based on Case I heat loads for a 7ACA m Me evaporation loss during the occurring just af ter the lith refueling.) above 11 hours is less than 3000 gallons or approximately four inches of fuel As the cooling water becomes available, component cooling water posi level. can be diverted to fuel pool cooling heat exchangers to reduce and maintain the fuel pool water temperature below 130F. r 9.1.3.2.2 Fuel Fool Cleanup System Em,u Y ha cleanup. portion of the Spent hel hol Q)olin,g and Cleanup Systea seassene- __2 ...... f... $ _ g g;;; $=; e. i...v.t.. g g --e2 a g _ _. - 7. 2.. =fr:lin;; ::ri ti: i; ni;;;isi ; ci;-ity . r' i, t'-- - *r '*r

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th :::r-frel ; :1 r:t : 2d th: ::5niin ;; 1 ;;;:. i p. 1;;i De cleanup systannalso remove K ' r:# ri:::xtir: cf ':__;::: :;;;;ti ::. dissolved fission products frou the refueling water storage tank.~ The cleanup Each train loop consists of two parallel trains of cleanup equipment.esitains one pool clea Most of the cleanup flow fuel pool filter, and one fuel pool ion exchanger.io drawn from below the surface rafueling pool, while a small fraction any be drawn through the surface okimmer to remove surf ace debris. Siphon breakers and anti-siphon holes are provided in the suction and discharge piping in the cleanup system. The basket strainer is provided in the cleanup suction line to remove anyRe. spent f relatively large particulate matter. by the pump through a filter, which removes particulates to five microns, and =-- H e purified water is through an ion exchanger to remove ionic material. roturned to the pool. Amendment No.1. (10/82) I 9.1-17

.r Question No. INSERT A 281.6 Respons3 ... is designed to provide a proper environment for fuel assem-(cont'd) bly materials, maintain pool radiation levels within acceptable limits and maintain pool clarity at levels sufficient to support refueling operations (primarily feel handling). Table 9.1.3-2 identifies the pool water chemistry limits associated with these The pool chemistry limits are periodically design criteria. evaluated througn the u:e of grab samples as discussed in FSAR table 11.5-2. The actual frequency of grab sampling will be addressed in the plant procedure manual. The filter condition will be determined by pressure drop. I I l l

16191i-2 WNP-3 Cw.h L. TABLE 9 1 3-2 =. 5N SPENT FUEL POOL 1ATER CHEMISTRY pH (777) 4.5 to 10.2 ct Boric Acid, Maximum,VtJt'd !!"6 A.T Anmonia, Maximum,Xl ppm 50 i Lithium, Maximum,# ppm 2.5 Dissolved Air, Maximum Saturated Chlorida, Maximum, ppa 0.15 Fluoride, Maximum, ppa 0.1 C la riff ecJ:a k b oels 4 2.s ra%/sr

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== Particle Bentention siza 5 micron (982) ASEE VIII N Code ~ Design Pressure, psig 200 Design Temperature, F 200 300-400 now, gym ressure Ioss, Clean, psi 5 21/2 we Z Boric Acid Solution nuid Materials in Contact with Fluid Stainless Steel Type of Elements Esplaceable Cartridge NHS Safety Class None Seismic Category swX.%ue Pre s s kit dicp,d.rh, p, gap gg 7. Fuel Pool Ion Exchanger 2 Quantity Mixed Bed Disposable Resin Type Code ASHE VIII 200 Design Pressure, psig De sign Temperature, F 250 300-400 . now, gym 3 90 Resin Volume (Useful), f t Stainless Steel Material 21/2wc: Boric Acid Solution nuid NNS Safety Class None Seismic Category -9.1-27

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t Question No. Ycu state that your analysis for aircraft hazards is forthcoming. 311.3 (3.5.1 4) Provide a schedule for furnishing this information.

Response

The subject analysis, in the form of a draft amendment to Sub-section 2.2.2.5 and 3.5.1.6, is attached. The results indicate that aircraft do not pose a credible hazard to the safe opera-tion of WNP-3 due to the low probability of impact and struc-tural design margins and features applied to safety related The FSAR will be amended to include this plant structures. information.

ggh,[f, 1803W-2 WNP-3 FSAR 3 / f, 3 2.2.2.4 Waterways l-Although an authorized Federal channel exists on the Chehalis River from Grays Q Harbor to Montesano (River Mile 13.5), the channel is not maintained to the authorized depth of 16 f eet and width of 150 f eet. The existing channel is used mainly for gravel barges. The NSSS components f or WNP-3/ 5 were barged to about River Mile 16, however this was a special one-time operation which was dependent on river flow and tide stage. In the vicinity of the site (River Mile 21), the size and capacity of the river make barge traffic impracticable. Pleasure craf t use the river as there are five public boat t-ramps within five miles of the project (Reference 2.2-6). l: 2.2.2.5 Airports and Airways The closest airport'is the Elsa Airport located three miles northeast of the reactcr/ (see Figure 2.2-2). The 2100-f t(runway is oriented east-west and is y paved and lighted f or use by light single and twin-engine fixed wing aircraf t A 1000-f tggrass taxieay lies parallel to the east half of and helicopters. Ianding maneuvets are standard lef t-hand approaches which take / the runway. the aircraf t about 1/4 mile south, of the runway on the downwind leg Fresterly wind) at an elevation of about 850 f tqSL. The airport is r :::;..d_Q g records on runway usage are not maintarned. p.: 19S rr:n 12: f:: th: cir;:r f::::x: if,500 y

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h Also in the vicinity of the site are two private grass airstrips located 1.2 miles NNW and 4.2 miles ENE of the reactors. The former s trip has not been used in recent years although the owner intends to rehabilitate it and base a l few small private airplanes. The other grass field is used very infrequently (Raference 2.2-9). The next closest airport of note is Bowerman Field in Hoquiam, 22 miles west of WNP-3/5. The runway is capable of handling aircraf t up to the DC-9 and 3-7 37 category. Although the airport has no scheduled commercial flights, a commuter service is expected to be established in late 1981. The airport master plan prepared in 1979 projected 125,000 operations f or the year 2000. 2 criterion of Regulatory Guide 1.70 This is considerably less than the 1000d (References 2.2-6 and 2.2-10). As shown on Figure 2.2-2, two f ederal airways used by instrument flight rules (IFR) traf fic are located in the site vicinity. Airway V204 between Hoquiam and Olympia passes directly over the site with a minimum altitude of 4500 f tp MSL. Airway V27 between Hoquiam and Seattle passes within eleven statute Traf fic on miles of the site with a minimum enroute altitude of 3200 f rqSL. airways V204 and V27 averages five and seven flights per day, respectively.g The FAA has no plans to change the routes (Reference 2.2-11). E0stest h h ? 7~ ia VZ7 a., r,. sam.,6 -Ad (y-

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F lAL M Y. w.s 1803W-3 WNP-3 FSAR In addition to the commercial traffic on federal airways, there is military y traffic originating at Ft. Lewis Army Base (42 miles ENE) and McChord Air Force Base (48 miles ENE) that could overfly the site. Nearly all the flights from Ft. Lewis in the vicinity of WNP-3/5 are IFR training flights on V204. These amount to one or two flights per month by OH-1, OH-60, or CH-47 /f helicopters (Reference 2.2-12.). McChord AFB originates about 280 flignes per year by C-130 or C-141 aircraft that manuver in ene vicinity of the site. These aircraft fly under visual flight rules (VFR) along predetermined routes that pass within tour miles of the plants. No or$ance is carried on the )( military aircraft (Reference 2.2-13). 2.2.2.6 Railways A track maintained and used by the Union Pacific Railroad passes about 1-1/4 miles north of the reactors along the south bank of the Chehalis River. Another east-west freight line, owned by cne Burlington Northern Railroad, passes within three miles of the project in the vicinity of the town of Satsop. 2.2.2.7 Projection of Industrial Growth Tnere are no specific proposals or plans for new or expanded industrial activities in the site area. Programs to promote commercial and industrial development in tne area have focused on Grays Harbor as a potential deep port and the exploitation of the agricultural, forest, end recreational resources of the region (Reference 2.1-2). 2.2.3 EVALUATIONS OF POTENTIAL ACCIDENTS '..sk Of the transportation route in the vicinity, the Union Pacific Railroad track is most significant as a potential source of accidents. There are no industrial facilities whi~ch pose a threat to plant operation. 2.2.3.1 Determination of Design Basis Events Explosions - For most of tne route near tne project, the railroad tracks are 350 feet below plant grade and shielded from the plants by the natural about topography and a screen of forest. The distance (approximately 7000 ft.) exceeds that specified in Regulatory Guide 1.91 for a credible rail car load of explosives. Explosions are not a threat to plant safety-related facilities. Flammable Vapor Clouds - The postulated rupture of a 85-ton rail tank car of propane was analyzed by methods specified in Regulatory Guide 1.78 using very conservative assumptions (e.g., Pasquill F and 1 m/see wind speed). The model f assumed the instantaneous telease and vaporization of nalf the centents of the tank car and the boil-of f of cne residual over a one-hour period. Tae analysts ignored the ef fects of topography (including tne 350-ft vertical elevation ditt'erence) and dispersion througn the dense timber and assumed that 2 the plant receives plume centsrline vapor concentrations. Witn these conservative assumptions, the grouno-level concentrations at the plant are below the minimum concentrations of propane gas in air necessary for detonation (2.4 percent by volume). Toxic Chemicals - Ine worst case exposure of a toxic enemical would result from the catastrophic rupture ot a 90-ton railroad tank car carrying liquid chlorine. Inis is tne largest railroad tank car commonly used for shipping 9 9.1 a...J-... M. 9 /19 /09) -n

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REFERENCES:

SECTION 2.2 2.2-1 Le tter, J. A. Durre'll, Thiokol/Ventron Division, to C.G. Ayers, WP PSS, January 9, 19 81. 2.2-2 Personal Connounication, J.P. Chasse, WPPSS, with Ron Lund, Western Washington Welding, Inc., May 14, 1981, 2.2-3 Personal Communication, J.P. Chasse, WPPSS, with Robin Amlin, Coastal oil Dis tributors, May 12, 1981. 2.2-4 Personal Communication, J.P. Chasse, WPPSS, with Robert Fawbush, Fawbush oil, May 12, 1981. 2.2-5 Personal Comununication, C.G. Ayers, WPPSS, with L.E. Anderson, Cascade Natural Gas Corporation, March 18, 1981. 2.2-6 Personal Communication, G.G. Ayers, WPPSS, with Stan La ttin, Port of Grays Harbor, December 17, 1980. 2.2-7 Personal Communication, J.P. Chasse, WPPSS, with Gordon Flemming, Elma Airport Advisory Board, May 7 and June 11, 1981. YY '2.2-8 Personal Consnunication, J.P. Cha, WP PS S, w i th E d C _ _ _ r, Elma "- r -';. 11 Airport Operator, Hehentos. S*, M S 2.2-9 Personal Communication, J.P. Chasse, WPPSS, with Mark Aarhouse, Crays Harbor PUD No.1, June 11,1981. 2.2-10 Le tter, LeMoine Stitt, Washington Department of Transportation, Division of Aeronautics, to D.E. Dobson, WPPSS, January 8,19 81. 2.2-11 Personal Coneounication, J.P. Chasse, WPPSS, with Jerry Wilson, Federal Aviation Adrainis tration, Nov,e $ ber81; liS'3, 12, 10"1. l 2.2-12 Personal Communication, J.P. Chasse, WPPS S, with Ma jor Klina, Ft. Iawis Army Base, May 12, 1981. 2.2-13 Letter, Colonel Cary H. Mears, McChord Air Force Base, to Washington Public Power Supply System, January 7,1981. 2.2-14 Preliminary Sa fety Analysis Report, Skagit Nu clear Power Pro ject, Docket Nos. SIN 50-522 and 523, prepared by Puget Sound Power & Li gh t, p. 2. 2-10e. 2.2-15 J Wing, Toxic Vapor Concentrations in the Control Room Following a Pos tulated Ac cidental Re lese, NUREG-0570, U. S. Nuclear Regulatory Conunission, June 197 9. w c,... : 4, J.E Cluna, u W 5> " Y U N Y 2.z-z n., H fr Q Ai, &46y 6v= G, A da.4 m3-q 2.2-7

1546W-10 5 3#I*3 WNP-3 FSAR 3.5.1.5.6 Military Facilities No military f acilities exist within 20 miles of the site. Consequently, i neither military aircraf t nor projectiles are considered credible missiles for the site. 3.5.1.6 A).rcraf t Ha za rds i :: dir;; : ?"?"C 75/09', S 2 9 :d ?.ed = ?12, S:t::: i: 2.5.1.5, firc :f: Hazards, an estimate on the aircraf t crash probability was made f or the loca air traf fic of Elma Municipal Airpo (about three miles f rom the site) .he p robability was f ound to exceed 10~ytper year. Accordingly, an air . c is a potencial missile. Characteristics of impact on the plant structure will b nalyzed f or some chosen aircraf t types obtained f rom the master plan Elma Airport. The most critical type will be studied. The methodology lized by Seabrook Station /Public Service Company of New Hamps e is employed. This method requires inf ormation on the weight dist ution of each aircraf t which is forthcoming. The resulting impact be compared to that of the worst 'to rnado. This section will be nded at the completion of the study program as detailed above. p he method employed 'T af ford protection due to aircraf t impinging on WNP-3/5 safety-related seismic Category I structures, systems and components will be consiste with the CE interf ace requirements identified in CESSAR-F Subsect s 3. 5. 3.1, 5.1. 4.B.1 through B. 6, 5.1. 4.D.1 and D. 2, 5.4. 7.1. 3. B.1 the B. 3, 5. 4. 7.1. 3. D.1 t hroug h D. 3, 6. 3.1. 3. B.1, 6. 3,1. 3. D.1, 7.1. 3. 4, l Q g 0...'.0.0.1, 0.2.'.0.0.1 C....;h 2.2 c.4 0.2.'.i.2. I \\5TM l V 3.5-%

x Ouestion Ng. INSERT A .311.3

Response

Based on observations of airport activity (References 2.2-7, .'tCont'd) 2.2-8 and 2.2-16), annual operations are estimated to be in the range of 4,000 to 10,000. 0ver 98 percent of these operations are by single-engine planes (Reference 2.2-8). In late 1983, thirty-five planes were based at the airport. The operators have near-term plans to lengthen the runway by 600 feet and double the hangar spaces from the present eighteen (Reference ~ 2.2-8). INSERT B 3.5.1.6 Aircraft Hazards _ 3.5.1.6.1 Federal Airways and Airport Approaches As described in Subsection 2.2.2.5, two federal airways are located in the rite vicinity. No VFR airport approaches are close enough to the plant to be of concern. A conservative estimate of the annual probability of an aircraft striking the plant was made using the method suggesteo in NUREG-800 based on the following information: Federal airway width is 9.2 statute miles. Airway V204 is 1. centered over the site and the centerline of V27 is eleven miles from the site. Traffic on V204 and V27 averages five and seven flights per 2. day, respectively. The enroute accident rate for commercial air carriers is 3. approximately 4 x 10-10 per mile (Reference 3.5.1-6). Safety related plant structures whose failure could lead, 4. directly or indirectly, to off-site radiological conse-quences are the reactor building, reactor auxiliary building, fuel handling building, refueling water storage tanks, condensate storage tank and dry cooling towers. Assuming an impact angle of 45 degrees, effective target 2 area (plan area and shadow area) is 239,000 ft, The probability of an aircraft impacting critical plant 5. areas is given by (Reference 3.5.1-6): PFA = CNA/w where: j -.._-4 .m----, ,,__.r,_.--_,,,,.,_,y,_,.,y., __..--n,c,,

Question No. INSERT 8 Continued... 311.3 Respc.1se C = aircraft accident probability per mile of flight (Cont'd) M = number of aircraft per year traveling on airway A = effective area of plant w = airway width (plus twice the distance from airway edge to site if site is outside the airway) With the above bases, the probability of a crash from aircraft on either of the airways in the site vicinity is: 10 yr-l FA ((V204) = 6.8 x 1010 yr-l P FA V27) = 4.0 x 10-P Therefore, traffic on designated airways is not a hazard to plant operation. 3.5.1.6.2 Airports The closest airpcrt capable of handling aircraf t in excess of 12,500 lbs gross weight is Bowerman Field in Hoquiam, 22 miles west of WNP-3. As noted in Subsection 2.2.2.5, project operations are well under 1000d2 where "d" is the distance, in miles, from the site to the airport. Elma Municipal Airport is located about three miles northeast of The plant is 2.8 miles from the end of the 2100-ft. the site. As noted in runway at an angle of 51' from runway centerline. Subsection 2.2.2.5, greater than 98 percent of the operations at l the airport are by single engine aircraft with a gross weight of less than 2,500 pounds and all are under VFR conditions with a l minimum visibility of one mile. Takeoff and approach maneuvers occur within about a mile of runway over the valley floor and As was noted away from elevated areas such as the plant site. in NRC staff testimony during hearings on the Diablo Canyon Nuclear Power Station Unit Nos. I and 2 (Reference 3.5.1-7): " Light general aviation aircraft are not considered a significant hazard to nuclear power stations because of their low airspeeds, short distance landing capability, high Plant maneuverability and low penetration capability. protective features against tornado missiles, the inherent strength of the systems and structures, as well as the diversity and redundancy of plant systems reduce the potential hazards to the facility from light aircraft oprations to acceptably low levels." l

~, Question so. INSERT B Continued... 311.3 ' ~

Response

With respect to WNP-3, the Kinetic Energy of the 2500-lb air-(Cont'd) craft impacting at 125 fps (140% of stall speed; Reference 3 3.5.1-8) is comparable to that of tne 4000-lb car at 90 fps which the concrete and steel barriers can accommodate with ample margin (ses, Tables 3.5.3-1 and 3.5.3-2). Nevertheless, the probability of a light airplane striking the plant during maneu- _ ~ vers at the Elma Airport is conservatively estimated based on the following information: Airport approaihes and departures (i.e., total operations) 1. are assumed equally divided between the east and west end of With total runway operations conservatively the runway. assumed to be 10,000, the relevant numbers for twin and single-engine aircraft-are 100 and 4900, respectively. The crash probability for general aviation maneuvers 2.8 2. miles from the runway is taken as 6.2 x 10-8 per square mile per movement (Ref+rence 3.5.1-6). This number is conservative on at le:st three accounts. First, it is based on fatal crashes of general aviation traffic in the years 1%4 through 1%8 (Reference 3.5.1-9), a period when such In 1979 through crashes averaged greater than 800 per year. 1981, when the total operations should be considerably greater than during the mid-1960's, fatal crashes averaged 685 (Reference 3.5.1-10). Secondly, the basis for this probability includes crashes under IFR conditions and those which occurred at airp6tts with a much greater traffic Thirdly, the number is based on distance density than Elma. from runway but does not reflect, as might be expected, decreasing probability with increasing angular deflection s from the runway extended. For twin-engine plane's, the effective target area is assumed 3. ft, per Subsection 3.5.1.6.1. Tnis is 2 to be 239,000 conservative since the few twin-engine operations at Elma are by planes of approximately 6,500 lbs, for which the vulnerable area would be less than estimated here. Notwithstanding the above comparison with design basis 3 missiles, the target area for single-engine planes is taken ~as the missile grating over the dry cooling towers, diesel i storage tanks, condensate storage tank, and steam tunnel on 2 the north side of the RAB. These areas total 42,000 ft, The probability of an aircraft impacting critical plant 4. areas is given by (Reference 3.5.1-6): PA = CNA

Question No. INSERT B Continued... 311.3

Response

(cont'o) where: 0 = aircraft accident probability per square mile N = number of aircraft operations per year that might affect the plant A = effective area of plant The probabilities of operations at Elma Airport contributing to an uncontrolled release of radioactivity from WhP-3 are: yr -1 A ((twin) = 5.3 x 10-8 single) = 5.6 x 10-/ yr -l P PA As with Subsection 3.5.1.6.1, these are the conservative probabilities of aircraft striking safety related plant structures; these values are not_ probabilities of exposures in excess of 10CFR Part 100 guidelines. In summary, operations at area airports do not present a hazard to safe operation of WNP-3. 3.5.1.6.3 Military Airspace Usage Military air traffic totals about 300 flights per year on pre-Using the same parameters as Subsection determined routes. 3.5.1.6.1 (assuming military aviation in-flight accident rate equivalent to civilian aviation), the probability, P, of a M military aircraft impacting the plant is: PM = 1.1 x 10-10 yr-l Military aviation is not a hazard to operation. 1

Question No. The FSAR states that the HVAC system, with the exception of some 410.40 (9.4.1) ductwork located in office areas and emergency living quarters, is designed to seismic Category I requiremens. Verify that the Control Room HVAC air intake chlorine and radiation monitors are seismic Category I. As noted in FSAR Sections 6.4 and 9.4.1 the Control Room

Response

Ventilation System, with exceptions, meets the criteria for Seismic Category I equipment. The control room air intake chlorine and radiation monitors are it,Junted inside the Seismic Category I air intake plenum duct. In compliance with the requiremer.;s of GDC-19, the control room air intake radiation monitors have been-designed and procured to meet Seismic Category I requirements as discussed in FSAR Subsection 12,3.4.2.3.1.2. Regulatory Guide 1.95 reconnends that, in order to comply with GDC-19, the chlorine monitors in the control room outside air intakes be qualified in accordance with Seismic Category I At WNP-3, these monitors were designed to remain requirements. functienal following a safe shutdown earthquake. As discussed in FSAR sections 2.2.3.1 and 2.2.3.2, an analysis of the worst case chlorine release was performed. This analysis was done in accordance with the recommendations of Regulatory Guides 1.78 and 1.95 and the standard review plan Section 6.4. The results of this analysis indicate that the combination of system detection limit, response time and redundancy, together with a routine maintenance program and other provisions for control room habitability discussed in Section 6.4, provide an adequate basis for concluding that WNP-3 meets the requirements of G0C-19. The FSAR will be updated as shown to indicate this exception to the recommendations of Regulatory Guide 1.95.

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m 4 uur-3 Feast TaeLE 1.5-3 Mut0E - 900g 8 m stageale REVIHF flas i "M see as are mar /=- usancs unim g 4.1 centret seen area veettlessen eyesen see. 3 - Jety test j ACCirtauCECaliigl'4 Accept. llity of the CahW5 design, se descelhed in the appilcent's safety l onelple report (5AA), le based on specific generet doelen criterte and recusatory guldet. The desty of safety-related pertlene of the Chews le acceptable if the latograted design of the system is in accordence with the fellenleg criteries as related to the systes W a6 capable of a g General Design Criterlea 2 i 1 witt.:tWlagtheeffectseIearthquetes. Acceptance Is hosed on asetlag t tae guldence of Aegulatory Emide 1.23, peef tlee C.I for safety related pertlees and peellten C.R for monstfoty-related partions. 3 I General Besly Celterlen 4. with respect to eststeletas emelrenmore A 2. canditless in the centrol rees cespetfble silth the deelge lletts al essentle) equiposat laceted therela durise eeruel, tremelent and stridset conditlene. a e meneek (s) (1) there ere me shared erecome er same e. General testy Celterten S. es related to abered systene and genponsets et uur-3. 3. ? leportant to safety. General testy Cetterien 13. as reisted to proelding adspete protecties ' s l 4. to paroit accese end occweacy of the centrol rees under accident consittens. Acceptance is based en meetlag the guldsace of Aeguietory Smide 1.73 reistlag to lastroentatlea to detect and alare any 'heaeresas chselcel release le the plant etclelty and reletlag to the systems capability te feelete the l control rees free such releases and see systems cepettlity to aset the slagle fellere criterles peeltlees C.S. C.7. and C.14, respectleel N

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Beguletwy Estee 1.35 reIotleg to the systems cepebility te llelt ab; an eccouletsen of chterine withse the centrol reen and the systems cepehility le fellere criteries._assittene C.4e and C.ed. to meet the ad M *M l I-@ S, les.Get to the sys cepebility to I the begulate. a"e en.0'"el" e "1"to .sedi:::"C :' '".."a'*"/ _.) No., v a/"M.. (w.d./.9s-e 7 a G.o-rai sesis. Crii.ri-u. as reioud.e,,e,. s.ce,.n 'f,e os t. e 3 s. s.ent. ause.t-e se me. sed. nsee.s radioactioe e3.u. e. rena. si."testi:tiistie. e.te.e.nce c,Tteru, ,e,.. e.d i,cear.i reisese of ne we.co of ee,.a'".e e eesi . wee fnt,t.es. d.ela eette. e. e,.tes al ,h-e cu-..yet-ord -i e. Sad edsorgtles units of ilght weter-cooled auclear e plante. .I eesillen 6.2, and posittens C.1 and C.2, respectlee

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/ mg. sf o FSAR \\ Inad[itiontothecommercialtrafficonfederalairways,thereismilitary ij g= E . Z g-traffic originating at Ft. Lewis Army Base (42 miles ENE) and McChord Air Force Base (48 miles ENE) that couldoverjlythesite. Nearly all the flights y ~=~ from Ft. Lewis in the vicinity of WNP-3M are IFR training flights on V204. I E Tnese amount to one or two flights per month by OH-I, OH-60, or CH-47 helicopters (Reference 2.2-11). Mcchord AFB originates about 280 flights per year by C-130 or C-141 aircraft that manuver in ene vicinity of the site. E, E These aircraft fly under visual flight rules (VFR) along predetermined routes [ that pass within four miles of the plants. No ordinance is escried on the military aircraft (Reference 2.2-13).. g E E 2.2.2.6 Railways Er A track enintained and used by the Union Pacific Railroad passes about 1-1/4 E, miles north of the reactors along the south bank of the Chehalis River. [ Another east-west freight line, owned by tne Surlington Northern Railroad, g iE passes within three miles of the project in the vicinity of the town of Satsop. 23 2.2.2.7 Proiection of Industrial Growth c There are no specific proposals or plans for new or expanded ip'2strial ). activities in the site area. Programs to promote commercial and industrial [ development it. tne area have focused on Grays Harbor as a potencial deep port i and the exploitation of the agricultural, forest, and recreational resources of the region (Reference 2.1-2). r 2.2.3 EVALUATIONS OF POTENTIAL ACCIDENTS [ d ? ---,. Ofthetransportationroutefinthevicinity,theUnionPacificRailroadtrack is most significant as a potential source of accidents. There are no industrial facilities which pose a threat to plant operation. 2.2.3.1 Determination of Desian Basis Events } Explosions - For most of tne route near tne project, the railroad tracks are about 350 feet below plant grade and shielded from the plants by the natural topography and a screen of forest. The distance (approximately 7000 ft.) exceeds that specified in Regulatory Guide 1.91 for a credible rail car load of explosives. Explosions are not a threat to plant safety-related facilities. Flammable Vaoor Clouds - The postulated rupture of a 85-ton rail tank car of propane was analysed by methods specified in Regulatory Guide 1.78 using very conservative assumptions (e.g., Pasquill F and 1 m/see wind speed). The model assumed the instantaneous release and vaporization of nalf the contents of the Tae tant car and the boil-off of the residual over a one-hour period. analysis ist.ored the effects of topography (including the 350-ft vertical 2 elevation dif ference) and dispersion through the dense timber and assumed that Wita chose the plant receives plume centerline vapor concentrations. ( conservative assumptions, the grouno-level concentrations at the plant are l below the minimum concentrations of propane gas in air necessary for --. detonation (2.4 percent by volume). 7tW,4 d Toxic Chemicals

1. The worst case exposure of a toxic chemi~ cal would result t

jg from the catastrophic rupture of a 90-ton railroad tank car carrying liquid ~.E"~ Tnis is the largest railroad tank car commonly used for shipping chlorine. l 2.2-3 Amendment No. 2, (12/82) -,e- ,n

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1. b 1803e6 WNP-3 4/f 40 75AR bs less enan'ene design capacity of the drainage basin (100 year rainfall).

c-1 ~"~~ Sicce the drainage flow capability of the basin is greater than assumed, any ptstulated tank discharge will be confined and not overflow the drainage basin. Th2 evaporation rate of the. ammonia, its airborne transport, and buildup in the Control Room was evaluated using the methodology in HUREG-0570 (Reference 2.2-15) and the physical properties of 29 percent aqueous solution i of ammonia in Reference 2.2-16. Ammonia concentrations inside the WNP-3 Control Room were calculated for a variety of meccorological conditions insluding the five percentile meteorological. conditions for the WNP-3 site. .h Ts ovaluate the impact on control Room habitability, the calculated concentrations 'were compared to the "Immediately Dangerous To Life Or Health" (IDLE) concentration of 500 ppe, which represents a maximum level from which one could escape within 30 minutes without any escape-impairing symptoms or eny irreversible health effects (Reference 2.2-17). Th2 results of the analysis indicate that the IDLE value is reached in the Control Room at the earliest approximately 270 to 280 seconds following arrival of the gas at the air intake. Since ene odor of ammonia is easily usticeable at a concentration of 20 ppa (Reference 2.2-18), the analysis indicates approsimately four minutes would be available to isolate ene Control Ro:s and don breathing masks whien are stored in the Control Room. Since two ninutes is considered sufficient time for a trained operator to put a sslf-contained breathing apparatus into operation, there would be sufficient time for the operators to take protective action and assure Control Room /. habitanility. Chierine Th3 potential effects on Control Room habitanility of chlorine releases I idsntified in Subsection 2.2.3.1, were evaluated by methods described in dagulatory Guide 1.95, Revision 1. Although the probability of the accident identified in Subsection 2.2.3.1 occurring adjacent to WNP-3/5 is very saali, the Station has installed l This equipment l aquipment capable of coping with the resultant concentrations. censists of chlorine detectors and charcoal filters. As described in Subsection 6.4.4',2, redundant chlorine detectors are located in the Control Room air intake. These detectors are capable of detecting a minimum level of one ppm chlorine and will provide signals to isolate tne l Contrcl Room. Including detector delay and valve closure time, the total isolation time is about eight se'conds. As indicated in Section 6.4, the maximum normal air intake rate into the Control Room envelope is 4600 cfs, and ths total infiltration rate into an isolated Control Room is expected to be For the net Control Room locs :. nan 250 cfm at one-eighth inch water gauge. 344.000 cu. ft., this corresponds to a normal and isolated cavalope volume of air exchange rate of 0.81 and 0.044 volume changes per hour, respectively. Scsed on thes's enaracteristics, the Control Room corresponds most closely to a Type I Control Room in Table 1 of Regulatory Guide 1.95 Revision 1. 74 'iClen L%lT l W aia),6wd7coc[+. is asGLufk14 i+W*%""*' **Nf- (OM ' "*'*d*/l 2 .C.. Assu ing that thy dis nce tween it po t of rel se inf the Co rol R a 'f , 'yy vs ilat' n inJ kes s 7. O fee the r concen acion 4esulti from 4 UP* imum ad owed b Reaul or ill 90 Jr6ns enlorine i ess t an the Id5* uide /.05.i Goosequently, tne WNF-3# Control Room design provides adequate protection 3mf+offsitereleasejofchlorine. p Amendment No. 2, (12/82) 2.2-6

1. ~ 3 1356W-3 FSAR 4 fro. @

s ,y.j 6. 2.2 Ventilation Svstem De sira Subsection 9.4.1 provides an overall description of ene Control Room Area Ventilation System which includes two engineered safety feature (EST) Atmosphere Cleanup Air Filtration and Adsorption Units. These Units will be referred to as the Filtration System. A system air flow diagram is presentad on Figure 9.4.1-1. Figure 6.4-2 is a plot plan showing the plant layout, including the location of onsite potential radiological and toxic gas release points with respect to the main control room air intakes. S The type of system provided for WNP 3Yis zone isolation, with filtered recirculated air and capability f or pressurization. Divisional separation is provided for redundant components including the missile protected outside air intakes located at opposite sides of the Reactor Auxiliary Building. All components, instruments and ductwork are classified as seismic Category I, Quality class I and Safety Class 3, except for a non-seismic portion of ductwork within the envelope which serves emergency living quarters Aush die rt. a da.4 4*t s Redundant ^ :lig 01 2 : ', ::i:i f: 6..., hlorine detectors and redundant G8adM W f 6*d**4%I)*. lass IE radiation monitors located inside the outside air intake plenum will provide individual signals to isolate the Control Roon envelope by closing the outside air intake isolation valves GPV-M61SA or 3PV-516133 and 3PV-B0635A or 3PV-B163SB) and exhaust air isolation valves (3PV-5060SA or 3PV-B160SB and 3PV-B062SA or 3PV-516255) for the respective operating train in the event of a chlorine release or radiation emergency. Radundant Class IE thermocouples are located at each outside air intake and automatically isolate # the control room when temperatures o f 127 are sensed. The Control Room Aree Ventilation System is automatically switched to the isolated mode which includes recirculation of all air which is filtered through the filtration system. The Filtration System includes two-20 percent capacity redundant EST

• Atmosphere Cleanup Units described as tne Fission Product Removal System in Subsection 6.5.1.

A comparison between the existing design of the Control Room ESF Atmosphere Cleanup Units and the positions of Regulatory Guide 1.52 Rav. 2,1978 is shown in Table 6.5.1-2. The' modes of operation of the Control Room Area Ventilation System and considerations for habitability in the Control Boca during a design basis accident are described in Subsection 9.4.1.2. The Control Room Area . Ventilation System component design data is presented in Table 9.4.1-3. 6.4. 2. 3 Leak Tisheness Table 6.4-1 summarizes the infiltration analysis perf orned to decernine ~ The leakage analysis is based on Control Room Envelope un' filtered inlaskage. a 0.25 in. water gage differencial pressure and is in accordance vien AEC R & D Report NAA-SR-10100, " Conventional Buildings for Reactor Containment," dated Ma y 1, 19 65. The Control Room envelope is provided with self-closing, single and double air lock doors. The unfiltered inleakage (10 cfm at 0.12 5 v.g. differential pressure) resulting from controlled egress and ingress has been estimated based on Scandard Review Plan (3RP) 6.4, Rev. 1. h, M 2 4* 6.4-3

u..- .a-a n -..--..u a_ _. -- -- -_ M9 b. 4356W-5 y;p,3 N'O ' O TSB Room back to the operating air handling unit and exhausts the balance of air t The flow between the air ventilation rate and the toilac exhause flow rate. jij toilet exhaust f an discharges air diently to the atnesphere durinC normal operation. The emergency mode of operation is automatically initiated by any one of the fallowing: n) Containment Isolation Actuation Signal ~ Redundant radiation monitors located inside the outside air intake b) plenus sense high airborne radioactivity and isolate the Control Roc.. If the sensed radiation levels exceed 10CTR20 linics an alarm is activated in the Control Room. c) Redundant ::'- ': r,;::7. chlorine detectors located within the outside air intake planum are sensitive to 1 pga. A chlorine gas concentration exceeding 1 ppa will initiate a signal to close the outside air and exhaust air isolation valves and activate an alare in the Control Room to alert operators. Redundant thermocouples are located within the outside air intake d) Upon sensing a 120F sir temperature or greater, a signal is plenus. initiated to close the outside air and exhaust air isolation valves and an ' alarm is sounded in the Control Room. The redundant subsystem and both air cleaning units CU-1A and CU-13 will start All outside air intake and exhaust and their associated dampers will open.The Control Roon Area Ventilation and air isolation valves are closed. Filtration systems operate with the recirculation and filtering of air through HEPA filters and charcoal adsorbers while the Control Room remains isolated The Control Room can be kept isolated f or 175 hours without the need to replenish oxygen and maintain an eccupancy of 12 persons only f or protection of toxic chemical release. Through this period the Control Room operator can admit outside air through air cleaning unic CU-1A (or CU-13) prior to discharge into the Control Room envelope to maintain an acceptable concentration of oxygen and carbon dioxide for occupants. 1 During a radioactive release the Control Room envelope is assumed to operate in an isolated mode from 0-20 minutes at which time the Control Room operatorThe pres can manually initiate the pressurized soda of operation. will support the Control Room personnel (12 persons) plus the technical i support center personnel (25 persons) f or a period of 30 day:s within the nare are no limiting factors for oxygen depletion and limits of GDC 19. carbon dioxide buildup in the pressurized mode of operatior.. Manual initiation of emergency operation can occur whenevee the Control Room operator determines its necessity. f.p M $5

f. tl. f Q) gg. I,(le/S A)

1336W-6 3 Qas Ndb_. mo.@ l 6.4.4 DESIGN EVA1.UATION l 6.4.4.1 Radiological Protection The evaluation of the radiological exposure to the Control Room Operators is presented in the Control Room ace,ident dose analyses given in Chapter 15, Appendix 151 shows the doses f allowing the design basis accident (LOCA) and demonstrates compliance with CDC 19. Table 6.4-2 is a summary sheet of the Control Roon HVAC System parameters used in the Main Control Room dose analysis. Figure.6.4-2 is a plot plan showing the plant layout, including the location of onsite potential radiological and toxic gas release points with respect to the main control room air intakes. Elevation and plan drawings showing building dimensions are given in Section 1.2. Potential sources of toxic gas release are identified in Section 2.2. A description of system controls and instruments is provided in Subsection 6.4.6. Redundant Class IE radiation monitors located at the outside air intakes are discussed in Subsection 12.3.4 6.4.4.2 Toxic Cas Protection a) Protection f rom Chlorine A1 MS.j L i._;. L..,,..,. chlorine detectors located in each outside air 1 intake of the Control Room Area Ventilation System are capable of detecting a minimum level of one ppa chlorine and will provide signals f or isolation valves to start closing within five seconds from detection of chlorine. The delay time f oc automatic isolation of the ' Control Room is eight seconds. This delay time includes detector [.isar

2. response time and valve closure time.J The Control Room operator will assure tnat at least one of the two air cleaning Units is operating in the recirculation mode throughout the accident period.

k, sad N s ' b) I:arbon Dioxide Generation and Oxygen Depletion The f ollowing assumptions were used f or determining carbon dioxide generation and oxygen depletion of the Control Room during a postulated accident. 1) The number of personnel in the Control Room envelope during a toxic gas accident is conservatively selected to be twelve. 2) The Control Room HVAC system is in the isolated mode with both outside air intakes closed to prevent infiltration of toxic gas into the Control Room envelope (i.e., valves 3PV-8061SA or 3PV-B161SB and 3PV-8063SA or 3PV-B163SB are closed upon actuation signals addressed in Subsection 6.4.3). .Y Amendment No.1, (10/82) 6.4-6

Ouestion No. 410.40 INSERT 1

Response

(cont'd) Sodium Hypochlorite is used at WNP-3 for water treatment as discussed in FSAR Section 10.4.5 and thus onsite chlorine gas mleases art not a concern. INSERT 2 A routine maintenance program will be perfomed to confirm system operability. l INSERT 3 The Supply System has taken exception to portions of the reconnendations of Regulatory Guide 1.95 as noted below; 1) C.4.d(4)- The chlorine detectors do not meet the criteria for Seismic Category I equipment. They were designed to remain functional following a safe shutdown earthquake. As discussed in FSAR Sections 2.2.3.1 and 2.2.3.2 an analysis of the worst case chlorine release was perfomed. This analysis was done in accordance with the recommendations of Regulatory Guides 1.78 and 1.95 and Standard Review Plan Section 6.4. The results of this analysis indicate that the combination of system detection limit, response time and redundancy, together with a routine maintenance pmgram and other provisions for Control Room habitability discussed herein provide an adequate basis for concluding that WNP-3 meets the requirements of GDC-19 "as it relates to providing adequate protection to permit access and occupancy of the control room under accident conditions".

Question No. In FSAR Section 9.5.4.3, you state that an automatic fire pro-430.23 taction system is installed in the fuel oil tank structures. (SRP 9.5.4) The fuel oil transfer pumps are also located in these struc-Consider the design basis seismic event with resulting tures. inadvertent operation of the nonseismic fire protection systems Show that operation of the in both storage tank structures. fire protection systems will not impair the safety functions of the fuel oil transfer pumps ans asscr.iated controls and alarms. The fuel oil transfer pumps, valves and filters are mounted on a raised pad and located in a watertight, walled cubicle within

Response

The 00ST room is the Diesel Oil Storage Tank (DOST) rr m. protected by a non-seismic fire proti. tion system consisting of three nozzles in the D0ST room and ow nozzle in the pump cubicle. The design basis seismic event with inadvertent operation of the fire protection systems in both fuel oil tank rooms will not impair the safety functions of the fuel oil transfer pumps system for the following reasons: Fire Protection System Piping is seismically supported and 1) will not fall on the equipment. Fire Protection System discharge in the fuel oil tank 2) structures cannot enter the watertight compartments. Fire Protection System discharge inside the watertight 3) pump cubicle is limited to 20 gallons per minute from the The resultant discharge requires a period single nozzle. in excess of 56 minutes to reach the fuel oil transfer Operator action from the main control room, pump base. upon receipt of the deluge system waterflow alarms will close the system control valves located outside of the fuel oil tank structures, and cease flow within 30 minutes, precluding impairment of safety function due to submergence of equipment. The transfer pump motor is a totally enclosed, fan cooled 4) motor controlled from a motor control center in the diesel Any spray from the nozzle in the pump generator room. cubicle on the pump system will not affect motor operation l

Question No. FSAR Subsection 6.5.2 describes the operation of the Containment 450.1 In this Spray System as a fission product removal system. (SRP Section, the FSAR states that a nitrogen cover gas is provided as 6.5.2) "the driving head for the tank contents to the containment spray This subsection also states that the NaOH injection rate lines." is controlled by the flow rate of the containment spray pumps. Describe in greater detail the NaOH Injection System including: the mixture weight percent of NaOH, the design overpressure of the nitrogen cover gas, the method by which the pump flow rates con-trol the injection rate (including any operator actions necessai'y to manipulate the control valves), the time delay expected from automatic initiation of the spray system until the preset minimum flow is established and chemical addition begins, and how a Con-trol Room operator can determine if the Na0H addition rate will adjust the spray pH within the prescribed pH range.

Response

The Supply System has not completed the response to this question A complete response will be available by March 1984. as yet. i i

9 Question No. As per Regulatory Guide 1.70 indicate whether, and if so how, the 471.1 guidance provided by Regulatory Guide 1.97 has been followed (12.3.4) concerning area radiation and airborne radioactivity monitoring instrumentation. Reference or provide this information.

Response

Refer to the response to Question Numcer 471.23 for a complete ~ discussion of Wif-3 compliance with the guidance of Regulatory Guide 1.97. l t i

Question No. 10CFR20.103(b)(1) specifies that the licensee shall use process 471.12 or other engineering controis to the extant practicable to limit the concentrations of airborne radioactive material to below that specified in Appendix 8. Table 1, Column I for any room, enclo-sure, or operating area; or when averaged over the number of hours in any week during which individuals are in the area,'ex-coeds 25 percent of the amounts specified in Appendix 8. Table 1 Coluan 1. Table 12.2.2-2 of the FSAR lists the fraction of Maxi-mum Permissible Concentration in air for "those areas normally occupied by operating personnel" as.52 MPC for the Fuel Handling In addition, Su11 ding and 184 MPC for the containment building. Table 12.2.2-3 lists 43 areas with airborne concentrations i greater than MPC. Describe why additional process or engineering controls are not used to lower these concentrations in each of these areas as required by 10CFR20.103(b)(1). (See SRP Refer-ences, Sections 12.3-12.4.11.3.)

Response

FSAR Tables 12.2.2-2 and 12.2.2-3 of the FSAR indicate airborne i activities in areas normally occupied by operating personnel as These well as those where the potential for exposure exists. airborne concentrations have been calculated based on very con-servative assumptions and operating conditions which do not nec-For example, the essarily represent normal plant conditions. activity concentrations of the leaking fluids are based on a pri-mary coolant concentration assuming operation with a 1 percent fuel cladding defect. Further, the leakage rates, assumed con-In addi-stant, are representative of design basis conditions. tion, equilibrium conditions are considered to be reached instan-taneously in the areas with the leaks regardless of leakage rate, l ventilation rate or' duration of plant operation. space volumet As a result the airborne concentrations and dose rates presented 1 in the tables are upper bound values which are at least four times as high as would be if the recommended SRP 12.2 value of 0.25 percent and the NUREG 0017 Value of 0.12 percent fuel cladding defect were used. The airborne concentrations for the areas with C/MPC values Table greater than I have been reduced by a factor of 1/.12. shows these C/MPC values. The areas, with concentration 471.12-1 ratios greater than 1 are located in radiation Zones IV and V. The information included in the tables, regardless of occupancy requirements, indicates areas with potential for airborne contam-As discussed it has been derived from conservative ination. The results, therefore, do not assumptions and methodology. necessarily reflect the need for additional process or engineer-ing controls to lower these concentrations but rather point to the need for performing calculations based on more representative plant operational data on reactor coolant concentration and equipment leakage. t --w-n,- %--*-7gw,- ,,y,- 99-wm,---v--

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TABLE 471.12.1 RADIATION ZONES WITH r. APCS GREATER THAN 1 BASED ON 0.12% FUEL FAILURE I Location C/MPC Radiation Zone 2.208(1) Containment Charge Pump #2 1.404(0) IV* Charge Pump #2 Val Gal 2.328(0) IV Charge Pump #3 Val Gal 2.256(0) IV Charge Pump fl 1.404(0) IV Charge Pump #1 Val Gal 2.832(0) IV N-2 Recycle Ink Val Gal 1.139 (1) V N-2 Comp Val Gal 4.500(0) V 2.064(0) V N-2 Compr Gas Analyz Phy Val Gal 4.356(0) V** Waste Gas Compr #1 VG 3.516(0) V Waste Gas Compr il 1.404 (0) V Waste H-2 Recombiner 1.224(0) V Gas Recomb Compr Val Gal 2.808(0) V 4.020(0) V Gas Recomb Compr Waste Gas Compr #2 Val Gal 4.356(0) V Waste Gas Compr #2 1.404(0) V Gas Surge Tnk Val Gal 5.028(0) V 1.404 (0) V Gas Surge ink 1.704 (3) V l Resin Mtrg Pump 1.872(1) V Dewatering Tank l

• Zone IV: Max Dost Rate = 100 mren/ hour Design Base Occupancy less than 1 hour / week Posted zones and controlled entry
  • Zone V:

Unlimited Dost Rate Positive access control Controlled entry and occupancy l _..-,_r...

Question No. As per Standard Review Plan, NUREG-0800, commit to the implementa-471.23 tion, or provide a description of your alternative approach, for the following: 1) Regulatory Guide 1.97, as it applies to providing radiatiori monitoring instrumentation following an accident. 2) sigulatory Guide 8.12 and ANSI N16.2-1%9, as they relate to ti:s requirements for a criticality accident alarm system. 3) Regulatory Guide 8.19 as it relates to your method of per-forming assessments of collective occupational radiation dose as part of the ongoing design review process so that expo-sures will be ALARA. 4) ANSI /ANS-HP SSC-6.8.1-1981 as it relates to the criteria for locating fixed continuous area ganna radiation monitors, and for design features and ranges of measurement. 5) Stone and Webster Topical Report, RP-8,1974, as it relates to methods of analysis employed in determining shielding requirement. 6) Regulatory Guide 8.14 as it relates to use of personnel neu-tron dosimeters where exposure to neutrons occur.

Response

The WNP-3 design for post-accident area radiation and airborne radioactivity monitoring instrumentation complies with USNRC Reg. Guide 1.97 Rev. 2 requirements with the following exceptions: Instruments monitoring airborne radiation exposure rates in-la) side buildings or areas which are in direct contact with pri-mary containment, where penetrations and hatches are located (Type C variable) will not be supplied. An increase in radi-ation levels in these areas would be due primarily to stream-ing through the penetrations or to direct shine from the con-tainment caused by elevated exposure rates inside the con-Under these conditions, any additional increase in tainment. the radiation levels as a result of airborne radioactivity leaking from the containment would not be discriminated from streaming or direct shine by the area radiation monitors reconnended by R.G.1.97 Rev. 2. The four Auxiliary Building Airborne Radiation Monitors provided in the Reactor Aur.iliary Building can detect airborne radioactive material leaking from the containment.

=. Question No. 471.23 Response (Cont'd) In order to comply with the requirements for Area Radiation Ib) monitoring (Type E variables) in areas where access may be used to service or operate equipment, monitors have been placed at locations of WW-3 which were identified and anal-yzed as vital areas in the WNP-3 TMI Shielding Study (See FSAR APP.12A). Allmonitorssuppliedhavethgdynamicrange R/hr to 10 R/hr) recommended in RG 1.97 Rev. 2 (10-2 with the exceptien of the monitors supplied in the two valve Operating Enclosure areas at elevation 351.00 ft of the RAS, and the monitor supplied in the post-accident Containment Atmosphere Sample Panel area at elevation 362.5 ft of the RA8. According to the TMI Shielding Study, the maximum dose rate expected after an accident in the two Valve Operating Enclosure areas is not more than 5 R/hr and in the post-acci-dent Containment Atmosphere Sample Panel area.is not more than 0.7 R/hr; therefore, the supplied monitors for these 4 R/hr. three areas have a dynamic range of 1 mR/hr to 10 Regulatory Guide 1.97 recomends that a High Range Circulat-l Ic) ing Primary Coolant Monitor (Type C variable) be provided for As discussed below, detection of fuel cladding breaches. WNP-3 does not strictly conform to this recommendation, but does meet the intent of the Reg. Guide. l During normal operation and anticipated plant transients, WNP-3 employes a process radiation monitor for detection of cladding breaches -- see Subsection 9.3.4.5.6.1 of the FSAR for further discussions. Monitoring for detection of cladding breaches during and fol-lowing accidents is accomplished primarily through the use of the Inadequate Core Cooling instrumentation which constitutes a defense in depth, for informational sources relative to the approach of cladding breach -- see appendix B of CESSAR-F for further discussion. Grab samples are used to obtain more specific information on the radiation concentration in the reactor coolant, as discussed in Sections 9.3.2 and 9.3.5 of the WNP-3 FSAR. The above methods are considered to provide plant personnel with all necessary information relative to fuel cladding j breaches, while avoiding inadvertent or intemperate contami-nation of reactor coolant processing systems and equipment. l

Question No._ 471.23 Response _(Cont'd) In place of the airborne monitors recommended for "All Other id) Identified Release Points," area radiation monitors will be provided with ranges of 10-1 to 104 mR/hr. The "other" identified release points are the Gland Steam Condenser Exhaust and the Steam Generator Blowdown Flash Tank Vent anu No flow measurement Auxiliary Condensate Flash Tank Vent. devices exist on these effluent release paths since any release will be slightly above ambient pressure. In place of the Radiation Exposure Meters at Fixed Locations le) of the Plant Environs (Type E variable), Thermoluminescent dosimeters (TLD) will be provided. A number of the R.G. 1.97 radiation monitoring variables re-If) The sensitivities quire Health Physics laboratory equipment.of the WNP-3 previously in response to NRC Question 471.6 (See Letter G03-83-893, dated November 22,1983). WNP-3 does not plan to place a criticality alarm monitor near A criticality safety analysis was 2. the new fuel storage area. performed using industry codes and the results indicate that the geometric design of the new fuel storage area precludes Furthermore, under worst case conditions, the criticality. geometric spacing limits the effective multiplication factor, < 0.80 An Keff, to less than 0.98; calculated valueexeatption from requirement of 10CFR 70.24(d) in the new fuel storage areaG. D. Bou was requested in letter G03-82-1324, Knighton, dated December 28, 1982 Criticality alarms in accordance with the requirements of Regulatory Guide 8.12 and ANSI N16.2-1969 (now ANSI /ANS-8.3-1979) do exist near the refueling canal and spent fuel pool. Refer to FSAR Table 1.8-3, page 1.8-446 for the Supply System 3. position regarding Regulatory Guide 8.19 Refer to FSAR Table 1.8-3, page 1.8-446 for the Supply System 4. position regarding ANSI /ANS-6.8.1-1981. Refer to FSAR Table 1.8-3, page 1.8-447 for the Supply System position regarding Stone and Webster Topical Report'RP-8. 5. \\

Question No. 471.23 Response (Cont'd) l The guidance given in Regulatory Guide 8.14 " Personnel Neu-6. tron Dosimeters" has been followed except that an alternate accuracy ;equirement has been substituted for the Position 2 The alternate requirement used AmBe accuracy requirements. instead of Cf-252 because of its longer half life and lower gamma dose rate. The alternate requirement is: "When exposed to either unmoderated or moderated (4 Poly) neutrons from an AmBe source, the average accuracy of a set of 10 dosimeters exposed in the range of 100 mrem to 3 rems should b _+50%." All remaining guidance and the alternate accuracy require-ments have been incorporated in TLD purchase specifications and acceptance criteria, and will be followed in TLD and neu-tren survey instrument calibration procedures, neutron survey procedures, and dosimetry procedures. The FSAR will be revised to reflect the response to this question. i l l

'! !i.

j; 55 WN P-3 f

FSAR TABtX 1.8-1 (Cont'J) l \\ Exceptions FSAR 3 Iso Yes Section Besarke Regulatory Date Culde No Rev Issued Title 1.92 1 2/76 Combiatag Model Respassee and I 3.7.3.1 slote 2 See CESSAR-F Spottet Componente in seisele Response Analyste a I Table 8.1-2 1.93 0 12 /71 Awa11eh!!1ty of Electric rower Sources I 17.2 alote 1 l3 1.94 1 4/76 QA Regelrements f or Imetettetten. Inspection and Testing of Structural Concrete sad Structural Steel ~l durlag tbs Ccestruction Phase of I seeclear tower Plante 1.95 1 1/77 Protection of Isueleer Feuer riant I 6.4.5 Table 7.1-3 Control Room Operetton Agelast j en Accidental Qalortee telease a 1.96 N/A (SWR) ) X 6.2.5.5 Isote 2 1 1.97 1 8/77 Instrumentation f or Light-Weter-Tables 7.1-1 Coeled thseleer rower Fleet to .ll@l; b e l y 2 12/80 Astees Fleet Conditione during Fw e nes ylic.is O and following en accident See !. bs.tel c.s t-1. 3.'t 1.98 N/A (BWR) l See CESSAR-F Note 2 l 1.99 1 4/77 Ef f acts of Realdual Elemente on Fredicted hadletion Desage to l teactor Vessel Meterials l i . g 4 h 1 y e ? 7 1 3 P I-i e e n 0 i a e r* *.-t r >t

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'l 1: kI a;: ' : ii ' 8 War.3 r5&e Tate I. l.8-1 pusEd - news s pec sT416ABB REvlsW flg vcs um M/A 6Dr/Acctt14aect C817tetA 9.1.3 Procese And Poes.Aceteent saartlas Systema Ree, 3. July 39e3 l 7, (cont'd) / $Ag., f e% r k(t) (th h $M $.g f,*,,, gg 3* Q 7q 2 To meet the requirements of GDC 1 and 2, the selselc design and qualitylines, components and instruments for both l 4. classificatten of samptf e the classificetten of the system to iAlch cuy 1 <ss to A (,,/u (;,,;.(, //j7 gr3e15 and PAS should confere hne and cesponent is c.onnected (e.f..,a sampli:fd.ine. con-l h s ste. sh. e sign.d il each sampl.l t,G,o...nd seis ic Catege g Qg 9.y to Quality Greg A and setselc Category I clasqs fication), in accordance n-ted ie ulth regulatory posittees C.1, C.2, e4d C.3 in Seguistory Guide I.N refer-and C.4 in Regulatory Gu a 1.29 't ence 6), regulatory peeltlens C.1, C.2, C.3(reference i), and the guidelines of Segulalery Guide 197 (reference 8). toepenents sad ofpfag downstrees of the second fseletten valve eey be designed to Quality Greap 8 and nonselselc Catesery I requirements, inaccordance ulth regulateiy pesillon C.3 in Re

• g p t fe % fk,)

ence 6). The post-accident sampilng system and operational procedures should meet the guldeilnes of item 11.8.3 la sluBEG-0737 (reference 9) and of Begula-tery Guide 1.97 (reference 8), and t'e 5. n t lt To meet the requirements of 60C 13 and 14 la Appendia A Ia 10 Cft I Part 50, if cheelcal analyses shou that chlerte concentratten in a. the reacter coolant esceeds the Technt:al Specificaties llatts, then y verificatten that the disseleed saygen concentratten is belou theVerificetten of Technical Specificatten lletts alls he aandatory. hydrogen residual in escess of le cubic centimeters (at standard temperature and pressure) per &llegram of reacter coolant util beacceptable in llou et direct analysis of disselved I see sa=ese (2) (2) h-I s.e e.. sheti s.s,.,i,y s e nos. d r,r sh, s,u e., To meet the eequirements of GOC 60 in Appendia A to le Cf3 part 50jf on-line,)as threestegraphy is used for r C**I**' ***8 F 6. i ag) should be f l special provistens 4.a., pressure relief end pt. en entering the available to prevent hS-pressure carrier gas reacter coelant. ,,""'**'"Ug 8 3' '**

• 8 " 8 8 " * ** 8 's e e e,.

e s t s u. ,,,,,,,,,4 ) le meet the requirements of GOC 60 la Appendia A to 10 CTR Part 50, 8 8 ' *"' * *"'.8 " s e=i t.e, 88 84 rearsar e tana e,.. e.,. passive flew restrictf ans in the sampling Ifnes may be repleted by 7, c. '=rsa*e ee es.e r redundant, fully quallfled, remotely operated Isolation valves toThe automatic contain-A d limit potential leakage free the sample lines. l E ment isoletten valves should close en containment isolatten signa s g $ All remotely operated valves should have %. D er safety injection signals. assured power suoplies and control se that trity can be reopened after Valves which are p ) an etcident without clearing the isolallen signal. Inaccessible during an er.cident should be environmentally quellffed V8 ~ to ensure operability under accident conditlens. p 1 l l

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1 T&EE l.8-3 e Merd - 9000 BC Stestam MTIst FLAN C00fLIAIICE g g gg serfeccretamct centssia Frecess And Ef floset Bodleteelsel steettering Isetreestatles And Seselles Spesens J 43.3 see 3 - July 1988 l accarlAssca Calltsla Elst acceptance criterle for the process end ef fluent redlelegical meelterlaglastrumentatten and semptlag systems are based i seats of the felleming regulations: It Cf B Port te $20.106 as it relates to redleactivity usetterlag of I t A. offluents to unrestricted erees. General Oestga Criterlen 60 es it estates to the redleective weste money-I j samt systems being designed to control release 47 redleective meterfele or s..tu I,,, ,og.1 er, 8.

• ee et es see 8

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Gener68 Bestga Criterle 63 and 64 as they relate to the tedlesctive weste senegement systems heleg des!gned to esatter redletten levels and leakage. k hk MM(7 {.Y l C. bIedt lj 7 Specific criterte necessory to meet the relevant regstressets of the Commission W Nd*sbTus 46 replettens teentlfled ebeve ere: QhQ [.3 Prowlslems should be made for the lastrumented esatterlag er for the sempitag I

• ~

end emelyseg of all merest and potentiel effluent pathways for release ofredleactive meterials to the environment to me 1 N b 4 a ese essers (t) Islems of To meet Criterten 64. the design of systees shewld meet the latory teguletory Guide 1.28 (Positten C and Appendia A) (Gef. 2), llcohle) Ref. 3), Calde 1.97 (Pesillen C and lette 1 or feele 2end Segulatory Culde 4.15 (Posttlea C) (Gef. I)es es, I the poseems sad llemld process streens or effluent release pelats } shouid be maaltered and sempled accordlag to lebles 1 and 2. e. l j I f or both Swas and PWRs. liguld westes and confland volmes of geseems weste should be sampled batchwise peter to release, la accordance h. Centlauses geseems effluent mealters t with Beguletory Guide 1.31. are met regelred for eyes structures such es.PW turtles b.uild.lage q 7 .r eie. spheric mis f.e u id.ste temas c ni.iaing tr. tedfor Ilquid r l' I processed figuld weste and located outside of buildlags.end geseews efflueets that commet be practicably esaltered or seepted [ k bettemelse one of the following methods of representative seepitag l4 Q I. should be provided: I See asesrb (2) p) temleessel eengsesg., ena on,,ggag 7-j y A centinuous propertioning sospilog system with et leest tue a a ia + e n - e se the system should be designed to p '"'*d""'*-'"- p { (1) fie. rene of the s~e conecie. re-4

• e-semple c,ellectien tents Le wide. i.e4 e,. es.,e m se.pi.d stre.e disen.re..r. cit.ra.tl.ely a

J 'j i I See seeerb f 33 A perledic automatic grab sospling system with et least f.we semple cellectica tents. the systee ehem14 he desfoned to (2) i 1 +. =e. e;* *

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1 T& ALE l.8 3 pepsG - 9000 e E eTasR>Aes atelaf 73.ag T" Yt ser/AqtrTaurr CetTsetA Preseos and Rifloosa Bedialeeleet tenetsostee Isottometettee And Samp!!ag eyeteos ees. 3 - Jety 1988 31.5 iceas'd) cellect e fleed eelume of semple et a rete proporttemet to the sessured flew to the 6 espied streen discherge. 3.dgeleg eel enestees., d ,g ges of ligned g E pee sneert (t) (3) Radleective meterials other them asble gemee la geseems effluents. and geocene efflesese se preended betehwies. A centinuous sempting system with replaceable partlcelete Illter (3) The system shem14 he designed te outenet* and redleledlee edsorter. fcelly tene semples at a fleed er seesered flee rette of the temple throughput to the sempted streen dinherge flow. for latermittently operettag effluent release pelats, the system should E e designed to estematically tote seselst whenever there is flew le j the effleent strees. / 3 for all of the above seseles, e periodic emelysis frequency for thecollected semples should es specified le the technical specificellees. I bl 97 ) a see senerk (2) (2) tar 1 g ,f gg3 previstens should.e ende for the lastrumented smelterlag of. er the g.,, gg ,g,g,g g gg,y ,g*3, 1 ggy perledic or continuses semplies and smalyels of, redleettive mestees they relate to redle- _& a 1c, p3 g n cg**t ay y 2. To meet Criterle SS and 63 process systees.ective weste systems and detectlen of escesslve redlettom levels and g3. e.d ide i.,.ees should .g ,.u. i f syst -e tellen o.f appropriate, safety ac.tlene. the.d,es gn o e,.iet.r, ei.,,se.tle.>s.asto .se.si-lait seide t. cr.sitte.of 4, e.di. ii....thie J. te su.ee f g, gig er,e.s appilcohien;c c. mi.e r ie n. es appi ak'N r 3 ,,e isie.s smid Le ess.re.re,resenteu.e se.pIl f,e,.e = c 4e.ts..oo m.iet e. re,ie-u..,r.cet s stre-es. t. s. s f.or ue.e.ld.eeste,eo ts < ceil-tl.a er se,ie t.st tem.ieu.getarateof tiesstw t e c., of smid for gaseems lige 4 process tuo tank solumes le elebt beers sample streas seagles. proelsleas shesid be nede for Islees lines and for retalag plateemt la sample lle 6.for gaseous samptlag vres kts and stects should be a egreement 1 with Aln:lilL1 (sef.5) 2 Were practicable, proelsions should be nede te collect samples free process easte streens at centrol sample stations to redece 6. leenage, dance wit $ and redletten espesures to aperetlag perseemel spillaes fw 3 h sap Sectlen 9.3.2. N la uter [ le streams back z 4 4 Pr6elstens should be made to purge and drela s e treatment D to the systee of origle or to en appropriate we c. )J y a systee. Preelstens should be nede for adelaistrative and precedural centrolfor necessary sosillary or ancillary egelpment. and P'. q 3. k i ,iii g....oggg . c., geg 4 y g g g g e =w "d hehe aha he g$ fg. g ,g

~ ~ " ~ ~ ~ ~ ~ ~ ~ ~ * * " ~ "' .... N........ m..,.~ss.l...~..-~~~~~--~-""~~----"~ l 1 } i 1 Mar-3 F8aB TaEE 1.5-3 unsas - Seet WC stameAD MTIEW fLaf l ~ cneDLt#FE van n s/a i ser/mernura cetrasta e remen ame sistwat meesetestal an tseelg 3aessementesten one semptsee yesen.see. 3 - Jety 1904 } es en.t enesden. aierta t.es. nee seese-la.s . e . (ins me.411 he emeloped peer se Smet seed (i.e. Isless-8C***'e) is,esentlei -.fn.ent.at6s. tne desi.m of.,ste.s one.u ette and immes-4737. Isen II. F. 3). est samfS-4737 (tofs. $ and 83) seet the previstens of le 6 4-8734 is 11.5-A (thss SN secties) g g Q A rz.% 4 -t 1 and 3 fltees II.F.1. Attachsea Golde1.I( ltlen C. and table 1 er gg7 "{ gg gp_"2 LSet.11). and begole.s'Isof. n. rg p lai. a..s e. isces j Provistens steeld 6e made for the festrimented anniterleg er semplieg a gm and emelysis of identitled if wid effluent petts la tAs esset of ale meet Criterien 64. as it relates te posteleted 5 eccid.nts one identified ligste effluent paths, the design of plantliquid weste collectfen and erecessfag streams postuteted eqcident. g lines saferenced la $ar lectlens 9.3.3 and 11.2 and, fa additles. the fall.= lag conditions should he met: asser se ee esessee 31.3 et este aar eestsee ser n (3) 3 ese same d (3) deseen of semptsamme. f Adel61strative centrols and precedures la confermance with subsection II.3 of this 68P sectlen are to te la effectto slatelse landverteet er acefdental releases of I a. sofer se senesessen 31.1 et ekte so eestsee ser i ligulds, and a see semest ( n (3) { degree et g ttense. tiguld effluent redlelestcal sentters are to be provided forthe entsmetic levefneties of releases la the avant that effluen S b. as provided la subsection II.I of this SAP section. I and as esleblished in the plant techalcel specificetfees, are setpotats esteeded. I l 1 5 I .i .g h, 4 b'

• ~

i >2 I E VI p' i i I v a 88

9 r WNP-3 FSAR thele 1.8-3 uunte - Song lWC 67409409 OcVISIs ft8df creeLlasiCE M W SIA $3r/AccfrTAsiCE CBittela 12.1 - 12.4 Bodletten Protecalee Destge Features bee. 2 - July 1991 (Cent'dl 4. Regulate Guidel.SL,gesign gstt and 9telatenpace Criterte for F~est Aic neeree-tere eisture e cieenig _sy Hi n ir Tilliallii~iE( rgrp(lie ~Bil s e TM _ NGenee PiseT Plan [s es it reiTe es to redlettee protectlen consiliialliis for Elf ( [g g Q[*m QQ simesphere cleasug systems operable esador postuteted 084 condillons, to be designated as primery systees." ti C)CL t A h,e n s. W 0 d' V / / y"k f,p gg,, + seaulaterv selm 1.59 *ceacrete medletten sbleids for mucine pe=e z 5. Plealf.'esItrelateIInterealmgresiileandrecommendedpracti con-lilied la AN51 bl41.5 - 1972, *contrate Bodlettee Shields' es acceptable I far constructlen of facilities, applicable to occiqpettenal redletten protection shleidtag structures for aucteer power plants. K sec h tk(si -water-ce muci Ps si plEs.1@E.& = wtrumentation for tlNTiilo7r "ny(ri1ber segulate s. ~ refil t - las n sen iirsiartiaisirtii4 a. is niildisi " n Ir.stes' m to snound.ccepteE1de ' E stirrier capsy-Ine.Hrm cm.l resui ttom te proel e lu tr tette for h redletl .e iterl., folie l.a en occident la a lisht. ster ceeied clear peuer plaat. c le f i.e precuces la n.dietle. sulle s.]2T3Perri,r a=1.lserf.itiscisTeaTiti.n.eatreil.a

i. see let*

• is 1rr sisiranifer siinitiii programs or edelaistrative perseasel.

!!!!!! E,y, v. u n i. c isst .a r ,.tlens.It. ra ,d t.....i M 4 " !; E E t W M "'~' '~ .rm part 20,1(c), concernlag the radiation protection informellen to be l supplied in Settlen 12 of Safety Analysis Reports shout actlens te6en durlag design, constructlen, operation, and decomenssioning to metatela p 2 occupallenal redletten espesures as low as is reasonably achievable. -t b 3, 4 A eircuns. es irihellenal for gestatelal_ag Dec HeAatery(culdee10'aperatingiillrisures.a iocins niin. l p W t. riisue iru .:ty-r non nous i.ie,,uunnsraarurar T co issio s o regulattens with re ed to le cf B part 29.l(c), concernfag the commitment p by management and e glience by the Redletten protectlee tsaaeger and the V g redletten protectlen staf f to maintain occupational radiatten espesures 2 as low as is reasonably act:levehle. t

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g ll; liNP-3 FSAR TAeLE l.a 1 \\ enetc ONul peC $148SAelt sEelEW PI g creeri t Aucs

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its le@ M/A ser/ Af f E FTAserg gret e 12.1 12.4 medlett.a Protecslon sealga resteses see. 2 - July 1984 (C as'd) m s ees := e..-eas 6 see. eee e.ned se,is.e-3.

• celltceilt g A cipes per.aitems." n it relates ses lesser soni-er-lize.

3 s a see me.e.* m r seguulerr_Sai_d'.s. it le a syslea accepiaileDilme sfi;f f lor meeting Ge comeTeston's require-no. seats for a criticality acideot stars system. l

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Dese Assessment la il 4 Guide 8. l

11. Segi.leteEirrisil sa ir%

e senter conoure i,t.mr,iairi -occupaste al redietten dose es part er the o.eelns d. sten re,lew process se that such.. pes.ru will se AtAaA. (3)n.. *II.s..sie s.,iner.3 aces.. dis,i,m... i s a s.. n arh (a) anse(c-olos *standeed Tes.hnical_spes_lficetjene, for sabcock_and_ WilcesPensurlishiler Eiicle s;* n it'relitu le radistion prelutlen cea-cE es 12. ilderettiis Isliie epplicentllty, format, and laplementatten of the Setcett and Wilces techalcol Specificelles putage. e Oficettene for General lecti I see semark (3) esIrrilatestoredliife~n[$ri h

33. NuefG-8323. " Standard Techalcel len Selllag Water Reactorg [5dt's) Ilty, foreat, and Septementetten of the cidii^derellens~lillei spilicat

~ General Electric Techalcol Spuf ficetten pactoge. (1) Tece.ascas epec s essoal we s t 6, e,ses,a p.e., z see e sh (3) cificeMons for testustlea ingtliin t_o ri31311ea prilislien,meeM i 14. " Standard Techalce eine(G-ott? coa-e,. g,,,,,,,, g i,,., i PressurijdWilerliicl36;'lity format, and laplementation of the as re ildefillons li W applicatl toebustlea Inglaeortne Tuhnico),5pecificetten pectate. a see o m rk (t)

• Standerd Techalcel welficettene for Wesths, Instic-oe52 15.

P,esisiliin v.tiraisilagrir1T'iliter13Teatatria protistlea coa-i iwritliii'li'llsrrpilicebility, ternet, sad leptementottea of the Westinghouse Techalcol Speci8icellen package. I see o m rh (4) W II.r.t will se met pyl.e to fall pe.wes ree. ele.e enielc-orts and NuttC-GFU u they relate to laplementlag task Actica Plan P. 16. d so .s so.,,.cere s, esp. liiii'11:5.2 ind'1171(3) for constructlea peralt aad operettaa licease j opplicattens. 8 9 .r.. esene .. s be ....ee.len deff ele sh, same-d t Celterle for Ar n sedletten ^ W (5) ishe

• tecellen and Des &in irirmiies to I w w..s m rih 98*1 ANil/ Ale 5 HP55C-t 8.1-

....e... 17. iriteile for istdllisinist 'if'liiilliarfeT Flaisce;ntlavons eres samme e rusiirmere se ser 6..ee ee.cl.de i ene f. I radlellen monitors, and for destga features and reages of stesurement. - deel rigetta e a esmpit e with the Isolan's Q f J e ellene wil be provide by Jeane B

== ,4 AMil Wl).l*t969 "Gulde to Sampli Altborne Sedleactlwe festerIale g y 1

1. hocleir iacIlitlii7ii'it tilNei Etiii erliclilii ulilifisiily la 14.

i .7 iLiaisliipTelfd siiilu et alrt.orne redle ctive meterfels, and accept 6te g metheds and materfels for ses and perticle saapilas. ,,,,m,....,,.,,.,...,,,,,, _, _ y, m.,,,,

UNP-3 FSA1: Tass.t s.e.1 l pupsc. gaae ISC ST4fspaSP OEvgew ff AM cunrtiasers un so ela r -- ser/arcernuce carwara .r.3 - tr.4 s.asu s r,seeense. ee.s. rees. u se,, a. s.,3,gt e troe 'd> i I See Bemesh (t) (3) Specific esenet tee bee been respeessed eer wur *. 8 Aaril N16.2- "Criticajjty Accident Alare Systees." as it relates to Kli Wi Te_1969,Wiiliin of crIIIGlily acclassis in the hendtlag, 13. see 3.sse, pyg.st-t}}i. r (Ee storlag. processing, and transporting of fissionable motorials. ,1 Astilbl01.6;}977,'tencrett Bedletten Shie1dg.* es it relates to roepstre-a 10. seals e^iriecemien3id procLITerfor const_uctien of concrete radletlen r shielding structures. f ineers." m. n. Schaeffee telter-l or muctor i

21. *se n ter shletal 05At 51 ~,DF as if Fiistes to the ihteld Tesigni'ag

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i -ae aa-usaersii mas,- niw ve ir iti n e i de;a"',,,e,,6--,, y ;',,,g' '.,, ;>,,,, 3 ribe !g u i;miricisiwirewwin.imiira u.ra. iiaiosining s,uint aiei.rg., e. = ing gyei, sis e.co,ed in .e .o he... h .e es 1. faCattiv st5ica ieAtuais @ Q(,q /,f,,3 g y-V o.e;4.se.4 / A.or..iaisity or the secule, sign entwo in se sued e,r. e.ede.co that 3,s,% ;j cg,,, g(,y wg,,.Q j,g u., .ucot e.n suunied the ese uniu re<iverm.ts. a cea cert n.wi. 20.1o3, and to.noe, n ti n t u radiation protectlen aspects er senere it6hAbl**f RNP 3. gf.3sgo,j,/j g. N Whl & #4LN6*lb #4 44s f4.w.g [g gp**M. i This includes evidence that a Design Criteria 19 and 63, and le Cf3 60. H.functlens (meistenance, refuellag, redleective JO4 fg gg g., e sajor espesure accumulett Q3".*, e/to1 9m saterial handling, process ag, etc.. In-service inspectles, callbretten. dece.misstening, and recovery f ree occidents) have been tensidered le plant e s> (,,; p,j 4% d,4.j % { g ,( g, design and that potential redletlea espesure free these activities will be / AM dd M.s !..dg.$ g%( /44cgj. trpt AtAAA la accordence with Is tra Part 20.l(c) and Regulatory Evides 8.g (y vjg< ~..g y,.( g g h end 8.lo by radiatten protectlen features lacerperated la the deef. $ech J y ,4, features may include (t) ease of accessibility to work and Inspect on sad s g g.s.g*N jA; 4ag*.8% 4 s..o,,n, a.m.,in u.e.n n,i.o ud ce s.ource i.ie su,. in. sign -as-es eig** * "'f s******. g t is d,, u,,, e.c o. disi, o ea, e-rete.o. er - o.eied cortesi. products. (4) the acility to reduce glee re< pelted in rodf atten (felds, and (S) (- 8 % t FJr 'As /c I.

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f / l werP 3 FSAE "j TAR.E 1.0-3 q a.mts - 8000 N asc st^-"" eEVIEnf PLAff ---a tr. L jjl!@ ires.A _ WIA seFfACCEPTAIICE CRITESIA Dev. 3 - JelF 198I pedtatten tretectlee Seelt9 feet ges 33,3 13,4 (Caet'd) f l

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will met create en additlenet redlettes heaerd to personnel ee atefa agAcceptability et the weattletten system, idence er the.e in %djecent accepted etees. re7etive to reJioactive gases and, articulates will else be based en evthat the appilcent has applied the guldence of Segul t acceptable alternettres move been'preposed. Segulatory Sufde 1.52, particular2r Sectlene C.4 and $, frevides guldence that a sen be is.ed in this review, althawsti the g4lde le weltten with regard toSeed practice la that I altipattag accidents involulag alrherme redleettivity. i ity regste is appilceble to no eat opere.ttaa es well slace relee.e of redleect v

m. netty different only la quantity from la morn.el operettenal occurrences 1 some of the accident cases.

7.) AA(A RASIAlleet Af8 Alatomet likel0ACTIVliY IWell10Alldi $Y5f tJIS 8 (1) uueRo 0737 Teek Acates Ptee seen 114F.1 shall be met yster l 4. 3 ses 8emask (O The aree endletten mentterlag systees will be acceptable if theg. ,, g,,,,,,,, g og, seet the proutstes, of le Cte Part 29.383. Segulatory Guido 0.1Begulatory Gelde 3.97 and IIUREG-4737, lesk Ac ( e. 4 / g end the following triterfer, f Prlaclpel protectica egelett latete of redleective seterleir le 3 provided by eastneerlag contrefs. y*~ " -MI (3yrts.. rser es est t.; -- ene Inst sees + sterte eatst e 1. a ee he oc aseee s The detectors are located la steen iAlch may be eereolly SeF be for s tudtag t the -3 I actupted without restricted eccess and iAlch may beve eefelds la encese el the redlet'les tems ei aien.. in ue-t. e e b j re essene I h.. seed by Jane Its, patential for.redlettee..im. isoa.a i. s-o.e u.s.n.i.c s.r.r..

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accordance with Alt 51/ Alt $-IIP 55C-5.8.1. go# ym g pg4 g. 1.,.0.. e,.,,evide e.. ue ro ui.g.,., *se.. e u.et fey 7 m,uieu.,e-1. 7 raa ci.,. t,e.. es. ..e,ete euct,eto e,-ano

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e WNP-3 FSAR Taett B.Sa3 _MIRFC '= tettle WC 51as*AllI8 DLVIEW liaW O!EIIM J. m get n/A ser/AarrTAtsrE 6:alTEst A l2.1 - 12.4 Badiatles Fe.tecslen peelen Features Bew. 2 - JalF 198I ((' ens'd) The in plant accident redlettee moniterlag sys:;as will be acceptable (l) c. If they meet the felleulag criterfe: ef 3. personnel have the capattlity to assess the redletten heterd.la a see s, ,6 til til T k art e a rien s eem s s.F.s f suesr; serir wa s he re pa s.,

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la accordance with the criterle of Iten II.T.1 of IRIG(4-0734 pcb. ' O.3.Y and 8737. and seculetory Guide 1.97, (4) & y,.m ,.a g a. ~,. , ort le 1 stre.este t. .se. la the even.t e, e-l.e.t s l. l.co. se as to - re ll... ell. l. t. rs.n I respondlag to en emergency. 3. (mergency peuer sheeld be provided for lastelled accident eentterlag systems. The accident menttering systees should have essable ranges which 3 4. T* laclude the aeslew conculated accident levels; and should be J designed to operate p. q*ely le the environment caused by the occlaset. 3 S. Applicants for cps and Ots should provide tue high-range redle-a see p sh (2) (2) thse save shall be e pas., se sual p uer preses. ties onetter systees in centelament iAlch are documented to meet the retreirements of Table II.T.1 of hum (G-elle end $737. 3 Regulatory Guide 1.25, 4pendia A. Provides esseful guidance about d. ef fluent monitettae. that is applicable to the e4;teptability of alreerne radioactivity mealtering in plant. Seguletory Guide 4.2 lacludes guldence en surveys to evaluate radiatten heterds. American hattenal Standard ANSI all.l-1969 provides detalled asid-ance en sampling altborne radioactive materials la nuclear facillttee and may be used for acceptance criteria en the actual sampling process and certeln techalques involved. Regulatory Guide 8.s g provides further guldence en monitoring systems. a, Instrumentellen to monitor for accidental criticality tellt be aC*;ept* I See Bene h 41) (1) Spectifs e mesetlea free the reestremessa of 80CFR 7e.24 has 3 able if it meets the criterla of le CIB Part 70.24 (e)(1). Segulatory s. ,,,.,a n, per.3. see s eser era 1 et-lite. e. Guide 8.12. and ANil $tenderd kl6.2. S. D05E A$ittistlist p The dose assessment util be acceptable if it documents la apprepelote a N 'E J detall the assumptlens made. calculatleas used. the results for each N O M radiatloa sene. lactuding must.ers and types of worters frvolved la each. espected and design dose rates, and pro etted persen-ree deses, la g auerdance ulth neeuteter, celterla e t. v' .1 i ) Ij E d' ll 8 * "9 ' f t*14 t't t ? E 5.- ? ?:'

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Re$ lid nidae perrernance.criterie ver.everal congeries of redlettens 5. le 1 folieulag supe. ore under pecified condfttees. y )"::;',-':;;;,:,=:o;';; 2,gJg:;;;.. n'il'nWin.w f.il.2.e.*n;'U.ni.tn.i.e!!$Wininif.#eki$he'!" ' '- - a W d ......,1,. l a.i-ter. e i.e rete -e.-e o te.  :>a;;',;;;;;.;;'; ;;d :: =:.=;g N n'il'.n.W,iM.E.eu*4,',"i"fu.I4'n!Nitilii.iill'.iM!IlikeniF ' ' - ' - ~ ' ' ' > ,l "n w i.,a.. ,,,,,,,,,,,.... a... .e, -.,,u..... e.,,. ou.i ,ier t m au - ew -..-e e.ie , inn,. tin _e 't"i e.pe.-e.f swi.i sue...e.e,.-e :-- s,.m.- w* u.uauH*.no.ei.wieu. i. i.ier = --r.e as n> a. r-irn-ei ew tie cowsue= e ar si ne.ei re.di. -c 4 which the espe -e may occur, J tlenei Beaulaterv Guide S.S. Inter.stlen Relevant to fa.orlag_that Occ pi31su i. Tae M ure, et aucieer reuer StaTTin will se as le_w e. ea l 9. _ 1 Eiiii.451i't 25mcT5,." e. ss revete. to siillig ini-rsiguTre.eET.3 pre.Idlas redlettea pret-Lla lafo*"** p-te . of i EnliuEli t com.trucuen. epetellen, and 18~CFR'Per l 180 o actlen. taken darlas the de.lga deco.sts.lenlag to a..ure that occupatlenei tedletten empo.-e. cre tapt I t 1 1 ALARA. 7"' Eguet and m ni.imetienr.sutde s.9, Ac_ceptable concest. 8tedei._,le mon _len.ili'cencept., seplatory E 30. tspra Misa.. ens i y'u7sp 1en

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1632W-5 burtfie5 ~ ni.0 This monitor is located on the 362.5 ft. level of the Reactor Auxiliary Building. It takes a sample of the water discharged from the gas (= E stripper and returns it to the same line. Physically it is a fluid MdE stream monitor as described in Subsection 11.5.2.3.2, and it requires a sample pump. In this case the microprocessor has been removed from the [ skid and placed on a nearby wall to protect it from possibly high ?" radiation fields present near the skid. The measured activity levels are automatically transmitted to the system computer where they are recorded and available for display through the system CRTs. If the activity. exceeds pre-established setpoints an annunciation is made through phe system CRTs and event typer. The receipt of these alaras will alert the operator so that additional radiation surveys, sampling and analysis can be effected in order to determina the cause of the problem. The alara setpoints are i to be set between the seasured activity levels of the degasified reactor coolant and the maximum level of contamination permissible in this system. The setpoints any he adjusted continuously over the entire range of the monitor. The range of the' monitor is from 10-4 to 10 Ci/cc which is the practical range of interest for normal power operation using a simple single detector radiation monitor. e) CVCS Letdown Radiation Monitor The CVCS letdown radiation monitor will alert operations personnel to an increase in the radioactive contamination of reactor coolant as =- quickly as possible.

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__t n. 1 1his monitor is located on the 373.5 ft. level of the Reactor Auxiliary l Building. It receives a continuous sample of the CVCS letdown, in l parallel with the boronometer and is upstream of the purification filter. System process travel time delays the sample for approximately two minutes delay to allow activation products with short half-lives, particularly N-16 to decay. Physically, it is a fluid stream monitor a described in Subsection 11.5.2.3.2, which does not require a sample In addition, two modifications have been made, first the pump. microprocessor has been removed from the skid and placed on a nearby wall to protect it from possibly high radiation fields present near the skid, and second a removable attenuator has been provided which may be manually placed between the sample volume and the detector. The measured activity levels are au:omatically transmitted to the system computer where they are recorded and available for display through the system CRTs. If the activity exceeds pre-established N :;.;;.. 11.5-14 A:nendment No. 4, (12/83)

g go,flM, 1647W-1 WNP-3 FSAR' ~ 471.2.3-12.3.4 AREA RADIATION AND AIRBORNE RADI0 ACTIVITY MONITORING INSTRUMENTATION The area radiation and airborne radioactivity sonitoring instrumentation are a part of the General Radiation Monitoring System. The balance of the General Radiation Monitoring System is described in Section 11.5 " Process and Ef fluent Radiological Monitoring Systen." The Radiation Monitoring System is a digital computer based system that communicates with various nonitors located throughout the plant. The system is described in Subsection 11.5.2.1 and the system block diagram is shown on Figure 11.5-1. r The area radiation and airborne radioactivity monitoring instrumentation furnish information to plant operators concerning the radiation dose rates and the radioactive airborne concentrations for selected sections of the plant. These monitors are designed to assist the plant operators in maintaining a radiation exposure to plant personnel that is as low as reasonably achievable and in minimizing release of airborne radioact.ive materials to the environment. The area radiation and airborne radioactivity monitoring instrumentation follows the recommendations of Regulatory Guide 8.8 (R3) in that: a) Monitors have readout capability at the Health Physics Office b) Area monitors have been placed for optimum coverage A c) Component failure is indicated d) ' Monitors have local and Control R6os readout and alara e) Monitors have clear readout f) Monitors have a five decada readout range as a minimum g) Readings are recorded in the computer The instrumentation is capable of supporting an administrative program as described in Regulatory Guide 8.2 " Administrative Practices in Radiation Monitoring," February,1973 (Rev. 0) by providing background and airborne monitoring per ANSI N13.2-1969, and provides monitoring histortcal recceds in accordance with Regulatory Guide 8.2 (Rev. 0). Regulatory Gaide 1.21 and ANSI N13-1-1969 are addressed in Section 11.5. The monitors are calibrated and maintained on a routine schedule. As a minimus, the monitors are calibrated annually or during outages and after any maintenance work is performed on the detector. Subsection 11.5.2.2 describes the displays, alarm measurement logs and controls of all radiation monitors. Those monitors which are described as being seismic Category I, and meeting Class IE requirement are physically and electrically separated f rom each other in accordance with the criteria set f forth in IEEE 279-1971 and IEEE-384-1974 The monitors are qualified in l1 accordance with IEEE-323-1974 and IEEE-344-1975. Refer to Sections 3.10 and l 3.11 for further discussion on qualification of Class IE equipment. =f } (Z,,sulA I t_ 12.3-27 Amendment No. 1. (10/82) .. ~ _ _ _

Question No. 471.23 INSERT A Response (Cont'd) The WNP-3 design for post-accident area radiation and airborne radioactivity monitoring instrumentation complies with USNRC Reg. Guide 1.97 Rev. 2 requirements with the following exceptions: la) Instruments monitoring airborne radiation exposure rates in-side buildings or areas which are in direct contact with pri-mary containment, where penetrations and hatches are located (Type C variable) will not be supplied. An increase in radi-ation levels in these areas would be due primarily to stream-ing through the penetrations or to direct shine from the con-tainment caused by elevated exposure rates inside the con-tainment. Under these conditions, any additional increase in the radiation levels as a result of airborne radioactivity leaking from the containment would not be discriminated from streaming or direct shine by the area radiation monitors reconsnended by R.G.1.97 Rev. 2. The four Auxiliary Building Airborne Radiation Monitors provided in the Reactor Auxiliary Building can detect airborne radioactive material leaking from the containment. lb) In order to comply with the requirements for Area Radiation monitoring (Type E variables) in areas where access may be used to service or operate equipment, monitors have been placed at locations of WNP-3 which were identified and anal-yzed as vital areas in the WNP-3 TMI Shielding Study (See FSAR APP. 12A). All monitors supplied have the dynamic range recommended in RG 1.97 Rev. 2 (10-2 R/hr to 105 R/hr) with the exception of the monitors supplied in the two valve Operating Enclosure areas at elevation 351.00 ft of the RAB, and the monitor supplied in the post-accident Containment Atmosphere Sample Panel area at elevation 362.5 ft of the M8. According to the TMI Shielding Study, the maximum dose j rates expected after an accident in the two Valve Operating l Enclosure areas is not more than 5 R/hr and in the post-acci-dent Containment Atmosphere Sample Panel area is not more than 0.7 R/hr; therefore, the supplied monitors for these three areas have a dynamic range of 1 mR/hr to 104 R/hr. ic) Regulatory Guide 1.97 recommends that a High Range Circulat-ing Primary Coolant Monitor (Type C vartiable) be provided for detection of fuel cladding breaches. As discussed below, WNP-3 does not strictly conform to this recommendation, but does meet the intent of the Reg. Guide. f

= t Question No. INSERT A 471.23 Response (Cont'd) During normal operation and anticipated plant transients, Wif-3 employes a process radiation monitor for detection of cladding breaches -- see Subsection 9.3.4.E.G.i of the FSAR for further discussions. Monitoring for detection of cladding breaches during and fol-lowing accidents is accomplished primarily through the use of the Inadequate Core Cooling instrumentation which constitutes a defense in depth, for informational sources relative to the approach of cladding breach -- see appendix 8 of CESSAR-F for further discussion. Grab samples are used to obtain more specific informaticn on the radiation concentration in the reactor coc %nt, as discussed in Sections 9.3.2 and 9.3.5 of the W 3 N R. The above methods are considered to provide plant personnel with all necessary information relative to fuel cladding breaches, while avoiding inadvertent or intemperate contami-nation of reactor coolant processing systems and equipment. In place of the airborne monitors recommended for "All Other Id) Identified Release Points," area radiation monitors will be provided with ranges of 10-I to 104 mR/hr. The "other" identified release points are the Gland Steam Condenser Exhaust and the Steam Generator Blowdown Flash Tank Vent and Auxiliary Condensate Flash Tank Vent. No flow measurement devices exist on these effluent release paths since any release will be slightly above ambient pressure. I In place of the Radiation Exposure Meters at Fixed Locations le) of the Plant Environs (Type E variable), Thermoluminescent dosimeters (TLD) will be provided. If) A number of the R.G. 1.97 radiation monitoring variables re-The sensitivities quire Health Physics laboratory equipment. of the WNP-3 Health Physics Laboratory Equipment are provided in Section 12.5.

Ouestion No. In accordance with Regulatory Guide 1.70 " Standard Format and 471.25 Content of Safety Analysis Rep >rts for Nuclear Power Plants," Rev. 3. Section 12.3.1, it is our position that the FSAR should The include plant layouts showing shield wall thicknesses. shield thickness of major radioactive equipment can be provided in a separate table. Subsection 12.3.2 of your FSAR should be revised to comply with Section 12.3.1 of Regulatory Guide 1.70. Respons'e In accordance with Regulatory Guide 1.70 " Standard Format & Con-tent of Safety Analysis Reports for Nuclear Power Plants", Rev 3, Section 12.3.1, the shield thickness of rooms containing major radioactive equipment in the Reactor Building, Reactor Auxiliary Building and Fuel Handling Building are provided in the new FSAR Table 12.3.4-2 (attached) with the corresponding plant layout in the new FSAR Figures 12.3-34 through 39 (attached). FSAR Table 12.3.4-2 consists of the room number, as depicted on the FSAR Figures 12.3-34 through 39, followed by the room description. The following columns describe the thickness of the vertical walls (ft) in each cubicle (North, South, East and West) as well as the ceilings, floors, labyrinths and shield doors (inches) where appropriate. On those walls which are constructed of concrete block, it is in-Where " MAT" is indicated dicated as such by an asterisk (*). under " FLOOR", this refers to the building base mat. FSAR Section 12.3 will be revised to include the aforesaid information. l l

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.m SPAN-4(10), is used. This program calculates the dose rate at a point from any number of sources having complex geometry and complex shield The geometry of the sources-and shields are described by configurations. suitable intersection of quadratic surfaces. ISOSHLD, SPAN-4, and Rockwell's method approximate scattering ef fects in the shields by using appropriate buildup factors. Concrete gamma scattering probless are treated through use of the Chilton-Huddleston dose-albedo met' hod.(ll) This method finds use, for example, in calculating scattered dose over partial height walls and through cubicle doorways. None of the above-mentioned methods co'nsiders the energy degradation ("sof tening") of the source energy spectrum a's the radiation penetrates the shield, and thus each predicts conservative values of the dose rate at ?he point of interest. Whenever neutrons are involved, or whenever scattering ef fects are expected to For example, be important, more advanced, transport theory codes are used. the shielding design for the neutron streaming shield configuration (see Subsection 12.3.2.3.2) in the vicinity of the reactor vessel employed these state-of-the-art methods. Output from the two dimensional, discrete-ordinates DOT transport code was conver gg into input for the MORSE Monte-Carlo The MORSE code solves the transport code by the DOMINO program. neutron or gamma ray transport problem in realistic geometries by following a sufficient number of particle random walk flight paths through the system. Various sampling and biasing techniques are used to reduce the number of The combination of i histories required to produce results of a given accuracy. DOT-DOMINO-MORSE allows the accurate calculation of primary source neutron and N-gamma radiation transport, as well as secondary gamma production and transport. Comparison of the near.ured dose rates at operating plants, both BWR and PWR, with corresponding calculated dose rates, indicate that the models and method of calculation used predict higher dose rates than actually observed. Therefore, shielding calculations based on such models and methods are i conservative. 12.3.2.3 General Shielding Design ZWwRT' _ JWe 12.3.2.3.1 Reactor Building Primry Shield The primary shield consists of reinforced concrete with a ninimum thickness of The annular cavity between 6 f t. - 6 in. which surrounds the reactor vessel. the primary shield and the reactor vessel is air cooled to prevent overheating This annular cavity also contains two and dehydration of the concrete. In neutron streaming shields which are discussed in the following subsection. addition to being a structural member of the Reactor Building, the f unction of the the primary shield is to serve as the first barrier to attenuate and limit radiation that emanates from the reactor vessel. i k ~a ) 12.3-11

6t Di.W vaC3 Outstion No. 471.25 Response (Continued) INSERT-1 The shield thickness of rooms contining major radioactive equipment in the Reactor Building, Reactor Auxiliary Building and Fuel Handling Building are provided in FSAR Table ".3.4-2 with the corresponding plant layout in FSAR Figures 12.3-34 through 39. .i 1

i i Q q 7L Af 2793-1 TABLE 12.3.4-2 SHIELD WALL THICKNESS SIssL AUXILIARY BUILDING CONCRETE THICKNESS (PT) THICKNESS EDOMS WALLS CEIL-FLOOR LABY-DOOR ROOM # DESCRIPTION N S E W ING RINTH (INCHES) A101 Floor Drain Tank A 6 1 5 2 2 Mat 2* A102 Floor Drain Pump A 6 1* 1* 5 1 Mat 1* A103 Floor Drain Valve Gallery 6 1.83* 2 1* 1 Mat 1* A105 Floor Drain Tank B 1.83 2 5 2 2 Mat 2* A106 Floor Drain Pump B 1* 1* 2 1* 1 Mat 2.17* A107 Floor Drain Valve Gallery 2.17* 1* 2 1* 1 Mat 1* A108 Floor Drain Hester 2 2.5 5 2 2 Mat 2* A110 Floor Drain Recirculating Pump 2.5 2.5 5 2 2 Mat 2* Alli Concentrate Storage Tank 2.5 1.33 5 1.33 2 Mat 2 A112s Concentrate Storage Pump 1.33 5 5 1 2 Mat 1 A112b Concentrate Storage Pump 1.33 5 1* 1 2 Mat 1 A113 Spent Resin Storage Tank 3 5 3 3 3 Mat 3* A114a Spent Resin Transfer Pump 1.67* 5 3 1.67* 1 Mat A114b Spent Resin Dewatering Pump

1. 67* 5 1.67* 2 1

Mat 5.75 A115 Spent Resin Pipe Chase 1.67* 1.67* 3 2 1 Mac A116a Secondary High Purity Wst TK A 1 5 1 1 3 Mat A116b Secondary High Purity Wst TK B 1 5 1 1 3 Mat A119a Secondary High Purity A Valve Gal .67* .67* 1 .67* 1 Mat .67* A119b Secondary High Purity B Valve Gal .67* .67* .67* 1 1 Mat .67* A120a Secondary High Purity Pump A .67* 5 1 .67* 1 Mat A120b Secondary High Purity Pump B .67* 5 .67* 1 1 Mat A122 I.C.W. Pump A .67* 5 1 .67* 1 Mat 4 A123 I.C.W. Pump B .67* 5 .67a 1 1 Mat A124 I.C.W. Valve Gallery A .67* 67* 1 .67* 1 Mat .67* A125 I.C.W. Valve Gallery B .67* .67* .67* 1 1 Mat .67* A126a I.C.W. Tank A 1 5 1 1 3 Mat A126b I.C.W. Tank B 1 5 1 1 3 Mat A127 Secondary Particulate Waste TK A 1 5 1 1 3 Mat A128 Secondary Particulate Waste A Pump .67* 5 .67* 1 1 Mat A129 Secondary Particulate Waste B Pump. 67* 5 .67* 1 1 Mat A130 Secondary Particulate A Valve Gal .67* .67* 1 .67* 1 Mat .67* A131 Secondary Particulate B Valve Gal . 67 * .67* .67* 1 1 Mat .67* A132 Secondary Particulate Waste TK B 1 5 1 5 2 Mat A137 Shutdown Cooling Heat Exch B 3 5 2 2 2 Mat 1.17* A138 Aux Feedwater Pump B 3 5 2 2 2 Mat A139 HPSI Pump 2 5 2 5 2 Mat A140 LPSI Pump 2 2 3 2 2 Mat A141 LPSI Pump 2 2 3 2 2 Mat A142 Charging Pump 2 2 5 2 2 Mat 2 A143 Charging Pump Valve Gall 1.5 2 5 1.5 2 Mat 2 A144 Charging Pump Valve Gall 2 2 5 2 2 Mat 2

• denotes concrete block

2793-2 G.47).h~ TABLE 12.3.4-2 (Cont'd) STEZL AUXILIARY BUILDING CONCRETE THICKNESS (PT) THICKNESS BOOMS WALLS CEIL-FLOOR LABY-DOOR ROOM # DESCRIPTION N S E W ING RINTH (INCHES) 1.75 A145 ICW Heater B 2 2 2 5 2 Mat A146a Charging Pump Valve Gal 1.5 2 5 1.5 2 Mat 2-A146b Charging Pump Valve Gal 1.5 2 5 1.5 2 Mat 2 7 A147 ICW Vapor Body 2 2 2 5 2 Mat - A148 Charging Pump 2 2 5 2 2 Mat 2 A150a ICW Bottoms Pump 1 2 1 1 2 Mat 1 A150b ICW Bottoms Pump 1 2 1 5 2 Mac 1 A151 ICW Bottoms Tank i 1 1 5 2 Mat A154 Reactor Drain Pump 1.5 2 3 1.5 2 Mat 1.5 A155 Reactor Drain Pump 2 1.5 3 1.5 2 Mat 1.5 A156 Equipment Drain Tank 2 2 3 5 2 Mat 2 7 A158 ICW Vapor Pump 2 2 2 5 2 Mat A159 ICW Heater 2 2 2 5 2 Mat 1.75 A160 Containment Spray Pump A 2 2 2 2 2 Mat A161 Containment Spray Pump B 2 2 2 2 2 Mat A162 HPSI Pump 5 2 2 5 2 Mat A165 SECO High Purity Sep1 Tks & Pumps 5 Open 1 5 2 Mat A166 Holdup Tank E 5 1 1 1 2 Mat A168 Holdup Tank E Valve Gallery 5 5 1 1 Grat's Mat A171 Holdup Tank C & D 5 5 1 1 3 Mat A173 Holdup Tank Valve Gallery 5 5 1 1 Grat's Mat A175 Holdup Tank A & B 5 5 1 1 2 Mat A176a Holdup Recire Pump 5 Open 5 1.17* Open Mat A176b Holdup Recire Pump 5 1* 1.17* 1.17* Open Mat A176c Holdup Recire Pump 5 1* 1.17* 1 Open Mat f A180 Boric Acid Tank A 1 1 5 1 2 Mac 1* A181 Boric Acid Tank B 1 6 5 1 2 Mac A182a Boric Acid Tank Pump 1* 1* 1 1* Open Mat 1* A182b Boric Acid Tank Pump 1* 1* 1 1* Open Mac 1* l A186 Shutdown Cooling Heat Exch A 5 3 2 2 2 Mac 1.17* l A187 Aux Feedwater Pumps A 5 3 2 2 2 Mac

• d2 notes concrete block 1

f

j: 2793-3 QLl71.15 i TABLE 12.3.4-2 (Cont'd) i STEEL CONCRETE THICKNESS (PT) THICKNESS AUXILIARY BUILDING ROOMS WALLS CEIL-FLOOR LABY-DOOR ROOM f DESCRIPTION N S E W ING RINTH (INCHES) A201 Floor Drain Vapor Body 2.5 2.5 5 2 2 2 A202 Radioactive Pipe Chase #7 1 1 1 1 2 1 A203 Radioactive Pipe Chase #9 1 5 1 2 3 1 A204 Radioactive Pipe Chase #1 1 1 1 1 3 1 A206 ICW Heater 3 2 2 2 5 2 2 A207 ICW Vapor Body 2 2 2 5 2 2 A209 Valve Operator Enclosure 2 1 2 5 2 2 A212 Equipment Drain Tank Valve Gall 2.17 2 2.17 5 2 2 2,17 A213 Equipment Drain Tank Valv s Gall 2 2.17 3 5 2 2 2.17 A214 ICW Vapor Body A 2 2 2 5 2 2 A216 Valve Operator Enclosure 1 2 2 5 2 2 A217 ICW Heater A 2 2 2 5 2 2 A218 Radioactive Pipe Chase #2 1 1 1 1 2 1 75 A219 Floor Drain Pumps Valve Gall 5 5 1 1 3 Grat 's A220 Radioactive Pipe Chase il 1 1 1 1 2 1 A221 Holdup Tank Valve Gallery 5 5 1 1 3 Grat'g A222 Floor Drain Heater 2 2.5 5 2 2 2 A226 Radioactive Pipe Chase #5 1 6 1 1 2 1 A226a Radioactive Pipe Chase #6 1 6 1 1 2 1 A227 Radioactive Pipe Chase #4 5 1 1 1 3 1 A228 Radioactive Pipe Chase #3 5 1 1 1 3 1 I l l

2793-4 097lM IABLE 12.3.4-2 (Cont'd) SIssu CONCRETE THICKNESS (PT) THICKNESS AUZILIARY BUILDING EDOMS WALLS CEIL-FLOOR LABY-DOOR ROOM f DESCRIPTION N S E W ING RINTE (INCHES) A301 Floor Drain Condenser Area 2.5 5 5 Open 2 2 A303' Floor Drain Vapor Body 2.5 2.5 5 2.5 Grat's 2 2.67 A303n Floor Drain Trap Valve Gall 2.17* 1.67* 1.67* 5* 2 2 1.67* A303b ' Secondary Part Trap Valve Gall 1.67* 2.17* 1.67* 5* 2 2 1.67* A304a Decontamination Sample Tk Pump 1* 0.67* 5* .67* 2 2 .67* A304b Decontamination Sample Tk Pump 1* 1* 5* 1* 2 2 1* 2.17* 2.17* 2.17* 5* 2 2 1.67* A305a Floor Drain Trap A305b Secondary Farticulate Trap 2.17* 2.17* 2.17* 5* 2 2 1.67* A306 Seal Inj Filter 2.25 5 2.25 2.25 3 3 A307 Seal Inj Filter Valve Gall 1.67 5 1.67 2.25 3 3 4 A310a Flush Tank Valve Gall 1 1* 1* 1* 3 3 4 A313b Fuel Pool IE Valve Gall 2.5 1.67* 1.67* 1.67* 3 3 A312a Fuel Pool Filter Valve Gall 1 5 1.5 1 2 3 A312b Fuel Pool Filter valve Gall 1 5 1 1.5 2 3 A312e Floor Drain Filter Valve Gall 1.5 5 1.5 1.5 2 3 A313 Fuel Pool Filter 1.5 5 1.5 2 2 3 A314 Fuel Pool Filter 1.5 5 1.5 1.5 2 3 A315 Flush Tank 5 1 1 1 3 3 A316 Flush Tank Pump 1 1 1 1 3 3 A317 Seal Inj Filter 2.25 5 2.25 2.25 2 2 A318 Seal Inj Filter Valve Gall 1.67 5 2.25 1.75 2 2 A319 Floor Drain Filter 1.5 5 2 1.5 2 3 A320 Fuel Pool Ion Exch A 5 2.5 2.5 2.5 3 3 A321 Fuel Pool Ion Exch B 5 2.5 2.5 2.5 3 3 4 A322a Fuel Pool Ion Exch Valve Gall 2.5 1.67* 1.67* 1.67* 3 3 4 A32*.D Boric Acid Conds Valve Gall 1 1* 1* 1* 3 3 A323 Boric Acid Conds Ion Exch 5 1 1 1 3 3 A324 Floor Drain Demineralizer 5 1 1 1 3 3 4 A325a Floor Drain Demin. Valve Gall 1 1* 1* 1* 3 3 4 A325b Radwaste Polish Domin Valve Gall 5 1.67* 1.67* 1.67* 3 3 A326 Radwaste Polish Demineralizer 5 2 2 2 3 3 A327 ICW Filter Valve Gall 1 5 1 1.5 2 3 A328 SHP Filter Valve Gall 1 5 1 1 2 3 A329 ICW Filter Valve Gall 1.5 5 1.3 1.5 2 3 A330 SHP Filter Valve Gall 1 5 1.5 1 2 3 4 A331a SHP Domin Valve Gall 1.5 1* 1.67* 1 3 3 4 A331b ICW Demin Valve Gall 2 1.67* 1.67* 2.17* 3 3 A332 SHP Domineralizer 5 1.5 2 1.5 3 3 A333 Seal Inj Heat Exch 2 5 1 5 3 2 1 A334 Boric Acid Filter Valve Gall 1 5 1 1 2 3 A335 Reactor Drain Filter Valve Gall 1.5 5 2 1.5 2 3

4 1219W-li WNP-3 FSAR 6.2.2.4 Tests and Inspections ~ i Preoperational testing of the CSS in conjunction with tne chemical additive subsystem is presented in Subsection 6.5.2.4. In summary, these tests will n functions. demonstrate that the systems art capable of fulf,illing their desig & : w a-f& P orp roof e f HemJ : The in-servi inspection of the spray nozzles shall be limited to visual examination enty, Since the size of tne nozzles (3/8 inch orifice diameter) is sunstantially larger than ene largest particle in the system. (.09 inch) clogging of the nozzles is unlikely. Periodic testing and inspection of the system active components, i.e., pumps, valves, etc. will be performed in accordance with the in-service inspection requirements of the ASME Code Section X1. Tne in-service inspection and testing of the CS pumps will be performed by using the minimum recirculation line. A flow meter in ene line is provided to check pump performance. Operability indication of each component during testing will demonstrate the continued state of readiness of that component to perform its design function. The detailed program for initial performance testing is covered in Section 14.2. 6.2.2.5 Instrumentation Requirements The CSS is automatically actuated by the Containment Spray Actuation Signal (CSAS). The CSAS is iniciacea in two independent actuation channels A and 3 with the instrumentation and controls for the equipment in loop A physically and electrically separated from the instrumentation and controls fer the The instruments and controls are designed in general in loop 3. equipment accordance witn IEEE Criteria and are qualified to withstand post-LOCA l The instrumentation associated with the CSS environmental conditions. provides information to the operator in the Main Control Room for monitoring all modes of operation. Refer to Section 7.3 for further discussion. Justification and selection of instrument setpoints are discussed in Operation of the CSS is fully automatic. Suosection 7.3.1.2.2d. Instrumentation is provided to enable the operator to assess the status of the system in ene standby or operation mode and verify proper operation of the System parameters that enable the operator to monitor the system performance are flow, pressure, temperature and position indication of all system. remotely operated isolation valves. All instrumentation is redundant to satisfy single failure criteria. Y* l Amendment No. 2, (12/82) 6.2-93

1736W-3 WNP-3 FSAR The Containment Integrated Leak kate Test will be scheduled to the extent practicable, during a period of forecasted constant meteorological conditions. Prior to the start of the test, containment test conditions of temperature, pressure and humidity will be monitored for a period of about four hours to ensure stabilization of containment conditions. The leak rate test period shall include a period of 24 hours at a peak accident pressure. If it can be demonstrated to the satisf action of the Commission that the leak rate can be accurately determined during a shorter test period, the agreed upon shorter period may be used. 14ak rate tests shall not be started until essential temperature equilibrium (less than 0 2F/hr) has been obtained within the contained volume. ( Test instrumentation is to be specified in the test procedure. This includes instrumentation to measure containment pressure, temperature, and humidity as well as local leakage. An error analysis of these instruments is made to balp establish the validity of the Type A Test. Accuracy of the Type A test is verified by a supplemental test. The supple-mental test method selected is conducted for suf ficient duration to establish accurately the change in leak rate between the Type A and supplemental Test. Results of this supplemental test are acceptable provided the difference between the supplesental and Type A test is within 25 percent of the allowable If results do not meet the acceptance criteria, the reason shall leak race. be determined, corrective action taken and successful test performed. ( The measured leak rate (L during the Integrated Leak Rate Test shall be less I than 0.75 of the allowabl$") leak rate (L ) at peak accident pressure (P ) of 39.4 l The contairement vessel and its pInetrations are designed to liaft leakage psig. (L to no more than 0.2 weight percent of the internal net free volume of approx-ial)ely 3.218 cubic feet in 24 hours at a pressure of 39.4 psig. / t ,0cro If two consecutive periodic Type A tests fail to meet the applicable acceptance erl.teria, a Type A test shall be perf ormed at each plant shutdown for refueling, or approximately every 18 months, whichever occurs first, until two consecutive Type A tests meet the acceptance criteria, af ter which time the original ratest schedule may be resumed. 6.2.6.2 Containment Penetration Leak Rate "est Containment penetrations whose design incorporates resilient seals, gaskets, or seal compounds including air lock door seals, equipment and access doors with resilient seals, and other penetrations designed to be tested shall be leak tested in accordance with 10CFR50, Appendix J, Type B tes e and the 45.4 - 1972. guidelines for local leak testing procedures established in ANSI N These procedures may include testing by use of the following methods: j a) Pressure Decay Test. The test chamber is pressurized with compressed leakage is determined by measuring the rate of pressure air or gas. loss for the volume being tested and applying the Perf ect Ga s Laws. Continuous or Intermittent Pressurization. leakage is detected by means b) of a temporary or permanently installed system with provisions f or continuous or intermittent pressurization and monitoring makeup gas flow to maintain the pressure, or by intermittent pressuri:ation and knowing the volume, determining the leakage by monitoring pressure decay. anos 6.2-173 -~ -.

WNP-3 [ 1424W-5 PSAR k '/5'C,19 [ ~E 14 2.12.2.27 SHIEI.D BUILDING VENTILATION .. = 1.0 CHective To demonstrate the proper operation of the Shield Building Ventilation Systeax === e = d a i - - 4.1. ,be+Qf;D$dC$urua), k erify that the filtration systen meets design specifications. V g 4.2 p Verify that the system can maintain the annulus area at negative @J.>nch W g or les sf ts. A 539'5 4(,w,g g, g, 0.~ L ya s.I ib ta, cco se fin. o

== 5.5 14.2-53

Question No. Containment leakage fractions, described in Table 6.5.3-1 480.24 are divided into three parts: to the annulus (0.4), to (6.2.6) controlled ventilation areas (0.52), and bypass leakage (0.08). Discuss and justify the bases for obtaining these fractions.

Response

Containment leakage fractions and bases for obtaining these fractions have been previously addressed in the responses to Question Nos. 450.4, 450.7 and 450.9. I t I

Question No. Item No. 7.Q.2 of Table 1.9-1 states that WNP-3 complies with 480.26 CESSAR-F, Section 6.2.1.5, regarding the minimum containment pressure analysis. However, the response to Q. 480.3 states that the analysis in CESSAR-F is not applicabic to WNP-3 and Therefore, that a plant specific analysis will be performed. Taule 1.9-1 should be updated accordingly.

Response

The WNP-3 FSAR Table 1.9-1 will be updated as indicated in the attached. In addition, the FSAR will be changed to reflect the completion of the containment pressure and reflood hydraulics calculations. + ' " ' " ' - m-w_.m,,,

hl bb g I. 7 1539W-13 y,,,3 rsAR I TABt.E 1.9-1 (Cent'd) I Referenced CESSAR Onepitence Status FSAR anference Sectica g g Encepttee Secties Interf ace Bogstrement (CESSAR Secties) 7. Inseleer Stees supply System (5.1.4) (Osat'd) P.2 BCrs deelan requiremente 5.1.4 I 5.1 9.5.1 3 1.4 I F.3 Fire protection requiremente 5.2.5 F.4 Beactor coolant leek 5 1.4 I detectiem 10.4.8 F.5 Safety-related centret ear 3.1.4 I i system 9.3.1.1 F.6 Instrument air for RCS ve1=s 5.1.4 I operetten F.7 Centelament structure 51.4, Figures I 1 2.2 5.1.3-1 and 5.1 3-2 q. mistroenentet .l y 41 acs enstreneentel qualift-3.(1 I 3.11 e cettene y j /0 3 [ 6.2.1 q.2 Centstaneet treestente 5.1 4, 6.2.1.5 I reesttias free LOCA Table 3.1.4-2 ed 3 q.3 centstament cooling heat need 5 1 4, Ta bles I 5.1.4-2 sed ,1 5.1 4-3 I R. Mechanical lateracties Bettseen empements A f 3.9.3.1.1, 3.9.3.1.2, I R.1 acs coepenente destaned for 5 1.4 I E J 3.9.3 1.3 D C DBA l 3.9.3.1.1, 3.9.3.1.2, y ? i I R.2 acs structure deelsned to 5.1.4, 3.9.3.1 I { 3.9.3.1.3, 3.8.3.1 g* restrain end support BCS p components i

1. 7

"'b ?. 1203W-7 WNP-3 j FSAR f qgp,y, 7 + 6.2.1.3 Mass and Energy Release Analyses for Postulated less-of-Coolant w Accidents See CESSAR-F Subsection 6.2.1.3. 6.2.1.4 Mass and Energy Relerose Analysis for Postulated Secondary System g Pipe Ruptures Inside Containment 9 See CESSAR-F Subsection 6.2.1.4 Minimum Containment Pressure Analysis for Performance Capability ~ 6.2.1.5 Studies on Emergency Core Cooling Systes g lJ - (A alaisua e ainment pr e analysis has performed for and the i 2 l resulta pressure posse dc's not any pe the CESSAR-F pr ure respons A specifi ressure response, uding a discussio methodolo andj i e paramet s, is presented as acussed in Sectio, .3. 6.2.1.6 Testina and Inspection Preoperational and periodic tests are conducted to insure the functional capability of the containment and associated structures, systems and Included are the Integrated Leak Rate Test (see Subsection components. 6.2.6), Shield Building Leak Rate Test and Operatio;.a1 Tests on mechanical equipment that are required to operate following a pipe break. A preoperational visual inspection and test shall be conducted to verify the u. structural integrity of the Shield Building. The test is conducted by ventilating the Shield Building at a flow rate of 10,000 cfm or less and insuring that the differential pressure between the Shield Building and the The results of the outside atmosphere is a negative 1/4 inch vg. or greater. visual inspection, any repairs made thereof, and of the test shall be recorded in acc adance with the test procedure. Visual inspections,'preoperational and periodic tests are performed at the frequency described in and to the requirements and heceptability of the Technical Specifications (Chapter 16). 6.2.1.7 Instrumentation Application Pressure sensing instruments monitor the containment atmospnere and initiate CIAS, SIAS and CSAS according to the logic discussed in Section 7.3. Radiation monitors which monitor containment atmosphere and isolats select containment penetrations are discussed in Subsection 7.3.1. Instrumentation applications for the various engineered safety features such as the Containment Heat Removal System associated with the containment, and the Combustible Gas Control System, are discussed in Subsection 7.3.1.

.=

~ Amendment No. 2, (12/82) 6.2-20

%c,h%ulu. n quo. w ln su h l__ 6.2.1.5 Miniw> Containment Pressure Analysis for Perfornance Caoability Studies on Emeroency Core Cooling System A minimum containment pressure analysis has been perforned using the same NRC approved methodologies and assumptions as CESSAR-F..Jhe input parameters remain the same as CESSAR-F except.the initial containment internal. conditions, 4 active and passive heat sir.ks, and Ke*t free containment volume. These differences are identified as follows: 1. Initial containment Internal Canditions The initial containment conditions which have been used for WNP-3 plant specific analyses are: 'Tseperature 75'F (minimus) Pressure 13.7 psia (minimum) Relative Humidity 741 (maximum) For each parameter the conservative directkon with respect to minimizing

: the containment pressure appears in parentheses.

2. Containment Volune The "* 3 ana' ysis is 3,414,000 ft 3. Active Heat Sinks The assumed operating parasaters for containnent i~. sprays are as fonows: Flow rate, 11,150 gal / min (maximum, total all pugs) 4. passive Heat Sinks The surface areas and thickness of all exposed containment passive heat sinks are listed in Table 6.2.1-7. ~ ' M N limiting large break LOCA,1.0 x DEG/PD, the minimum containsent F pressure response to be used in analyzing the effectiveness of the ECCS is shown in Figure 6.2.44 It is noted that the resultant pressure response is lower than the CESSAR-F pressure response. The responses of the containment atmosphere and containment sump temperatures are shown in Figures 6.2.45 and 6.2.46, respectively. O, r

• =

e ee----w=---mm.v-vw-e-- ,,-v--ww-r- =--e---

---

w-----=------,-----mir

--,.----m- -e e-e---e--m-,w-m-ww-+=emeem ae--

t CONTAINMENT E " "' " ".'w b .e- -., : m - 4fc ) Dt. bi/.-r..- i i.. e g g &' - e *'*' m"je i At D L. i O i, --- WH P-3 --- CESSAR-F -m.non { I CL WWW e i i 1 } I i l e i i, i c m. n.., n. i l l l i i i. l t i i i i sew+.n.n i e .g i. .i 1 i i i i i i i I i e i i i g. i, ".I d't'. O. 9 5 i, 4.-- y, 1 I i i l i i 'A i r e i t i I ( 3-.. n.. n. n s I i 1 i i j i i 2 l l i i I l i .,., n. n. i .i I. i i j 8 i e i i. i i. ,i l l l '1 0 000 E e o C i o e o C e. s e, o c C C C C e N v c C C 3 l m m w c o 1 i i 5 f-tu I r,M. .C.!,i F, B t t R K, r 1 Assas t

g (Qce b'n54b-pCCj nnn as s .L,. v ', i r& rg.CCO 4/ $0. Ab I Aw.= 1 ,v .,mec..- 1, - = - e- ,.,i. ! g.- - WN P-3 l

---

cssssa p i

I ..s a.- c. m.. ..i I if I i e ,r l // } i s p t g I i i .. s.. 3 .I g / j j 4 i t

I.

• r t

t ,.i i j e.f e ,+

2..._?.w...

• s I

r. /

, r,' i e i t i l. i. l e i i

.e m, n s..

-- -5 , /. t g s-t 8 e t e-i I i I 1 i ..,I..e.p/ ...!........ li...... - - i. i. , --m, g I 6 /,. 6 i.< i 1 t I I. i. i $4MA% g i 9 s I i,- i l I i i. c. t i ~' - a w ca-2 o. Q S.. 4 O. c) O C C C-C- C u w. O c o cv rrr v c_ C I i g q.g7. a t- - .y.r. m---m. -eqb NE =b U Ob! I b 1 N O

< C.,. L: ~ L,. ~ 4eo.xs Flauns.6.2.44 _ 1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CONTAINMENT PRESSURE 60 - i .n ~ 50 40

s _..

q E! ~ g 30 =Mg l + l 20 ~ 10 0 O 120 240 360 480 600 TIME AFTER BREAK, SECONDS l

i h l,%sk aso.n FIGURE 6.2.45 1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CONTAINMENT ATMOSPHERE TEMPERATURE oggP ~ 240 200' l 160 g-g_. "g 120 - 5 >== 80 1 40 C .0 - 120 240 360 480 600 TIME AFTER BREAK, SEC -. -. - - - - - - - -, _ _ - _. ~...,, _ - - _. m_,

i sueel, sac. '1SD X" FIsuas 6.2.46 1.0 x DOU3LE ENDED GUILLOTINE BREAK -~ IN ? UMP DISCHARGE LEG CONTAINMENT SUMP TEMPERATURE 240 200 160 8-g__. ar. y '120 . LaJ B 80 40 / - 0 O 120' 240 360 480 600 TIME AFTER BREAK, SEC --v-- w-- w----m-w-- - - -w

---

w-

--.m-e ,-w-w- -,-,e--,---wy--

---

,---,-w-wew--

e- --*--~,ww--w---

1713%ii-1 3 h k,*4. yso..w 6.3 EMERGENCY CDIE CDC.DG SYSTM CE5511-F Subsectida 6.3.3.2.4 allows reference to the results of the ICCS analysis performed therein if a higher miniaun containment prorsure than CESSAR-F Figure 6.2.1.5-3 can be demonstrated. As discussed in Subsection 6.2.1.3, the W-3 minimum pressure e=1-=1= tion does not meet 4his requiremnac. A W-3 specific ECCS analysis will be prepared to demonstrate 2. ---14= e with 20CFE50.46. Ezeept for the result's of the ICCS analysis the existing CES.SG-F descriptions of plant systems, EC3 dalculations, methodologias, and analytical assumptions are applicable to W-3 and' are je referensed herein. The W-3 *Wm **M%t liase. b 'E N app 14 ram 14ry of CESSAR-F will then be re-evaluatedemw-re d ens d WL e. and any differences ,,,4,d,,j

• % identified on the W -3 docket. Phst experience has snown that plant specific demonstrations of comp 1t=~e with.10CFR30.46 acceptance criteria do not result in changes to ECG system designs.

It 1.s therefore expected that the W-3 re-saalysis will confirm the adeenacy of the existine desif ew. .::i t,.s..ne. wap-s ... mN c.*=hin < h p.<ssere.,J reR ual omfe l t j

a. :

e ey rs<...cE.s sM-r a aL.,,,,.,,..i ,,e. .l .m +a.>;i-4.. ve u s.na.y

• h
• f e?l1:.

WNP-5 A. t sy-<.if e.. I' a . t. < J e$ cL.t., te-,e.,-.-~t. aaah- - i .+ 4 u .p .'ara .Tg* ~~= G ~ =!!* pale. p<.f;} /~ -I y.aak I.,n ' h e<+ ed< M ** Y " ~ fll* nee saHL jo g o, org u a,,,p n a s c o. ~ < *.'x f The Emergency Core Coolias Systiain REC 3) or Rafety Injection System conforme to the CE interface requirements as described in CESSAR-F Subsections 6.3.1.3 itan A through 6.3.1.3 item q.1, and Section 1.9 of this FEAR. Figures 6.3U through 6.3-lh shoii the Qiaracteristic Curves for the.Righ Pressura and the Low Pr' essure Safety Injecg.4 Pumps (EFSI and LFSI):. QQ f l l.e

• DEG/PD kw s

.wa % c +=t ,4a p I'*if. e f,se e.y;we ~ h =f4. AM M. ' j.g

I dttc5 bha b. 49 0. D Fisuns 6.3-li- _1.0 x DOUBLE ENDED GUILLOTINE BREAX IN PUMP DISCHARGE LEG CONTAIHMENT PRESSURE 60 i i s ;, '

50....

~ ' ' ~ ~ 1 go.

== g -.. =- g 30 - R, u E 20 10 1 1 0 O 120 240 360 480 600 TIME AFTER BREAK, SECONDS 9

$t.LC.*>N+*sN. w o.A6 FIGURE 6.3-lj . _.l.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG MASS ADDED TO CORE DURING REFLOOD 120,- l 100 x 80 ~ - - - h v 60 12! a TIME (SEO fLTt00D RATE f 0.0 - 7.0 3.26 IN/SEC 40 h 7.0 - 47.2 1.10 IN/SEC 47.2 - 600.0 0.645 IN/SEC 20 l 0 O 120 240 360 480 600 TIME AFTER CONTACT., SECONDS -,w e-e,, --r----,.-.,,r.- ,-------.-,,,,,ne,. ,,,.,r-.,-,-. w.,, ,,,wm,,~m,. -,-...-.,-w

i Question No. Provide the Procedure Generation Package'(PGP). The PGP should f 640.16 be comprised of four parts. I 1. Plant-Specific Technical Guidelines (PSTG) Since WNP-3's E0Ps will be based on CEN-152, " Combustion Engineering Emergency Procedure Guidelines," the PSTG should consist of: Description of the process used to develop the PSTGs a. from the generic guidelines (as opposed to providing the PSTGs themselves). A list of deviations from the generic guidelines that b. may be necessary because of different plant designs, operating philosophy, or operator needs. This list of deviations should also include a technical justifica-tion (engineering evaluation or analysis) for each deviation, along with a statement of the safety sig-nificance of the deviation. A description of the analysis performed to identify c. the operators' information and control needs, and the method for determining the adequacy of existing in-strumentation to meet these needs. This effort is normally referred to as a Task Analysis." 2. Writer's Guide Provide the actual writer's guide to be used by plant per-sonnel in ensuring the E0Ps are usable, accurate, complete, readable, convenient to use, and acceptable to control room The Writer's Guide should, as a minimum, personnel. address the goals and intents provided in Section 5.0 of NUREG-0899 " Guidelines for the Preparation of Emergency Operating Procedures." 3. Validation / Verification The PGP should contain a description of the program for validation and verification of the E0Ps. The description should include what methods are to be used to accomplish each objective stated in Section 3.5 of NUREG-0899, and a description of how those methods will meet the specific objectives.

Question No. Containment leakage fractions, described in Table 6.5.3-1 480.24 are divided into three parts: to the annulus (0.4), to (6.2.6) controlled ventilation areas (0.52), and bypass leakage (0.08). Discuss and justify the bases for obtaining these fractions.

Response

Containment leakage fractions and bases for obtaining these fractions have been previously addressed in the responses to Question Nos. 450.4, 450.7 and 450.9. J

Question No. Item No. 7.Q.2 of Table 1.9-1 states that WNF-3 complies with 480.26 CESSAR-F, Section 6.2.1.5, regarding the minimum containment pressure analysis. However, the response to Q. 480.3 states that the analysis in CESSAR-F is not applicable to WNP-3 and that a plant specific analysis will be performed. Therefore, Table 1.9-1 shculd be updated accordingly.

Response

The WNP-3 FSAR Table 1.9-1 will be updated as indicated in the attached. In addition, the FSAR will be changed to reflect the completion of the containment pressure and reflood hydraulics calculations.

l gl l il il i .i i-i i. !.f e 4 y,,,3 153W-13 FSAR TABLE 15-1 (Cent'd) f Referenced CESSAR thepliance Status FSAR Bef orene e Section Yee N/A 8eceptise Secties Interf ace Seguirement (CESSAR 8ecties) Inselear Steen supply System (5.1.4) (Osat'd) 7. 5.1 3.1 4 I F.2 BCr5 deelga requiremente 9.5.1 5.1.4 I F.3 Fire protectine requiremente 5.2.5 F.4 Beector coolant leek 5.1.4 I detection 10.4.8 5.104 I F.5 sefety-related centret air egetem 9.3.1 1 F.6 Isotrument ear for act valve 5 1.4 I operation 1.2.2 31.4 Figures I F.7 Centsimment structure 5.1.3-1 and 5.1,3-2 I q. sir - etsi q.1 BCS enstressental queltff-3.L1 I 3.11 cettene g [ 6 2.1 b3 3 1 4. 4.2.1.5 I y 42 Coatsinseet tremetente resulting free IMCA I Tebte 5.1.4-1 and 3 q.3 cents 1meent coettag beat need 51.4 Tables I 5.1 4-2 and i 5.1.4-3 1 Mechseleet lateracties netween thepeneste A 5' R. 3.9.3 1.1, 3.9.3.1.2 F4 v 3.1.4 I 3.9.3 1.3 D C R.1 BCS componente deetsne4 for DBA T 3.9.3.1 1. 3.9.3.1 2, y % a.2 aCs structure deelaned to 5.1 4. 3.9.3.1 2 3.9.3.1.3. 3.8.3.1 D p. restrain and empport RCS coeponente

b#' 1203W-7 WNP-3 l q co.2Jo -l FSAR 7, a-Jii ~ /

6. 2.1.3 '

Mass and Eneray F.elease Analyses for Postulated Loss-of-Coolant i: ; Accidents ..ul 3 See CESSAR-F Subsection 6.2.1.3. i Mass and Emeray Relosse Analysis for Postulated Secondary Systes 6.2.1.4 j Pipe Ruptures Inside Containment I.- See CESSAR-F Subsection 6.2.1.4. Minimus Containment Pressure Analysis for Performance Capability ~ 6.2.1.5 [ Studies on Emergency Core Cooling Systes ggl A minimum e ainment pr e analysis has performed for and the r 2 g does not ene pe the CESSAR-F pr ure respons A specifi ressure response, uding a discussio methodolo and ) i resulta pressure j .3. i e paramet s, is presented as acussed in Sectio, 6.2.1.6 Testina and Inspection Preoperational and periodic tests are conducted to insure the functional capability of the containment and associated structures, systems and Included are the Integrated Leak Rate Test (see Subsection components. 6.2.6), Shield Building Leak Rate Test and Operational Tests on mechanical j equipment that are required to operate following a pipe break. A preoperational visual inspection and test shall be conducted to verify the u structural integrity of the Shield Building. The test is conducted by ( ventilating the Shield Building at a flow rate of 10,000 cfm or less and insuring that the differential pressure between the Shield Building and the The results of the J outside atmosphere is a negative 1/4 inch wg. or greater. visual inspection, any repairs made thereof, and of the test shall be recorded in accordance with the test procedure. Visual inspections,'preoperational and periodic tests are performed at the frequency described in and to the requirements and acceptability of the Technical Specifications (Chapter 16). 6.2.1.7 Instrumentation Application w Pressure sensing instruments monitor the containment atmosphere and initiate CIAE, SIAS and CSAS according to the logic discussed in Section 7.3. Radiation monitors which monitor containment atmosphere and isolate select containment penetrations are discussed in Subsection 7.3.1. Instrumentation applications for the various engineered safety features associated with the containment, such as the Containment East Removal System and the Combustible Gas Control Systesa, are discussed in Subsection 7.3.1. .CEe

.~

Amendment No. 2, (12/82) 6.2-20

%.,h%.ilu. quo.w Lv1W N I 6.2.1.5 Minime-containnent Pressure Analysis for Perfernance Caoability Studies on Emergency Core Cooling System A minimum containment pressure analysis has been perforned using the same NRC approved methodologies and assumptions as CESSAR-F..Jhe input parameters remain the same as CESSAR-F except.thg initial containment internal. conditions, active and passive heat sinks, and Et free containment volume. These differences are identified as follows: 1. Initial Containment Internal conditions The initial containsent conoitions wnten have been used for WNP-3 plant specific analyses are: ' Temperature 75'F (minimum) sia (minimum) 13.7 p(maxinum) Pressure Relative Humidity 745 For each parameter the conservative direction with respect to minimizing

the containment pressure appears in parentheses.

2. Connainment Votume The net Me cdainant Muse ash fe Ws 3 ana'ysis is 3,414,000 ft 3. Active Heat Sinks The assumed operating parameters for containment i'. sprays are as foH ows: 11,150 gal / min (maximas, total all pumps) Flow rate 4. passive Heat Sinks The surface areas and thickness of all exposed containment passive heat sinks are listed in Table 6.2.1-7. For the limiting large break LOCA,1.0 x DEG/PD, the minimum containment pressure response to be used in analyzing the effectiveness of the ECCS is shown in Figure 6.2.44 It is noted that the resultant pressure response is lower than the CESSAR-F pressure response. The responses of the containment atmosphere and containment susp temperatures are shown in Figures 6.2.45 and 6.2.46, respectively. 8 9 9 r m, m _..-.-_.-._,-,_,-w-e-,c----m e----- ---wse-t-W**"fsw m, un-w w- --w-euy---'rv g-u,--v--w,rg-N-w --w-t---t-W9---mywc-ww-- -ug-ew = v

---

,.,------m

E""' b""'b CONTAINMENT 4fC. Au l m., q o,v. :. m, m a ...-.v i.,.i. ~ -,..,Di_.,... - W8 P-3 I --- CEssAR-F -,u. n,n e aa l i e i i i. i i i i i i. l i I i i I e. i ,,,... n. n. i i. 'l [ g n i i i i i + a i l-i i i l l ts' men u 1.. i i. r,, l i i i i '5 I 8 e i k_ 0 4 w..o.ne Sa l ww I I i I i I i i i l I i 8 l i, i l. i ] j .m . nn

• ,-<w

,j .l j l-8 { f i I i t i I C. Cn.0 e e e a e C C. C C C =; .c C C C C I c. n w w C C s j N m v c c I e n-r. s TrM: M r,i. R

BRtRK, Stu 1

Fis0 45 l

0 ("4 (.ti$ DI Q (( j.L,.f s.. v, I O'. e(. 7$ h 1 2 '.s. i-wv

g =.=%f% t% e. *%

1 m

• rs? v 6 ! g =.

m. .p - m e-3 -.-- css,44.p s 8 s .*>~.*i l l .4...w i g , f e/ /. 6 / l i / .g q / p I f 1 i i e a .I l / s I /.. e s s# i j p

1. /

l / -, ga.nes l

r,-

... a.a... e r.* I I e.- l t t t l t i i i 1 ! s **. i a e s a e s l l I .:.9 9 % % ..e=* ~ ~ ~ ~ _ _ ~ ~.* ,rg g w I g t. i em-s 6 .I 6 .e f.' e . )l., # ..'.......b... i 4 ....g a s. ,,,5m i 8

w...

I a I ? /. ' r l s .f g

s..

I. e i e 2 i

• • S a a *.

-c.. e e .e..es o e l o = e I. l l. l l 's t i i I l I .,= 9 r='.. a g g C-R 'w e N. N O. D. - k. a go c c C-C> C N M. C C C = N A T.- C. c ,..r- [nN7.nyi ar e - ! AMP M. r. = g y e _v ,t ~ Fii:, o a. 2.

I A.:.L w il,. ~ wo.xe FIsunE 6.2.44 1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CONTAINMENT PRESSURE e e 60 .-i 50 e ,8 e 40 5 -. p s: g 30

=

g 20 10 O e 0 O 120 240 360 480 600 TIME AFTER BREAK, SECONDS ~ -- ,,,,,r--

---

,,,,-----n----en--

,,-w ,,--,,m-,,,,,---,-,----.----,,-,,-,-,,,-w .,en----em,--,--- e

] h<, N 4G,. -i wso.xo FIGURE 6.2.45 1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG ' ~ ~ CONTAINMENT ATMOSPHERE TEMPERATURE 240 200 16g LL. ~ 'g -- - g 120 E 80 ~. e 40 C 0-120 240 360 480 600 TIME AFTER BREAK, SEC ___,__....._-,_-__.,__.,_,_,7,-

A e l,% do. ygo.Ac. ~ FIsuas 6.2.46 i 1.0 x DOUBLE ENDED GUILLOTINE BREAK ' ~ IN ? UMP DISCHARGE LEG CONTAINMENT SUMP TEMPERATURE 240 200 ISO lii! _. t 120 l ~ q 0 O 120' 240 360 480 600 TIME AFTER BREAK, SEC c. e ---_,_.,,_,----__y 7 --.w,,, y-,.-,w ,,,.,,,,,,..,_,--c,

I 1713*J-1 3 G w kom is. 450.16 i 6.3 EMEEGENCY COE COCK.ING SYSTDi CESSut-F Subsectida 6.3.3.2.4 allows ref arance to the results of the ECC5 analysis performed therein if a higher mininum contairnment pressure than CESS &R-F Figure 6.2.1.5-3 can be demonstrated. As discussed in Suhsaction 6.1.1.3, the W-3 =4=4== pressure calculation does not meet this requirement. A W-3 specific ECCS analysis will be prepared to demonstrate 2 compliance with 20CFE50.46. Escept for the results of the ECC5 analysis the existing CE55AE-F descriptions of plant systems, ECC5-(alculations. methodologies, and analytical assumptions are applicable to M-3 and' arm it refaranced hernia. The W-3 aia:% c=A .e=t-bet b 'I applicability of CESSAR-F will then be re-evaluatedemww. mad <=,4d WL ,j e. and any differences e. h idaaetf'ied on the talP-3 docket. Phat experience has shown that plant specific demonstrations of compM-a with 10CFE50.46 acceptance criteria do not result in changes to ECCS system designa. It is therefore expected that the W-3 re-enalysis will confirm the adecuacy of the azistine deaf en. S t <.a..n. wng-s E.. m N u d e.'n m<~h p <ss am -, J rer7 ual ste. t 1 e n,.p. ,e,.j &;

d. - ~.st e & r4<..

.CE.S SAA~r p4sse r*de.,, .l* me$< y u y 6.fe $, a-WNP-5 A t sf -: e.. ]' re 1 . k*t. col P <s 2.. claJ., t< n_ h.m..= f. nda5=~ <=<l sn-4 .'r4,e y .gN ~ ~u='~ L - %.Lle. 9,e-.ltay. /. < I,o.s-k I. nee

• 4 <<4

~~ +c- ^ * ~ 4 ~ L -.g - u.. y. w ia c; g s e a ,..s.. a. c 4 :n. The Emergency Core Cooling Systsim TEcc5) or Rafety Infection System conforms to the CE interfaca requirements as described in CESSAR-F Subsections 6.3.1.3 iten A through 6.3.1.3 item Q.1, and Section 1.9 of this FSA1. l. Figuras 6.N.a through 6.3-lh sho the Characteristic Chrves for the.J11gh w Pressura and the Low Pr' essure Safety Injection Pumps (EPSI and LPSI)'. %- u (.. s - u - t. c 3-g.,L.m f., u it a,, q, . J t.ca,, 1 = x WM Hu- *:.c u u. a ~ raie. n<,. w.1). ~ l l l 1 \\ 1 i

i Asb6 yso.:t.e FIGURE 6,3. 1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CONTAINMENT PPISSURE 60 i l o 50 - N.

40..

1_. E f 30 - ~ S w 20 10 0 O 120 240 360 480 600 TIME AFTER BREAK, SECONDS he

sa., Hn.4 ngo.a.s FIGURE 6.3-1J 1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG MASS ADDED TO CORE DURING REFLOOD l 120 1% ~ E x 80 .%m -- v 60 IEl a { TIME (SEC) RCFLOOD RATE 0.0 - 7.0 3.26 IN/SEC 40 h 7.0 - 47.2 1.10 IN/SEC. 47.2 - 600.0 0.645 IN/SEC 20 0 O 120 240 360 480 600 TIME AFTER CONTACT, SECONDS

i Question No. Provide the Procedure Generation Package'(PGP). The PGP should 640.16 be comprised of four parts. i 1. Plant-Specific Technical Guidelines (PSTG) Since WNP-3's E0Ps will be based on CEN-152, " Combustion Engineering Emergency Procedure Guidelines," the PSTG should consist of: Description of the process used to develop the PSTGs a. from the generic guidelines (as opposed to providing the PSTGs themselves). A list of deviations from the generic guidelines that b. may be necessary because of different plant designs, operating philosophy, cr operator needs. This list of deviations should also include a technical justifica-tion (engineering evaluation or analysis) for each deviation, along with a statement of the safety sig-nificance of the deviation. A description of the analysis performed to identify c. the operators' information and control needs, and the method for determining the adequacy of existing in-strumentation to meet these needs. This effort is normally referred to as a Task Analysis." i 2 Writer's Guide Provide the actual writer's guide to be used by plant per-sonnel in ensuring the E0Ps are usable, accurate, complete, readable, convenient to use, and acceptable to control room The Writer's Guide should, as a minimum, personnel. address the goals and intents provided in Section 5.0 of NUREG-0899, " Guidelines for the Preparation of Emergency Operating Procedures." 3. Validation / Verification The PGP should contain a description of the program for validation and verification of the E0Ps. The description should include what methods are to be used to accomplish each objective stated in Section 3.5 of NUREG-0899, and a description of how those methods will meet the specific objectives. - _ _ _.}}