ML20065Q347
| ML20065Q347 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 10/22/1982 |
| From: | Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Kerrigan J Office of Nuclear Reactor Regulation |
| References | |
| GO3-82-1085, NUDOCS 8210270012 | |
| Download: ML20065Q347 (300) | |
Text
{{#Wiki_filter:. . . . . . . - . - - . - . .. Washington Public Power Supply System P.O. Box 968 3000GeorgeWashingtonWay Richland, Washington 99352 (509)372-5000
- Docket 50-508 October 22, 1982 G03-82-1085 Ms. Janis D. Kerrigan, Chief Licensing Branch #3 (Acting)
Division of Licensing US Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
WASHINGTON NUCLEAR PROJECT 3 I RESPONSES TO NRC ACCEPTANCE REVIEW QUESTIONS
References:
a) Letter D. G. Eisenhut to R. L. Ferguson, dated 8/20/82 Reference a) transmitted a set of questions generated during the NRCs accep-tance review of the WNP-3 Operating License Application. These questions addressed both the WNP-3 Final Safety Analysis Report (FSAR) and the Environmental Report (ER). Responses to the questions associated with the FSAR are enclosed in Attachment 1. Responses to the questions associated ) with the ER will be transmitted under separate cover. Several of these questions address items which are still being evaluated by the Supply System. In those cases where our evaluation is not yet complete, we nave provided a schedule detailing when we will be able to present the results for review. l In those cases where it is considered necessary or desirable to amend the l FSAR due to our responses, we have provided marked up FSAR pages which show, l in detail, what our intentions are. It is hoped that this will allow early
! review by the NRC prior to actual incorporation of the revised information into the FSAR.
r, 8210270012 821022 PDR ADOCK 05000508 A PDR 4( $.4 .--_.
Ms. Janis D. Kerrigan Page 2 October 22, 1982 WASHINGTON NUCLEAR PROJECT 3 RESPONSES TO NRC ACCEPTANCE REVIEW QUESTIONS If you require additional information or clarification, the Supply System point of contact for this action is Mr. K. W. Cook, Licensing Project Manager (206/482-4428 ext. 5436) G. D. Bouchey, Manager Nuclear Safety and Licensing AJM/tn cc: D. J. Chin Ebasco, NY0
- N. S. Reynolds D&L E. F. Beckett NPI J. A. Adams - NESCO l D. Smithpeter - BPA Ebasco - Elma WNP-3 Files - Richland l
l l l l
-1 2266A Question No.
100.1 Table 1.8-3 addresses conformance and exceptions to the Standard (1.8.3 and Review Plant (SRP), NUREG-75/087. You have stated that WNP-3 other will be reviewed and evaluated relative to NUREG-0800. This sections) review and evaluation should conform to the following:
- 1. Applications for light water cooled nuclear power plant operating licenses docketed after May 17, 1982, shall in-clude an evaluation of the facility against the Standard Review Plant (SRP) in effect on May 17, 1982, or the SRP revision in effect six months prior to the docket date of the application, whichever is later.
- 2. The evaluation shall include an identification and descrip-tion of all differences in design features, analytical tech-niques, and procedural measures proposed for a facility and those corresponding features, techniques, and measures given in the SRP acceptance criteria. Where such a difference exists, the evaluation shall discuss how the alternative proposed provides an acceptable method of complying with those rules or regulations of Comission, or portions thereof, that underlie the corresponding SRP acceptance criteria.
- 3. The SRP was issued to establish criteria that the NRC staff intends to use in evaluating whether an applicant / licensee meets the Comission's regulations. The SRP is not a sub-stitute for the regulations, and compliance is not a re-quirement. Applicants shall identify differences from the SRP acceptance criteria and evaluate how the proposed alter-natives to the SRP criteria provide an acceptable method of complying with the Comission's regulations.
In addition, a schedule should be provided for completion of the review and evaluation.
Response
The Supply System has initiated a program to provide the re-quired Standard Review Plan (SRP) compliance review in accor-dance with recent modifications to the Comission regulations. This review and evaluation process is being conducted based upon the acceptance criteria contained in NUREG-0800. The initial portion of this program was culminated by changes in the WNP-3 FSAR provided in Amendment 1. The changes are sumarized below:
- 1. Insertion of revised material in Subsection 1.8.3 and Table 1.8-3 (consistent with NUREG-0800 review rather than NUREG-75/087).
Attachment 1-2 2266A Question No. 100.1 The revised Table 1.8-3 provides a listing of SRP acceptance (1.8.3and criteria and indicates those for which the WNP-3 design is other in conformance and those for which differences exist. In sections) addition, a schedule for completion of the required evalua-(contd.) tions is included. In several instances the compliance review was not completed or a schedule for completion was not available prior to submittal of Amendment 1. These areas are expected to be completed and submitted in the next FSAR Amendment. The evaluation process to be utilized is described in Subsection 1.8.3 of the FSAR as modified in Amendment 1. 1 i
-3 2266A Question No.
100.2 The WNP-3 FSAR contains numerous references to the CESSAR-FSAR but does not specifically address the Safety Evaluation Report (NUREG-0852) for the CESSAR-FSAR. This Safety Evaluation Report (SER) imposes requirements on applicants utilizing the CESSAR-FSAR and identifies open items. The applicant should provide a plan identifying and addressing the interface between NUREG-0852 and the WNP-3 FSAR to assure that the SER requirements are addressed in the WNP-3 FSAR and are, or will be, incorporated in the design and operation of WNP-3. Provide a schedule for im-plementation of this plan. Types of information to be addressed in this plan are as follows.
- 1. Open items identified in the SER. This should include both items identified for final resolution by the licensee as well as those for Combustion Engineering resolution.
Although the latter items may not require specific licensee action at this time, licensee tracking is necessary to in-sure that any resolution is incorporated into the WNP-3 design.
- 2. Specific license conditions and technical specifications which are imposed by NRC on applicants referencing CESSAR.
- 3. Interface requirements identified by NRC which differ from, or are in addition to, those identified in CESSAR.
The following are specific examples of items from the CESSAR-SER which should be addressed.
- 1. The SER identifies, in Section 15.3.9, specific items which must be implemented by the licensee as an interim fix for anticipated transients without scram until rulemaking and formulation of final requirements are completed. These items are not discussed in Section 15.8 of the WNP-3 FSAR.
- 2. The SER requires, in Section 7.3.2, that control logic be configured such that an ESFAS signal will override MSIS.
This is not consistent with the statement in Section 7.3.1 of the WNP-3 FSAR which states that "there are no overrides on any MSIS actuated devices with the exception of the atmospheric dump valves".
- 3. The SER requires specific plant technical specifications in Section 5.2.2 which should be addressed in the WNP-3 FSAR.
Response
The Supply System is in the process of developing a plan for identifying and addressing the interface between NUREG-0852 and the WNP-3 FSAR. This plan will be provided by December 1982.
Attachment 1-4 2266A Question No. 100.3 (All) Describe your system for monitoring updates to the CESSAR-FSAR and SER and incorporating these updates into the WNP-3 systems, operations and documentation.
Response
Monitoring updates to the CESSAR-FSAR to assure the material is properly incorporated involves reviews by Combustion Engineering (CE), Ebasco Services (A/E) and the Supply System. For each snendment to CESSAR-FSAR, CE reviews the material for applica-bility to WNP-3 and provides their assessment to the Supply System and Ebasco for review and concurrence. For those areas where the update material is not applicable to WNP-3 a change notice is prepared providing WNP-3 specific input to replace the CESSAR-FSAR update. The change notice is reviewed in accordance with the FSAR anendment review procedures and incorporated into the WNP-3 FSAR in a subsequent amendment. Should the WNP-3 de-sign be modified to incorporate CESSAR design changes the WNP-3 FSAR will be amended, at a later date, to reference the CESSAR FSAR material. As discussed in NRC Question 100.2, a review of the CESSAR-FSAR Safety Evaluation Report (SER) has been initiated. Updates to the SER are expected to be in the form of supplements to NUREG-0852. These will be reviewed in the same manner as CESSAR-FSAR updates to determine applicability to WNP-3 and provide snendments to the WNP-3 FSAR where required. l 1 \ l l l l l
Attachment 1-5 2266A Question No: 100.4 Correct the following deficiencies in the General Information: (General a. 10CFR50.33(d)(2)(-1) requires " names, addresses and Information) citizenship" of the Directors. The tendered application gives the names and addresses but the citizenship is not included.
- b. 10CFR50.33(d)(2)(iii) requires a statement as to whether the corporation is downed, controlled or dominated by an alien, a foreign corporation, or foreign government and if so, give details." The tendered application is silent on this requirement.
Response
a) All Directors and Principal Officers of the Supply System are citizens of the United States, as are the Directors and Principal Officers of Pacific Power and Light, Portland General Electric, Puget Sound Power and Light Company, and Washington Water Power Company. b) Neither the Supply System nor any of the private owners of WNP-3 are owned, controlled, or dominated by an alien, a foreign corporation or a foreign government. The private owners of WNP-3 are: Pacific Power and Light Portland General Electric Puget Sound Power and Light Company Washington Water Power Company ( l l 1
Attachment 1-6 2266A Question No. 210.1 Provide a discussion of the extent of compliance with Subsection (3.8.2.4) NE of the ASME Code, Section III, Division I for the procedures used in the design and analysis of the steel containment.
Response
The steel containment vessel is designed and constructed in full compliance with Subsection NE of the ASME Code, Section III, Division I and will be ASME Code stamped as stated in Subsection 3.8.2.2.3 of the FSAR. i i
Attachment 1-7 2266A Question No. 210.2 Regulatory Guide 1.70 states that the description of the computer (3.9.1.2) programs used in dynamic and static analysis should include the extent of the programs application, and the design control measures employed to demonstrate the applicability and validity of each program. Reference or provide this information.
Response
The following tabulation provides a listing of those computer programs used in dynamic and static analysis of WNP-3 and reference to the appropriate subsection providing program application and the design control measures employed to demonstrate the applicability and validity of each program. Computer Application Validity Code Reference Reference PIPESTRESS 2010 3.9.1.2.2.1 3.9.1.2.2.1 PLAST 2267 3.9.1.2.2.2(a) 3.68.2 CALPLOT F 3.9.1.2.2.2(a) 3.68.1 CALPLOT FII 3.9.1.2.2.2(a) 3.6B.1 WHAMM0C II 3.9.1.2.2.2(a) 3.6B.3 RELAP 3 3.9.1.2.2.2(a) 3.68.1 RELAP 4 3.9.1.2.2.2(a) 3.68.1 NASTRAN 3.9.1.2.2.2(b) 3.7.2.1 The FSAR will be updated to reflect this response in Subsections 3.6.2.2.1 and 3.68.1. i A
WNP-3 1681W-14 FSAR ,1 If 2. c) Breaks are postulcted to occur either instantaneously, with a break opening time of 0 001 seconds or, with a break opening based on analytical or experimental methods to determine break opening time for a particular case. Blowdown forces are determined by either a simplified conservative method or a detailed computerized method as described in a, and b, below. The method to be used is determined by the forcing function input required by the methods of dynamic response analysis, which are discussed below in Subsection 3.6.2.2.2. a) The predicted blowdown force (FBD) on a pipe with flow area A fed by an infinite source volume can at pressure Po be described by FBD
= KPAo Where K is the thrust coefficient for the case of steam, saturated water choked steam flow is assumed and, based on this flow, the resulting steady state force thrust coefficient is K = 1.26.
The source pressure Po is taken as the maximum operating pressure experienced by the piping system and the resulting blowdown force is assumed to be a step function with zero rise , time. For the case of subcooled liquids, such as feedwater, the force time history will consist of two values; an initial X magnitude equal to (2.0)P Aoat time zero, decreasing linearly to (1.26) PsatA at a time, t = L/C, where L is the length of the line from the break to the reservoir or pump. C is the speed of sound in the fluid, and Prat the saturation pressure corresponding to the maximum operating temperature in the line. h Ymp' b) The following is an alternate less conservative computerized method for calculating blowdown force: Computer codes, Relap-3 (Ref.1) Relap-4 (Ref.2) or WHAMMOC II (see Appendix 3.6B) are used to determine thermodynamic properties for gpfgp(jg straight pipe lengths defining changes in direction. See Subsection
. 6.2.1.4 for typical examples of output for c main steam line break.
Ig,, jg From this data a -
-- - ' "' "' ^'-- -- ' - - been written to convert U bg ,.g the transient flow conditions calculated in a piping system by the M "** " WHAMMOCII, or the RELAP series of computer codes into tcansient forces CAPLOTFlT on the piping system. Specifically, CALPLOTFII calculates and plots G the forces on straight lengths of pipe between changes in pipe
[- (jg/e.ftr a, (*M direction (bends), or between a change in direction and a pipe break. For discussion of CALPLOTFII 4 see Appendix 3.6B. bklAL FL-or F J 3.6.2.2.2 Pipe Whip Dynamic Response Analysis Using the blowdown forcing functions, dynamic analysis is performed in order to determine: loads on restraints. loads on piping nozzles, loads on valves and maximum displacements of whipping pipe, unless studies show the whipping pipe is of no consequence. For this purpose these quantities can be determined by simplified conservative methods as described in a) below or by A t - Y 3.6-l'. p
WNP-3 146 9W-2 g' FSAR d 2 t&< 2. @ bj -J \ cal.f*l.oT F .% qs f r*cusei f lede. usad 4* caled,di /r-,.wl pigba s us. A Yi 4 ' I l j 3.68.1 A CALPLOTFII Tp c.d .( h As.W-J gu (s . The CALPIDTFII computer code is a post processor code used to calculate
, transient piping forces using, as input, the magnetic tape created by the
,S J RELAP4 Mod 6 or the WHAMMOCII computer code. This tape contains the calculated rmodynamic and flow conditions required by CALPLOTFII to calculate forces.
'@A more complete description of theqCALPLOTFIIo computer c de e-contained in 4 Appendix 3.6C of the Waterford SES Unit No. 3 Final Safety Analysis Report
( Do cke t No . 5 0-3 82) . p g 3.6B.2 PLAST The PLAST computer code uses the f orces calculated by CALPLOTFII to determine the pipe whip restraint reaction loads by performing a dynamic structural analysis on a lumped mass parameter of the piping system. A description of the PLAST computer code is contained in Appendix 3.6B o f the Waterford SES Unit No. 3 Final Safety Analysis Report ( Do cke t No . 50-382). 3.6B.3 WR AMMOCII A Computer Code f or Perf orming One of Two Phase Water Hammer Analysis. 3 . 6B . 3.1 Summary A computer code named WHAMMOCII has been developed to solve rapid fluid transients (acoustic hammer) ;roblems for water, water with dissolved air, steam, or two phase one or two component fluid networks. The transient fluid conditions, thermodynamic conditions, and the forces on a piping network can be calculated, f ollowing any initiated rapid transient. The WHAMMOCII computer code uses the method of characteristics to solve the continuity and momentum e quations a ssuming compre ssible, adiabatic, homogeneous one or two phase no slip flow, and includes the ef fects o f piping deformation and cavitation on pulse speed propagation. The pressure predictions of the WHAMMOCII computer code are compared with various experiments including two subcooled water blowdown test, and a rapid valve closure transient in a pipe containing water with dissolved air. The code's prediction of the forces generated by a subcooled blowdown are also compared. The nomenclature used in WAMMOCII equations is given in Table 3.6B-1. 3.6B.3.2 Introduction l j Fluid hammer phenomena have become a major concern of the utility industry for the design of piping networks. At present , no publicly available computer code can analyze the water hammer phenomena which results in a transition from one to two phase flow. Consequently, the computer code WAMMOCII(l) has been developed to analyze acoustic wave (hammer) problems f or one phase water ! or steam conditions, for one phase water with dissolved air, for two phase ! multicomponent (air-steam-water) flows, and f or the determination of phase The code is not intended to be used for the calculation of transitions. nonacoustic phenomenon ef fects such as s hock waves. WHAMMOCII can calculate the fluid thermodynamic properties, and the piping fluid forces following any 1 initiating rapid transient such as an opening or closing valve, a starting or tripping pump, a pipe break or any known time dependent pressure boundary c ond it ion. A one phase condition, a homogeneous two phase condition, or a column separation condition for one or two phase flow can be predicted by the code. 3.6B-1
i Attachment 1-8 2266A l Question No. 210.3 Either w e,,1y the information identified as later in Appendix (3.9.38) 3.9.3B or provide a schedule for submittal of this information.
Response
The information identified as "later" within Appendix 3.9.3B will be supplied by September 1983. The FSAR will be amended to reflect this. s l
. . . . .o g 9 , ty if d DESIGN / SEISMIC QUALIFICATICN' REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Valve Tag No. Manufacturer Report No.
2AF-VD035SA Target Rock Corp TRP-2341 Rev B 2AT-VD036SB Target Rock Corp - TRP-2341 Rev B 2AF-VD0395A Target Rock Corp TRP-2341 Rev 3 2AF-VD040SB Target Rock Corp TRP-2341 Rev B - 3AF-VD098SB Target Rock Corp - TRP-2341 Rev B 3AF-VD0995A . Target Rock Corp TRP-2341 Rev B JAF-VD1005B Target Rock Corp TRP-2341 Rev B 3AF-VD1015A Target Rock Corp TRP-2341 Rev B 23D-VE0375A Borg-Warner Corp
- 2BD-VE038SA Borg-Warner Corp
- 2BD-VE0395B Borg-Warner Corp *
\
2BD-VE040SB Borg-Warner Corp
- 2BD-VR033SA Borg-Warner Corp
- 2BD-VR034SA Borg-Warner Corp
- 2BD-VR035EB Borg-Warner Corp
- 2BD-VR036SB Borg-Warner Corp
- 2BD-VR0775A Borg-Warner Corp
- 2BD-VR078SA Borg-Warner Corp ,
1 l 2BD-VR079SB Borg-Warner Corp , 2BD-VR080SB Borg-Warner Corp
- 2BD-VR089SA Borg-Warner Corp
- 2BD-VR090SA Borg-Warner Corp *
- Later 3.9.3B-1
WUP-3 . FSAR 2 DESIGN / SEISMIC QUALIFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Valve Tar No. Manufac turer Report No. w 2BD-VR091SB BorB-Warner Corp
- 2BD-VR092SB BorB-Warner Corp -
* /
- 3CC-3006SA Litton Contromatics No. 14225-M Rev 1 3CC-3005SB Litton Contromatics No.14225-5A Rev 1 ~
3CC-3066SN Litton Contromatics - No.14225-5A Rev 1 3CC-30675N Litton Contromatics No.14225-5A Rev 1 3CC-B068SN Littoir Contromatics No.14225-5A Rev 1 3CC-30695N Litton Contromatics No.14225-5A Rev 1 3CC-55075A Litton Contromatics No. 14225-2A Rev 1 3CC-5508SA Litton Contromatics No. 14225-2A Rev 1 3CC-B50953 Litten Contromatics No.14225-2A Rev 1 3CC-5510SB Litton Cocitromatics No.14225-2A Rav 1 3CC-5511SA Litton Contromatics * .
/
3CC-B512SB Litton contromatics . 3CC-B513SA , Litton Contromatics No. 14225-2A Rev 1 3CC-5514SA Litton Contromatics No. 14225-2A Rev 1 l ' l 3CC-5515SB Litton Contromatics No. 14225-2A Rev 1 3CC-B516SB Litton Contromatics No. 14225-2A Rev 1 2CC-B 521SB Litton Contromatics ' No. 14225-3A Rev 1 2CC-5522SB Litton Ccatromatics No. 14225-3A Rev 1 2CC-5523SA Litton Contromatics No. 14225-2A Rev 1 2CC-B524SA Litton Contrematics No. 14225-2A Rev 1
- Later 3.9.3B-2 l __
WNP-3 TSAR I f DESIGI/ SEISMIC QUALIFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's V,g,1ve Ten No. Manufacturer Report No, 2CC-352553 Litton Contramacies No.14225-2A Rev 1
~
2CC-552653 Litton Contromatics No. 14225-2A Rev 1 3CC-8531SA Litton Contromatics No. 14225-2A Rev 1 3CC-5532SB Litton Contramatics No. 14225-2A Rev 1 3CC-85395A Litton Contramatics
- 4
. l 3CC-554053 Litton Coacromatics *
~
3CC-VE255SB Target Rock Corp
- TR 2684 Rev 3 3CC-VE535SA Targee Rock Corp TR 2684 Rev 3 3CC-VE53653 Target Rock Corp TR 2684 Rev 3 3CC-VE638SA Target Rock Corp TR 2684 Rev B 2G-VP011 SBR Borg-Warner Corp
- 2CH-VP0295AR 3org-Warner Corp
- I 2CH-VP030 SBR 3org-Warner Corp * '
2CH-VSO41SA Borg-Warner Corp
- 2CH-VSO42SER Borg-Warner Corp ,
2CH.VSO44SA Borg-Warner Corp
- 2CH-VSO4733R Borg-Warner Corp
- 2CH-VWO375AR Borg-Warner Corp
- 2CH-VWO40 SBR Nissho-Iwai American Corp
- l 2CH-VW401 SBR Atwood & Morrill No. 62 Rev 0 2CH-VW404SAR Atwood & Morrill No. 62 Rea 0 l
- Later 3.9.33-3 l
- -- . y-. v-- - - - , - - . . . .-
UNP-3 FSAR DESIGN / SEISMIC QUALIFICATION REPORIS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Velve Tam No, Manufacturer Report No. 2CS-300lSA McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2CS-8002SE McNally Pittsburg Mfg Corp
- SRV-1502 Rev 1 2CS-VQ005SAR Target Rock Corp TRP-2341 Rev B 2CS-VQOO6 SBR Target Rock Corp TRP-2341 Rev B 2CS-VS015SAR Atwood & Wrrill N'o. 34 Rev B -
2CS-VS016 SBR Atwood & Morrill - No. 34 Rev B Atwood &'Morrill 2CS-V30175AR
- No. 31 Rev C 2CS-VS018 SBR Atwood & brrill No. 31 Rev C 2CS-VS021SAR Atwood & Morrill No. 32 Rev A 2CS-VS022 SBR Atwood & herill No. 32 Rev A 2CS-VSO23SAR Atwood & Morrill No. 34 Rev B 2CS-VSO24 SBR Atwood & brrill No. 34 Rev B 2CS-VS079SAR Atwood & Morrill No. 31 Rev C ,,
2CS-VS0805BR Atwood & Morrill No. 31 Rev C 2CS-VS O91SA - Atwood & brrill No. 33 Rev B 2CS-VSO92SB Atwood & hrrill No. 33 Rev B 7 2CS-VUO55SAR Borg-Warner Corp l l 2CS-VUO56 SBR Borg-Warner Corp
- 3EC-8001SA Litton Contromatics No. 14225-3A Rev i 3EC-8003SA Litten Contromatics No. 14225-3A Rev 1 3EC-3004SA Litton Contromatics No. 14225-3A Rev 1 3EC-300755 Litten Contromatics No. 14225-3A Rev 1
- La t er 3.9.33-4
WHP-3 - FSAR ( DESIGI/ SEISMIC QUALIFICATION' REPORTS FOR A/E SUPPLIID ACTIVE VALVES Valve Taz No. Manufacturer's Manufacturer Report No. 2EC-800953 Litton Contr:anatics No. 14225-3A Rev 1 2EG-5010SA Litton Contrennatics - No. 14225-3A Rev 1 2EC-B011SB Litton Contrcanatics No. 14225-3A Rev 1 2EC-B012SA Litton Contromatics No. 14225-3A Rev 1 ~ 3EC-B013SB Litton Contremstics - No. 14225-3A Rev 1 3EC-3014SB Litton contromatics No. 14225-3A Rev 1 2FP-VH005SB Borg-Warnar Corp .
- 2FP-VE006SB Borg-Warner Corp
- 4 2FP-VH00753 Borg-Warner Corp
- 2FW-VD021SB Atwood & Morrill 201-14225-00 Rev 1 2FW-VD022SA Atwood & Morrill No. 36 Rev A .
2FW-VD023SB Borg-Warner Corp * , 2FW-VD024SA Borg-Warner Corp *
. 2FW-VD02753 Atwood & Morrill 201-14225-00 Rev 1 2 W-VD028SA, Atwood & Morrill No. 36 Rev A 2FW-VD036SB Atwood & Morrill 201-14225-00 Rev 1 i
2FW-VD0375A Atwood & Morrill No. 36 Rev A 2FW-VD038SB Borg-Warner Corp *
/
1 2FW-VD0395A Borg-Warner Corp
- l 2W-VD042SB Atwood & Morrill 201-1422 Rev 1 2FW-VD043SA Atwood & Morrill No. 36 Rev A r
3.9.3B-5
- Later i
h WNP-3
- a. FSAR P""
DESIm/ SEISMIC QUALIFICATICN REPORTS 3 4 FCR A/E SUP"21.IED AC*IVE VAI?RS Valve Tag No._ Manufacturer's Manufacturer _ Recorr No. 2HA-VS0085RA Target Rock Corp TR 2684 Rev B 2HA-750095RA
'P Target Rock Corp 1
TR 2684 Rev B 2HA-VS01053A Target Rock Corp
- ' TR 2684 Rav 8 2HA-VS0275RB Target Rock Corp TR 2684 Rev B 2HA-VS028 SRB Target Rock Corp TR 2684 Rev B 2HA-VS0295RB Target Rock Corp ,y TR 2684 Rev B 2MD-VE00lSA Borg-Warner Corp .
2MD-VE002SA Borg-Warner Corp 2ND-VE003SB Borg-Warner Corp 2MD-VE004S3 Borg-Warner Corp 2MD-VE0875B
,, Borg-Warner Corp 2MD-VE088SB Borg-Warner Corp 2MD-VE089SA Borg-Warner Corp 2MD-VE090SA Borg-Warner Corp
- 2MS-P012SA l Control Coc:ponents Inc c/o Babcock & Wilcox 2MS-P016SB Control Components Inc c/o Babcock & Wilcox s 2MS-P021SB Control Components Inc c/o Babcock & Wilcox 2MS-P0235A Control Components Inc c/o Babcock & Wilcox 2MS-R006SA Dresser Industrial Valve & SQ-37-7 Rev 1 Instr.: ment Divisten i
o 1.ater
t . . WNP-3 - FSAR DESIGN / SEISMIC QUALIFICATION ' REPORTS 7 FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Valve Tar No. Manufacturer Reoort No. 2HS-10075A Dresser Industrial Valve & SQ-37-7 Rev 1 Instrusant Division 2MS-R008SA Dresser Industrial Valve & SQ-37-7 Rev 1 Instruinant Division 2MS-10095A Dresser Industrial Valve & SQ-37-7 Rav i - Instrument Division 2MS-R0105A Dresser Industrial Valve & SQ-37-7 Rev 1 Instrument Division 2MS-10115A Drester Industria1 valve & SQ-37-7 Rev i Instrument Division , 2MS-R01353 ~ Dresser Industrial valve & SQ-37-7 Rav 1 Instrument Division + 2MS-R01453 Dresser Industrial Valve & SQ-37-7 Rav 1 Instrument Division 2HS-R015SB Dresser Industrial valve & SQ-37-7 Rev 1 Instrument Division i 2MS-R01853 Dresser Industrial Valve & SQ-37-7 Rev 1 Instrument Division ' 2HS-R01953 Dresser Industrial Valve & SQ-37-7 Rav 1
. Instrument Division l
2MS-R020S3. Dresser Industrial Valve & SQ-37-7 Rav 1 Instrument Divistor> 2MS-VD001SA Anchor Darling Vc.1ve Camp E-6232-6/E-6210-1C 2MS-VD002SA Anchor Darling Valve Comp E-6232-4/E-6210-1C 2MS-VD002S3 Anchor Darling Valve Comp E-6232-6/E-6210-1C 2MS-VD00453 Anchor Darling Valve Comp E-6232-6/E-6210-1C l l 2NS-VE055SA Borg-Warner Corp *
*]
- Later l 3.9.3B-7 l
t
WNP-3 - FSAR DESIGN / SEISMIC QUALIFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Valve Tan No. Manufac turer Report No. 2H5-VE08253 Borg-Warner Corp
- 2MS-VE0915A Borg-Warner Corp -
l 2MS-VE09253 Borg-Warner Corp
- l 2M5-VE138SB Borg-Warner Corp * ,
2M5-VE1405A Borg-Warner Corp -
- 2NG-VE05755 .Borg-Warner Corp ,
2NG-VE058SA Borg-Warner Corp - .
- 2PV-B00lSA McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-80035A McNally Pittsburg Mfg Corp SRV-1502 Rev i l
2PV-B0043A McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-5005SA , McNally Pittsburg Mfg Corp SRVE1502 Rev 1 2PV-B0075A Litton Contromatics No. 14225-2A Rav 1 2PV-B009S5 Litton Contromatics No. 14225-3A Rev 1 , 2PV-50105A Litton Contromatics No. 14225-3A Rev 1 - 2PV-8014SA . ' McNally Pittsburg Mfg Corp ' SRV-1502 Rev 1 2PV-5016SA McNally Pittsburg Mfg Corp SkV-1502 Rav 1 2PV-B01733 McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-B01853 McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-B0195A Litton Contromatics N'o . 14225-3A Rev i 2PV-502353 Litton Contromatics No. 14225-4A Rev 1 2FV-5024SA Litton Contromatics No. 14225-4A Rev 1
- Later 3.9.33-8
- l. _ . . _ _ _ __ _ _ _ _ _ . _ . _ _
. . . . . . .. .. - n WNP-3 FSAR
( k ' DESIGN / SEISMIC QUALIFICATION ' REPORTS TOR A/E SUPPLIED AC*IVE VALVES Manufacturer's Valve Tan No. Manufac turer Report No. 2N-8025SA McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-50296A Litton Contramatics
- No. 14225-1A Rev 1 3PV-50335A McNally Pittsburg Mfg Corp SRV-1502 Rev 1 3PV-B034SA McNally Pittsburg Mfg Corp SRV-1502 Rev 1 3PV-8035SA McNally Pittsburg Mfg Corp
~
SRV-1502 Rev 1 3PV-5036SA - McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-80375A Litton Contronistit:s No. 14225-3A Rev 1 2PV-5038SA Litton Contromacies No. 14225-3A Rev 1 Litton Contromatics 2PV-5039SA No. 14225-2A Rev 1 2PV-B041SA Litton Contromatics
- 2PV-B042SA Litton Concroma.:ics
- 2PV-B043SA Litton Contromacies
- 2PV-8044SA Litton Contromatics
- 2PV-3045SA Litton Contromatics * '
i 2PV-5046SA . Litton Contromatics
- l l 2PV-50475A Litton Contromatics
- l 2PV-5048SA Litton Contromatics
- 2PV-8052SA Litton Contromatics No. 14225-3A Rev 1 2PV-8054SA Liccon Concromacies No. 14225-3A Rev 1 2PV-5055SA Litton Contromatics No. 14225-3A Rev 1 3PV-80605A Litton Contromacies No. 14225-1A Rev 1 l
3PV-5061SA Litton Contromatics No. 14225-1A Rev 1
- Later 3.9.33-9
WNP-3 FSAR
/0 DESIGN / SEISMIC QUALIFICATION REPORTS PCR A/E SUPPLIED ACTI'E VAL'ES Manufacturer's Valve Tea No. Manufs ecurer Report No.
3PV-5062SA Litton Contromatics No. 14225-1A Rev 1 3PV-50635A Litton Contromatics , No. 14225-1A Rev 1 2PV-5064SA Litton Contromatics No. 14225-4A Rev 1 2PV-5101SB McNally Pittsburg Mfg Corp S17-1502 Rev 1
- 2PV-B103SB McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-B104SB McNally Pittsburg Mfg Cor,. SRV-1502 Rev 1 2PV-5105SB McNally,Pittsburg, Mfg Corp SRV-1502 Rev 1 2PV-B10753 Litton Contromatics No. 14225-2A Rev 1 2PV-51095A Litton Contromatics No. 14225-3A Rev 1 2PV-B11055 Litton Contramatics No. 14225-3A Rev 1 2PV-B111SB McNally Pittsbu'rg Mfg Corp SRV-1502 Rev i s
2PV-B112SA McNally Pittsburg Mfg Corp SRV-1502 ' Rev 1 2PV-Bil3h5 McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-8114SB McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-B12353 Litton Contromatics No. 14225-4A Rev 1 2PV-B124SA Litton Contromatics No. 14225-4A Rev 1 2PV-5125SB McNally Pittsburg Mfg Corp SRV-1502 Rev 1 2PV-B12953 Litton Contromatics No. 14225-1A Rev 1 3PV-B133SB McNally Pittsburg Mfg Cor? SRV-1502 Rev 1 3PV-5134SB McFally Pittsburg Mfg Corp SRV-1502 Rev 1 3PV-5135SB McNally Pittsburg Mfg Corp SRV-15 02 Rev 1 3PV-B136SB McNally Pittsburg Mfg Corp SRV-1502 Rev 1 l [
- Later i
3.9.3B-10 i L _ _ _ _ - - - -
WNP-3 FSAR Il r 1 DESIGN / SEISMIC QUALIFICATION REPORTS FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's Valve Tar No. Manu facturer Reporr No. 2PV-313753 Litton Contromatics No. 14225-3A Rav 1 2P7-B13853 Litton Contromatics No. 14225-3A Rev,1 2PV-514053 Litton Contromatics No. 14225-2A Rev 1 ,a , 2PV-B14153 Litton Centramatics *
. 2PV-B142SB Litton Contromatics
- _
o
. , 2PV-514353 Litton Coctromatics * ..
2PV-B144SB Litton Contromatics
- 2PV-B145SB Litton Contromatics
- 2PV-316653 Litton Contromatics
- 2PV-B147S3 Litton Contromatics
- 2PV-B14833 Litton Contromatics
- 2PV-B15253 Litten Contromatics No. 14225-3A Rev 1 2PV-B154SB Litton Contromatics No. 14225-3A Rev 1 2PV-B155SB .Litton Contromatics No. 14225-3A Rev 1 3PV-B160SB Litton Contromatics . No. 14225-1A Rev 1 3PV-3161SB Litton Contromatics No. 14225-1A Rev 1 3PV-B162S3 Litton Contromatics No. 14225-1A Rev 1 3PV-8163!3 Litton Contromatics No. 14225-1A Rev 1 2PV-8164SB Litton Contromatics No. L4225-4A Rev i 1SI-VP0915AR Atwood & Morrill
- ISI-VP092SAR Atwood & Morrill
- ISI-VP09753R Atwood & Morrill '
- Later 3.9.3B-11
WNP-3 - TSAR
/E DESIGN / SEISMIC QUALIFICATION REPORIS FOR A/E SUPPLITD ACTIVE VALVES Manufacturer's Valve Tar No. Manufacturer Report No.'
1SI-VP098531 Atwood & Morrill
- ISI-VP106SAR Atwood & Morrill .
- ISI-VP1075AR Atwood & Morrill
- ISI-VP110531 Atwood & Morrill * .
ISI-VP11353R Atwood & Morrill . ISI-VP185 SBR Dorg-Warner Corp
- ISI-VP18653R Borg-Warner Corp .
2SI-VQO11SAR Borg-Warner Corp
- 25I-VQO27SER Borg-Warner Corp
- 2SI-VQO33SA Borg-Warner Corp
- 2SI-vQO35S3 Bors-Warner Corp
- 25I-VQ046SA Borg-Warner Corp
- 2SI-VQ04953 Borg-Warner Corp
- 2SI-VQ0875A Target Rock Corp * ,
2SI-VQ088SB ,
' Target Rock Corp TRP-2341 Rev 3 2SI-VQ115SA 3org-Warner Corp
- 2SI-vQ170SA Target Rock Corp -
- l 2SI-VQ171SB Target Rock Corp
- 2SI-vQ171SAR Target Rock Corp TRP-2341 Rev 3 2SI-VQ173S3R Target Rock Corp TRP-2341 Rev B 2SI-VQ201SAR Targe't Rock Corp TR 2684 Rev B 2SI-VQ202S3R Target Rock Corp TR 2684 Rev B 2SI-VQ2MSA Target Rock Corp TR 2684 Rev B
- Later 3.9.3B-12
BNP-3 - FSAR / 13
'. a DESIGN / SEISMIC QUALIFICATION REPORTS
~
FOR A/E SUPPLIED ACTIVE VALVES Manufacturer's valve Taz No. Manufacturer Report No. 2SI-VQ206SB Target Rock Corp TR 2684 Rev 3 2SI-VQ207S3R 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-VQ210$A Target Rock Corp - TR 2684 Rev B 2SI-VS054SAR . Nissho-Ivai Americha Corp
- 2SI-V5060 SBR Nissho-Iwni American Corp w 2SI-VS131SA Target Rock Corp TR 2684 Rev B 2SI-VS134SA Target Rock Corp TR 2684 Rev B 2SI VS1375A Target Rock Corp ,
TR 2684 Rev B 2SI-VS140$A Target Rock Corp TR 2684 Rav B i 2SI-VS 193SB Target Rock Corp TR 2684 Rev 3 2SI-VS194SB Target Rock Corp TR 2684 Rev B , 2SI-VS195SB Target Rock Corp TR 2684 Rev B 2SI-VS 196SB, Target Rock Corp TR 2684 Rev 3 2SI-VU001S E Atwood & Morrill No. 39 Rav A 2SI-VUO21 SBR Atwood & Morrill No. 39 Rev A 2SI-VUO55SAR Atwood & Morrill No. 38 Rev B 2SI-VUO61SER Acvood & Morrill No. 38 Rev B 2SL-VP100SABR Borg-Warner Corp
- 2SL-VP 101SABR Borg-Warner Corp
- I 2SL-VP102SABR Borg-Warner Corp
- s.
l
- Later -
3.9.33-13
~
WNP-3 , FSAR j i4 l DESIGN / SEISMIC Qt:ALIFICATION llEPORTS FCR A/E SUPPLIED ACTIVE VALVES Manufacturer's Valve Tar No. Manufacturer Recort No. 2SL-YP103SABR Borg-Warner Corp
- 2SL-VP104SABR Borg-Warner Corp .
2SL-VP10$SAAR Borg-Warner Corp * ( 1
- Later
. 3.9.3B-14 l
l . _ . . _ - , - - - - - _ .- - --
Attachment 1-9 2266A Question No. 210.4 The SRP (NUREG-0800) contains the following requirements: (3.9.3.4) All safety-related components which utilize snubbers in their support systems should be identified and tabulated in the FSAR. The tabulation should include the following information: (1) identification of the systems and components in those systems which utilize snubbers, (ii) the number of snubbers utilized in each system and on components in that system, (iii) the type (s) of snubber (hydraulic or mechanical) and the corresponding sup-plier identified, (iv) specify whether the snubber was con-structed to the rules of ASME Code Section III, Subsection NF, (v) state whether the snubber is used as a shock, vibration, or dual purpose snubber, and (vi) for snubbers identified as either dual purpose or vibration arrestor type, indicate if both snubber and component were evaluated. Provide or reference this material for snubbers utilized on all safety-related components.
Response
Information regarding snubbers on safety-related components re-ouires input from both the A/E and NSSS vendor, as well as the Supply System. This information is being assembled now and will be available December 1982. l
-10 2266A Question No.
210.5 Provide a schedule for submittal of the in-service testing (3.9.6) program.
Response
The In-service testing program will be submitted one year prior to fuel load. This committment will be included in a subsequent amendment to the WNP-3 FSAR as indicated on the attached markup of FSAR page 3.9-154.
+
/
~
h &lLLsNEh A lC. b~ ,p 3 9.6 IN-SERVICE TESTING OF PUMPS AND VALVES This section addresses the program for inservice testing for operational readiness of ASME Code, Section III, Class 1, 2, and 3 pumps and valves. The detailed program, indicating valves and pumps to be tested and the test specifics, will be submitted as a separate documentxa Tne pronram will incorporate the requirements of the latest approved edition and addenda of ' ASME Section XI, Subsections IWP and IWV at the time of its preparation. As required by 50.55a(s), the program will be revised prior to commercial operations to incorporate the operational readiness requirements of the ASME Code Edition and Addenda in effect one year prior to date of issuance of the operating license. W jest f ro*er h hell-&e J . O A {r 3.9-154
1 I Attachment 1-11 2266A Question No. 210.6 Provide in Table 5.2-1 or reference the code requirements (class, (5.2.1.1) edition and addenda) for the RCS pumps and RCPB valves (A/E).
Response
The Code requirements for the RC Pumps and valves (A/E supplied) are as follows: I Addenda Class Edition Unit 3 Unit 5 RC Pump 1 1974 None None RCP8 Valves (A/E)
- size 2-1/2 in, and larger 1 1977 None None
- size 2 in. and smaller 1 1977 Summer 77 Summer 77 FSAR Table 5.2-1 will be amended to reflect this information.
1 [ l
.~ VNb3 1A440-1 FSAR -
6 AIO. h e TAILE 5.2-1 (ll MACTDI COOLANT SYSTEM FMSSURE BOUNDART CDDE MQUnmerS ASME 301 3 1 & PRESSURE TESSEL CODE SECIION III - NUCUAE FOUE1 MAlff _ COMPONENTS , QMIFOMNT AIMEMDA CLASS EDITION UNIT 3 UNIT 5 MACTD1 VISSEL 1 1971 513851 73 WIMER 73
~
e SIIAlt GENIEA201
- FRIMAI! SIDE 1 1971 3133E173- WINIER 73
- SECONDART SIDE 2 1971- SIDGER 73 WIM ER 73 FRESSUIZZER I 1971 StagsEn 73 WINTER 73 n 9o mF- _ l tMe neue . ua_n t . ._22 1 QNE. ANT FIFE 1 1971 IIDAIER 73 WINTIE 73 VALVES FRESSURIZER SAFETT 1 1974 S13081 75 StaeER 75 FRESSURIZE1 SPEAT 1 1974 WINZER 75 WIMIX 75 McIOR OFTRATED 1 1974 WINTER 75 WIETER 75 FIEineN 1 1974 WINIII 75 WIE ER 75 CONIICL EUMEIr .
DRIVE MECEANISIt 1 1974 WI5IER 75 WIETER 75 - RCPS FIFE (A/I) 1 1974 SmtsER 76 SUlstER 76 IRCPS VALVES (A/3) 1 (IN ACCORDANCE WITE TEE CODE EDITION AND ADoENDA In EFFECT AT TIME . OF FURCIASE ORDEE) _. p.c.P 5 V At.VEI (A/E) Neu c. j 1917 WSNC Wuc 28g= omel lqer S * "" 'I _ y g , q, m d s=.ller i 19't 7 . Soame8 71 N >=s-. , . , l- - 5.2-16 . l 1
Attachment 1-12 2266A Question No. 210.7 Subsection 5.4.1.1.2 references Table 5.4-1. Either supply (5.4.1.1.2) the information identified as "later" in Table 5.4-1 or provide a schedule for its submittal.
Response
Attached are the Certified Material Test Reports provided by the material manufacturer. The CMTRs are indexed by the Pump Serial Number. FSAR Table 5.4-1 will be amended to reflect this information. I l l l l l
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THRU SUMMER 1974 ADDENDA L CODE CASE 1356 u-1160 E/4/76 w.m iru 0.K. ..w.o.1- n u st _ _- SHEET 1 OF 2
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. EXHIB T NO itREAK. --- N.DlT. 15'-50eri OR DELOW<
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Attachment 1-13 2266A Question No. 220.1 As per Regulatory Guide 1.70, summarize for each safety-related (3.4.2) structure, system, and component that may be so affected, the design basis static and dynamic loadings, including considera-tion of hydrostatic loadings, equivalent hydrostatic dynamically induced loadings, coincident wind loadings, and the static and dynamic effects on foundation properties. Provide or reference this material.
Response
Seismic Category I structures do not require flood protection as described in Subsection 3.4.1. As a result design basis static and dynamic loadings, including consideration of hydrostatic loadings, equivalent hydrostatic dynamically induced loadings and coincident wind loadings resulting from flooding conditions are not applicable. The treatment of groundwater induced hydrostatic loading is de-scribed in Subsection 2.4.13.5 which is referenced in Subsection 3.4.1.2.3. The FSAR will be amended to reflect the aforementioned response.
1465W-3 WNP-3 FSAR 6ardh N* I For the portio 3s of the vertical collection system on the wall adj: cent to the Turbine Building, the vertical stand pipes will be four feet above the ground floor EL. 390 feet (see Figure 3.4.1-2). This elevation is above the highest level (3 f t) that circulating water has been estimated to rise in the event of a CWS break inside the building (see Subsection 10.4.5). Thus any water released into the Turbine Building will be prevented from directly entering the drainage collection system. A portion of the standpipe is designed to be seismic Category I (see Figure 3.4.1-2) and as such will resist the passive pressure of the weathered sandstone in the embedded portion. Pipe straps are used to restrain the standpipe above the Turbine Building ground floor slab to resist the peak seismic acceleration of t he RAB at grade elevation. The RAB response spectra and ground acceleration are described in Subsection 3.7.1. A one inch gap is provided between the Turbine Building grou3d floor slab and the pipe, and is filled with a flexible compound. Watertight seals are provided on all below grade penetrations of the RAB to prevent groundwater seepage into the RAB and to stop water from a pipe break in the RAB from entering the groundwater drainage system. The GWDS is not classified as seismic Category I except for the manholes at the corners of the RAB and the portions of the vertical collection system on the wall adjacent to the Turbine Building. Subsection 3.8.4 provides additional discussion of the design of the Category I manholes. This classification is adequate since a failure of the GWDS (clogging of the drain pipes) during a seismic event would not cause an appreciable rise in the grcundwater level for a minimum of 115 days (see Subsection 3.4.2.1). In addition, the CWDS will be inspectable to assure proper functioning at any time, including af ter an earthquake. Partial plugging of the GWDS will not reduce its ef fectiveness. Each horizontal pipe has a full flow capacity of at least 385 gpm and can accept flow in either direction so that the drains will function satisf actorily if one outlet should become temporarily plugged. Furthermore, I ' the vertical drains are six inches in diameter to minimize the chances of plugging. In the unlikely case of a complete blockage of the system during the interval between inspections, the groundwater has been calculated to rise to an elevation of 365 feet above MSL in a minimum of 115 days during maximum recharge conditions. (See Subsection 3.4.2.1). The walls and mat are designed accordingly to withstand the resulting hydrostatic load (see Subsection
-2.5.4.10). ^-
;t.4.13.5' Any blockage in the horizontal or vertical drains can be removed by using power augers or flushing with a high pressure stream of water. Blockage in the drainage tunnel can be removed by either of the above two methods or by digging.
l The vertical manholes at each corner of the building are designed to seismic l Category I classification to provide access to the horizontal headers and the drainage tunnel in the event of an earthquake (for seismic criteria see 'tw 3.4-3
Attachment 1-14 2266A Question No. 220.2 Regulatory Guide 1.70 states that the basis for any response (3.7.1.1) spectra that differ from the spectra given in Regulatory Guide 1.60 should be included in the FSAR section. You state that the vertical design response spectra does not comply with the recom-mendations of Regulatory Guide 1.60 but do not provide a basis for this divergence. Reference or provide this basis.
Response
The procedure for development of the vertical design response spectra was based entirely on the recomendations of Newmark, Blume, and Kapur (Reference 3.7.1-1) and was previously accepted by the NRC as documented in the Construction Permit SER. As indicated in Subsection 3.7.1.1.2, the vertical design re-sponse spectra differ from those given in Regulatory Guide 1.60 only in the frequency range higher than 33 cps representing the maximum (peak) vertical ground acceleration. The peak vertical ground acceleration was chosen to be two-thirds of the peak horizontal ground acceleration based on the historical evi-dence. The available earthquake data (Reference 3.7.1-1) indi-cate that the ratio of the former to the latter ranges from 1/2 to 2/3. Subsection 2.5.2.6.2 contains further discussions of earthquake ground motions. FSAR Subsection 3.7.1.1.2 will be changed to reflect the basis for divergence.
.'100 71>2 8WP-3 a
FSAE, . M MO,1 The Category I structures are founded on sandstone; hence, the peak rock " .. acceleration determined in Section 2.5 represents the peak acceleration . associated with S E. The peak horizontal acceleration associated with SE for ' the Category I structure is 0 32 3; that associated with OE (1/2 SSE) is 016
- g. -
Eerisestal design response spectra for spectral desping ratios of 0.02, 0.04 0.07, and 010 for SSE are presented on Figure 3.7.1-3. Borisontal design respesse spectra for spectral damping ratios of 0 02 and 0.04 for OE (1/2 SSE) are presented on Figure 3 7.1-4. Mese design response spectra are i
' defined for the free field and are applicable at the plant grade level (EL.
390.0 ft). 3.7 1.1 2 vertical Design Essponse Spectra
. < *, i Me smeeth vertical design response spectra were developed following the procedure ointlined by Neumark,11tme, and Espur (Enference 3.7.1-1), which is illustrated on Figure 3.7.1-5. 21s procedure is based on modifying the herimental design spectra as follows:
a) Se vertical response spectrue la drawn as indidated by the dashed lines on Figure 3.7.1-5, by using two-thirds of the horisontal design spectrue free very low frequencies (i.e., long periods) through points D' and C', both of which pass through the same frequency as points D (frequency = 0.23 herts) and C (frequency = 2.3 herts). b) Line D'C' is extended to point C* at which the vertical design spectrum becomes equal to the horizontal design spectrum. , c) The complete vertical design spectrus is given by the line D'C" EA and then merges into line C, which represents the peak vertical ground acceleration. The point of intersection on line C is approximately at a frequency of 30 herts for all spectra with different spectral damping . ratios. , The vertical design response spectra does not comply with the recoueandations of Begulatory Qaide 1.601.e. , for frequencies higher than 33 cps the vertical design response spectra values do not follow the maximum ground acceleration line het decline to 2/3 of the maximum ground acceleration value between 33 and 30 cpe and then ins at 2/3 of that value for frequencies highe than 30 ep The pende.
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aoc am 3 7.w deste R=d *Lc4aler=N pa ae s kseiWe vertical des gn response.tspectra de rm%. e u io %. . constructed in accordance witM the 4= 2/3 - above procedure for an SSI peak vertical ground acceleration of 0.22 g and spectra damping ratios of 0.02, 0.04, 0.07, and 0.10 are pre sented on Figure 3.7.1-6. Mese design response spectra are defined for the free field and are applicable at the plant grade level (EL. 390.0 ft). 3 7.1.2 Desian une HLatory I 3.7-2
=
. -15 2266A Question No.
220.3 Regulatory Guide 1.70 states the applicant should indicate the (3.7.2.6) extent to which the procedures for considering the three compo-nents of earthquake motion in determining the seismic response of structures, systems, and components follow the recommenda-tions of Regulatory Guide 1.92. Reference or provide this information.
Response
As discussed in Subsection 3.7.2.6 the procedures used for com-bining the three components of earthquake motion in the seismic design of structures are in conformance with Regulatory Guide 1.92 except for the mat design. The reason for the exception and the alternative method used in the mat design are also presented in Subsection 3.7.2.6.
, i Attachment 1-16 2266A Question No. 220.4 Pro /ide an FSAR Subsection 3.7.4 on seismic instrumentation as (3.7.4) outlined in the Standard Review Plan (NUREG-0800) and Regulatory Guide 1.70.
Response
The Seismic Monitoring System is discussed in FSAR Subsection 7.6.1.1.7. FSAR Subsection 3.7.4, Seismic Instrumentation, as required by Regulatory Guide 1.70 will be added to the FSAR and will reference the Seismic Monitoring System. In order to meet the requirements of NUREG-0800, Inservice Surveillance capability has been provided in the Seismic Moni-toring System, such that the system instruments can be demon-strated to be operable through the performance of channel check, calibration and functional testing operation. The frequencies at which these operations will be performed will be provided in Chapter 16. The FSAR will be revised to reflect the above response.
. l 176W-7 wr.3 da2oA ysA1 3,[3 d) the containment foundation.
ry ~ i (R/v-12A/Pst-1200) - Support Systema "x *
- Ar;ohw.d^T't Auxiliary are systems supporting the operation of the seismic mo to stes Woninterruptible 1207 vital ac system.
7.6 1 1.8 Loose Parts Monitoring System The Loose Parts Monitoring Systen bas the function of indicating to the operator describedthe presenes of4.4.6.1. in Subsection loose metallic parts in the primary coolant loop as these loose parts before they cause damage to the primary system.The systes The display instruments utilised by the operator for monitoring loose parts are 7.6-3.found on the loose parts detection panel CP-14, and are listed in Table Figure 7.6-20 provides the general systen arrangement of the local sensors and i preanplifiers and of the equipment and indication available on CF-14,in the Main control Roos. Figure 7.6-21 deplets general sensor location. ( 7.6 1.1.9 Tornado Protection Yalvas J The Tornado Protection Valves (TF7) are provided on various safety-related NVAC Systems and are described in Section 9.4. The T77's are self-contained cod close automatically as described in the EVAC System descriptions. The Tornado Protection Valves (TFY) are capable of automatic return to normal . operating ctuospheric position following the restoration of normal outside ambient conditions. Operation of the valves during a , tornado do not inhibit torna do. the valves from subsequently performing normal functions followinge th of the design quantity of normal sirflow.The valves remain open during normal oper The description of the dispisy instrumentation is found in subsection 7.5.1.4.8 and it is listed in Table 7.5-17, 7.6.1.1.10 Safety injection Systen Tank Interlocka Suetto Yalv4 Isolation Yalve and shutdown Cooling km d. Saf ety injection Tonk (SIT) isolation valve interlocks and shutdown Cooling System ($DCS) suction lita valve interlocks are described and analyzed in CISSA 1-7 Subsections 7.6.1 and 7.6.2. Final control viring diagreas, for the abote valves and their interlocks are provided on Figures 7.6-22 and 7.6-13. 7.6-7
~r' 17699-13
/ . *
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g1 d) IEEE-279 Section g r '* i bypasses on this system.4.11, 4.12, 413 and 4 14 - Ther e are no aparatlag e) IEE!*279 Sestion 4.15, 4.16 and 4.17 - not applicable to th f) ese sys tems, IEZI-279 Seetton 4.20 - Diepiaya systems are described in Tables 7.provided 6-1,2 and in4. the control Room for abase 7.5.2 144 t RC lask Detection, seismic and loose Parts Monitori ng Systems
~
loose Parts Monitoring System are Chaldes onitoringdesigned Systaa, and in ac 1.45,1.12 and 1 133 respectively and since e th y with Regulatory systems the criteria sections are discussed below: of IEEE-279 is not totally applicable are not Protective
. Applitable a) 4.1 Autountie tat tletion" I
these systems are continuously monitoring plantons conditi for conditions which could have an adverse effect on continued plant operation. Subsection 7.6 The1. parameters monitored by each are discussed in t b) 4.3
" Quality Centrol of Components s'nd oMudes" l Re fer to subsection 7.3 2.3.lc.
t) 4.4 "Ecul peen t Cualf fication" Re fer to subsection 7.3.2.3.14. d) 4.8 " Derivation of_ input a" . Ra for to subsection 7.3.2.3.1b a) 4.9. 4.10 n -- - Capability and Calibration"of Sensor cheeks and Caoability for Test t The ins tru:nentation required by thes e systems is capable of bei
@testeds ec tionand s 7. 2.1.1 calibratedan d 4.4.in 6.1.accordance with scribed the methods a kls de hp f) 4.18 , ,-
"Ac cess to se tpoint Ad ius tments and Calibration" Re for to Subsection 7.3.2.3.1r.
af . 4.19 "Lientifteetion of operation"
- 4. 2 0 _"In f ores tion Re s dou t"_
t 7.6-13
3@ INSERT "I" I Inservice Serveillance Capability _ Capability has been provided in the seismi the system instruments can be demonstratedc monitoring system, such that performance cf channel check, calibration and futo be operab.1, throug nctional testing operations. For the frequencies at which these operati to Chapter 16. ons are to be performed refer l _ INST.RT "II" 7.6.2.1.4 e) 4.9, 4.10 " Capability of Sensor Checks n adC apabiliev for Test and Calibration" The instrumentation for RC Leak n, Detectio I.oose Parts,and Seismic Monitoring Systems s i capable of being tested and calibrated as in Subsections e describ 7.2.1.1, 4.4.6.1, and 7 .... 6117 i respectively, m 9 i7 2266A Question No. 220.5 In Subsection 3.8.2.5.1 you refer to Tables 3.8.2-3 and 3.8.2-3a. (3.8.2.5) These tables have not been provided. Provide these tables or a schedule for when they will be provided.
Response
The reference to Tables 3.8.2-3 and 3.8.2-3a is incorrect. The correct reference is to Table 3.8.2-2. The FSAR will be amended to reflect the response to this question.
I .o - ,. ,.. WNP-3 1576W-1 y3Ag g QC,6 { 3.8.2.5 Structural Acceptance criteria if / . i 3.8.2.5.1 Benispherical Head and Cylindriesl Shen I a.a.2-2 ; the load combinations for the containoast are specified in Subsection 3.8.2.3 and summarized in TabigM 2-2 d L" 2t. The allowable stresses for each of these load cases are summarized in 'Dable 3.8.2-11. Tables 3.8.2-12 through 3.8.2-17b compare the calculated stresses with the nuovable stresses for those critical sections of the containment shown on Figure 3.8.2-18. Primary membrane enouable stress levels conform to the general design rules l as specified in ASME Subsection NE 3131. Imed combination cases eight and nine include an accident condition with seismic loads (CEE or SCE) and pipe rupture loads. As specified in Subsection 3.8.2.4, pipe rupture loads are investigated as local effects. The various load combinations and stress limits for each type of penetration are given in Ta bles 3.8.2-4, 3.8.2-5, and 3.8.2-6. Imad combination cases 10 through @ l'1.- include pipe rupture, accident condition with seismic loads (OBE and SSE) and , jet impingement loads. General design rules and stress anowables are defined in ASME Subsection NE 3131.2. The load combinations are given in Table 3.8.2-2.
, The snowable buckling stresses were determined by the fonowing methods:
' a) Allowable BGekling Stresses for Unstiffeued Hemispherical and Ellipsoidal Esads .
Compressive stress resultants in the top head are compared to the allowable stresses obtained from the paragraphs entitled "Biazial Compression-Equal Unit Forces
- and *Eiazial Compression-Une Forces" of the Welding Research Cbuncil Bunetin No. 69(6). qual Unit
' Using these snowables for the spherical does is based on the assumption that the done acts as a cylinder with the radius equal to the radius of the
- done.
Three cases are considered (refer to Figure 3.8.2-16):
/
, 1) for a uniazial compressive stress resultant and for biaxial l unequal tensile and compressive stress resultants:
l '
$4 allowable = 1.8 x 106 g
.__ _ . ___ _ _. 1 where: t = shall thickness = 1.181 in.
1 = does radius = 900 in. Therofore Se alloveble = 2362 psi
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3.8-35
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Attachment 1-18 2266A Question No. 220.6 As per Regulatory Guide 1.70, Revision 3, provide a discussion of (3.8.3.3, the extent of compliance in the indicated sections with the 3.8.3.4 following: 3.8.3.5)
- 1. ACI-349, " Code Requirements for Nuclear Safety Related Concrete Structures".
- 2. AISC, " Specification for Design, Fabrication and Erection of I
Structural Steel for Buildings".
- 3. Subsection NF of the ASME Code, Section III, Division 1.
Response
- 1. ACI 349 was not the design code for WNP-3. The concrete internal structures were designed in accordance with the provisions of ACI 318-71 as indicated on FSAR pages 3.8-84 and 3.8-95.
- 2. The design of steel internal structures is in compliance with the requirements of AISC. As indicated on page 3.8-90 the elastic working stress design methods of AISC Part I were used.
- 3. As described on page 3.8-94.
The design and analysis for the NSSS component supports is within NSSS supplier's scope of responsibility. Attaching structures designed to mate with the component supports and allow transfer of loads to the building structure are within AE's scope of responsibility. The ASME Code jurisdictional , boundaries have been established such that these attaching ! structures are outside ASME Code jurisdiction and hence are designed in accordance with the provisions of AISC. Connection bolts between NSSS supports and attaching structures are in l accordance with ASME Code, Section II and the additional I requirements of Subsection 3.8.3.6.3(e).
Attachment 1-19 2266A Question No. 220.7 Provide your schedule for furnishing design information for (3.8.4.1.3) the Ultimate Heat Sink - Dry Cooling Towers.
Response
The following is the design information for the Ultimate Heat Sink (Dry Cooling Tower) Structure: Dry Cooling Tower Structure The Dry Cooling Tower Structure houses the Ultimate Heat Sink (UHS) components (Trains A and B) which provide heat rejection from the Component Cooling Water System (CCWS). The descrip-tions of the UHS and CCWS are given in Subsections 9.2.5 and 9.2.2, respectively. The Category I Structure is located to the east of the Reactor Auxiliary Building and occupies an area of 520 ft by 85 ft, having its longitudinal axis oriented in the north-south direc-tion as shown on Figure 1.2-2. It is predominantly a reinforced concrete structure supported on a 4-ft thick foundation mat. It extends from the top of the mat at EL. 390.00 ft to the roof grating level at EL. 437.33 ft. A wind wall that extends 17 ft above the roof level is provided for each longitudinal exterior wall. Within the structure, the heat exchangers are located at EL. 417.5 ft and supported by cross walls which are spaced ap-proximately 49 ft apart. A fan deck is located at EL. 428.50 ft. Air flow to the tower train is from each side - the east and west sides, and the concrete labyrinth or maze type design that , provides exterior missile protection by the line-of-sight ap-l proach is provided for.the air inlet. Air exhaust from the l tower train is through the missile protection roof grating. Wind walls are provided to minimize exhaust air recirculation. An electrical control building is located between two redundant tower trains and is an integral part of the tower structure housing Train A. The tower structures including the foundation mat for Trains A and B are separated by a 3-in gap to permit an unobstructed seismic displacement during an SSE. The control building is a two-story Category I reinforced concrete structure 85 ft by 24 ft in plan with an overall height of 54 ft. It houses station service transformers and HVAC equipment on the ground floor, and motor control centers on the second floor at EL. 416.5 ft. l
Attachment 1-20 2266A Question No. 220.7 There are two Diesel Oil Storage Tank Buildings located adjacent (3.8.4.1.3) to the Dry Cooling Tower structure, one at the north end of (contd.) Train A and the other at the south end of Train B. Each tank building is a Category I reinforced concrete structure housing a 92,500-gal capacity diesel oil storage tank and transfer pump located at the ground level. It is supported on a separate foundation mat 41 ft by 36 ft in plan and 4 ft thick. The roof is at EL. 437.33 ft. Loads for Dry Cooling Tower Structure Snow or Ice Load: 30 psf per Horizontal Plane Projection. Maintenance Load: 175 psf on Mat 250 psf on the Fan Deck Static Analysis for Dry Cooling Tower Structure The Dry Cooling Tower structure static analysis is performed by the finite element method using the MSC/NASTRAN program. The analyses provide stress resultants (shears, moments, and axial forces per unit length) and deformations for each element under each individual load specified in Subsection 3.8.4.3.1 as well as the load combinations required in Subsection 3.8.4.3.2. A static analysis for earthquake is performed for each of the three orthogonal directions, two horizontal and one vertical, for which the dynamic seismic loads have been separately estab-lished. For the superstructure, the SRSS method is used to ob-tain the combined effects of the three directions of seismic motion occurring simultaneously. For the mat analysis, the same approach as described in the analysis of the Category I Tank foundation mat is used. l FSAR Subsections 3.8.4.1.3, 3.8.4.3.1 and 3.8.4.4.1.2 will be amended to reflect the information of the Dry Cooling Tower Structure. i l l I
1565W-5 wwF-3 TSAI,
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Ihb crack. De portions of high enerE7 piping systems within the steam tunnel satisfy the requirements of Subsection 3.6.2.1.4 and therefore 4S are not postulated to rupturr. However, the tunnel is designed for a pas compartment pressure. associated with a hypothetical crack in the main 90 steam line. For additional discussion see subsection 3.6A.2.3.2. 3.8.4.1.3 " ' . . ' _ n t - *8 ^ %" ry moling Tower [ Sfruc3ure. 3.8.4.1.4 Category I Tank Enclosure Structure He Tank teclosure Structure is located as shown on Figure 1.2-2, and serves
" to support and enclose two 445,000 gal. Refueling Water Storage Tanks and one 340,000 gal. Condensate Storage Tank.
Me structure consists of a reinforced
,, concrete foundation mac (175 ft long, 66 f t wide and five it thick) and *
.. reinforced concrete enclosure walls ranging in height from 22.5 feet to 53 feet. Me top of the sat is at EL. 390. Inclosure walls for the Refueling Water Storage Tanks form an open roof, box type structure in order to retain liquid in the event of tank rupture. Enclosure walls for the Condensate Storage Tank extend to EL. 443.0, and together with the steel missile shield
~ roof, constitute missile protection for the tank. See Figures 3.8.4-2 through 3.8.4-5 for the masonry drawings of the Tank Enclosure Structure.
l e e l t l r O b* i 3.8-131
~
I 2 \ Insert i l f 3.8.4.1.3 Dry Cooling Tower Structure The Dry Cooling Tower Structure houses the Ultimate Heat Sink (URS) components (Trains A and 3) which provide heat; rejection from the Component Cooling Water System (CCWS). ThedescriptionpftheCBS and CCES are given in Sectiorg9.2.5m.wl %t.2 reepaakively. The Category I Structure ia located to the east of the Reactor A=-414=ry Building .and occupies an area of 520 ft by 85 ft, having its longitudinal
- axis oriented in the north-south direction as shown in figure 1.2-2.
It is pred-inantly a reinforced concrete structure supported on a 4-ft thick foundation aat. It ertends from the top of.the sat at Elevation 390.0cfc to the roof grating level at Elevation 437.33 ft. A wind wall ', that extenda 17 ft above the roof level is provided for each longitudinal exterior us11. Within the structure, the best exchangers are located at Ilevation 417.5 ft and supported by cross walls which are spaced approxi-mately 49 ft apart. A fan deck is located at Elevation 428.50ft. Air flow to the tower train is from each side - the east and west sides, and the concrete labyrinth or mese type design that provides exterior t missile protection by the line-of-sight approach is provided for the air inlet. Air awhminat from the tower train is through the missile protection roof grating. Wind walls are provided to minimize exhaust air recircula- , tion. l
'( An electrical control building is located between two redundant tower .
trains and is an integral part of the' tower structure housing Train A. The tower structures including the foundation mat for Trains A and B are separated by a 3-in gap to permit an unobstructed seismic displace-ment during an SSE. The control building is a two-story Category I reinforced concreta structure 85 ft by 24 ft in plan with an overall height of 54 ft. It houses station service transformers and HVAC equip-ment on the ground floor, and actor control centers on the second floor '. at Elevation 416.5 ft. There are two Diesel 011 Storage Tank Buildings located adjacent to the Dry Cooling Tower structure, one at the north end of Train A and the other at the south and of Train B. Each tank building is a Category I reinforced concreta. structure housing a 92,500-gal espacity diesel oil storage tank and transfer pump located at the ground level. It is supported on a separate foundation mat 41 ft by 36 ft in plan and 4 ft thick. The roof is at Elevation 437.33 ft. l
. n.
15739-4 WHP-3 5 FSAE
.r The following live loads are considered:
N Shield kilding ,, Snow, ice or construction loads on dose: 30 pef per horizontal plane jrojection. l Annulas negative pressure: 0.5 poi during normal operation Beactor Auxiliary Building Boofs: 30 pef except roof at EL. 495 for which the design load is
. 50 pef. .
. Floors: 100 pef or the equipment load in a designated area, l
whichever is greater. l Tank Enclosure Structure i 4#> l Snow or ice load: 30 pef per horizontal plane projection. 5 * *' 'l. Maintenance lood: 100 pef or 40001ho, whichever is greater. l T,= Thermal effects and. loads during normal operating or shutdown conditions, based on the most critical transient or steady state condition. Thermal loads are induced by the thermal gradient existing l across wall's between the building interior and the ambient. external ( environ 4ent during normal operating or shutdown conditions. Bo th winter and summer conditions are considered as follove:
- 1) Shield kilding Summer Winter Shutdown '
! Temp , F Temp, F F - Interior face 120 70 50 Exterior face, minimaa 7 day average - 19 19 Exterior face, maximum 7 day average 72 - - Exterior f ace sheltered in RAB 104 70 50
- 2) Esactor Auxiliary Building Summer Winter Shutdown Temp , F Temp, F F o' medoor, air 7 day average 72 19 -
Outdoor, rock below EL. 361.5 50 50 - Outdoor, rock from EL. 3 61.5 to grade Varies linearly from 50 at 361.5 to 72 at grade 1 A-lig e---
(; . I M T. Dry Coolina Tower Structure ey Snow or Ice Load: 30 pef gHorizontal Plane Projection. M acanace Load: 175 pef on Mac . l . 250 pef on the"F m Deck 9 0 4 G l l
. .. * ~
. 1379t>6 ggy.3 g FSAR 7 p y orf c %
e Q 3) r t h Tower Structure t:---' ."--
- 4) Category I Tank Enclosure Structure '
2e Category 1 Tank nelosure Structure is analyzed by the finite element method using the MSC/NASTRAN computer program. The analysis provides in plane stresses (shears, forces per unit length) and deformations in each wall and sat element under the influeene of each individual load specified in Subsection 3.8.4.3.lc, as well as the specific combinations of loads as required in Su?section 3 8.4.3 2. A static analysis for earthquake ir, performed for each of the three orthogonal directions wealdered in.the dynamic analysis. 7br the enclosure walls, the SRSS Method is used to account for all thase directions , of seismic estion occurring siamitaneously. Por the foundation sat analysis, the SRSS Nethod of combining three component seismic actions is not used because of the non-linear nature of the foundation resistance. Here a slaultaneous application of two-component seismic actions, one horizontal and one vertical, is perfonned with a load multiplier of 1.2 for the horizontal motion. See Subsection 3.7 2.6. l The enclosure walls are analyzed for flexure ~and shear either as one mer or two-wey slabs according to span ratios, and for in plane shear occurring primarily from the earth l 1he base of the enclosure walls is assumed fixed. quake loading. The foundation having no tensilenat is treated as a slab on elastic foundation resistance. The Category 1 Tank Enclosure Structure is investigated for stability with respect ta sliding and overturning in the event of SSE. , b) Dynamic analysis . Analytical techniques for t.he seismic dynamic analysis are described in kbsection 3 7.2. c) Missile analysis
- The method for missile analysis is described in Section 3.3.
3 8.4.4.2 Structural Steel Framing The portions of other Category I structures consisting of structural steel framing are analyzed for the effects of load combinations given in Subsection 3.8.4.3, using conventional procedures of structural analysis which are well i documented (References 3.5-6 and 3.8-7). 1he Reactor Auxiliary hilding has 1 1 3.8-146 l w -, ,n. - - _ .
% *9
- Insert 3 b 0
{ (3) Dry Cooling Tower Structure . The Dry Cooling Tower structure static analysis is performed by the finite element method using the MSC/NASTRAN program. The analyses provide stress resultants (shears, moments, and axial forces per unit length) and defor , nations for each elanant under each individual load specified in subsection 3.8.4.3.1 as well as the load combinations required in subsection 3.8.4.3.~2. A static analysis for earthquake is performed for each of the three ge orthogonal directions, two horizontal and one vertical, for which the dynamic m saf -f c lod = ' W arstely established. For the superstructure, the SRSS method is used to obtain the combined effects of the three directions of seismic action occuring simultaneously. For the mat analysis, the same approach as described in the analysis of the Category I Tank foundation met is used. . ( 9 9 1
- _ _ - ..~. . .
Attachment 1-21 2266A Question No. 220.8 Provide a Subsection 3.8.4.8 that discusses the effects of (3.8.4.8) masonry walls on other structures in accordance with SRP 3.8.4 (NUREG-0800).
Response
The response to this Question is being prepared in conjunction with the response to the request for additional information con-tained in Enclosure 4. item #6. The complete response will be available by December 1982. l I
Attachment 1-22 2266A Question No. 220.9 The SRP (NUREG-0800) and Regulatory Guide 1.70 state that a (3.8.5.1) description should be provided of the relationship between adjacent foundations, including any separation provided and the reasons for such separation. Reference or provide this informa-tion. Also provide the dry cooling tower information identified as later or provide a schedule for its submittal.
Response
l The location of the foundations for Category I structures is shown on Figure 1.2-2. The Category I foundations are: the common mat, the UHS Cooling Tower Foundations, the Condensate and Refueling Water Storage Tank Enclosure Foundation and the Diesel Oil Storage Tank Foundation. A separate reinforced concrete mat is provided for each Dry Cooling Tower trairr structure. A foundation mat which is 272 ft long; 85 ft wide, and 4 ft thick supports both the Train A structure and Control Building. The foundation mat for Train B structure measures 248 ft long, 85 ft wide, and 4 ft thick. i These two mat foundations are physically separated by a 3-in l isolation joint and are both founded entirely on weathered sand-stone. The Dry Cooling Tower foundations are located 50 ft from the common mat structures. Two separate foundations are provided for Category I Diesel Oil Storage Tank enclosures. Each foundation measures 41 ft long by 36 ft wide and 4 ft thick. These foundations are adjacent to the Dry Cooling Tower Train A and Train B foundations. They are l physically separated by a 2 in. isolation joint and are founded entirely on weathered standstone. The Condensate and Refueling Water Storage Tank foundation is located 41.5 feet south of the common mat. The drununing station foundation is located between the two. A 1/2 thick isolation gap is provided between the tank foundation and the drumming station foundation and the common mat. The separations between adjacent foundations have been provided to insure that there is no interaction under seismic conditions. FSAR Subsection 3.8.5.1.1 will be amended to reflect t;o. response.
WNP-: 1580u-1 FSAR g gOA 3.8.5 FOUNDATIONS 3.8.5.1 Description of the Foundations M I 3 8.5.1.1 Foundations for Category I Structures L--> A common foundation mat is constructed to support the Reactor Building and the Reactor Auxiliary Building / Fuel Handling Building Category I Structures. This foundation mat is a continuous reinforced concrete structure which is 298 feet wide, 310 feet long and nine feet in thickness. It rests entirely on the fresh sandstone of the Astoria Formation at EL. 326 feet MSL. Ho rizontal shear loads are transfered to the sandstone by friction at the bottom of the mat and by bearing against the sandstone vertical walls. IIn dit n to e bove en oned mmon at, t re a ind id 1m ty e fo da ons su p ting e HS Coo ing wer St etu and e . ate ry I ank
. E lo re St c re. Ih Categ / I anks a the two R fu ing ater 8"S j S or e Tanks nd the Co densat Stor ge Tan .
The reinforced concrete mat for the Tank Enclosure Structure is 175 feet long,
. rg344 66 feet wide and five feet thick. The top of the foundation mat is at elevation 390 feet and the mat rests entirely on weathered sandstone.
The in ation of the foundations for Category I structures and their relationship with foundations for non Category I structures are shown on C Figure 1.2-2. See Figures 3.8.5-1 through 3.0.5-6 for the typical casonry n,
- U plans and sections of the Category I foundations.
The cylindrical wall of the Shield structure is supported directly on the common foundation mat. A typical reinforcing patern for the wall-mat junction is shown on Figure 3.8.5-7. The primary shield wall, secondary shield wall and refueling cavity wall of the Reactor Building internal structure , and the 5 teel Containment Vesse l (descriptions of which are given in Subsections 3.8.3 and 3.8.2 respectively) are supported on concrete fill which transfers the loads, by bearing , to the common foundation mat below. For the Reactor Auxiliary Building (and Fuel Handling Building), load transfer to the mat is achieved through the walls and columns. The typical reinforcing pattern for the wall-mat junction is shown on Figure 3.8.5-8. . A drainage system, described in Subsection 3.4.5, is provided and no water I proofing membranes are utilized. Piling or similar foundation concepts are not used. The URS Drv Cooling Tower Lat e r C 3.8-154
.a
D h![$N. Nh!.fl 3 .. . , h.M .P. -. INO..N,!.T.h.. M'd!?-$
- A@i w .; a . ;;@? :'.N;. . $*.,j@. */7.. 7 9ii.i.4. .T i -@.sn ...o:-n:e..%
;E'.
.,.,,..,u .w.
- . . a 2-
~
6l270.9 13 ay k$ y The location of the foundations for category I strutturea is shown on Tigure 1.2-2. The Category I f oundations nre: the een:en =st, the Uns cooling Tower Foundaticas, the Condensate and Tsefueling WatergTank Enclosure Foundation and the Diesel Oil Stora;;e Tank Feundation. s94. b s er t 'l A separate reinforced concrete mat is provided for each Dry Cooling Tower train structure. A foundation mat which is 272 ft long; 85 ft vide, and 4 ft thick supports both the Train A structurc and Control Building. The foundation mat for Train B structure measures 248 ft long, t:5 ft wide, and 4 ft thick. The r.e two 34t foundations are physically acparated by a 3-in inclation joint and are both founded entirely on venchered sandstone. The Dry Cooling Tower foundations cre located 50 it f rom the cot =on mat struotcres. Ja sev4 ~3
'No separate foundations are provided f or Category I DicscI 012 Storage Tank cnclosurec. Each foundation acasures 41 ft 3cnf by 36 ft vide and 4 ft thick.
These foundations are adjacent to the Dry Coeling Tower tra :. A and train 3 foundations. They are physically separated by a 2 in. isole.tlon j oint and cre counded entirely on vaathered sandstone. l fe..j4 The Condensate and Refueling Water Storarc Tank foundation in located 41.5 feet south of the cot ==on mat. The dru=:ing station f oundation is located between
- the tve. A b thich isolation gap is provided between the tank foundation and the l dru=.ing station foundction and the co
- =:n =at. B 1
l 1 hseA 5 . The separr.tlons'between adjacent fou.,datiens hartbcen provided to insute that I 1 there is oc interaction under seit ic conditions. l l
- 1 M M NY rrf..e t uiw.r .y e .
- ._-~%, .~ . wm _ .e, . ~ ___ m... .,..y_
Attachment I h23 2266A Question No. 241.1 In your write-up of FSAR Subsections 2.5.4 and 2.5.5, there are (2.5.4 and a number of omissions of symbols and functions used in your 2.5.5) mathematical expressions. Please thoroughly proofread your text and correct it so that it can be reviewed.
Response
The errors and omissions in Subsections 2.5.4 and 2.5.5 have been corrected. The attached pages identify the corrections. The FSAR will be amended to reflect the response to this question. l l l l l I 1 l l l
> WNP 0831W-4 FSAR g
fAgL lkWEl A-series, B-series, D-series, L-series and CT-series borings; and results are presented in Appendix 2.5A. Typical values for the index properties of the fresh sandstone are sum =arized in Table 2.5-16. 2.5.4.2.2.1.2 Compressive Strength The compressive strength of f resh sandstone specimens was determined in the laboratory by uniaxial compression tests. Tests were perf omed in general accordance with ASni Standard D2938-71a. The test sa=ples had a length to diameter ratio ranging f rom 2.0 to 2.5 to assure a f airly unifom stress distribution through the sample. The rate of loading was controlled to minimize the effects of rate of loading on the strength of the sample. The compressive strength of f resh sandstone was calculated by using the empirical relationship presented in ASTM Standard L2938-71a for specimens with L/D ratio equal to 2. 0.889 o,
'c 0.778 + 0.222D L
where 'm = compressive strength for L/D at failure
'c = compressive strength for L/D = 2 L = length of the specimen C D = diameter of the specimen.
Results are summarized in Table 2.5-16 and details of test procedures are discussed in Appendix 2.53. 2.5.4.2.2.1.3 Static Deformation Properties Deformation properties determined include tangent modulus E7 and Poisson's ratio /7 Tangent Modulus E T
- The tangent modulus E7 was determined by =casuring incremental axial strains (l'a) and corresponding incremental axial stress (a"a) in uniaxial compression tests. The tangent modulus E7 was computed at 50 percent of ultimate uniaxial strength as:
ETs ca ar a Results are summarized in Table 2.5-16 and details of test procedures are discussed in Appendix 2.5B. Poisson's Ratio 9 7 - The Poisson's ratio 47 was deter =ined by measuring incremental axial strain (ara) and incremental radial strain (arr) in C 2.5-149
WNP- 3 0831W-5 FSAR 1 uniaxial compression tests. The Poisson's ratio 87 was computed at 50 percent ultimate uniaxial strength as: T*- S'r ata Results are su=marized in Table 2.5-16 and details of test procedures are discussed in Appendix 2.5B. 2.5.4.2.2.2 Dynamic Properties Dynamic properties were investigated through field measurements and laboratory testing. The field investigation consisted of cross-hole shear-wave velocity measurements, and seismic-vave refraction measurements to determine compressional-vave velocities. The laboratory test program consisted of sonic velocity measurements to determine compressional-wave and shear-wave velocities on rock specimens. 2.5.4.2.2.2.1 Field Geophysical Measurements Co=pressional-Wave Velocity (V p )f - Results of compressional-vave velocity measurements f or the f resh sandstone are summarized in Table 2.5-16 and details of test procedures are presented in Subsection 2.5.1.2.11 and Appendix l 2.5D. Shear-Wave Velocity (V 3 )f - Results of shear-wave velocity measurecents f or the f resh sandstone are sum =arized in Table 2.5-16 and details of test procedures are presented in Subsection 2.5.1.2.11 and Appendix 2.5D. l Dynamic Deformation Procerties - Based on elastic theory, in situ dynamic tangent modulus Et and dynamic Poisson's ratio /f can be determined from the typical values for the compressional - and shear-vave velocities and for therockmass[as(ASTMD2845-61):
/f (V - 2V3 )/[2 (V - V )]
2 2 ' E' = S+( P~ S} 2 2 v p -V g Results of calculations are presented in Table 2.5-16. 2.5.4.2.2.2.2 1.aboratory Sonic Velocity Measurements Sonic velocity measurements were conducted on saturated specimens of fresh sandstone axially loaded at 200 lb/in.2 The axial stress was so selected that the mean stress [(jf a -fa )/3] in the specimen was similar to the M 28-l50 g , --
0831W-6 W1.t'- 3 FSAR 3 C calculated stress in the field. Compressional-Wave Velocity (V p )1 - Results of compressional-wave velocity measurements on f resh sandstone specimens are summarized in Table 2.5-16 and details of test procedures are presented in Appendix 2.5C. Shear-Wave Velocity (V3 )1 - Results of shear-wave velocity measurements on f resh sandstone specimens are summarized in Table 2.5-16 and details of test procedures are presented in Appendix 2.5C. Dynamic Deformation Properties - Based on elastic theory, dynamic tangent modulus El and dynamic Poisson's ratio ui can be determined from the representative values for the compressional- and shear-wave velocities and for the selected rock f mass using the equation presented in Subsection 2.5.4.2.2.2.1 2.5.4.2.3 Properties of Weathered Sandstone The weathered sandstone below the plant grade is of low to moderate hardness, moderately weathered, yellow brown, coarse to fine grained; see Appendix 2.5A for classification criteria. Le mineralogy of the weathered sandstone is described in Appendix 2.5I. 2.5.4.2.3.1 Static Properties Static properties determined include index properties, compressive s errangth, and deformation properties. 2.5.4.2.3.1.1 Index Properties Unit weight, saturation water content, specific gravity, core recovery, and RQD were derived by employing the procedures described in Subsection 2.5.4.2.2.1.1 and are presented in Ta ble 2.5-15. 2.5.4.2.3.1.2 Compressive Strength The compressive strength was derived by e= ploying the procedures described in Subsection 2.5.4.2.2.1.2 and is presented in Table 2.5-15. 2.5.4.2.3.1.3 Static Ce formation Properties ne tangent modulus ET and Poisson's ratio uywere derived by employing the .);lc procedures described in Subsection 2.5.4.2.2.1.3 and are presented in Table 2.5-15. O 2.5-151 . - _ = _ _ . . _ _ . _ _ _ _ --- -- -
- h
0832W-2 WNt - a FSAR y 2.5.4.2.4.2.1 Field Geophysical Measurements Tuff Bed 1 could not be identified during seismic-vave refraction meas-urements. Shear-wave velocity of Tuff Bed 1 was measured at two locations. Results are summarized in Table 2.5-17, and details of tes t procedures are des {cribed in Subsection 2.5.1.2.11 and App. 2.5D. 2.5.4.2.4.2.2 Laboratory Sonic Velocity Measurements Compressional and shear-wave velocities (Vp )1 and (VS )1, dynamic tan-gent modulus Ei , and dynamic Poisson's ratio ui were determined using the procedures described in Subsection 2.5 ' 2_2.2Tnd are presented in Table 2.5-17. A. f.4. 2. 2. 2.1 2.5.4.2.5 Properties of Class I Soil-Cement Fill Material 2.5.4.2.5.1 General Soil cement was used for the Class I fill to backfill the access ramps leading
! .o both Reactor Auxiliary Building excavations. In addition, the soil-cement was used to stabilize the north face of the excavation for Reactor Aux-iliary Building No. 5. Figure 2.5-117 shows the locations of soil-cement.
Based on a review of the unconfined strength and shear modulus of fresh sandstone previously measured (refer to Appendices 2.5B and 2.5C re-spectively), values of unconfined compressive strength of 600 lb/in2 and shear modulus of 400,000 lb/in2 were selected as representative of the f resh sandstone. A laboratory testing program was performed to determine a blend of processed concrete sand, Portland cement and water which would closely resemble the static and dynamic properties of the fresh sandstone. 2.5.4.2.5.2 Specimen Preparation and Testing Representative samples were obtained of the Class I concrete sand from the site batch plant. Standard Proctor Compaction Tests ( ASTM D-698-78 or ASTM D-558-57) were performed o~n the Class I concrete sand with cement contents of 0 percent, six percent, eight percent , 10 percent and 12 percent to obtain maximum densities and corresponding water contents (optimum water content). Subsequent to determining the optimum water content, additional compaction samples were prepared by varying the percent compaction for each cement content. These samples were then cured for seven, 28 and 101 days prior to further testing. Sonic Velocity and Unconfined Compression Tests were performed on the cured samples. In the Sonic Velocity Tests compression and shear wave velocity mea-surements were made with a load of 16 kg applied to the sonic velocity heads. Unconfined Compression Tests were performed on specimens typically loaded using two dif ferent strain rates. Initially, a strain rate of about three percent / hour was applied until the strain reached between 0.05 percent O l i 2.5-153 I \ h
0834W-3 WNP- 3 FSAR 'f C = cohesion q = surcharge pressure Ny = tan6(45 + 4/2) + 1 NC
=
7 can4 (45 + f/2) Nq = tan6 (45 h/2) 4 = rupture angle
- 2.5.4.10.2 Averags. Displacement
- The average displacement of the mat under combined static and dynamic loading was found by using the equation derived from Boussinesq for the normal displacement of a semi-infinite elastic solid under the action of a normal load (Staff and Zienkiewicz 1969) and can be expressed as:
2) u 3 , @iii P (1 - EE (1 + i B where: S = average displacement lii = displacement coefficient dependent on the shape loaded surf ace and the distribution of load (Stagg and Zienkiewicz 1969) 9 = Poisson's ratio E E
=
design tangent modulus P = total load applied D = depth of embedment B = width of foundation A = area of foundation 2.5.4.10.3 Subgrade Modulus The subgrade modulus K is calculated as: I K" where: P = total load applied A = total area on which load is applied O l 2.5-166 l , 4 _;__- --.-- __ _
0834W-6 WNP-3
! PSAR h
1 C 2.5.4.11.2 2.5.4.11.2.1 Design Parameters for Presh Sandstone Static Parameters The required design parameters are grouped into four categories: in situ index properties, in situ compressive strength, in situ static deformation properties, and ir. situ permeability. ' 2.5.4.11.2.1.1 In Situ Index Properties The design water content, total and dry unit weight, and laboratory < compressional and shear-wave velocities of fresh sandstone are obtained based a the statistical laboratory; method using values determined on rock specimens in the see subsection 2.5.4.2.2.1 1 The design RQD is obtained based on the statistical method using the RQD recorded on the summary logs. values for the field compressional and shear-wave velocities are selected inThe design Subsection 2.5.1.2.11. 2.5.4 11.2.1.2 In Situ compressive Strength the representative value of compressive strength of the fresh sandstone cores is obtained based compression on the tests; see statistical method using results of the laboratory Subsection 2.5.4.2.2.1.2. Using a reduction factor RF, ' the representative compressive strength is reduced to incorporate the effects of in situ geologic discontinuities (Deere et al 1968). The design compressive strength is, thus, stated as: 5
*E " #cs (RF) where: ag =
design compressive strength
=
'e s statistically selected value of laboratory compressive strength RF =
l [(V)f(V)1]2 p,/ p l The reduction factor is obtained using the design values for the field and laboratory compressional-wave velocities. 2.5.4.11.2.1.3 In Situ Static Deformation Properties Two parameters are required to define the in situ static deformation properties, the elastic modulus E frequency and nature of geologicEdiscontinuities and the Poisson's ratio / E. fThe are significant actors in [ determining the properties of fresh sandstone, and their effect can be i incorporated Young 's modulus.in the design equations by use of the reduction f actor applied to C 2.5-169
m -- y Y 7 0836W-5 WNP- 3 FSAP, selection of soil strength parameter f or use in the analysis of the man made slopes is discussed in detail in Subsection 2.5.5.2.3.3. The values of elastic modulus Ei and Poisson's ratio)J are selected based on laboratory test re sults presented in Appendix 2.5B and summarized in Table 2.5-14. M The cyclic strength is graphically expressed in terms of the variation of cyclic strength versus the number of cycles to cause 5 x 10-2 s trai n. This curve is based on laboratory results discussed in Appendix 2.5C and presented on Figures 2.5-101, 2.5-10 2, and 2.5-103. Values of dynamic shear modulus are graphically ex;ressed in terms of the variation of shear modulus versus shear strain. This curve is based on field measurements and laboratory test results discussed in 1 Appendix 2.5C and presented on Figure 2.5-104. The damping ratio A versus shear strain curve based on laboratory cest results described in Appendix 2.5C is pre sented on Figure 2.5C-13. 2.5.5.2.1.2 Weathered Sandstone The properties of weathered sandstone were selected based on field and laboratory test s. A total of 40 index property tests were made on specimens from the plant area. The results of these tests are discussed in Appendix 2.5B and presented in Table s 2.5B-1 and 2.5B-2. Based on these results, a saturated unit weight of 130 lb/f t3 and a submerged unit weight of 68 lb/f t3 were selected. Static strength parameters were selected based on results of 13 uniaxial compression tests on saturated specimens taken from borings in the plant area. l g These results are discussed in Appendix 2.5B and presented in Table 2.5B.2. Based on these results, the ultimate streng th of the wea thered sandstone varies betwgen 325 lb/in.2 and 800 lb/in. 2 (i.e. , between 4 6.8 k/f t2 and 115 l k/ft ). The grain size, nature of cementation, and the fracture pattern in l uniaxial compression tests indicate that the weathered sandstone is a C-0 material. If 0 = 0, the above range of ultimate strengths could be ) interpreted to correspond to cohesion ranging between C = 23.4 k/f t,2 to 5 7.6 t k/ft 2. Fracture planes in the uniaxial compression tests indicate inclinations between 550 and 7,00 to the horizontal. Based on these - inclinations, the sandstone could be interpreted to have an angle of internal l friction 0 ranging between approx. 200 and 500 This range of angle of ) internal friction 0 is in agreement with published data f or sandstone (e.g , Stagg and Zienkiewicz 1969). Based on these values, a cohesion of 23 k/f tg and an angle of internal friction 0 = 160 were selected for analysis of the natural slopes. Analysis of man made slopes that could have influence on the Category I structures were performed using the more conservative parameters of cohesion equal to 23 k/f t2 and internal friction 0 = 00 All of the above values are conservative since both the C and 0 selected represent the minimum values interpreted from the tes t data. In addition, the range of the tested strength values is very similar to the strength range f or fresh sandstone. Dynamic properties of the weathered sandstone were selected based on field shear-wave velocity measurements. Below EL. 390 (i.e. , below the plant grade) I the weathered sandstone has a shear-vave velocity greater than 3000 f t/sec; see Appendix 2.5D. Above EL. 390 the shear-wave velocity varies between 3000 f t/see and 2300 f t/sec. Hence, below EL. 390 the shear modulus, damping ratio, and '3 variation of shear modulus and damping ratio with shear strain are considered to be such as to produce no site-dependent ef fect s. Above EL. 390 the shear moduli l
- 2. i. 179 t
@QU6W-8 WNP-3 FSAR Y 2.5.5.2.2.2.2 Lowe's and Taylor's Methods The modified Swedish Method suggested by the Corps of Engineers (1970) considers that the direction of forces between the slices remains constant throughout the force polygon. To ascertain the effect of change in the direction of these forces, the slip surf ace with the smallest f actor of safety in Profile I was analyzed using a method proposed by Iowe (1967) in which the forces between average chord. the slices are considered to be in the same direction as the insignificant change (9.9 instead of 10).For the surface The same analyzed, slip surface was alsothe factor of safe analyzed using the Friction Circle method (Taylor 1948). The analysis gave a somewhat lower factor of safety (9.2 instead of 10) but does not produce a significant change in the results. The procedure utilized in Iowe's method ,
and Taylor's method is shown on Figure 2.5-108. 1 2.5.5.2.2.2.3 Wedge Method The stability of Profile 1 was analyzed. For this purpose wedges bounded by the surface slip surf aceof the considered. were weathered sandstone, the excavation surface and a planar Three planar slip surf aces were selected based on proximity to the Category I structures 0 , 60 and 120 to the horizontal; see Figure on a common mat with inclinations of 2.5-109. All planar surfaces intersect the interpreted weathered rock surface; thus the wedges defined by them do not encounter passive resistance on the downslope side. This is conservative because inclusion of a passive resistance in analysis would increase the factor of safety. Results of the analysis shows.that there is 34).an ample factor of safety (FS) with respect to wedge failure (FS = 16.7 to 2.5.5.2.2.2.4 Sensitivity Analysis To determine the effect of variation in slope geometry and material properties on static slope stability, a sensitivity analysis was made on Profile 1. Calculations based on the contours given on Figure 2.5-100 indicate that the slope of the sandstone surface varies from 190 to 430 The interpreted slope of Profile 1 is 2601 For sensitivity analysis, Profile 1 was readjusted to a surface slope of 450 In the above analyses, the shearing strength of weathered sandstone and fresh sandstone was defined by a cohesion C = 23 k/f t2 (minimum laboratory value for 9 = 0) and an angle of internal frictiong = 160 (minimum value based on fracture pattern). For sensitivity analysis, a cohesion C - 23 k k/ft and an angle internal friction 9 = 00 were selected. The profile was analyzed using both modified Swedish Method and the wedge method. Results of the analyses are shown on Figure 2.5-110. Based on the modified Swedish Method, the smallest factor of safety is 6. Based on the wedge analysis, the smallest factor of safety is 8.5; i.e., in both cases there is ample factor of safety. 2.5.5.2.2.3 Dynamic Slope Stability Analysis 2.5.5.2.2.3.1 General Considerations
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Profiles 1, 2, 3, and 3A were analyzed. The dynamic analyses were based on l 1 the same general considerations as the static analyses (given in Subsection 2.5.55262.1) except for the magnitude of the lateral force applied to the wall
- 2. 182
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FSAR l of the excavation. Where applicable, the lateral f orce is a dynamic f orce generated by the relative motion of the Category I s tructures on a common mat. Its magnitude is obtained by a soil-structure interaction analysis described in Subsection 3.7.2.4 In the analysis the f orce is considered to be distributed uniformly over the vertical f ace of the excavation. 2.5.5.2.2.3.2 Modified Newar k Me thod The method proposed by Newark (1965) was applied to Profiles 1, 2, 3, and 3A. On each profile slip surf aces with the minimum static factor s of saf ety were a nalyzed . As an example, Figure 2.5-111. s hows the f orces considered and the procedure used f or slip surface on Profile 3A. The f orce NW is the f orc e applied through the center of gravity of the slip surf ace. For a minimum + value of N, the force NW is oriented in a direction perpendicular to the line joining the center of the slip surf ace with the center of gravity. If the static and dynamic shearing resistance are selected approximately equal, the approximate minimum value of N f or a circular slip surf ace is given by N = (FS - 1) sinf FS = static f actor of safety
@ = angle between vertical and the radius to the center of gravity N = a coefficient when multiplied by the weight of the sliding mass gives the total dynamic f orce required to cause movement.
O The minimum values of N were calculated and compared with the resultant value o f N f or S SI (N = 0. 3 9) . 3, and 4 Minimum values of N were determined f or Profiles 2, The Newmark me thod canno t be directly applied to Profile 1 because of the lateral dynamic f orce that is a function of N. Minimum Static Pr o file , Factor of Safetv Value o f Coef f!.cient N 1 , 10 NA 2 16 1.8 3 10 3.3 3A 9 2. 5 As can be seen from the above table, the value of N required to cause movement in Profiles 2, 3 and 3A is greater than the value of N f or SSE. Therefore, there is an ample factor of safety. 2.5.5.2.2.3.3 Pseudo-Static Analysis The dynamic stability of Profile s 1, 2, 3 and 3A wa s also calculated by making pseudo-static analyses on the same slip surf aces. The procedure utili:ed is illus trated on Figure 2.5-111 f or a slip surf ace f or Profile 3A. Ihe procedure is s uggested by the U.S. Corps of Engineers (EM-1110-2-1902, April 1970). To each slice the hori:ontal seismic f orce was applied in a direction out of the slope, and the vertical seismic f orce was applied in a direction up l C 2.5-183 __ w
0837W-3 WNP-3 IO i FSAR Profile 7 shows the thickest extent of residual soil underlying a cut slope (V and failure. the lowest safety factors against a pseudo-static slip circle type For these reasons, this profile was subjected to a dynamie' slip circle type analysis as described in Subsection 2.5.5.2.3.8, using the average soil strength properties described in Subsection 2.5.5.2.3.3. The location of the (Figureworst circle along with the minimum factor of safety is chown on profile 7 2.5-114). Further evidence that the residual soil slopes will remain stable during and af ter the SSE can be seen by reviewing the residual strength of the residual soil materials. Figure 2.5-119 is a composite plot showing material shear strength under undrained conditions [( a1- a 3)/2] versus axial strain for the various unconsolidated undrained triaxial compression tests that were performed on the material (see Figures 2.5B-15 through 2.55-18). These curves generally show a peak strength at small strain levels followed by a subsequent leveling off residual or decrease strength in resistance until a more or less constant undrained is reached. ! Also shown on this figure is the average static shear stress induced along the slip circle sliding surf ace for profile 7 having the lowest factor of safety under static conditions. (Figure 2.5-114 circle "c" e = 260 e = 750 psf r)(= 1.4 ksf. ) . Comparing the induced shear stress against the material residual shear strength indicates that the M residual shear strength is higher than the induced stress with a f actor of safety of 1.4. Therefore, the material has adequate undrained strength to resist sliding during the SSE as shown in previous paragraphs of this section, as well as adequate residual strength to resist sliding following the seismic event. W 2.5.5.2.3.5 Static Slip Circle Analysis The static stability of the permanent man-made alopes against a slip circle type failure was determined using the ICES-LEASE-1 computer program which employs the Fellenius slip circle method. LEASE-1 is a subsystem of ICES designed to perform stability analyses of arbitrary slopes by the method of slices. The f ailure surf aces are assumed to be arcs of circles. This computer program locates the radius and center having the minimum f actor of safety by starting at a specified trial center and using a random search technique. i The static stability analysis was performed on each of the five profiles discussed in Subsection 2.5.5.2.3.4 using each of the three sets of residual ( soil strength parameters discussed in Subsection 2.5.5.2.3.3. Input data to the ICES-LEASE program included the following: (a) properties of the soil,friction, internal including etc., submerged weights, pore pressure, cohesion, angle of (b) water level data; (c) total weight of the overlying mass of soil and/or water; (d) the inclination of the failure surfaces at the bottom of the slice. The factor of safety for the static analyses is defined as the ratio of the moment of the available shearing forces on the trial failure surface to the net moment of the driving force. The minimum allowable safety factor is 1.5 {w static. 2.5-187
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Attachment 1-24 2266A Question No. 241.2 You have indicated that, during excavation, you discovered (2.5.4.5.4) ground cracks adjacent to the excavation of Reactor Auxiliary Building (RAB-3). Provide plans and cross-sectional details clearly identifying the location of these cracks and the corre-sponding location of sets of pins with strain gauaes that you installed to monitor the cracks. Provide the time-displacement monitoring records of the installed strain gauges. Also, dis-cuss in detail, using appropriate drawings, the rock joint and dip pattern in this area. Also discuss if the cracks observed during excavation could pose any landslide or slope stability problems in the future during the plant operation.
Response
The locations of ground cracks and associated sets of pins with strain gauges are shown on Attachment I, Information Drawing - Instrumentation Locations. The results of all the time-displacement monitoring records of the installed strain gauges are shown on Attachment II, Crack Monitoring - Unit No. 3. The ground crack monitoring program was initiated in order to provide early warning of rock instability or hazardous condi-tions as part of the overall Workman Safety - Related Instrumen-tation Program. Their development was caused by tensional stresses resulting from the excavation of the adjacent sandstone for the RAB-3 foundation. The excavation mapping program sum-marized in FSAR Appendix 2.5F indicated that extensive cracking did not occur on either the sandstone floor or walls of RAB-3. The excavation for RAB-3 lies roughly 2,000 feet stratigraphi-cally above the base of the Astoria Formation. The general bed-ding attitude of the Astoria Formation at the site in N 75 E, 12 N as indicated on FSAR Figure 2.5-53 Site Locality, Geology and Structure map. The Astoria Formation which is predominantly sandstone is typically thick-bedded and massive to faintly cross bedded. The distinguishing of true bedding from cross-bedding in many sandstone outcrops is difficult and results in some locally erratic bedding attitudes. As a result, the stability of the slopes adjacent to the seismic Category I structures were analyzed using conservative material properties under both static and dynamic loadings as described in FSAR Subsection 2.5.5. The results of the analyses indicate that the slopes adjacent to the seismic Category I structures are stable. . - - - . - _ - - - . , ~ , , , , . _ , , _ - _ , _ , , ,_y_, -
Attachment 1-25 2266A Question No. 241.2 As a result of the time-displacement monitoring records, the (2.5.4.5.4) extensive geologic mapping program and the stability analyses, (contd.) it was concluded that the " hairline" cracks adjacent to the excavation of RAB-3 could not pose any landslide or slope stability problems in the future during plant operation. l l
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-26 2266A Question No.
241.3 Provide in tabular form, the as-built dimensions of various (2.5.4.10) seismic Category I structural foundations (length and width), foundation elevations, description (and thickness) of soil / rock characteristics between the foundation base and fresh sandstone, area foundation loads for static and dynamic conditions, corre-sponding bearing capacities, and resultant factors of safety.
Response
Attached in tabular form is the response to your question. e
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Attachment 1-27 2266A Question No. 241.4 From your FSAR write-up, it is not clear to the staff whether (2.5.4.10) you are currently monitoring settlement of various Category I structures or not. Modify your FSAR to identify the settlement monitoring program and give reasons if you are not currently monitoring Category I foundation settlements. Provide location drawings of settlement monuments along with the time vs. settle-ment plots that include up-to-date rebound and settlement data obtained for all Category I structures where settlements have been monitored.
Response
Since all seismic Category I structures were founded directly on either the fresh or weathered sandstone settlement and rebound wa: not a factor in the design of the plant. The average dis-placement of the Category I strectures on a common mat was due to the recompression o' the fresh sandstone underneath the foun-dation mat placed at EL. 326'. This displacement was caused by the gradual reloading of the fresh sandstone during construction to the bearing stresses of approximately 81 lb/ft2 . This dis-placement was elastic in nature and most of it took place during construction. The post-construction displacement is computed using the equation presented in Subsection 2.5.4.10.2 to be less than one-quarter inch and the differential displacement will be even smaller. Therefore, displacement is not a factor in the design of the plant and no instrumentation is required for sur-veillance of foundations for safety-related structures.
l Attachment 1-28 l 2266A Question No. 241.5 You mention that the slopes selected for stability analysis (2.5.5.1) were those bounded by the interpreted surface of the weathered sandstone, shown on Figure 2.5-124. You have not included this l figure in the FSAR. Provide the figure or a correct reference j to its location in the FSAR. 1 Response The correct reference is Figure 2.5-100. FSAR Subsection 2.5.5.1 will be amended to reflect the response to this question. l I
- . , . _ . _ . ~ . . -
0836W-1 WNP- 3 FSAR k(dt,5 N'I I 2.5.5 STABILITY OF SLOPES The plant grade is established at EL. 390 through a cut and fill operation; see Subsection 2.5.4.5.1. As shown in the excavation profiles presented on Figure 2.5-9 7, there are natural slopes in the north-south direction and man-made excavation slope s in the eas t-west direction. To analyze the static and dynamic stability of natural and man-made slopes, the following procedure wa s utilized: a) Select representative slope at the plant location. b) Select applicable material properties. c) Examine available analytical methods f or static stability and select one method. d) Analyze the static stability of all slopes by the selected method and identif y the slope with smallest f actor of safety. e) Analyze the static stability of the slope with smallest f actor of safety by other analytical me thod s. If required, also analyze other slopes. f) As required, make a sensitivity analysis on the slopes analyzed in (e) to study the effec t of variation in slope geome try and material properties.
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g) Examine available methods f or dynamic stability and select one method. f& h) Analyze the dynamic stability of all slopes by the selected method and identify slope with smallest f actor of safety.
- 1) Analyze the dynamic stability of the slope with smallest factor of safety and, if required of other slopes, by other analytical methods.
j) As required, make sensitivity analysis on *he slopes analyzed in (1) to s tudy the effect of variation in slope geometry and material properties. 2.5.5.1 Slope Characteristics 2.5.5.1.1 Selection of Slopes Na tural Slopes - To select profiles f or natural slope stability analysis, the areas north and south of plant location below the plant grade (EL. 390) were considered. Below this elevation, materials generally underlying the slopes to the north are Helm Creek Deposit, residual soil, weathered sandstone and fresh sandstone. Generally, the materials underlying the slopes to the south are residual soil, weathered sandstone and fresh sandstone. On both slopes the thickness of residual soil below EL. 390 is relatively small ((10 f t) and would not af f ec t the Category 1 s tructures; therefore, it is not considered in stability analysis. The Helm Creek Deposit was more than 100 f t. away from the edge of the excavation, and its dynamic properties were not ascertained. Therefore, its properties were not considered in analysis. Thus, the slopes { ""# selected for stability analysis were those bounded by the interpreted surf ace 'of the weathered sandstone. This surf ace, interpreted from the borings at the plant location, is shown on Figure 2.5j)S4C Contour s on Figure 2.5-100 ref er to
-10 0 2,5-175
Attachment 1-29 2266A Question No. 241.6 Your bases for selecting the critical cross-sections for slope (2.5.5.1) stability analyses of natural as well as man-made slopes are not adequately justified. Provide sufficient details of your reasons for the selection of critical slopes for the staff's independent evaluation.
Response
The bases for selecting critical cross-sections for slope stability analyses are explained below. Attachment I shows the locations of critical cross sections se-lected for detailed slope stability analysis, superimposed on a contour map showing the top of the weathered sandstone. Since the shear parameters of the weathered sandstone, fresh sandstone and tuff are the'same, the stability of the natural slopes is a function of the steepness of the surface of the weathered sandstone.. As a result the locations of Profiles 1, 2, 3 and 3A, shown on Attachment I, were selected for stability analysis. Man-made cut and fill slopes lie primarily in sandstone and only partially in residual soil. All cut and fill slopes are three horizontal to one vertical. The stability of the slopes in residual soil were judged to be more critical than the cuts in rock and were analyzed under both static and dynamic condi-tions. The location of the residual soil and rock cuts are shown on Figure 2.5-117. Attachment I shows the location of the Profiles 5, 5A, 7, 7A and 78, judged to be the most critical slopes, and were selected for detailed stability analysis. A detailed discussion of the stratigraphy of the profiles is pre-sented in Subsection 2.5.5.2.3.4. i All factors of safety exceeded the minimum allowable values I recomended by the Corps of Engineers Manual EM 1110-2-1902, April 1970. The slope stability analyses were performed using an assumed Reactor Building and Reactor Auxiliary Building common mat load-ing selected during the PSAR stage of 8.0 ksf. However the loadinghassincebeenincreasedtoslightlylessthan14.0ksf. In order to verify the results of the stability analyses pre-viously performed additional analyses using the 14.0 ksf common mat loading are deemed necessary. l
Attachment 1-30 2266A Question No. 241.6 A preliminary review of the sensitivity of a 6.0 ksf loading (2.5.5.1) increase on the Reactor Building and Reactor Auxiliary Building (contd.) common mat indicates that the resulting factors of safety against a stability failure will not be altered significantly and will still exceed the minimum values recmnnended by the Corps of Engineers. A verification of critical cross-sections for slope stability analyses of natural as well as man-made slopes will be provided subsequent to performing the above analysis. The amended re-sponse to Question 241.6 will be provided on January 15, 1983. l l
e
.e*
s,
+W
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gsss, #3:= s 4
+3 a / f ,
ffo / ' A ,..N k WY $'* ' 3
.,__ WNP-34.*.7
^5 4**',*58 84 6 to?%.,h 16 4 .or 2*a' b 5i SNh+...24...(fi4[, e
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- T2,55
- EDGE OF PLANT ISLAND f['3 ,,,4 % .e*
m ,, u
\s \ \
)
, _s** g
\ \ 3. , -5f8 p N ... L 3as
% . N
.. h % .l" EXPT.ANATION oz_
n ~4 2 1 ) 4 Contour, elevation top of
- E 33 D f' (\ +3,
/ ** weathered 5
sandstone,in feet y "o d . terior = constructione ~ c > soi + g
%5 E @ =-a 2 ser
' 4 Boring loc'stion showing g z -f g rn 1 y elevation top of weathered -
sandstone, in feet a@ @ ( g - z o .- _ o eo zoo .oo soo soo p . yE S
- At exarnpf etion of NOTE: This contour map is interpreted ITO'n Presently availatAe geologic O E: Contour Intervos to feet construction, elevation boring data and topographic maps.
z> of plant island is +890
Attachment 1-31 2266A l Question No. 241.7 You mention that strain dependent damping coefficients used in (3.7.1.3) the deconvolution analysis are shown on Figures 2.5-87 and 2.5-88. The figure numbers are incorrect.
Response
The strain dependent damping coefficients used in the deconvolu-tion analysis are shown on Figure 2.5-122. The incorrect reference has been corrected. FSAR Subsection 3.7.1.3 will be amended to reflect the correct reference. l l l l l
- .. WNP-3
. ,10 0N-6 -
ygAg .
. G M I. 7 IhI Cooperison of the spectral values of the vertical design spectra with t the corresponding response spectra of the artificial accelero gram is made at the same period intervals used f or the horizontal component presentud in subsection 3.7.1.2.1..
b) At youmlation Level To obtain a vertical acceleration ties history to be used at the base of the rock-s tructure interaction system (khsection 3.7.2.4), a deconvolution analysis of the vertical design time history at the grade level wea perfarued as described in Appendix 3.7.A. f 3.7.1.3 Critical Despina values j i
^
e The damping ratios, expressed as percent of critical damping, which are used in the analysis of Category I systems and components, are presented in table 3.7.1-1. These damping ratios are from the recommendations by N.N. Newark, J.A. Blume, and*K.K. Espur (Reference 3.7.1-1) and are in accordance with Regulatory Guide 1.61. As indicated in Table 3.7.1-1, the two sets of damping rctios are specified to be used respectively for the ogE and SSE seismic cnalysis. 122 . The desping coefficients of foundation rock for the site are taken to be otrain level dependent as shown on Figurei 1.5 ame : " O d and are used to define desping f or the rock elements in the deconvolution analysis ( Appendix 3.7.A) and rock-etructure interaction analysis (Sabsection 3.7.2 4) . I 3.7.1.4 supportina Media for seismic Category I Structures All seismic Category I structures are supported on either fresh or weathered sandstone. The following seismic Category I structures are supported on an embedded '. common mat founded on fresh sandstone: , , c) Imactor Auxiliary Building
*b) puel Bandling Building c) Shield Building d) Internal Structure o) Steel Co'neainment Vessel l
3.7-6 . l i 1
Attachment 1-32 2266A Question No. 241.8 You have stated that the engineering properties of fresh and (3.7A.2) weathered sandstone used in the deconvolution analysis are pre-sented in Tables 2.5-7 and 2.5-8. These table numbers are in-correct. Provide the correct reference.
Response
The engineering properties of weathered and fresh sandstone used in the deconvolution analysis are presented in Tables 2.5-15 and 2.5-16, respectively. FSAR Subsection 3.7A.2 will be amended to reflect the correct reference. i
,. 16119-1 vn?-3 -
\
75A1 . N
- I APPENDIZ 3.7A DECDNVOLUTIDN OF FREE-FIELD ACCELERATION TIME HISTORIES 701 ROG-S"JLUCTURE INTERACTION ANALYSIS ,
3.7A.1 IN3t000CTIDN This section describes the deconvolution analyses performed for the finite element rockstructure interaction analysis (Sabsection 3.7.2.4) of the Satsop Site. Se purpose of these analyses is to determine the base level motion which, when input to the base of the rock column representing the rock characteristics of the site, will produce a time history at the plant grade elevation whose response spectra are virtually. identical to the design , ! response spectra defined for the free field at the grade level. i l Me procedure used in decernining the unknown base level motion is based on the conciaucus solution to the wave equation adapted for use witti transient
- motion through the Fast Fourier Transform techniques. Me nonlinearity of the shear modulus and damping is accounted for by the use of equivalent linear ,
rock properties using an iterative procedure to obtain values for shear modulus and damping compatible with effective shear strain at the middle of each layer. The design time histories at the grade level as discussed in Subsection 3.7.1.2 are used. Separate deconvolution analyses are performed for both the horizontal and vertical components of the free-field SSE and 1/2 SSE. 1he rock characteristics, mathematical model and description of the analyses are discussed below. 3.7A.2 103 ,CHARACTERISTI CS The rock on which all Category I structures will be founded is fresh sandsc'one which has a shear wave velocity ranging from 3000 f t/sec. to 4300 f t/sec.'- Weathered sandstone overlies the fresh sandstone up to the plant grade elevation. The straindependent rock properties of the fresh and weathered sandstone were obtained from laboratory tests described in Section 2.5 Figure s 2.5-121 and 2.5-122 present the dynamic shear modulus and damping ratio, respectively, as function of shear strain. The engineering properties of (8EssTand.Eveathered) sandstone used in deconvolution analyses are presented in Tables 2.5-6 and 2.5-6(The coefficient of earth pressure at rest is taken to be 0.6. 15 16 mspeduly. 3.7A.3 MATHDIATICAL MODEL
-- The mathematical model used in perforseing the deconvolution analyses of both l horisontal and vertical design time histories consists of one-dimensional i
system of homogenous, viscoelastic layers of infinite horizontal extent with l rigid base defined a t 570 f t below the plant grade elevation. Each layer is characterized by the thickness, h, mass density p elastic modulus, and critical damping ratio, S . I 3.7A-1*
Attachment 1-33 2266A Question No. 241.9 Provide the thickness of the various soil and rock layers, their (3.7A.3) assumed or measured mass densities, shear wave velocities, moduli, and damping values for the model used in your deconvolu-tion analysis. What is your basis for selecting a 570 ft depth of rock column for this model?
Response
As discussed in Subsection 2.5.4, the subsurface materials below the plant grade at EL. 390 ft involve essentially the weathered and fresh sandstone. Therefore, the deconvolution analysis model consists only of rock layers. As shown in Table 3.7A-1, the rock column, that measures 570 ft from the grade to the base, consists of five 14 ft thick layers and twenty 25 ft thick layers. An average unit weight of 128 lb/ft3 is used for all rock layers since the range of variation is small as shown in Tables 2.5-15 and 2.5-16. The final iterated values of shear modulus and damping ratio for each layer are presented in Table 3.7A-1. The shear wave velocity associated with each rock layer can be determined from the rock density and the shear modulus of each layer. The average shear wave velocities for the SSE and OBE conditions are 3,848 fps and 3,883 fps respectively. The basis for selecting a 570 ft depth of rock column for the de-convolution analysis is given in Subsection 3.7.2.4.1.
.A l _ _ _ - _
l Attachment 1-34 2266A Question No. 241.10 Describe in detail the procedure you used for calculating the (3.8.4.4) subgrade stiffness that was used in MSC/NASTRAN for the analysis of the Category I Tank Enclosure Structure, and provide the values of the geotechnical parameters for staff review. Also provide the foundation loading results and factors of safety with respect to sliding and overturning of this structure for SSE conditions and reference results to Subsection 2.5.4.
Response
The subgrade stiffness for the analysis of the Category I Tank Enclosure Structure was calculated using (1) Ks = 2.16G Rd (1- ) A where: Ks = Modulus of subgrade reaction G = 330 Ksi from Figure 2.5-121
= 0.40 from Table 2.5-16 A = Area but to a maximum of 10m2 Rd = reduction factor which varies from 5 to 20 to convert from laboratory data to field data.
Equation (1) was obtained from " Dynamics of Bases and Founda-tions" by D. D. Barker, McGraw Hill, 1962. Substituting the indicated values into Equation (1) and using the maximum reduction factor for conservatism, a value for Ks of 500 #/in3 was determined and used in the static analysis for , the Tank Enclosure Structure. l The factors of safety against sliding and overturning for the Tank Enclosure Structure and respectively, 1.6 and 1.5 (see Subsection 3.7.2.14) for SSE condition. The maximum bearing pressure (also due to SSE) is 6.0K/ft2 The bearing capac-ity of weathered sandstone o ture is founded, is 50K/ft E(nwhichthetankenclosurestruc-factor of safety is 8.3). l l l l
Attachment 1-35 2266A Question No. 241.11 The type, location, and purpose of each instrument used for (2.5.4.13) surveillance of foundations for safety-related structures should be presented.
Response
Since all seismic Category I structures were founded directly on either the fresh or weathered sandstone, settlement was not a factor in the design of the plant. Excessive hydrostatic pres-sure development will be guarded against through an in-service surveillance program providing for periodic inspections of the vertical drains, horizontal headers and drain tunnels as de-scribed in Subsection 3.4.2.2. Therefore, no instrumentation is required for surveillance of foundations for safety-related structures, l 1 l 1 I
Attachment 1-36 2266A Question No. 250.1 If applicable for the case of turbine destructive overspeed, an (3.5.1.3) analysis should be presented justifying the assumption of only one disc failure. Turbine overspeed acceleration characteris-tics, statistical distribution of destructive overspeed failure speeds, and related information should be considered in the evaluation of the probability of second wheel failure during the interval of physical disassembly caused by the first failure. Provide or reference this information.
Response
The analysis of the WNP-3 turbine for the case of destructive overspeed is presented in Westinghouse Electric Corp., Turbine Missile Report CT-24903. Subsection 3.3.2 of that report con-tains the rationale for considering only one disc failure.
Attachment 1-37 2266A Question No. 250.3 Is Subsection 5.4.2.2, Steam Generator In-Service Inspection, (5.4.2.2) intended to replace Subsection 5.4.2.2, Description, in CESSAR-F7 If this is the case, provide a description of the WNP-3 steam generator. If Subsection'5.4.5 of CESSAR-F is ap-plicable to WNP-3 this should be indicated or the appropriate information should be provided for this section.
Response
l Subsection 5.4.2.2 " Steam Generator In-Service Inspection" presented within the WNP-3 FSAR is not intended to replace Subsection 5.4.2.2 " Description" presented within CESSAR-F. In addition Subsection 5.4.5 of CESSAR-F is applicable to WNP-3. As a result of the aforementioned NRC Question (250.3), WNP-3 FSAR Subsection 5.4.2.2 will be amended.
i G A50 5 5.4.2 STEAN GENERATORS 2 Re for to CESSAR-F Subsection 5.4.2. In addition, the following components and 1 features are used in conjunction with the condensate and feedwater system I (Subsection 10.4.7) to maintain proper steam generator water quality. a) Steam and feedwater materials; subsection 10.3.6 b) Gland steam condenser Subsection 1034 , c) Condensate desinaralizar 10.4.6 d) Steam Generator Blowdown System subsection 10.4.8 5.4.2.1 Desizu nasis - Re fer to CESSAR-F Subsection 5.4.2.1.
'5.4.2.2 Stesa cenerator In-service Inspection
$lhh$!N$W,a.a . - . .A A /I AA Al AA gig A 41s @ h w u.aff5K " " '
5.4.2.2.1 In-servbe Inspection of ASME III, Class 1 and 2 Parts of the Steam Generators All Class 1 parts of the steam generators subject to examination are listed in Table 5.2-11 Section C and consist of all pressure-retaining welds, pdssure-retaining bolting, integral 1 'w'el'ded'aupports, 7 and components cladding which forms part of the primary pressure boundary. . . t All Class 2 parts of the steam generators subject to e==iastion are listed in Table 6.6 under Class 2 Pressure vessels, and consist of all pressure-retaining welds which are gross structural discontinuities, nosale,-towessel welds, integrally welded supports, and pressure retaining bolting which form part of the secondary pressure boundary. 5.4.2.2.2 In-service Inspection of Steam Generator Sabing This subsection addresses the preservice and in-service, inspections of steam
~
generator tubing as required by ASME Section XI i97T74ition with Addenda'-
) through ~Sumer 1978. The preserVice'insgegtion of steam generator tubing vjl1 be done in accordance with Regulatory Qaide 1.83 and Appendix IV of ASME Section I_I 1977 Edition with' Addenda through Summer 1978. The extent of -9 in-e.orvice examinations of steam generator tubing will be determined by the code in effect at the time of program submittal.
a) Areas Subject to F===In= tion The areas subject to examination consist of the entire length of all . the U-bend tubes in the steam generators. l
*e.
5.4-4 l l _- . . - . ._
~
* ~
- ~. 2 ^-
h&3
. FSAR i 1669W-9 i .
i indications such as dents which do not preclude probe passage, scratches, be reported.and variations in permeability will be recorded but will not
- of the tube wall thickness will be reported.All relevant indications which e All tubes' fouiid iio
~
contain defects, or tubes containing imperf ections that equal or exceed the plugging limit, will be plugged.
,,.. 'In accordance with the recousandations of Re_sulatory_ Guide .1.83, Ecv.1, the NRC will be informed if, during any inspection, it is determined that more than 10 percent of the total tubes inspected have detectable wall ~
penetrationexceed inspected greaterthe than 20 percent, plugging limit. or more than three, of the tubes f) Systes laakage and Hydrostatic Pressure Tests
- The system 1,eakage and hydrostatic pressure tests conducted during i
fabrication and preservice testing are important in determining the initial quality of the steam generator tubing. The staas generator tubing will*, throughout the service life of the staaa generator see ' i r. periodic 5.2.4.7. systen leakage and hydrostatic tests as defined in Subsection During plant operation, the secondary system will be periodically monitored,for radioactivity and the presence of boron j using off-line analysis techniques t's detect sessa generator tubing leakage.
- 3) In-service Inspection Commitment t
The preservice inspection of the steen generator tubing is conducted in
.g accordance Susaar 1978.with ASME Section II 1977 Edition with Addenda through Subsequent in-service inspections will be conducted as indicated in Table 5.2. Mis program of inwervice inspection is based on the requirements of ASME Section~ XI 1977 Edition with Summer 1978 Addenda.
Se ingervice inspection progras for the steam generator tubing will be updated as required. p h) Steam Generator hbe Flugging Limit W gg
- The Steam Generator Tube Flugging Limit will be determined by the code referenced in the ISI program. -
p 1 .) [
*- o'l In addition, the main stesa system valves and arrangement are described #
in Subsection 10.3.2. 3 j discussed in Subsection 3.9.3. Main steam line isolation system operability is y 5,{ ; .
.g g d
}
5'.4.6 f I e f I REACTOR CORE ISCLATION CDOLING SYSTEM
~
V at i u. g a Refer to CESSAR-F Subsection 5.4.6 t E ln 5
- l }us , j t ve }a, 5.'4.7 RESIDUAL HEAT RINOVAL SYSTEM .
g; , Refer this to CISSAR-F Subsections 5.4.7 through 5.4.7.1 1 and Subsection 9.2.10 of FSAR. ' g S.4.2.3 6cowwen- 1;,egien gde. h.g h k k 6- cam A-F M .s b.w.* ion s.92 3 "9d S*l 2A h s. G 4 e n w er- Mc,teeAis '
' ar.Ase e. us144-i: .3.,l secei m S.4 2 8 S.9.Q,f Tses e+4 repec6A=4 g g g.g 3%g.,, s.9.Q [
Attachment 1-38 2266A 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 quarterly to incorpo-rate information as it becomes available. The first submittal shall be issued on December 1982 with an overall completion date scheduled for June 1984. f e I
j Attachment 1-39 2266A Question No. 270.2 Figures 3.11-14 and 3.11-15 have not been submitted. Provide a (3.11.1) schedule for submitting these figures.
Response
Figures 3.11-14 and 3.11-15 will be submitted by January 1983. The FSAR will be amended to reflect the inclusion of the two figures.
r d 170 . 2- ,. l L. ' ( - l e i
. LATEM l
l l WASHINGTON PUSLIC FIGURE POWER SUPPLY SYSTEM CCWS ORY COOLING TOWER Nuclear Projects 3 & 5 ELECTRICAL EQUIPMENT ROOMS 3.11 14 FINAL SAFETY ANALYSIS REPORT l l -- - ( i _ -.
. l l .
1 e l l 1.ATI:R ' l i l
, WASHINGTON PUBLIC FIGURE POWER SUPPLY SYSTEM
' DIESEL GENERATOR FUEL OIL Nuefear Projects 3 & S STORAGE TANK ROOMS 3.11 15 i FINAL SAFETY ANALYSIS REPORT i
~ .
Attachment 1-40 2266A Question No.
. 271.1 Table 3.10-1 is not complete. Provide the missing information or (3.10) a schedule for providing it.
Response
Table 3.10-1 will be revised quarterly to incorporate informa-tion as it becomes available. First submittal shall be issued December 1982 with an overall completion date scheduled for February 1984. l l l l 3 $
Attachment 1-41 2266A Question No. 280.1 Table 9.5.1-2 provides a listing of unusually hazardous material. (9.5.1.1.6) As per Regulatory Guide 1.70, Revision 3, discuss the conditions under which these materials are to be used.
Response
As per Regulatory Guide 1.70, Revision 3, the conditions under which unusually hazardous materials are used is listed in Table 9.5.1-2. The FSAR will be revised for Table 9.5.1-2 to change the column headed " Storage Conditions For Use" to read "Condi-tions of Use".
i - FSAR l a M o.( , TABLE 9.5.1-2 Ih I ( UNUSUALLY HAZARDOUS MATERIAL CMih3, ofOse Espected - Time kaardous Approximate Fiant on kration Ma terial Amount Imeetion of Us e hee Bo flamsble liquide are used in the plant systems. .
- 2. COMBUSTI3LE LTCUIDS Diesel 011 92,500 gal. Yard Ambient constant (Storage) -
Diesel 011 1,100 DC-13 Jubient constant (Day 2nk) Diese2 011 1,100 DC-23 Anbient Constant (Day tank)
^
Diesel Oil 1,700 DC-11 mbient Constant (Engine Se ps) 0 Diesel Oil- 1,700 DC-21 Ambiest constant (Engine Susps) h rbine 25,000 T-20 Ambient Co nstant Imbe 011 - (Dirty) ,, Imbe 011 25,000 T-20 Juthient constant (Clean) Labe 011 15,000 T-21 Ambient Constant , (Nein Reservoir) Imbe 011 835 T-9.2 Am bient Cons tant (':DSCTU Pump hs)
~ ' '
I ~ 1L he Oi l F35 T-9.3 Ambient Constant (TDSCFW Pump Des) t 9.5A-42 w
y 1 TABLE 9.5.1-2 (Cazt'd) COM ICMS Gf OG
, . Espected
" tors Time Essardous Ap proximate Plant Omn ion Eharation .
Ma terial Amosse Imestion or se of Use
- 2. STit0NC GEIDIZING AGENT 5 Sodium Erpechloride 7500 gal. Yard Ambient Ttrice a
'Enaks I day 3s. CONRESSED CASES-Ft.APORELE ,
l Eydrogen 170,000 SCF Yard 100 peig Constant , hs . Cryogenic Rydrogen 76.200 SCF Yard 100 peig Constant Qas Cyltad=*s - Imv Pressure Bydrogen 7,245 SCF Tard 2,400 psig Conatant Qas Cylinders - Eish Fraseure ~ Bydrogen Supply System Bot Imb (EAR Ambiant Continuous (in bottles) EL. 402.00 and Cold lab ( Admin. EldE . -- EL. 405.00) l 3b. CDNPRESSED CASES - NON-Ft.AMMAELE Carbon Dioxide 3 Tons Yard 100 psig/ Intermittent 150F Cryogenic 1 Ritrogen 300,000 SCF Yard 700 psig/ Co nstant 120F Cryogenic Sulfur Dioxide 4 Tons Yard ' 35 psig/ Internittent 70F l Cylinders i l 9.5A-43
ucLU TA3LE 9.5 1-2 (Cont'd) "S
~pf OSe. ;
, Ekpected
- or time h aardous Apprezimate Flant Go ion Duration Macert al Amouse Imcation of Use Onysen 75,000 SCF 1 erd 150 psig Gr.neinuous nazism in procese-hbient ing one decay tank Nitrogen 1 bottle per RAB EL. 3 Ambient Monthly for (bottles) po st-accident 362.50 calibration Itrdrogen and contin-Analyser nous post- '
accident Aa. (DRROSIvt MATERIALS - ACIDS Sulfuric Acid 20,000 gal. Water Ambient In termittenc Treatment - (66 Be) kilding Sulfuric Acid 20,000 gal. Chlorination Ambient In termittent Placility ~ (66 Se) ~ Ab. CURROSIVE MATERTALS - CAUSTIC , Note: All water treatement - no threat to safety-related areas. Sodium Wydroxide 20,000 gal. Water Ambient 25termittent . (5CE Solution) Treatment Building ' hoonius Bydroxide - Tard Anbient Int ermittent (24 Solution) Rydrazine 550 gal. Turbine Ambient In t er31st ent (352 Solution Estimate Building 55 ga2. Bottles) Zone T-1 -- 9.5A-44
I , 1265b6 UNP-3 rsaa TABLZ 9.5 1-2 (Cont'd) Cou$Nicu5 ( hpected TLas . haardous Approziaste Plant en on kration Ma terial Amount Iscation, of Use Pydresine 100 sal. brhine Ambient @ntinuous (121/21 Selaties) Building Day Mak Zone T-1 Rydrasine 100 gal. hrbine Ambient Continuous (52 Solution) kilding hy hd h m T-1 Ammonium Bydroside 250 as1. brbine Ambient Continuous (12-152 Solution) Evilding Dmy Mak Zone T-1 Ammonium Rydroxide 200 gal. hrbine Ambient Continuous (12-15K Solution) Building Day Tank . Zone T-2 l Sodius Wydroxide 588 sal. Wrbine Ambient Int ermittent (SCE Solution) Building Doy Mak Zone T-2 Sedfue Rydroxide 250 gal. Reactor # blent Intem ittent (50I Solution) h:tillary Day Tank Building Zone RW-51 Sulfuric Acid 288 brbine Ambien t Int ermit tent (66 Be) Day Tank Building Zone T-2 i hirurie Acid 100 n .etor Ambient Int ermittent (66 Be) 2 Day Tanks Auxiliary hilding Zone RW-51 1 1 . ! 9.5A-45 l l
ttULA/ LuE)J FSAR TABLE 9.5.1-2 (Cont'd) C J oas
*
- 8" hpected Ste e Time Enzardous Approximat e Flant tion kretion Me teri al Assune In cation of Use Amnomine %droxide 100 gal. hrbine Achiest ' Intermittent (12-15%) Doy Taok Buildins Anzi11ary Boiler Zone AB-1 Selfuric Asad 100 gal. EVAC Miller hbient Intermittent (66 3e) Emom, Zoos , , '
T-1.2, hrbine - h11,1.,
- 5. EIFIDSIVES 03 NICRLY FLAfMAILE M&TERIALS None e
. O e t l
9.5A-46
Attachment 1-42 2266A Question No. 280.2 As per Regulatory Guide 1.70, Revision 3, include in the (9.5.1.3) evaluation of fire hazards in each zone, a discussion of the expected rate of fire development and maximum intensity, as these relate to fire detection response sensitivity and auto-matic and manual firefighting activities. Also, discuss the generation of smoke and other combustion products considering both the toxic and corrosive characteristics.
Response
The type of fire detection system provided for any given fire area is determined by the expected rate of development of the postulated fire in that area. Where a fast developing fire is postulated due to the nature of the combustibles present, a i thermal detection system is provided for prompt actuation of the i alarms (at local control panel and Main Fire Control Panel) and, when provided, either the pre-action valve, multi-cycle valve, or deluge system. Where a slow developing, smoldering fire is postulated due to the nature of the combustibles present, an ionization detection system is provided for early actuation of the alarms (at local control panel and Main Fire Control Panel) and, when provided, either the pre-action valve, multi-cycle valve, or deluge system. Provision of an automatic fire suppression system in any given fire area is determined based on fire hazard present in the area as well as the presence of safety-related equipment or cable. Fire areas considered high hazard are those in which the postu-lated fire is expected to have a rapid rate of development and/or a high maximum intensity. In area of low fire hazard, a detection system suited to the nature of the postulated fire and manual fire response equipment are provided on the basis of early detection system alarm and prompt fire brigade response, as detailed in Item 9 in the Fire Hazards Analysis. In areas of high fire hazard, an automatic suppression system, a detection system suited to the nature of tne postulated fire, and manual fire response equipment are pro-vided on the basis of quick actuation of the suppression system, early detection system alarm and prompt fire brigade response, as detailed in Items 8 and 9 in the Fire Hazards Analysis. In consideration of maintaining manual firefighting capability, the use of materials which generate smoke and other combustion prod-ucts having toxic and/or corrosive characteristics, especially halogenated plastics, is minimized in the plant design to the extent possible. Such materials are used in restricted quan-tities and only where non-combustible materials are not available. FSAR Subsection 9.5.1.3.1 will be amended to reflect the response above.
, - 163gw-31 WNP-3 PSAR kMA c) Containment and consequences of the fire within the considered fire area, and/or its effect on other fire areas.
d) Provision of properly located detectors to sense area fire or smoke
, conditions so that prompt fire control response can be made.
e) Effective see of manual fire control equipment and backup systems. f) Adequate smoke removal to permit personnel to enter the fire area, assess the fire condition, and use manual equipment.
- 3) Effects from the postulated fire on required operation of essential
' equipment in. the area. '
h) Protection of redundant systems, equipment or trains, if located in the same fire area, to maintain operability. Separation or isolation of redundant equipment. The fire hasard analysis was initiated by establishing the fire areas listed in Dhle 9 5.1-1. Dese are delineated in Figures 9.5A-1 through 9.5&-29. Boundaries for these areas are based on the nature of occupancy of the plant space, the amount and distribution of combustible materials, within the area, . and the location of safety-related systems and equipment. Plant areas important to the plant's capability for safe shutdown, such as electrical penetration area, cable spreading rooms, diesel generator areas, switchgear and battery rooms, are designated as fire areas. Other areas of the plant are considered as fire sones within fire areas. Fire areas are bounded by usils, floors, ceilings and penetration seals that - provide a minisue three-hour fire resistive rating except where they are shown to be unnecessary. Pire sones within fire areas may be bounded entirely or partially with barriers having a thredour fire resistive rating or less or any be defined by the area limits of fire suppression systems or of occupancies of different nature based upon the results of the fire hazard 4 analysis. l For each of the designated fire arose listed in Table 9.5.1-1, the fire hazard analysis, as detailed in Appendix 9.5A covers: identification, occupancy, boundaries, systems present in fire area, combustible loading, supplemental building features, fire detection, fire suppression, access and initial response, analysis of effects of postulated fire, equipment lists, assessment of postulated fire consequances, safety evaluation. De Fire Ba zard Analysis has not yet been completed. D e safery evaluation is scheduled to be complete in the fourth quarter of 1982 and Ar 2ndix 9.5A vill then be modified to describe the plants ability to achieve safe shutdown as l~~ '- well as any "other changes to design which may occur as a result of this , review. In response to the NBC staff request for information, the safety 1 evaluation will consider the criteria.of 10CyR50 Appendix 1 as well as plant design criteria. D e amendment to Appendix 9.5A will provide a discussion of compliance with plant design criteria (BTP APCS 3 9.5-1 and it s Appendix A) as I well as a comparison of plant design versus 10CyR50 Appendix R in response to the information request. l l 9.5A-31
l . ', A 1++ad I - The type of fire detection systsa provided for any given fire area is determined by the expected rate of development of the postulated fire.in that area. Where a fast developing fire is postulated due to the nature of the combustibles present, a thermal detection system is provided for prompt actuation of the alarus (at local control panel and Main Fire Control Panel) and, when provided, either the pre-action valve, multi-cycle valve, or deluge system. Where a slow developing, smoldering fire is postulated due to the nature of the combustibles present, an ionization detection system is provided for early actuation of the alarus (at local control panel and Main Fire Control Panel) and, when provided, either the pre-action valve, multi-cycle valve, or *- deluge system. t Provision of an automatic fire suppression system in any given fire area is determined based on fire hazard present in the area as well as ths presence of safety related equipment or cable. Fire areas considered high hazard are those in which the postulated fire is expected to have a rapid rate of development and/or a high maximum intensity. t . In areas of low fire hazard, a detection system suited to the nature of the postulated fire and manual fire response equipment are provided - on the basis of early detection system alare and prompt fire brigade response, as detailed in Itas 9 in the Fire Hazards Analysis. In ' l areas of high fire hazard, an automatic suppression system, a detection l system suited to the nature of the postulated fire. 'and manual fire response equipment are provided on the basis of quick actuation of the suppression system, early detection system alarm and prompt fire brigade response, as detailed in Items 8 and 9 in the Fire Hazards Analysis. In consideration of maintaining manual firefighting capability, the use of materials which generata smoke and other combustion products having toxic and/or cotrosive characteristics, especially halogenated plastics, l 1s minimized in the plant design to the extent possible. Such materials are used in restricted quantities and only where non-combustible materials are not available. 1 l l 1
Attachment 1-43 2266A Question No. 280.3 As per Regulatory Guide 1.70, Revision 3, where automatic fire (9.5.1.3) suppression systems are installed, include an evaluation of the effects of the postulated fire both with and without actuation l of the systems.
Response
The effects of a postulated fire in an area protected by an automatic fire suppression system will be limited to fire damage in the imediate area of inception with only limited propagation of the fire through the area. If the automatic fire suppression system has not actuated automatically, the postulated fire might involve cabling and equipment adjacent to the point of inception and the fire could propagate throughout the fire area. The ex-tent of damage beyond the fire area will be limited by the three hour fire rated barriers enclosing the fire area. However, even without actuation of the automatic suppression system in the area, the fire will be sensed by the fire detection system which will alarm fire condition in the Control Room. The Control Room operator will dispatch the Fire Brigade for prompt assessment of the situation and initiation of effective manual firefighting through the use of portable fire extinguishers, hose lines and/or manual actuation of the automatic fire suppression system thus reducing the potential for the fire spread. FSAR Subsection 9.5.1.3.1 will be amended to reflect the response provided above. l I
FROM 5780 '92.10.21 13803 e 1688W-31 I (, c) Containstent and consequences of the fire within the considered fire area, and/or its effect on other fire areas. d) Provision of properly located detectors to sense area fire or smoke conditions so that prompt fire control response can be made. e) Ef factive use of manual fire control equipment and backup systems. f) Adequate smoke removal to permit personnel to enter the fire erva, assess the fire condition, and use manual equipment. { F.ffects from the postulated fire on stquired operation of essential 3) equipment in the area. h) Frotection of redundant systems, equipment or traina, if located in the same fire area, to maintain operability. Seperation or isolation of radundant equipment. The fire hesard analysis was initiated by establishing the fire areas listed i n ' table 9 5 1-1. nese are delineated in Figures 9.5A-1 through 9 3A-29. Boundaries for these areas are based on the nature of occupancy of the plant space, the amount and distribution of combustible materials within the area, and the location of safety-related systems and equipment. Plant areas important to the plant's capability for safe shutdown, such as electrical penetration area, c3ble spreading rooms, diesel generator areas,
, switchgear and battery rooms, are designated as fire areas. Other areas of the plant are considered as fire zones within fire areas.
Fire areas are bounded by walls, floors, ceilings and penetration seals that - provide a miniaua three-hour f tre resistive rating except where they are shown to be unnecessary. Fire zones within fire areas may be bounded entirely or partially with barriers having a three-hour fire resistive rating or less or may be defined by the area limits of fire suppression systems or of l occupancias of different nature based upon the results of the fire hasard 9 analysis. g4 For each of the designated fire areas listed in Table 9.5.1-1, the fire hazard analysis, as detailed in Appendix 9.$A covers: identifiestion, occupancy, boundaries, systems present in fire area, combustible loadin6, supplemental i building features, fire detection, fire suppression, access and initial response, analyals of effects of postulsted fire, equipment lists, assesement of postulated fire consequences, safety evaluation. he Fire Ha zard Analysis has not yet been completed. W e safety evaluation is scheduled to be complete in the fourth quarter of 1982 and Appendix 9.5A will then be modified to describe the plants ability to achieve safe shutdown as well as any other changes to design which may occur as a result of this review. In response to the NRC s taf f request f or inf ormation, the saf ety evaluation will consider the criteria of 10CFP50 Appendix R as well as plant design criteria. De amendeent to Appendix 9.5A will provide a discussion of coc pliance with plant design criteria (BTP APCS 3 9 5-1 and its Appendix A) as well as a comparison of plant design versus 10CFPJO Appendix R in response to (. the i nf on:a tion request . 9.5A-31 l
enon 5780 '82.10.21 13:04 GMo. s y 2. Iss.A i The effects of a postulated fire in an area protected by an automatic fire suppression system will be limited te fire damage in the immediate area of inception with only limited propagation of the fire through the area. If the automatic fire suppression system has not actuated auto-matically, the postulated fire might involve cabling and equipment ad- . jacent to the point of inception and the fire could propagate throughout the fire area. The extent of damage beyond the fire area will be limited by the three hour fire rated barriers enclosing the fire area. However, even without actuation of the automatic suppression system in the area, the j fire will be sensed by the fire detection system which will alarm fire ! condition in the Control Room. The control room operator will dispatch the Fire Brigade for prompt assesament of the situation and initiation of effective manual firefighting through the use of portable fire ex-tinguishers, hose lines and/or manual actuation of the automatic fire suppression system thus reducing the potential for the fire apread. e e l l
'~' ~~~ _ _ _ _ . . . _ _ _ _ *mm- ~+ .m ,
Attachment 1-44 2266A Question No. 280.4 Many items in Appendices 9.5-1 through 9.5-21, " Fire Hazard (9.5.1.3) Analyses by Fire Areas" are marked "Later". In the proposed completion of these items, they should be evaluated against the Standard Review Plan, (NUREG-0800).
Response
The items marked "Later" in Appendices 9.5A-1 through 9.5A-21 will be completed according to the following schedule: Clarification of "Later" for cable trays: The function of cables in cable trays will not be given. Cable trains SA and SB will be protected. Accordingly, all "Later" designations for cable tray functions will be deleted. The function of cable trays will be marked "not applicable". Function of equipment now designated "Later" will be completed for inclusion into the " Fire Hazard Analyses by Fire Areas" by April 1983. These items will be reviewed and evaluated against NUREG-0800. FSAR Appendices 9.5A-1 through 9.5A-21 will be amended to reflect this. l I
17719-6 WNP-3 hS$0*Y
- 9. Access and Initial Response (Cont'd) i f3 Fire Adjacent Areas Enarest '
Zone Zone Corridors Stair Tower Elevator Other Access C-3 . C-5 Euee 2 4 hoe Via ladder C-4 C-5 Bone 1 4 Mone C-5 C-4 Bose 2 4 Mene C-6 C-4 Bonn 1 4 Bone C-7 flater Esse 1 - 4 Bona C-6 Later Bone 2 4 Eone i C -, (later J se 1, 2 4 e .- .
- 10. ' Analysis of Ef fects of Postulated Fire
- a. Combustibles - In Fire Area C, fire hazard combustibles include cabra insuistion, charnaal, labe oil and grease internal to the equipment.
- b. Control of Essards - The quantity of combustible material which may be involved in the postulated cable fire, and consequently, the magnitude of both the fire and the resultant damage to p1 ant familities, is reduced by the use of IZZZ-383 cables, by the confinement of re1 eased combustible liquids through provision of drainage of released oil to area simps and oil disposal systems.
The introduction of transient combustibles is not considered likely because the Beactor Bailding is not occupied during normal operation. However, transient combustible materials may be brought l into the fire area for maintenance and repair or during plant . shutdown. The introduction of transient combustibles resulting from normal maintenance and operation activities is controlled through administrative procedures.
- c. Damste Limitation The extent of damage within and beyond the fire zone on fire is limited by removal of hast, smoke and other products of combustion through use of the Containment Purge System, by partial structurat barriers and separation within the area and by three-hour fire rated barriers enclosing the fire area.
, 9.5A-57 - - - - - , - - . ,y.--
17719-10 WNP-3 FSAR 2.
- 10. Analysis of Ef fects of Postulated Fire (Cont'd)
Iquipment or ID Eo. and . Component Division Function Cooling Units 2CC-RS24SA @ntaiment Isolation - 2CE-VP025 NA Containment Isolation 2CR-VP0295AR Qantaiment Isolation EC-3012-SA Hydrogen 2 Purge Supply 3C-3010-SA %drogen2 hrge Supply Transmitter FT-268 Easctor Drain Tank Pressure 2hermocouple TE-221 14 actor Beat Exchanger Intdown
- Ostlec Pressure Cable Tesys P27-EA (Iater) d[4 F29-EA *
(Later) g C69-5A (Iater) ' C71-EA (Istar) ' -
- I.35-sA (Inter) y F28-MB (Intar)
C70=NE (later) 5 L34r-MB (In ter) L3M (Inter) l Fire Zone C-2. Steam Generator 1-SA Area Ptamps 1A-SE Esactor Coolant Rump 13-SM Leactor Coolant Pump Steam Generator 1-SA Steam Generation meat Exchangers 1-SA Steam Generator Elowdown 3-SA Steam Generator Blowdown Valves 131-VP0925AR Shutdown Cooling System *a olation 2FT-5010-SA Mydrogen2 krge Supply 2FT-3016-SA Hydrogen 2 krge Supply Fire Zone C-3. Steam Generator 2-SB Area haps - 2A-SN Essetor Coolant Ramp 25-55 Reactor Coolant Pump Steam Generator 2-$3 Steam Generation Heat Exchangers 2-6B Steam Generator 31owdown 4-63 Steam Generator 31owdown Valve ISI-VP0985BA Shutdown Cooling System Isolation 9.5A-59 e m,, v,m--,--- -
1771W-11 FSAR k
> 10. Analysis of if facts of Postulated Fire (Cont'd) .
Equipment or ID Mo. and Division
~
Component Function Fire Zone C-4, Containment Fan Cooler Arms Unit Cooler DC-5 2DK Unit Q>olar i Fan Cooler FC-1A (Ia .ar) FC-1C (Iater) Faas UC-5F3 (Lacar)\ DC-5FB (Iater) Tanks 3-SA/B Safety Injection 4-5A/3 Safety Injection Valves 13I-VFLD4 EAR Safety Injection Tank Isolation 151-V7107841 Safety Injection Tank Isolation 181-VF1235B1 laakage Ealief of Valve VF106EA/BE ISI-YP12teEBR 1aakage Relief of Valve VP1035A/BE ELE-dV47025 Puel Pool (C) Transmitters FT-100Z Pressuriser Pressure Osatrol FT-100T Pressuriser Pressure Control FT-101A Pressuriser Pressure Protective FT-1013 Pressuriser Pressure Protective o FT-101D Pressuriser Pressure Protective - FT-102A Pressuriser Pressure Protective FT-1023 Pressuriser Pressure Protective FT-102D Pressuriser Pressure Protective FT-103 Eastricted Bange Pressure FT-104 Restricted Range Pressure PT-106 Eastricted Range Pressure FT-199A Pressurizer Pressure (Supplementary Systes) 9.5A-60
WNP-3 1771W-12 g Y
- 10. Analysis of Ef fects of Fostulated Fire (Cone'd)
Equipment or ID Me. and . Component Division Function FT-1995 Pressuriser Pressure (Supplementary Protection System) FT-199D Pressuriser Pressure (Suppiamentary ~ Protection System) LT-1101 Pressuriser level control LT-110T Pressuriser Level Control PT-1013A Steas Generator Pressure Protective . PT-10135 Stama Generator Prossure Protective FT-1013C - Steam Generator Pressure Protective FT-1013D ' Steam Generator Pressure Protective , i LT-1113A Sessa Generator Iavel Protective (Wide Range) l LT-11133 Steam Generator level Protective (Wide Range)
~ '
LT-1113C Steam Generator Iavel Protective (Wide ' Range) LT-1113D Steam Generator Iavel Protective (Wide
- Range)
LT-1114A Steam Generator level Protective LT-11143 Steam Generator Level Protective LT-1114C Steam Generator level Protective LT-1114D Steam Generator Level Protective cable Trays P31-SA (later) '" C73-SA (u ter) C75-SA (uta r) 8 L39-SA (Ia ter) . L41-SA (Laear) F30-SB (u ter) '. C72-$3 (later) C74-5B (La ter) '. L42-5B (Iater) '
- L44-53 (Laear 9.5A-61
__m ___
WRP-3 17719-13 PSAR T p 10. Analysis of Ef facts of Postulated Fire (Cong'd) ! 1 _. Equipment or ID No. and , Cosponent Division Faneeion - 1
~
Pire Zone C-5, containment Fan Cooler Area . Cost. Paa Choler PC-13 (later) Tanks 1-SA/B Safety Injection 1 2-SA/B Safety Injection Yalves ISI-VP110S EE Safety Injection Tank Isolation 1SI-VP113531 Safety Injection Tank Isolation i 151-7P121531 Leakage Balief of Yalve VP112SA/31
- 15I-V712231 TAshage Belief of Valve VP1095A/$3 1CH-VPm0SAR containment Isolation valve ICE-VP001331 Pressure Qmerol Valve Auxiliary 3 pray to Prasourizar 1CH-VF039 EAR Isolation ,
Velve ICE-VP009531 Safeey Injection Isolatioa Valve Transmittere FT-1023& Steam Generator Pressure Protective FT-10233 Steam Generator Pressure Protective FT-1023C Steam Generator Pressure Protective ' FT-1023D Steam Generator Pressure Protective LT-1124h Stesa Generator level Protective LT-11243 Stesa Generator favel Protective LT-1124C Steam Generator Level Protective LT-1124D Steam Generator Imvel Protective Differeatial Presoure PDT-240 Charging Baek Preasura Yalve cehie Traye P31-SA (Tater) W* C73-SA (14ter) C7 5-SA (facer) * . , L3 9-SA (Ister) ), i L41-SA (Iaeer)
- P30-63 (later)
C72-63 (Istar) }' C74-63 (Inter) . 142 43 (Tater) . L44-63 (fatar) J_ Fire Zone C-6, Pressurizer Boom Pressurizar SA/S3 h*1msey 7.oop Pressurization Volves 3CC-E0068N , Steam Generator Slowdown Best Exchanger 4
- Isolation Valve 3CC-3008SN Steas Generator Slowdown Beat Isolation Valve 9.5A-62 l
l
WNF-3 17 719-14 FSAR
, , I 4
- 10. Analysis of Ef facts of Postulated Fire (Lane'd)
Equipment or ID Es. and . Comoonent Division Function Fire Zone C-6, Pressuriser Enos (Cont'd) - 3CC-500955 Steam Generator Riowdown Beat Exchanger 2 Iselation Yelve 3CC-5010SN Steam Cenerator Elowdown Beat Exchanger 2 Isolation Valve 3CC-5012SN Steam Generator 11oudown Esat Enchanger 1
. Isolation Yalves 3CC-3015SN Steam Generator Elowdown Beat Exchanger 1 ,
Isolation Valves 3CC-3018SN Stesa Generator Blowdown Beat Exchanger 3 Isolation Valves 3CC-BC19BE Steam Generator Elowdown Beat Exchanger 3 Isolation Valves Fire Zone C-7, Head Assembly laydown Area and Upper Invel Valves 1EC-P5013A/RE Pressuriser Spay Line Isolation 11C-PS02SA/31 Pressuriser Spray Line Isolation 2FV-5010SA %drogen2 Purge Ezbaust l
'Dransmitter s FT-101C Pressuriser Pressure Protective FT-102C Pressuriser Pressure Protective FT-105 Bestricted Bange Pressure ~
FT-199C ~ Pressuriser Pressure (Supplementary Protection System) Boisting Winch (Iater) Hydraulic Power Pack (later) Fire Zone C-8 Upper Imvels Taaka 1-SA/B Safety Injoction 2-SA/B Safety Injection Faa Cooler 7C-1D (la ter) l Bydrogen Escombiner 53 (later) l Valves 2PV-B1095A Hydrogen Purge Exhaust 2PV-5064SA Containment Purge Exhaust 2FV-B112SA Containment Purge Exhaust . 9.5A-63 l l
r WNP-3 17715-15 FSAR P 10. Analysis of Ef facts of Fostulated Fire (Cont !d) Equipeest or ID No. and Component Division Function *
~
Fire Zone C-9. Enactor Pressure vessel Ares
- Esector Pressure Ve seal SA/SB Con *=i -ant Feas E-la Beactor Cavity Cooling Sapply Fane I-15 Emactor Cavity Cooling Supply Lifa Eig
- Beactor Drsin Tank Core Support Barrel Storage Stand (later)..
Yalve III-EV47 013 h1 bl Transmitters II-IV-670LAS Fuel Fool IE-EV-6702&S bl Boel IE-EY4701BS . Fuel Fool IE-EV-670 35 h1 Phol
- 11. Assessment of Postulated Fire Consequences ,
The above listed safety-related and non-safety equipment in Fire Area C may be damaged or lose function due to a fire. The fire in Zone C-5 is limited to that none because of the slow burning nature of the cable and the non-self propagating nature of IEEE-383 cable insulation. Eedundan;: equipment in Fire Zone C-4 is beyond the influence of the postulated fire and costianes to remain available. The postulated fire and its associated effects in Fire Zone C-2 is limited to the immediate vicinity of the arposed reactor coolant pump by the limited caount of fuel, the reinforced concreta cubicle walls and pedestal design. The other pump 1B(SN) ap;roximately 30 f t away and Steam Generator 15A, in Fire Zone C-2, and the redundant equipment in Fire Zone C-3 are beyond the influence of the postulated fire and continue to remain available.
- 12. Saf ety Evaluation The safety evaluation will be presented later as discussed in Subsection
( 9.5.1.3.1. Additional changes to design will be incorporated into the FRA at that time. l t 9.5A-64 8 9
Dh51) { 1773U-6 FSAR, l 7
- 10. Analysis of Ef fects of Postulated Fire (Cont'd) e.
Fire Effects - The fire is contained within the three-hour fire barriers ~ of Fire Area CSA. The ' redundant Fire Area CSB is enclosed and separated from CSA by three-hour fire resistive boundaries and is ' therm. fore not , esposed by the fire. f. Ecruipment in Area - Safety related and non safety-related equipment s're listed below. l Equipment or ID No. and Component . Division Function Fim Ama CSA. Cable Vault A ~ Damper D7A-SA Battery Boom & Switchgear Ventilation Cable Trays C39-SA CA1-SA (Lacer) (Later) h I' C43-SA Latar) , CAS-SA f ((Later) C37-SA f (Late r) ' i CA7-SA (Later) - C49-SA ; (Later) C51-SA (Later) f C53-SA (Later) L21-SA (Laeer) / ; L23-SA (Later) . L2!>SA (Later) C25-SC (Later) C27-SC (Later) i C77-SC { (Later) C5 9-NA (Later) l C61-NA (Later) C15-NA (Later) C25-NA (Later) l C6-NA (Later) C17-NA (Later) C5 7-NA { (Later) { C5-NA \ (Later)
. C19-NA (Later) f C21-NA (Later) (
C63-NA (Later) L3-NA (Lat er) L11-NA (Later) / L13-NA (Laeer) l' L15-NA (Later) L17-NA (Lat e r)
. ~
9.5A-77 a . _ _ - . _
WNP-3 1773W-7 F5AR 2 .
- 10. Analysis of Ef fects of Postulated Fire (Cont'd)
Equipment or ID No. and. Cosmoment Division Function Fire Area CSA. Cable Vault A (Cont'd) L33-WA (Later) b[d C1bB (Late r) ' CIS-MB (htar) C20-NB 1 (Later) l C22-NB l (Later) , C6-NB (Lata r) ' C28-NB (Latar) *
~
I C30-NB (Lasar) C32-MB (Lata r) ( C34-B (Latar) - L12-NB (hear) ' (Late r) I Li& n L16-5 (Late r) ~ L2FM (Later) I Fire Area CE!. Cable Vault B Damper D-75-85 Battery Boca & Switchgear Ventilation Cable Traye C36-55 (Later) 4/A C38-SB (Later) ,
- C40-SB (Later) .'
C42-5B (Lasar) CA4-83 (hter) C46-SB (Late r) / . C48-SE (Late r) C50-S B (Late r) / L26-65 (Late r)
~
L28-5B (Lacer) f CS 2-SD (Later) C78-5D (Lata r) L22-SD (Laser) { C25-NA (Leta r) f L33-NA (Lacer) l L3-MA (Later) L13-NA (Late r) L17-NA (Laee r) LIS-MA (Late r) < L1-NA (Late r) - CIS-NB (Laear)
. C20-NB (Lat e r)
C28-NB _( Lat e r) 9.5A-78 s -e e
'e---- --_s _ . _ y -__v___
WNP-3 1773W-8 FSAR IO
- 10. Analysis of Ef fects of Postulated Fire (Cont'd)
Equipment or ID No. and ^ Component Divi sion - - Function
~
Fire Area CSB. Cable vault B (Cont'd) C30-NB (Letar) C32-NB (Later) ' i C34-NB (Later) . C23-NB (Later) ( C14-NB (Later) L14-NB (Later) ,', t L16-NB (Later) . L2-NB (Later) j Lk-NB (Later) g L20-NB (Later) L12-NB (Later) ( L33-NB (Later) . L3-NB (Later) L13-NB (Laeer) (Later) L17-NB L15-NB (Later) ( L30-NB (Later) I L32-NB r)
- 11. Assessment of Postulated Fire Consequences N above listed safety-related and non safety equi' pment in the fits areas may be damaged or loose function due to a fire.
N fire in Fire Area CSA is limited to a small length of SA cable trays which ! cre damaged or lose function due to fire. N redundant Fire Area C5B contains the SB and SD cable trays and is beyond the influence of the fire so that it continues to remain available as required.
- 12. Safeev Evaluation N safety evaluation will be presented later as discussed in Subsection 9.3 1.3 1. Additional changes to design will be incorporated into the FHA at that time. .
1 l 9.5A-79
- - , - - -- - - - - - , ------,w-w--
WNP-3 1774W-10 FSAR ll 10 . Analysis of Effects of Postulated Fires (Cont'd)
-- ID No. and Eeuionant or Component ,
Division Function DG Intercooler Engine Driven Pump A-SA Supplies Cooling Water DG Air Intake Silencer A-SA DC Air Intake
,DG Cranlacsee Yacuum Fan A-SA DG Operation DG Crankcase Yacuum Oil Separator A-SA DG Operation o
DG Jacket Water Esat Exchanger A-SA ,DC Cooling Systen - DG Intercooler Best Eachanger A-SA DG Cooling System DG Lube Oil Esat Exchanger A-SA DC Lube Oil Cooling i DG Air Istake Filters A-SA DG Operation Esciter Enclosure A-SA Protective aleutral Ground Transfomer A-SA Electrical Islance Fire Zone DC-12. Electrical Eeuineent Room DG Control Panel A-SA (later) l 480Y scc A323-SA (late Cable Tray F9-SA f (later P d/A Cable Tray C53-SA (later) l MA i Cable Tray L21-SA (later) N/g 1 i Fire Zone DC-13. Diesel Day Tank Boom A-SA Diesel Day Tank A-SA DG Set Fuel Fire Are a DGB. Diesel Generator B-SB Fire Zone DC-21. Diesel Generator Room B-SB Diesel Generater B-55 Required for emergency power during accident l l DG Jacket Water Expansion Tank 3-53 DG Cooling DG Lube Oil Make up Tank A/B B-SB DC Set Lubrication i Starting and Control Air Tank B-SB DG Startup DG Engine Driven Fuel Oil Pump B-SB DC Oil Supply ? l .. t 9.5A-89 l [ ._ . . l _ _ . -. _
WNF-3 1774W-11 FSAR l2 10 . Analysis of Ef fects of Postulated Fires (Cont'd) ID No. and Eeulosent or Component Division Function ' DG Jacket Water Engine Driven Pump B-SB Water to Coolers DG Intercooler Silencer B-SE DG Air Intake DC Crankcase Vacum Fan B-SB DG Operation DG Crankcase Vacuum oil Separator B-SE DG operation DG Jacket Water Seat Exchanger B-SB
~
DG Cooling System DC Intercooler Seat Exchanger B-SB DG Cooling System DC Lube Oil Beat Exchanger B-SB DG Lube Oil Cooling DG Air Intaka Filters B-SB DC Operation Thermocouple - TE-BY-4921313 DC Boom Temperature Themosouple TE-HV-4921325 DC Room Temperature Thermocouple TE-BY-4921333 DG Roan Temperature The mocouple TE-BY-4921348 DG Room Temperature Valm 3CC-5540SB DG Room Isolation Vent Line 3WG-VE006SA DG A Intercooler Vent Line to Expansion Tank A Cable Tumy F10-SB (later) M/A I Cable Tray C42-SB (later) Cable Tray C42-SB (later) l Cable Tray L26-SB later) Fire Zone DG-22. Electrical Equipment Room DC Control Panel B-SB (lat e r) 480V MCC B-3 23-SB l Cable Tray P10-SB ,fA [(later)} I Cable Tray ,C42-SB (later) Cable Trey L26-SB acer) 9.5A-90
17759-9 WNP-3 FSAR l3
- 10. Analysis of Effacts of Postulated Fire (Cont.'d)
- Equipment or ID Es. and
- Component Divisi'on --
Function Unit coalers DC-1A-SA EFSI Pump Boom A. UC-2&-SA Containment Spray Pump Boom A UC-3&-SA LFSI Pump Boom A i UC-6A-EA w 14= y Feedwater Pump 1A . 2A ) l Eoom .. Valves 3AF-F023SA Motor Driven Auxiliary Paed Pump A l Discharge 3&F-VD0993A wit tary Feedwater System Isolation , Valve . 3&F-VD1015A Ansziliary Peeduster System Isolation *f Valve 3CC-VE535SA CS Pisap A Coolar Isometric Valve Control Valve FCV-AF-4321AS Motor Driven Auxiliary Feed Puay A Fire Zone EC-3 Mechanical Penetration Are's A Unit Cooler UC-7A-SA Mechanical Puestration Area Baergency Core Cooling System. Area Tank Z-SA/B Spray Chemical Storage Tank Pump 1 Spray hical Pump f Valves 2CC-RS2583 containment Isolation I 2CC-552153 containment Isolation 3CC-55075A Safety Injection Isolation 3CC-5508SA Safety Injection Isolation - Costrol Valve TCV-HV-4957AS Mechanical Penetration Area SA Tamperature Transmitter TE-HV-4957AS Mechanical Penetration Area 5A Temperature Cahlo Trays C43-SA hLacar) A L21-SA I (Lacar) ' P13-SA QLater) Fire Area EC3 Essential Cooline Ecuipment B Fire Zone EC-4 Shutdown Cooling Heat Exchanger B Beat Exchanger 3-33 Shutdown Cooling Unit Cooler UC-45-53 Shutdown Cooling Beat Exchanger 3 valves 3CC-B51253 Shutdown Hest Exchanger B Isolation valve 3CC-VE536SB Containment Spray Pump 3 Cooler Isometric Valve P' 9.5A-100 l -
1775U-u WNP-3 ysag . 14
- 10. Analysis of Effects of Postulated Fire (Cont'd)
- Equipment or ID No. and -
Component Diviaion -- PuAction Piro Zona EC-6 Mechanical Penetration Area B valves 3CC-50673N Men Nuclear Safety Enader Esturn Isolation 3CC-350953 Safety Irfection Isolation 3CC-B31033 Safety Injection Iso'Istion 2C3-VQ005SNE Seal Injection line Isolation outside Contaisneet , Unit Cooler UC-73-6B Mechazical Penetration Area Emergency Core 914 == System Area
, 2CE-VQ040$ME Charging Line Isolation ontsids Containment , 2CE-VPC11SEE Cone =ta= ant Isolation 2CE-VP0308BE Containment Isolation
- 2EC-300953 Containment Ventilation Eschange 2EC-B911SB Bydrogen Purge Supply Ibermocouples TE-EV-223 Emactor Beat hehanger Imtdown Outlet Temperature TE-EV-224 Ion Exchanger and Boronometer Inlet Temperature Control Valve TCV-EV-4957BS Mechanical Penetration Area 5B Temperature Transmitters TE-EV-4957BS Mechanical Penetration Area 53 Temperature FIS-CC-7241 Imtdown Heat hehanger Component Cooling Water Flow Cable Traya F13-SA N/A 32&SB Imeerh (14 car) i. -
327-5B (Later) C42-SB (Later) \ L2 bSB (Later) 1 ( P14-SB (Later)j CS2-NB \ ((Later)J
- 11. Assessment of Postulated Fire Consequences Ihe above listed safety-related and non-safety related equipment any be i damaged or lose function due to a fire and smoka. The postulated fire in Zone
! EC-3 will be evaluated later during the safety evaluation to' determine the l cffect on redundant equipment and safe shutdown capability. H-l 9.5A-102 _y,w , - - , , -- _ . - _ . -
WNF-3 177 9-16 FSAR g
- 10. Asalyeis of Ef feeea of Foetulated Fire (Co'nt'd) components within a single train encept as indicated in Section 3, and '
by thredour rated fire resistive barriers enclosing the fire area
, except the boundary between D-11 and D-21. -
- d. Fire Postulation - A fire is postulated at the boundary of Fire Zone :
D-11 from an exposure fire on the floor adjacent to the CEDM panels. i The fire could achieve a naziana development that directly exposes cable insulation inside the panels and non-eafety-eelsted cable trays directly over the panels. The cable fire could spread several feet in both directione along the trays and reach SA and SB cable trays whi.ch j enter Fire Area 13 from Zones D-11 and D-21. e
- e. Fire Effects - All fire sones in Fire Areas UA and EF5 are enclosed
! within three-hour fire resistine boundaries with the exception of Fire 2soes D-11 and D-21, and are not exposed by the fire. malf a limited portion of the low peramment combustibles in D-11 and D-21 is involved in the fire.
- Damage by fire to other safety-related equipment such as switchgar would be limited to one train because of the more than 50 foot separation between A and B trains in D-11 and U-21 and by enclosure -
of redundant trains in three-hour rated fire resistive walls in the other D fire semes.
- f. Equipment in Area - Safety-related equipment and non-nuclear safety-related equipment are listed.below:
Equipment or ID Ns. and Component Division Ftanetion Fire Area IPA. Electrical Penetration Area A Fire Zone U-12, Battery Boom SA Battery Encks SA Fire Demper D13A 4A Battery Boos ir Switchgear Ventilation Fire Zone D-13, Enttery Esom SC Bettery Eacka SC !ater l l l 9.5A-117
, _ _.- - - _ , . , - - - - _ . _ . _ , _ - , . - . . _ - _ , -__ _ - - - - - m- , _ . _ , _ - - - . . . - , _ - , . - - - _ _ _ _ . , - _ , _ _ . - . , _ . - - . - . -- , - _ . - - - - - - - - - - - - - - , , - - - - - -
WNP-3 1778-15 FSAR
-/6
- 10. Analysis of Effects of Featulated Fire (Cant'd)
EquiFeent or ID No. and ' Component Division . - Ehnetion Fire Zone D-14 Motor Generator A (Ia tar) Set Fire Zone U-15, Comunications Ecom Communications Fanels (1atar) *
, 480V Motor 2 n- Atll-SA (Inter) taol Center I A313-SA 123V DC Fanels A-6& Emergency Diesel Generator A 5&-SA Steen Turbine Auxiliary Feed Pump A
. A 4A g Eeoombiner ,
13.8 kV Switch- EF-SA Reactor Trip Saar 48V Fover A-11 (la ter) Cabinet Fire Zone D-16 Electrical Fenetration Area A . - Cable Traye L21-SA
- L23-SA (Ia tarh (Imeer) -
C23-SC ' (Ia ear) L25-SC (Iacer) 1 F55-Ma F57-#A (la ter) (14eer) d l 759-MA (Ister) Fire Zone D-21, Electrical Penetration Area A . Fane l, D Control Element Drive Mechaniss D Control Element Drive Mechanism Auxiliary Battery Quarger Al-SA (lat er) A2-SA (la eer) MUI 2 (Iatar) e 9.5A-118 e- , , , , , - - - ,w-,m -m , ,- m -_- - - - - - - - - . , - - - - - - - w -----mr --
WNP-3
, ,1775-16 FSAR
- 10. Analysis of Ef facts of Postulated Fire (Cont'd)
IquiFuese or ID No. and - Component Division -- Function Badiation Monitor ' E!-BY-67 01AS Airborne Effluent, hel Fool (C) IT-BY-6702AS Airborne Ef fluent. Fuel Poel (C) l IVAC Dampers D-185-SB Eattery Boos & Sritchgear Ventilation Cable Treys F13-SA (In ter) F15-SA (Iater) F17-SA (Ia ter) l' ! C51-SA (Inter) . j
. C53-SA (Ia ter) I '
C&3-SA (later) L21-SA (Is ter) g F9-SA (Iater) / Cal-SA (Iater) ' CA5-SA (Iacer) L23-6A (Is ter) C67-SA (Later) / CAS-SA (la ter) { C574A (Iater) C59- R . (In ter) I C61-WA (Inter) C63- h (Iater) l
, L13-W A (later) ,
L15-NA (la ter) f L17-NA (Iater) F19-NA (Iater) F 21-WA (!ater) 723-NA (la ter) F5 5-NA (Iat er) 757-E (La ter) F594A (Iater) F56-NB (la ter) ( PSS-NB (later) F60-NB (In ter) C34-NB (Iat ar) L30-NB In ter) Fire Zone EP-26, Access Area Radiation Monitor EE-HV5040A (Iacer) RE-HV5041A (la car) . ar 9.5A-119 m-m
- - - - ---w r ---'--
WNP-3 17769-17 PSAR la P
- 10. Analysis of Ef f acts of Postulated (Cont'd) 1 Equipment or ID Es. and .
I component Division ., h uction Csble Treya 79-SA - C33-4A (la ter) ,, L21-SA (Iater) , P21- R (Ia ter) i C6 M A (Iater) g L15-R (la ear) ' P2 M A (Iater) g C61-MA (Ia ter) Pire Zone EP-27, Electrical hope
- Sample Encycle Tonk later)
$seple Eacycle Pumps -
Fire Zone IP-28, he Iaboratory and Sample Boone Primary Sosple Penels A, 3 [(Inter) Secondary Sample Panels A, 3 ; (la ter) Endiation Monitor (Iater) Chiller Unit QInter) Tire Area EF3. Electrical Area Penetration 3 Fire Zone EP-11 440V Motor Con- 3111 4 3 (later) trol Center - 3313-53 L(Is ter)J 125V DC Panels 345 kergency Diesel Generator 3
. , 53-53 Steam hrbine Auxiliary Feed Pump 3 343 H2 Escombiner 13.8 kV Switch- EP-53 Esactor Trip gear 483V Pover 3-11 (Ia ter)
Cabinet Panels D Control Element Drive Mechanisa Omtrol Element Drive Meehanism 9.5A-120
\
WNF-3 17769-18 "
/1
- 10. Amelysis of Ef fects of Po stulated Fire (Cons'.)
Equipuest or ID No. and - Component Division . - h ection Bettery OnezBer B1-63 (Iater 32-$B (Ia ter) 15!1 3 Q1 ster FFE 1211 Power Source f or fotwrd, reurse speed Radiation B"-5747 0135 Airborne Effluent, hel Fool (c) l Monitor IT-NV-6702BS Airborne Ef fluent, . Fuel Fool (c) . EVAC Despera D-18L-SA Bettery Boca Switchgear Ventilation Cable Treys L28-SB khter) N[$ C3645 (Iater) C3845 (Iater) . C4 A-6B (later) . C4&SB (la ter) l L2 6-65 (later) CAS-S E (Iater) / C5043 (later) / F18-SB (Iater) f P16-65 (Ianer) C42-S E (Iater) j F14-6 B (later)
- F10-S B (Leter) -
' V443 ~(Inter) ~
F55-E (Is ter) - 25 7-N A (later) _s f-F59-R (Iater) - C63-N A (later) L13- R (Iater) . F 61-NA (Inter) F63-R (Is ter) ' F56 NB (later) F58-NB - (la ter) F60-NB' (Iater) C34-NB (Iater) C30-NB (Iater) F62-NB (leter) } F64-NB (Iacer) F24-NB (Ia ter) L32-NB (Iatar) F22-NB (lacar) '
~
9.5A-121
WNF-3 177 65-19 FSAR SO p 10. Aselysis of Ef fects of Postwisted Fire (Cont'd) . Equipeest er ID No. eed Component Division .. Function r2*Ns filat.rP PIA C32-NB (Iater) ' L20=NB (Iater) 1 C28-B Ister)J $ Fire Zone D-22. Bettery neos 53 Enttery Becke SB (later) Fire Zone U-23, leetery loos SD - l Bettery Rocks SD tar Fire Zees D-24. Motor Generator Set 4 (botrol heel hos Motor Generator 3 (Inter) Set
- Fire Zone U-25, Accese Area 4807 Power C . 323 (In ter)
Eadiation RI-57504(3 (later Noaitor 1E-5750415 ,(Iater) Qable Trays Y4-4 5 (TLeted N[A F10-SE (later) g C4 8-6 B (lat er) , C42-68 (Iater) i 126-45 (later) F22-NB (Iater) i P24-NB (Ist er) ( C32-NB (la ter) I C34-NL t (later) L32-NB ( QIaeer) l Fire Zone U-29 Elaetrieel Penetration Area 3 Cable Traye L2 6-63 (later) M L2S-SB (14 ter) i L2 2-6 D (14t er) i C24-5 D 3 (Ia ter) . F56-NB (lat er) F58-NB (Ister) l l 9.5A-122
17879-13 tuEB ' FSAR RI ?
- 10. Analysis of Ef fects of Festulated Fire (Cost'd)
Lysipeest or ID No. and
- Coopoment Divia f de Funetion The mocouples 7E-57-4946&13 Control A Qutdoor Temperature
, TI-EV-4946518 Control A outdoor Temperature i
TE-EV-4947AIS Control Boom Supply Air A Eigh ' Temperature TE-ET-4987&S Control Room Set Air A Temperature Temp control valve TCV-EV-4987AS Control Room Air A Temperature Control Cable Troya F52-SA (LaterM [A ' C115-SA (Later) 8 144-SA (Later) A., ' L J . l Fire Zone EVA-2. Control Room Fans A-SA
- Estura Air Faa E-37A Control Esom Ventilation Eshaust Air Faa. E-6A Control Room Ventilation valves 37V-5060-SA Control Room Isolation 3FY-B160-SA Control Room Isolation j 3FY-3034-SA Control Eoom Isolation 3FY-8134-SA Control Room Isolation .
Radiation Monitors EE-hV-4913A Plant Yeat 3 Particulate IE-EV-49144 Flant vent 3 Noble Despers D-84-SA Control Room Vent. Filt. and Cooling D-864-SA Control Room Vent. Filt. and Cooling
- D-%-SA Control Roan Vent. Filt. and Cooling Cable Trays F52-SA kLaterb N/A C115-SA (Later) A ,
14A-SA , Fire Zone EVA-3. Chillers and Pump Room A-SA Chilled Water Pump F1A-SA I I Water Chiller WC-1A(SA) Transmitter FDT-EC-4905AS Essential Services Chilled Water Chiller A Differential Pressure Pressure control Va've FCV-EC-4905AS 9.5A-141 l
WNP-3 1787W-16 FSAR i U
- 10. Analysis of Ef facts of Fostulated Fire (Cont'd)
Equipment or ID No.'and Cesoonent Division Function Fire Zone EVA-4. Switchaear AC Boom A-SA Air Conditioning Unit AC-3A* SA Esttery Enom and Switchgear Ventilation Air Cleanup Unit m-5A Energency Core Cooling System Area Exhaust . Espansion Tank hak A Milled Unter Erpension
. Exhaust Fans CUn-59-53 Emergency Core Cooling System Area Ishaust E-Sk-SA Esttery Boos and Switchgear Ventilation 9
l Estura Esa 1-E -SA Esttery Boos and Switchgaar Ventilation Valvea 2FV4Q29-SA ECC5 Area Ezhause 2FT-5134-55 ICC3 Area E=h==wt l 2FT-315 5-SE ECCS Area Exhaust 3FT-513 3-53 control Roon Isolation 3FT-5161-63 (batrol Eoom Isolation l l Eadiation Montiors RE-HV-4986-A15 Control Roon Air Intake la E!-HV-4986- A15 Control Boom Air Intake la II-IV-4986-315 Control Roon Air intaka 13 E!-ET-498 6-315 Control Boom Air Intake 13 Temp. Control Valve TCV-HV-4912AS Electrical and Battery Ecom A Escirculation %1ve 14 vel Transmitter LT-EC-4907A15 Essential Services milled Heter Chiller A !avel j l Despers D-3-SA Battery Boca and Switchgear Ve atilaeion l D=2A-SA Battery Boca and Switchgear Ventilation D-3A-5A Battery Boom and Switchgear Ventilation D-7 9&-SA Battery Boca and SwitcMeer l Ventilation l 3attery Boos and Switchgear i D-8 0A-SA Ventilation D-83A-SA Battery Boom and SwitcMeer I , Ventilation l l Cable Trays P33-SA klater) d/A , l CL16-5A (La ear) ' L45-5A QI4cer)} 9.5A-142 l 1 .
- w
1787W-13 F5AR b) I 1 m 10. Analysis of Ef fects of Poe'tulated Fire (Cont'd) l Equipment or ID Bo. and Comecoent . Division Function 1 Fire Zone EVA-5. Relav Room A . Centrol Room Relay A (Later) Equipa m Fire Area 375. Control Raam IVAC Area B Fire Zoos 375-1. Control Room AC and Cle===== Units B-55 . Air Conditicolag Unit AC-23-SE Control Room hatilation l Air Cleasey Unit CD-1B Filtration and Cooling i Thansosogles TI-EV-4986A28 Control Room B outdoor Temperature T5-67-4986523 Control Room B Outdoor Temperature T5-5V-4987513 Control Room Supply Mr B Eigh Temperature T4-5V-498738 Centrol Doom Set Air B Temperature Temperature Contral TCV-Ev-4987BS Control Room Air B Temperature " Valve Control Valves 3F9-8163-SE Control Room Isolation.
. 3F7-8136-55 Control Room Isolation 3CC-700535 Essential Cooling Water Chiller B Isolation valve I,
Despers D-825-SB Control Roon Vent Filtration { l D-843-SB Control Room hat Filtration D-855-SB Control Rose hat Filtration D-875-SB Control Room hat Filtration D-885-6B Control Room hat Filtration D-835-55 Control Roos Vest Filtration D-1883-SB Contral Room hat Filtration Cable Trays PSI-SB (Later) C114-SS , (Later) N/P - L43-SB (Later) - i Fire Zone dVB-2 Cont rol Roos Fans 5-6B asturu Air Fan R-375 Control Room Ventilation Bahaust Air Fan ., E-6B Filtration and Cooling Velns 3-PYB035-SA Control Room Isolation Valve 3-PTB135-SB Control Room Isolation Valve 3-PVB062-SA Control Roos Isolation Valve 3-775162-53 Control Roon Isolation Valve 9.5A-143 g - - , - , , - - - -, , , , - - - - - - - - , . , - - - - - - - - -
17 879-16 FSAR
- 10. Jealysis of Ef f acts of Postuisted Fire (Cont'd)
Equipuest or ID Me. and Componeet Division Function , i Endiation Monitora 31-E7-49133 Fiant hatilation 2 Btreiculate BI-ET-49143 Flant Ventilation 2 Noble Gas Dempers' D-83-6 3 Oratrol Been %ntilation D-6 S-S E Control Room Vestilation D-93-63 Control Boon Ventilation Cable P5143 (Iater) N/4 . L43-63 (la ter) Fire Zone IVB-3, Giller end Bump Boom 3-43 l Chilled Heter Ptamp F-13 43 hinter) l Water Chiller HC-13-(53) ((la ter) [ ' Treassitter IDT-EC-490 55 Issential Service QLilled Water ' QLiller 3 Differential Pressure Press. Control Valve FCV-EC-490 SS l Damper D-173-SE Battery Roon and Switchgear
%stilation Fire Zone E73-4. Switchgear AC Boom E-63 ,
Air Conditioning Dait AC-13-63 Battery Boos and Switchgear Vestilation Air Cleanup Unit CD-53 Energency Core Cooling System Area
- Enhaust Erpension Tank 1hak 3 Otilled Water Erpension Escurn Fan CUR-5A-SA Emergency Core Cooling System Area Exhanst I-93-53 Battery Boom and Seitchgear Veatilation Valves 2FV-305 4-SA Energency Core Cooling System Area Ethaust 2FV-305 5-SA Energency Core Cooling System Area Exhaust 2FT-5129-43 Emer5ency Core Cooling System Area hhaum g I
3FV-306 3-SA Control Roon Isolation l . . 3FV-5036-SA Control Roon Isolation Eadiation Monitors EZ-HV-498 6A2S Control Room Air Intake 2A l IT-HV-4986A2S Control Roon Air Intake 2A , EE-HV-4986525 Control Roca Air Intake 23 ,. IT-HV-4986325 Control Room Air Intake 23 9.5A-144 l
WNP-3 178N-17 FSAR g
? .
- 10. Analysis of Ef facts of Postulated Fire (Cont'd)
Equi,mest or ID'io. and Division
. Comoonent _
punction - 1has, matrol valve TCT-E74911BS Electrical and Battery Boom B Bacirculation Blve Iasel Transmitter LT-EC4907313 Essential Services milled Unter
) Qiiller B Iavel
(. Desspers D-13-63 Bettery Boos and tritchgear hat D-23-53 Battery Boos and Switchgear hat , D-33-63 3sttery Boos and,3ritchasar hat . D-7 5-83 3attery Bean and Switchgear hat D=403-63 Bettery Boom and 3ritchgest hat D-613-83 Battery Boom and Switchgear hat Onble Days FS2-63 kieter) Mp
. C115-53 l (Iater)
- 144-63 (Iater) !
Fire Zone 373-5. Belay Boom 3 matrol Boom 3 . (later) Balay Equipment
- 11. Assessment of Poseulated Fire consecuences
- The above listed safety-related and noir-safety equipment could be damaged or lose fuection due to a fire.
The fire and its associated effects are contained by the three-hour fire rated fire resistive barriers enclosing Fire Area BVA. Fire Area 375 is beyond the influence of the sans fire and continues to remain available as required.
- 12. Safety Evaluation the safety evaluation will be presented later as discussed in Subsection 9.5.1.3.1. Additional changes to design will be incorporated in the FEA s.t l
that time. 9.5A-145
.* * * = =
WNF-3 -
, 178,a# - 9 FSAR N , 10. Acalysis of Effects of Postulated Fires (Cont'd)
Equipment of Component ID Mo. and Division Function Fim Area MA. Miscellaneous Areas A . , j Fin Ana MA-1. Fuel Eendlina Buildina R&V Room Fuel Pool E-8 FEB Ventilation Eshaast Unit FMS EV Unit IV-1 (Later) FEB E-7A, E-73~ . (Lacer) ' Units 4807 MCC 231 (Later) Despers D-10&-SA EVAC Dampers D-105-SE Cable Trays F21-EA (Later) t C6FMA , (Later) g '* Fil-NA (Lacer) L15-MA (Later Fire Ione MA-2. Elect rical Ares Volws 277-501883 Cost. Isol. 2FV-501785 Cost. Isol.
- 277-301985 Cont. Isol.
2FV-3023SE Cost. Isol. 2F7-50243B Coat. Isol. i Transmitters II-MV-5071A5 Airborne Effl. Man. Fuel Fool FT-SI-0352AS Cont. Abs. Press. Quad 5 FT-SI-0332CS Cont. Abs. Press. Quad C . II-IV-5072AS Airborne Effl. Mon. Fuel Pool FT-SI-0351AS Cont. Abs. Presa. Quad. "A" Monitors EE-EV-5064A Plant Vent 4* RE-HV-5043A Plant Vent 4* 4180V NCC A-121 (Later) Cable Trays Fil-NA C63-NA '
%ter)]
(La ter) i F21-NA (Lat er) ,
" L15-NA (Later) J ,
k 3 9.5A-154 .-
* ~
m FSAE
- 10. Analysis of Ef fmts of Po stulated Fir ~s (Cont'd)
Q} Equipment of Co mponent g No. and Division - Function . Fire Zone Mk-3, hermency Diesel Generator R&Y hos A- - Supply Air thit EVE2ASA - (Inter) Dampers D-9 E-SA (Ia tar) D-9IA-SA (later) D-9 2A-6A (Ia ter) l D-4 3&-SA (Ianer) D-115-53 (Ia ter) < 1,0-24-SA (Iater) Esturn Fans F41A-SA (Im tar) . Exhaust I'ans E-la -SA (Tfacer) E-38&-6A (Iater)
- I-3 SA-SA l (Iater)
I-40&-4A ((Ia ter) A407 Brr C A-23 QIater[ Cable Trays Fil-NA F21-MA [(Iatar) [ (later) p , C63-RA I (14 ter) - LIS-NA '(Later) ' Fire Area MA-4, Shield Bailding Cleanup thit Shield Ild3 CU-34-SA Maintain shield 31d3 studer Clean Up thit negative pressure Emmhwat ha E-4A %drogen purge exhaust Valves 2FT-80 01-SA (Ia ter) 2FY-E03-SA (Iater) 2 FV-5004-SA (later) l 2FV-E05-SA (Iater) 2 Ff-5025-SA Iater) 2FV-B152-63 2FV-5001H53 k(Iaeer) Cont'ainment Isolation Csble Traya F9-SA (Iaear) d C33-SA (latar) A)p Mrs Area M11. Miscellaneous Area s 1 Fire 2.one MS-1, Solid Radvaste Pump Room B l Pump FSB YES Cond. Sample A,3 Peed "Jank Metering Resia Metering A,3 (budensate Sampling i 1 9.5A-155 1 - ~. . , . -
1788E - 11 3 gg p 10. Analysis of Effects of Postulated Fires (Cont'd) - Equipeest of . i component ID No. and Division Function Tank . Dewataring A3 Feed Cable Trays F26-NB (Lazar) l C34-E3 '(Istar) L32-R3 F10-N3 (later) '. (Iatar) . C42-u3 ' L(Later)J l
~
Fire Zone MB-2. Electrical 4rea 3 - Valves 2FY-31248A 2FT-3123SE m-u.4 - l (Inter) ) yt.t.,p Monitors - BE-EV-50443 Flaat Yeat.1 Noble Gas
- 33-37-50433 Flaat Vent. 1 Particulate
- 4807 Mcc 3121 -
Iata ) 4807 For C 312 (Later) In t. Backs BA-6235 (14 tar) BA-44 (14ter) Cahie Trays F12-NB (Later) F26-53 [M L (Latar) C34-53 L32-R3 (Later) 1-Later) l Fire Zone M b 3. Baertency Diesel Generator B&V Boom Air Dust BT-2F SB (Lazar) Beturn Fana 1-413-53 (Iacar) Exhaust Fana 3-103-53 . (later) ' t-393-53 later) l B-403-53 f ((Lacar) l ( Despera D-9CB-53 (later) l D-913-53 (Later) i D-923-53 (Lacar) D-933-53 (later) i LD-23-53 (Later) Cable Trays F12-NB (Later) F26-N3 (Inter) Mh C34-NB (Later) ; L32-NB s (later) / , 9.5A-156 ~- i 1 I . _ -
WWF.3 178 W - 12 PSag
, b
- 10. Analysis of Effects of Postulated Fires (Cont'd)
Equipment of - Component ID No. and Division Function
- Monitors EZ-HV-50 71BS Airborne Effl. Fuel Fool (FER)
- 32-5V-507258 Airborne Effl. Fuel Fool (FIB) .
Fire Zone MB-4. Shield Buildina Cleanup Unit 9 Shield 31d3 CD-33=S3 Maintain Shield 31 ding under clean Up Unit negative pressure Rahause Fan 3 Rydrogen purge sahaust Isolation valves 2Fv-5052-SA Shield Building Isolation 2FV-5101-SB (Later 2FV-B103-SE (Later) 2F7-5104-85 (Later) 2FV-5105-S3 (Later) 2F7-B125-83 (Letar) 277-5110-53 (Later) 2FV-511143 (Later) 277-5113-55 Later) Transmitters FT-SI-035135 Containment Absolute F:sseure Quad a FI-SI-0351DS , Containment Absolute Pressure Qead 3 Ff-514352BS Contaiment Absolute Pressure Quad B FT-SI-0352DS Containment Absolute Pressure Quad 3 Cable Tesys C42-SE (Later) [h F10-SE (Laeer) Fire Zone MB-5 VES Distillate Rev' r (Lacer) (Later) vas Concentrator (Later) (Later) VES Distillate Condenser (Later) (Later) VES Distillate Pump (Later) (Later) Sscondary Particulate Filter (Lacer) (Later)
. VRS GDS Sample Tk A (Later) (Later)
VRS CNDS Sample Tk 3 (Later) (Later) Her.nr erd-16 Contro11e (Later) (Later) 9.5A-157 l l so.
. - = +
' WMP-3 178 W - 13 FSAR 36
- 10. Analysis of Effects of Fostulated Fires (Cont'd)
Equipment of ~ Component ID No, end-Division Function (Later) (Later) 713 Rack #1 - VES Rack #2 (Later) (Later) SPF Inst Eack (Later) / (Later) l
.Iste Rack #2 (Later) (Later)
Inst Eack $4 (Later) (Later)
- Loss of information/ data on radiological measurements results from equipment damage.
- 11. Assessement of Fostulated Fire Consecuences safety-related and sou-safety-related equipment located in Fire Areas MA may be damaged or lose fuoscies due to a fire. The fire and associated effects cre sostained by the three-hour rated fire resistive barriers enclosing Fire Area MA. Fire Area MB equipment required for shutdown is redundant to Fire Area MA equipment and is beyond the influence of the postulated fire so that it continues to remain available as required.
- 12. Safety Evaluation
.The safety evaluation will be presented later as discussed in Sub'section 9.3.1.31. Additional changes'to design will be incorporated into the FRA at that time. -
4 e h 9.3A-158
WNP-3
. . 178 017-4 ISAI bl
- 10. Analysis of Ef facts of Postulated Fire (Cont'd) :
Introduction of transient combustibles resulting from normal amineenance and operation activities is controlled through administrative procedures.
- c. Demare Limitation - h extent of damage within and beyond the fire area is limited by controlled removal of heat,' smoke and other products of combustion through continued operation of the area ventilation systema, and by three-hour fire resistive barriers enclosing the fire area.
- d. Fire hetulation - A fire is postulated in Fire Area 13 resulting from an exposure fire.
The fire could achieve s ==wf == development that directly exposes the Eamote Shutdows knel, the cables within it, and the cable trays over the burn area. The Safety Train A and B cable trays are approzinstely four feet apart with no barrier between them. e.
- - Fire Effects - The three hour rated fire resistive barriers enclosins Fire Area R$ are fully capable of containing the postulated fire and its effects within the fire area.
f. Equipment in Area - Safety-related and non-safety-related equipment in the fire area are listed below: l Equipment er ID No. and j- Component Division haetion Fire Area RS - Remote h edown Pa =1 Remote Shutdesa Panel - Alternate shutdown of plant Cahie Traye C47-SA A (Iaeer) C41-SA (14 car) L23-SA (Iater) t C64-53 (later) i C36-53 l (Iater) ( L28-53 (14 ter) ;
- 11. Assessment of Postulated Fire Consequences 3e above listed safety-related and non-safety-related equipment could be damaged or lose function due to a fire. The Control Room, which is beyond the influence of the fire, continues to remain available so that the capability to shutdown the plant remains.
- 12. Safeey Evaluatioa Safety evaluation will be presented later as discussed in Subeaction 9.5.1.3 1. Additional changes to design will be incorporated into the FRA at that time.
9.5A-176
s 17819-11 WNF-3 FSAR
)l
- m. -
- 10. Analysis of Ef fects of PoseuIsted Fires (Cont'd) .
Equipment of Comoonent ID No. and Division Function Dalve FCT-CS-711138 metainment Spray Pump 3 Flow Control A (Iater) -
~
FD Tanks (3) (Ia ter) l 1 FD hop (!ater) Floor kein Sump 6 ,
- Floor Drain Sump 7 -
1kak (Iater) macentrate Storage
. Pompe (Later) ,
, Coscentrate Storage Transfer (Iacer) Spent Basia Ransfer i (14ter) Speat Imain Dewataring A,3 Secondary Egh krity Waste A,3 Inorganic Chemical Waste A,3 Secondary Particulate Weste 1 Tanks Later Spent Essia Storage A,3 Secondary Eigh krity Weste A,3 Inorganic Gesical Weste A,3 Secondary Particulate Weste cehie Trays PS2-B I(La ter)
CS4=EE (Iater) h/A C66-NE (Iater)
- C68-N3 (Inter)
- CS6-5 (Ia ter) '
IAS-53 Istar - Fire Zone EU-3 Accese and Boric Acid Sample
^
taasmitter LT-CBG108 hidup Tsak !avel Tanks A, 3 Boric Acid Condensate Sampling Pumps A,3 Boric Acid- madensate Sampling Instr Backs BA-17 Boric Acid Sampling Syst "A" Tas trument a EA-18 Boric Acid Sampling Syst "B" Inatrumenes l Chble Days P31-NA (Later) (Ia ter) i Ct 7-RA
- l L13-NA (Iat ar) '
l. l l 6 l l 9.5A-187
)
1781W-12 WNP.3 rsia e$ j
- 10. Analysis of Bf fects of Fostulated Fires (Coat'd) I
{ Equipmese of - - Component ID No. a d Division Fauction Fire Zone 19-4. Bedweste Boulement Yalva 2FY410755 Isolation Fuel Bandling Building
'Esaks (Iater) Decosem=dnation Sagle A. B Caustic Addition (Inter) Flusb Task
~ .
Pompe (Is ter) Decostamination fanple A45,345 Condensate Makeup (Inter) Caustic Addition Tank Notering (later) Acid Addition 1kak Matering (Iater) Flush Pump - Condenser (Iater) Floor !kain Distillate Filters (Iater) DeterBast vaste A. B Ibel hel (Ia ter) Floor Drain-(Iater) InorBamic mexical Haste (Ia ter) Secondary Bish Purity (Iater) Boric Asid , su Besetor Drain 14N,2-4N hrification 1-SA.1-83 Seal Injeetion
. Cable Trays F32-NB (Iater) N/A C64-III (Iater) -
CS2-IIB (Ister) - L12-NB (Iater) i F32-NB (Ister) C6443 (Iater) # C86-Il3 (Iater) p V4-S B (Iater) ' C48-43 Iater L2 6-6B Fire Zone 59-7, Badweste ?sIve Cs11erv Seal Injection Filter Batches Automated Filter Cartridge (Is tar)- 3ransfer Cart Cahie Trsys P32-NB (Iater) NA C64-NB (Iat ar) Mk 9.5A-188 l y --e c --
i 1781W-14 WNF-3 FSAR,
)4l I
- 10. Analysis of Ef facts of Postuisted Fires (Cont'd)
Equipense of . Conoonent D Bo. ead-Diwision Fuaction Pteps 1.2 SN 3eric Acid Makamp A,3 Floor Drein Sump 3 brganic manical Weste Condenser A,5 Inorganic mesical Weste Seal Water A,1 Taorganic manical Waste Recirculating B Inorsanic Chemical vaaee condenser Secondary Particulate Wasta Floor Drain Soap 4 ,
'! kaka 3 Secondary Particulate Wasta (Za cer)
(later) Inorsanic Genical Wasta Seal Water A,5 Taorsanic memical Wasta Bottoms Isorganic m anical Weste Condensate Cable Troys C66-EB (Iater) N C66-EB (Iater) * (Ister) CSS-EB , C86-MB (I4ter)
- F32-EB (Iater)
CB4-EB (In ter) g
- i 148-MB (Iater) .
i Fire Zone EN-26. Access Area s. EVAC Tank (Iater) Auxiliary Coodenaste Flash Ptape A,3 Auxiliary endensate Flash Fire Zone 19-27. Access Areas. EVAC Elevator Machinery (Iater)' (Iater) Fire Zone 1W-28. Radiation Monitors. Hydrogen Analyzer Y2.1ve l 3CC-VE638SA (later) l Monitors 1E-5V5020& (Ia ter) RE-HV50 21A (lat er) Fire Zone 1W-29 Radiation Monitors. Hydronen Analyser Valve 3CC-VE255SB Monitors RE-HV50205 Radiation RX-HV50213 la diseion 9.5A-190
, H 17819-15 UNP 3 Psia g
p - 1 l
- 10. Analysis of Ef fects of Postulated Fires (Cont'd) -
Equipseae of - Commonent I ID Mo. and Division Function '
'Beric Asid (Iater) kIater)
Concestrstor ceble Trsys Vir S3 (Iator) C&843 L2643 Fire Zone R9-51. Eadweste feuipment Valves 2CE-9QOO1551 Seal Injection Best Exchanger Isolation 2G-Vub01351 RW Storage Tank Isolation 2CI-W4048AE RW Storage Tank Isolation 2CE-TU4173EE RW Storage Task Isolation Transmitters EE-SD-001 Samp, Secondary Eigh Purity Discharge
- W-6D401 Sump, Secondary Egh krity EscharEe*
EE-wi>6106 Discharge to Neutralization Pond
- E4M-6106 Eacharge to Emutralization had*
PIT-201 Intdows Pressure Control TIC-213 Boric Acid Bacching Tank 1haperature TI-131 Seal Injection Tamperature Fire Zone 19-61. Access and Badweste Iquipment Transmitters LT-CE3276 hidsp 1hak Invol LT-CE277 Boldup Tank Level. LT-CE3278 hidsp 1hak Invel LT-6279 Holdup Tank Inval Valve FCV-CS-7111AS Contaiuneat Spray hop A Flow Orntrol Taakt: A,3,C,D,E Eoldup
- A,5 Secondary Egh krity Semple Pumps A, B Secondary High Purity Sample Fire Zone 1U-42. Radweste F4uipment Yalve 27Y-80075A (Ister)
Instrument Backa RA-9 Discharge to Neutralization Pond
- EA-10 EA-7 (Ist (Iater)er)]
i EA-4 (Inter) RA-15 (Iater) ) BA-16 (Inter w
~
9.5A-191
r 178LU-16 WNP-3 F5&R 3p
- 10. Analysis of Ef facts of Postulated Fires' (Cont'd)
Equipt of ~ 1
' Component ID No. and Division Ftanction Tasks Daeergent usaea (2)
Floor Draia Condensate (4) 1.2. 3. 4. 5. 6. 7. 8. 9(5N) es Decay Esat SE Nitrogen Encycle Osepressors 1-EN hate Qas SN Nitrogen Recycle Gas Analyser Package SN
.hIster) ',
Gas Stripper EN (la ter) metrol Penal SN Heste N2 Becombiner metrol Maste $drogen Escambiner SN { valve Back [(Iater) I Weste Nydrogen Gas SN. (Iater) Analyser Cable Trmya 73-NA (Iater) N[A G43-SA (Inter) , L21-5A (I4ter) . C8 M A (Iater) - CS 7-R& (Iater) mA (Later) CSI-NA (14 ter) C83-NA (Iater) f L13-R& (Ia ter) ' Mansmitters FIT-204 Process Monitor Flow FT-210X Emactor Makeup Water Flow Control FT-210T Boric Acid Makeup Flow 0:strol i l LT-226 Voline Control Tank Level LIT-227 21ume @ntrol Tank Inval Beat Exchanser' (La ter) Seal Injection Tanks (later) memical Mdition Boric Acid Bacching 3 Quatic Mdition B Acid Addition SN W1ume @n::rol Puspe (24 ter) i Q emical Mdition 3 Qustic Mdition S Acid Mdition (later) Inorganic menical Wasca Disi:illace I 9.5A-192 f __ _ _ _
17819-17 WNP-3 rsia o ' 10. Amelvsis of Effacts of Postulated Fires (Cont'd) ' Equipment of - Component ID No. and Division Function Qpodensers _ A.3 laerganic Chemical Weste Cable Traye C82-MB L12-B N/A 732-NB (Inter) , C64-B (later) ' CSD-R (Inter) CB6-B (Iatar) ' C84-MB (Iater) , 144-MB (Inter)
- C66-EB (Iater) .
C64=EB (Iater) ater) - Fire Zone 19-52, Radueste Oratrol Room
- Control Cabinet BCP-3 RYAC Cuatrol Multiples Penal 6 (later)
BVAC thit AC-3 ((Iater) Ceble Ttsys CSO-B CB2-MB
~ (Iater) h, CS6-IB (Inter)
CS4-MB (Inter) ' L12-MB (later) C64-NB (later) ' C64-MB (Iater) C66-EB (later) , L48-NB (later) CSI-MA (later) 1 C83-NA (later) C85-NA (lator) i C87-NA (Iater) (La ter) L13-NA (Iacer) Fire Zone R9-91, Radveste Storage and Decontamination Bridge Crane (Iater) Iater)\ 10 Ton Trolley (later) (Iater)y Pumpe A,3 Sump #15 Fire Zone 12-92, Radweete Storage and Decontamination Bridge Crane (Iater) [(Iater) 15 Ton Monorail (Ia ter) t Ia ter)
*Iaes of information/ data on radiological measurements results from equipment damage.
9.5A-193
17829-7 WWF.3 PSAa 2
- 10. Analysis of Effects of Postulated Fire (Cont'd) -
Equipment or ID No. and 1 Component Division Function I
)
Fire Area SVA. Esseneini Switchenar Area A-SA i Fire Zone SU-1. Zeeencial Switchgear Boom A-SA 4807 MCC 311-SA matrol Boom Isolation Valve 312-SA Shutdown Cooling Beat Exchanger A 321-5A W Hary Feedvater Fiamp 1A & 2A e: 322-SA Emergency Core Cooling System
- Esat Enchanger Isolation Dalvee 4807 FC A31-SA Fever to siiald Bailding -
A32-SA cleanup Dait 4.16 kV Segr A3-SA hver to Q1111er UC-1A-SA UFS Penal and Transformer A-SA Dainterruptible Power Supply , Switchgear East Cabinet A-SA Damper D-5A-SA (Iater) Cable Days C41-SA (Inter) , M[d C37-SA (14ter) , C43-SA (Iater) C49-SA (Inter)
- CSI-SA (Iacer) .
C53-SA (Iatar) L21-SA (Iaeer) ( F9-SA (later) i F13-SA (Iater) F15-SA (Intar) ) F17-SA (Iaeer) { V3-SA (later) V5-SA (Iater) l F47-SA (later) F49-SA (Iater) \ C105-SA (late r) C103-SA (Iater) F43-SA (later) F45-SA (Iacer) l C99-SA (later) / C101-SA (Iatar) F35-SA (later) F37-SA (Iate r) C91-SA (Iater) { C93-SA (Iater) F39-SA (late r) F41-SA } (Iaear) > C95-SA (late r) C97-SA (Iate r) I 9.5A-201
' 1783r-s WIF 3 q N .
- 10. Analysis of If fects of Postulated Fire (Cont'd) !
Equipasst or - - ID Es. and Component Division h uction Fire Zone 59 2. Beersency Elsetrical Boos C-SC, Enttery Marger SC-1 (SC) Garging Estteries SC-2 (SC) hast SC 25 Dit de hael UPS Panel and Transformer SC _ Uniaterruptible Power Supply Cable Tray CSS-sC (Iacer) M//4 Fire 4res 593, Essential Seitehneer Area B-6B Fire Zone 59-3. Es sential Switchaear Beam 3 4aDY ECC, 311-53 Contrel Beas Isolation VaIvo 312-s3 hutdown @oling hat Etchanger 3 321-53 Auxiliary Feedwater Pump 13 & 23 A322-43 hergency @re moling Syntaa Esat Nh==ger Isolation Valves 480V PC 331-53 Itnrer to hield 31ds cleanup 332-63 Unit 4.16 kV Swgr 23-43 Power to miller UC-1341 l DFS Penal and Transformer 3-63 (Inter) Stitchgear het cabinet 3-43 Bettery Boom and Switchgear hatilation Damper B-53-6B Cable Trsys F10-5B F14-65 f((Inter)La ter)T F16-48 (Imter)
- F18-45 .(Iater) -
i C50-43 (Iater) ; C48-48 (Istar) C4 6-43 ! (later) l C40-6B (Iater) L2 6-43 (Iacer) 74-5B (later) l P10-43 (Iater) . C42-45 (Letar) 1 C36-5B j (lat er) V6-4 B (Iacer) - ( . 9.5A-202 l
. 1782U-9 WNF-3
'5^^
/f0 P
- 10. Analysis of Ef feets of Postulated Fire (Cont'd) l Equipment or . ID No. and Component Divistors Function ,
Cable Trays (@nt'd) F46-53 kla F48-53 (Iatar) C100-53 (Iatar) - C102-83 (Iater) , 742-83 (later) F44-53 (later) ' C96-53 (later) g F38-53 (Iater) i F40-53 (later) $ C92-53 (Iater) C94-53 (later)
~
() 4 F34-53 (later) W . F36-SE (Iater) C88-53 (Iater) s C90-53 (later) - Fire Zone 59-4, Beersency Electrical Boca D-SD Battery marger SD-1(SD) marging Batteries SD=2'(SD) Fanal SC 25 Volt de Panel UPS Panel and Transfdamer Sc Onble Trays C32-SD - l 11. Assssement of Postulated Fire Consequeness The above listed safety-related and non-safety equipment in Fire Area SWA may I be damaged or lose function due to a fire. Se fire and its associated ( effects are contained by the three-hour rated fire resistive barriers enclosing Fire Area SWA. Fire Area SWB equipment is redundant to that contained in Fire Area SWA and.is beyond the influence of the postulated fire so that it continues to remain available as required for safe reactor shutdown.
- 12. Safety Evaluation The safety evaluation will be presented later as discussed in Subsection 9 5.1.3.1. Additional changes to design will be incorporated into the FEA at that time.
t. I s. 9.5A-203
- l. - .
. WNF-3 17839-17 FSAE P .
- 10. Analysis of Ef fects of Postulated Fire (Cont'd)
Equipment or ID No. and . Coevenest Division Function 125Y de Panels NA,53 (Inter [ 125Y de Distribution A (I4ter) Panel . Electro Magnetic Filter (Iater) (I4 tar) t Control Panel
~
Electro Magnetic Filter (latar) (Istar) Tastrument Panel Electro Magnetic Filter A, 3 (Iatar) ,
, Coil Cooling Units Amercap Ball Collectors (4) collectors Fans E-44A b(I4te E-448 (14 tar) -
Saltiplazar MDE-8 (Iater) Instrument Air Receiver (latar) I (Istar) Instrument Air Dryer (Iater) Cable Trays CS-NA Non-helaar Saf sty Systems Cables C7-NA Mon-Esclear Safety Systems Cables C9-MA Non-Naclear Safety Systems Cables Cl-NA Non-hclear Safety Systems Cables C3-NA Non-Nselaar Safeey Syseses cabias L1-NA Non-hclear Safety Systems Cables L3-NA Non-hclear Safety Systems Cables F3-NA .Non-hciear Safety Systems Cables V1-NA Non- h clear Safety, Systems Cables F1-NA Non-hclear Safety Systems Cables l C5-NA Non-hclear Saf ety Systems Cables C25-NA Non-hclear Safety Systems Cables C16-NA Non-Naclear Saf ety Systems Cables CIS-NA Non-hclesr Safeey Systens Cables C20-NA Non-hclear Safety Systems Cables
. C22-NA Non-hclear Safety Systems Cables C6-NA Non-Naclear Saf ety Systems Cables I L12-NA Non-Naclear Safety Systems Cables C15-NA Non-hclear Safety Systems Cables C17-NA Mon-Maclear Safety Systems Cables C19-NA Non-Nuclear Safety Systems Cables C21-NA Non-helear Safety Systems Cables Lil-NA Non-Nuclear Saf ety Systems Cables F Non-hclear Safety Systems Cables CS-NB
! Non-hclear Saf ety Systems Cables N C10-NB C2-NB Non-hclear Safety Systems Cables 9.5A-220 l 1 i
WNF 3 1783W-18 FSAR
- 10. Analysis of Ef facts of Postulated Fire (Cont'd) .
Epipment or ID No. and Component Division Function Cable Trays C&NB Non-helaar Safety Systems Cables C6-NB Non-hclear Safety Systems cables L2-NB Non-hclear Safety Systems Cables L4-5 Non-hclear Safety Systems Cab 1as Y2-NB Non-helaar Safety Systems Cables P2-NB Non-hclear Safety Systems cables , 74-NB Non-Maclear Safety Systems Cables . ! C28-NB Non-hclear Safeey Systems Cabias - Fire Zone T-1.1, Batter 7 Ecoes
=
125V da Betteries (Inter) (Istar) . 2507 de Rectaries (Istar) (later) Fire Zone T-2, Condensate Domineraliser Area Basia Strainer Condensate (Iater) (Iater)l J l Domineralizer Yessels , 1 thru 12
~
Tanks (Iater) Cation Eageneration Tank ~ l' (Iater) Anion Ingeneration Tank (Iater) Essia Storage Tank (Iater) Acid h k- . (Inter) Caustic h k (facer) Ammonia h k
- (later) Caustic Escovery Tank (Iacer) Ultrasonic Resia Cleaner Drain (Istar) Tank Fangs A,3 Sluice Ptape A, B Recycle Pumps A, B Service Water Pamps A,3 Floor Drain Sump f12 l IC C's A215 (I4 ear) i A223 (14ter)
Instrumane Back EZ-4 (Iater) Mattiplazer MUI 9 (14ter) D i.etau..e iir m .er Giem (1acer) h .1..ra u s.r Co..ro1 ,a.e1 g (teter) 1 l l 9.5A-221
WNP-3 1783W-19 FSAR 45 3 10. Analysis of Ef fects of Posenlated Fire (Cont'd) , Equipment or ID No.' and Component Division Function Cable Trays V1-NA Non-hclear Safety Systmas Cables 71-NA Non- h clear Safety Systems Cables 73-NA Non-hclear Safety Syntans Cables Cl-NA Non- k clear Safety Systems Cables C3-NA Norkelaar Safety Systems Cables L1-NA Non-hclear Safety Systems Cables C5-NA Non-hclear Safety Systema Cables *, ( V2-NB Non-h clear Safety Systems Cables . P2-MB Nea-helear Safety Systems Cables P6-MB Non-hclaar Safety Systems Cables C2-NB Non-hclear Safety Systems Cables C4-NB Non-helear Safety Systems cables L2-NB Non-helaar Safety Systems Cables C6-NB Non-Maclear Safety Systems Cables Fire Zone T-2.1. Unioading Eey Rosa Fire Zone T-3, Condenser Tube Pull Iron Pumps A. B C, D Mechanical Vacuum Pumpe g A, 3 Heater Drain Pumps-I A, B Floor Drain Sump #13 A, 3. C, D Amartap Eaiajection Ptusps Instrument Air Aftercooler (Iater) (Iater) h Separators A, B (Iater) Instrument Eack ET-5 (Iater) l Radiation Monitors RE-WM6213* Liquid Effluent, Waste Management System Discharge ET-WM6213* Liquid Effluent, Waste Management System Discharge RZ-NV1400* Condenser Mechanical Vacuus Pump 10 Toa Monorail Cranes (2) (14ter) Cable Trays F1-NA Non-Nuclear Saf ety Systems Cables l 73-NA Non-Maclear Safety Systems Cables l Cl-NA Non-Nacisar Safety Systems cables ! C3-NA Norhclear Safety Systems Cables L1-NA Non-hclear Safety Systems Cables V1-NA Non-hclear Safety Systems Cables
- P2-NB Non-Nuclear Saf ety Systems Cables 1
l 5
- Radiological sessurement data is lost if this equipment is damaged.
l 9.5A-222
' ^ ^ ~
WNP-3 . 1783r-20
- PSAR
/
/ <*... _ Analysis of Effects of Postulated Fire (Cont'd) l
- 10. .
Equipment or ID Bo. and Component Division , haction Cable Trays 74- 5 Non-helaar Safety Systems Cables C2-MB Moa-Ecisar Safety Systems Cables C4-MB Non-hataar Safety Systems Cables L2-EB Non-hclear Safety Systems cables V2-EB Moa-Inietmar Safety Systems Cables : hwi1f ary Boiler Desarator (later) 3 (Iater) Paape (Istar) A==41tary Boiler Recirculation hap
. hwitiary Boiler Peed hop Boiler Control Paeol (Iater) (14ter) 428 kW 13.8 kV (Iater) -(Iater) hw41tary Electrode Bailer (Iater) (Inter)
Fire Zone T-3.1. Unioading 347 Mona , Fire Zone T-4.1, Switchaear Boos A 13.8 kV Bos A1 (Iater) 4.16 kV Ems A2 (later) .
~
480V MCC A211 (later) A221 (Iater) 4807 FC A21 (Iater) A22 (14ter) l Panel NA Uninterruptible Power Source Cable Trays 73-NA Son-Naclear Saf ety Systems Cables C7-NA Non-hclear Safety Systems Cables C9-NA Non-hclear Saf ety Systems Cables Cl-NA Non-hclear Safety Systems Cables L3-NA Non-Naclear Safety Systems Cables 75-NA Non-hclear Safety Systems Cables i Fire Zone T-4.2, Switchsear Room 3 13.8 kV Eus 31 (Iacer)\ 4.16 kV Bue 32 (14 ter) l 9.5A-223 e -
WNP-3 17839-21 F5AR
, s. 10. Analysis of Ef fects of Postulated Fire (Cont'd) ,
F.quipeemt or . ID Nd; and Component Division Function 440V 3CC 3121 Iater) 3222 (Inter) , 480V FC 312 Panel NB Unistarruptible Power Source Cable Trays 74-NB Nochelaar Safety Systema Cables , C3-MB Non-belear Safety Systems cablas Cit > NB Nea-helaar Safety Systema Cables C2-MB Non-b clear Saleey 5yseesa Cablaa L4-NB Non-helaar Safety Systems Cables F6-55 Neo-b elear Safety Systems Cables Fire Zone T-5, Eshaster Drain Tank Area Tanks A1, 31 lat Stage Embaster Drain Tank A1, 31 2nd Stage Embaster Drain Tank
. A1, 31 Moisture Separator Drain Tank High Pressure Heaters . la (Iater) ,
13 (Iater)
~
011 Separa or Pumps (14ter)
- Waeer Puay l '
(Iacer) 011 Ptamp
. Gland Steam Condenser (Iacer) (Iacer) l 1hrbine Elsetric (14 tar) (14 car)
Hydraulic (later) (later) g l Fluid Unit (Iater) (Iater) I Accumulators (later) { (Iater) MCC A114 (Iater) Instrument Backs II-3 (Iater) IT-6 (Iater) IT-7 ' (14ter) EI-10 (Iater) Cable Trays C7-NA Non-kclear Saf ety Systems Cables C9-NA Non-% clear Safety Systems Cables Cl-NA Non-Welear Saf ety Systems Cabl.es L3-NA Non-Welear Safety Systems Cables P 9.5A-224
WNF-3 17839-22 FSAE
)
- 10. Analysis of Effects of Fostulated Fire (Cont'd) .
I F4 mipeans or ID No. and Component Division Function Cable Trays 73-NA Non-belmar Safety Systems Cables FS-MA Non- h clear Safety Systans cables CS-NB Non-betaar Safety Systems Cables C10-MB Non-hclear Safety Systems cables C2-NB Non-hclear Safety Systems Cables L4-MB Non-hclear Safety Systems Cables 76-NB Non-5,t aar Safety Systaas Cables . F6-NB Non-hclear Safety Systems cables Fire Zone T-6, hrbine Esheast Ares
- Tanks 32 1st Stage Babeater Drain Tank B2 2nd Stage Rehester Drain Tank 32 Meistare Separator Drain Tank hw Pressure Heaters SA, B , C - (Inter) 6A, B, C (hter)
Covered Besia Sepper (late (Inter) Generator Main Leeds (Inter) . (I4 tar) I Cooling Unit , l Lead Breaker cooler (Inter) (I4ter) Instrument Eack ET-20 (Iater) Cable Trays PS-MA Non-Maclear Safety Systems cables F3-NA Non-hclear Safety Systems Cables C7-NA Non-hclear Safety Systems Cables C9-NA Non-hclear Safety Systems Cables L3-NA Non-h clear Safety Systems cables FM Non-bclear Safety Systems Cables F4-NB Non-hclear Saf ety Systems cables CS-NB Non-hclear Safety Systems Cables C10-NB Non-hclear Safety Systems cables L4-NB Non-bclear Safety Systems Cables Fire Zone T-7, High Pressure Easter Area hups (Iater) Feed hap Webine Exhause A Feed Fump h rbine Izhause B High Pressure Heaters 1A, B .(tater) low Pressure Heaters 2A, B I Tanks A2 1st Stage Reheater Drain Tank j A2 2nd Stage 24 heater Drain Tank C A2 Moisture Separator Drain Tank 9.5A-225
'* WNF-3 17839-23 ysag , 47 10.
Analysis of Ef fects of Postulated Fire (Cont'd) g Iquipment or ID No and Component Division Function Generator Neutral (Inter) (later) Grounding Transformer
& Essister NCC 3114 (hter)
Instrument Backs IT-8 (Iater) IT-11 (I4ter) ,; l Gas Dryer (late (later) Ceneracor A=wiHary ((Inter)
\ (I4t'er) -
Penal -/ J' Cable Trays 76-NA Non-Natlear Saf ety Systems Cables 73-NA Norkelaar Safety Systems cables C7-NA he& clear Safety Systems Cables C9-NA Norkclaar Safety Systems Cables l L3-N& Non-h clear Safety Systems Cables PS-NA Neo-hclear Safety Systems Cables
~
F6-NA b rNaclear Safety Systems Cables 76-NA Nos-hclear Safety Systems Cables CS-NA Non-h eleer Safety Systems Cables C1(MIA Mon-Maclear Safety Systems Cables L4-NA Non-h clear Safety Systems Cables , Fire Zone T-8, Hrdrogen Seal Oil Unic Area Nydrogas Oil Unit er) (laeer Fire Zone T-9.1, Motor Driven Steam Generator Phap Area Pump (Iater) , Motor Driven Steam Generator Pump Cable Trays (later)) '(Later) s Fire Zone T-9.2, hrbine Driven Steau Generator Pump A Area Pump A hrbine Driven Steam Generator Pump Cabis Trays
- Fire Zone T-9'.3, hrhine Driven Steen Generator hap 3 Area Pump 3 hrbine Driven Stads Generator Pump Cable Trays (14 ter) (Later) { h 9.5A-226
l
. . wur-3 17839-24 FSAR h$
- 10. Analysis of Effects of Postulated Fire (Cont'd) ,
Fquipment or ID Mo'. and component
, Division Function Fire Zones 10.1 Thru 10.6, hrbine Bearius Areas
%rbine Bearings later) (Later Ere Zone T-11. Operating Floor Eish Pressure Turbine (later) (I4ter) ,
Iow Preneure hrhinee IA (14ter) 25 3C Moisture Separator A, 3 (Inter) Esheaters low Pvessure Besters 3A, 3, C (Latar) 4A, 5. C Air Bandling Units IAC-2A, 3, C, D (Iater) - Inatrument laek E -12 (14 tar) , E-13 (Inter) E -15 (14ter) E-16 (14ter) E-14 (Inter) E-17 - (14ter) MC C's A212 (laeer) A213 , (14ter) l A223 (14ter) 1 A224 (14cer) Rhc Control Panel ECP-5 (I4ter Cable Treys 73-NA Non-14aclear Safety Systema Cables C7-NA Non-&cisar Safety, Systems Cables L3-NA Nc" 4 c1use Saf ety Systens Cables P4-NB bw .? mitar Safety Systems Cables I C3-NB ebclear Saf ety Systems Cables L4-NB sh .a 4 ear Safety Systems Cables I Fire Zone T-20, Inbe 011 Tank Roca t N I De 14 val 011 (I4ter) (Iacer) } Purifier Unit Inte 011 Transfer (I4ter) (Later) Pump (Iacer) l 9.5A-227
WNP-3 qe 1783e-25 FSAE p . ~
- 10. Analysis of Ef facts of Postulated Fire (Cont'd)
Equipment or ID No. and Cossonest Division Function Dirty L 0 [(Later) (Later Batch Tank f Clean L 0 (Later) (hter) Batch Tank Cable Trays 73-EA Mon-BacIsar Safety Systems
- Cl-NA Cables Fire Zone T-21. Lube Oil Room Main Turbine Later) (Later)
Lube Oil Reservoir Phap (Later) 011 Fump Fire Zone T-22 Auxiliary Boiler Room Auxiliary Boiler Deserecor (Later , Pumpe (Later) ' Auxiliary Boiler Recirnulation Pump
. (Late r) Auxiliary Boiler Feed Pump Boiler Control Panel (Later) (hter) 428 Idi 13.8 kF (Later) (Later) i Anziliary Electrode Boiler Later) (Later)
Fire Zone T-23. El,ectrical Switchaear Room Electrical Switchgear Later) (Later)
- 11. Assessment of Postulated Fire Consequences The above listed safety related and non safety equipment in Fire Area T may be damaged or lose ' function due to fire. The fires and associated effects are contained by three-hour rated fire resistive barriers enclosing. Zones T-4.1 cod T.20. The equipment exposed by the fire is not required for safe reactor s hutdown. Any safe shutdown equipment in any adjacent fire areas is beyond the influence of the postulated fire and continues to remain available as re quired.
- 12. Safety Evaluation The safety evaluation will be presented later as discussed in Subsection 9.5 1.3.1. Additional changes to design will be incorporated into the FEA at
=' that time. 9.5A-223
' 3 1820EF-5 Q = .
- 10. Analysis of Effects of Poeculated Fire (Cont'd)
- f. Equipeone in Area - Safety a'ad non-safety-related equipment are listed below. ,
Equipment or ID Me. and - Component Division Function Fire Area Ulf-1 Electric Area A Fouer Center A331-SA Fower for CG Dry Cooling Towers A&3 Cable Traye later) . Fire Area U11-4 Electric Area 3 Power Center A332-SA Feuer for CW Dry Ceoling Towers Cahh Treye v
- 11. Assesement af Poatulated Fire Coasequeaces The above listed safety-related and non-safety equipment in each fire area l could be damaged or lose function .due to a fire. The fire and its associated effects are contained by the three-hour rated fire resistive barriers enclosing each fire area. Since- the fire areas contain redundant equipment safe shutdown will est be impacted by the same fire.
(
- 12. Safety Evaluation The safety evaluation will be presented latar as discusssd in Subsection 9.5.1.3.1. Additional changes to design will be incorporated into the FEA at that time.
l b 9.5A-233
1 WNP-3 1817W-9 FSAR L 10. Analysis of Ef fects of Postulated Fire (Cont'd) Equipseat or ID 5e, and Comoonent Division Function Fire Zone FE-7, Spent Riel Handling Area Transmitters RE-EV-5071AS (Iater) RE-HV-5072AS (Iatar) RE-HV-50 7133 (Iatar) RE-HV-5072P3 (I4eer) ILE-HV-50715 (Iater) ILE-HV-50723 . (Iater) . Power Center (Iater) (Iater)
" Spent Fuel (14ter) (la ter) -
knating Machine Jib Crana (Ia tar) (Iater) ) 5 Ton Trolley Hoist (Ister) (later) ,
- 11. As sessment of Postulated Fire Consequences The above listed safety related and non-safety equipment may be damaged or lose fumetion due to a fire. ' Die postulated fire and assceiated affects are contained by three-hour rated fire resistive barriers enclosing the fire area. No safe shutdows e'quipment is involved so that safe ghutdown may he achieved and there is no loss of control of radioactive release.
- 12. Safety Evaluation The safety evaluation will be presented later as discussed in Subsection 9.5.1.3.1. Additional changes to the design will be incorporated into the FHA at that time.
e W 9.5A-242 r ____ _ . _
WNP-3 1013W-4 FSAR 2 72- >- 10. Analysis of Ef facts of Postulated Fire (Cont'd) Equipesat or ID No..and t n==t Division Funetion F1-e Area FT. CCW Heat Exchanner and Pues fire Zone FE-1. CCW Pums Room A-SA Ptempo IA-SA CCW 2A-4A CW Unit Cooler UC-8k-SA CCW Esat Exchanger A and CCW Piaspe 1A ,
& 2A -
(CG Area A) Transmitters TE-EV-4963AS CCW Pump A Area Temp TCP RV-4963AS CCW Pump A Area Temp - Fire Zone FH-2. CCW Heat Exchanner Boce A-SA Heat Exchanger A-SA Remove Esat from CCW System Instrument Eack EF-35A/SN (later)
- 11. Assessment of Postulated Fire Consequergej The above listed safety and non safety equipment in the fire ar'es FT any be damaged or lose function due to fire. The postulated fire and the associated .
cffects are contained by the three-hour rated fire resistive barriere which cacices the fire area. Redundant safe shutdown equipment is not involved so that safe almatdown may be achieved and there is no loss of control of radioactive release.
- 12. Safsev Evaluation The saf ety evaluation will be presented later as discussed in Subsection 9.5.1.3.1. Additional changes to the design will be incorporated into the FRA De that Cian.
9.5A-248
U WNF-3 FSAE hO P .
. 10. Analysis of Ef facts of Postulated Fire (Cont'd)
- c. Deasse Ilmitation - Me extent of damage within and beyond the fire some is limited by fire resistive barriers which separata trains of safety related equipment.
- d. Fire Postulation - Fire Ares UH A fire is postulated from an esposure fire in the storage tank dike area. Me fire could achieve a nazimum development that directly exposes the diesel oil storage tank and pumping equipment.
e Fire Ef fact - Fire Area - UH The fire area is enclosed within three hour raced fire resistive barriers and therefore the redundant fire ' area (UB4) is not affected. .
- f. Eauinment in Area - Safety-related and non-safety related equipment are listed below: . .
Equipment or ID Bo. and Component Division Thaction Fire Area Astive Area A Fire Area Coolinn Tower A Fire Area UE-3 Diesel Oil Storate Tank and Puno A ' . Diesel Oil Storage SA Emergency generator A diesel Tank hansfor fue1 oil supply Pump A Cable Ttsys (Iater) , Fire Area Electric Area B Fire Area Drv Cooline Tower 3 Fire Ares UH-6 Diesel Oil Storste Tank and Puso 3 Diesel Oil Storage 33 Emergency generator and diesel T.sak and Transfer fuel oil supply Pump B
- 11. As sessment o f Postulated Fire Consequences ne above listed safety-related and non-safacy-related equipment in each fire area could ha (sungsd or lose function due to a fire. Me fire and its associated effects are contair.ed by the three-hour rated fire resistive barriers enclosing each fire area. Since the fire areas contain redundant equipment safe shutdown will not be affected by the same fire.
- 12. Safeev Evaluation 3
Safety evaluation will be presented later as discussed in Subsection
- 9. 5.1. 3.1. Additional changes to design will be incorporated into the FRA at that time.
9.fA-260
Attachment 1-45 2266A Question No. 280.5 Discuss the extent of conformance to the guidance given in NFPA (9.5.1.5) 27.
Response
At the time our fire brigade training program is finalized and training commences, we will be in full compliance with the Recommendations for Organization, Training and Equipment of Private Fire Brigades, NFPA No. 27-1975. We anticipate that there will be no exceptions required. i l
Attachment 1-46 2266A Question No. 281.1 For all postulated design basis accidents involving release of (6.1.1.2) water into the containment building, estimate the time-history of the pH of the aqueous phase in each drainage area of the building. Identify and quantify all soluble acids and bases within the containment.
Response
l A complete response to this question will be available by December 1982. I _- _ ___________________________ ____ ____________- - _ _ _ _ . - _ - - _ l
Attachment 1-47 2266A
)
Questinn No. 281.2 Subsection 9.3.2 does not include requirements to minimize, to (9.3.2) the extent possible, hazards to plant personnel; provide these requirements or a date by which they will be supplied.
Response
The following safety features described in FSAR Subsection 9.3.2 are provided to minimize the hazards to plant personnel:
- 1. Sampling system is designed to limit radioactivity releases below the 10CFR20 limits under normal and failure condi-tions. An example of this is the samples from the Reactor Coolant System hot leg that must be provided with a contami-nant delay time of 90 seconds to permit sufficient delay of nitrogen - 16. Due to the length of pipe inside containment and the sample flow rate, 135 seconds of delay will actually exist.
- 2. Instrument is provided to monitor the temperature and pres-sure of the samples to be taken. The instrumentation is provided to insure the samples are cooled by heat exchangers and their respective pressure reduced by pressure reducing valves to values less than 120F and 30 PSIG. This is to minimize the possibility of local airborne activity.
- 3. Samples are normally taken only when the sample hood fan is operating to minimize the possibility of local airborne activity. In addition, the Reactor Auxiliary Building Ventilation System provides a back-up means of maintaining i
low airborne activity levels.
- 4. The sample lines penetrating the containment are equipped with two isolation valves which close on actuation of the l containment isolation actuation signal. In addition, should any of the remotely operated valves fail to close after a sample has been taken, back-up manual valves in the sampling room may be closed to limit the spread of contamination.
Attachment 1-48 2266A Question No. 281.3 As per Regulatory Guide 1.70, Revision 3, include a discussion of (10.4.6) the control of chloride ions and other contaminants in the con-densate cleanup system.
Response
Subsection 10.4.6.1 (Design Bases) states the following: "The purpose of the condensate demineralizer is to remove dissolved and suspended solids from the condensate in order to maintain a ! high quality of the feedwater being supplied to the steam gener-ators under all normal plant conditions (start-up, shutdown, hot standby, power operation)." The phrase " dissolved and suspended solids" includes chloride ions. The WNP-3 condensers are cooled using fresh water with a maximum chloride ion concentration of 50 ppm. Should the plant be operating with a postulated circulating water inleakage of
, five gpm due to condenser tube leakage, the maximum expected chloride ion concentration in the condensate will be 10 ppb. As the condensate cleanup system will be in operation during a con-denser leak, any chloride ions in the condensate will be removed.
The FSAR will be amended to reflect the response to this question. I
__ _ . _ . .. .._-~r 17fitN-1 * ! e ggp.3 FSAR ggl } 10.4.6 /qf)( ,f_ C0h*DENSATE CI.EANUP SYSTDi
- \,, 10.4.6.1 Design Bases The purpose of the Condensate Domineralizer is to remove dissolved and suspended solids from the condensate in order to maintain a high quality of the feedvater being supplied to the steam generators under all normal plaat -
conditions (startup, shutdown, hot standby, power operation). A flow dia' gram - l for the Condensate Cleanup System is provided on Figures 10.4-2a through g. This is accomplished by directing the full flow of condensate to a set of mixed bed domineralizer units. Since the domineralizers need periodic resin regeneration, out of service.spara units are provided in the system to replace units taken condensate water, The system provides final polishing of the secondary cycle The system is designed to process the full flow of condensate corresponding to 100 percent power level with one heater drain pump out of service. This equals 28,506 gpe, and it is the highest condensate flow expected under all operating conditions. The flow is split into 10 parallel units of mixed bed domineralizers, each designed to process 2,851 gpa. Each domineralizar unit maintains the outlet water quality within the limits shown in Table 10.4.6-1 with a postulated circulating water in-leakage of five spo due to condenser tube leakage. Circulating water quality is also Set d 4 escribed in Subsection 10.4.5 and Table 10.4.5-2.
---t Two standby regenerated. units are provided to replace units under maintenance or being Since only one demineralizer unit is expected to be exhausted at one time, the regeneration system is designed to handle 'the amount of resin of one mixed bed vessel at one time.
Since the system is located on the discharge of the low pressure condensate pumps, all related equipment is designed to withstand the pump's discharge pressure at shuto(f conditions. The regeneration portion of the system is designed to withstand the maximum discharge pressure of the sluice water pumps, which are used to transfer resin to and from different tanks. The design temperature which is 150F is determined by the maximum condensate temperature expected during plant operation, which is 126.7F at 4.14 in. Hg condenser pressure for power operation and 134F at five in. Hg for the f eedwater cleanup ' cycle. Since supplying feedwater to the steam generators at all times is necessary to ensure continuity of plant operation, the system is provided with a bypass line which can be used to bypass all or part of the condensate flow when some condensate impaired. demineralizer units are not available or their operation becomes s 10.4-25
O
.Tos,J 1 The WNP-3 condensers are cooled using fresh water with a maximum chloride ion concentration of 50 ppm. Should the plant be operating with a postulated circulating water in - leakage of five gpm due to condenser tube leakage, the maximum expected chloride ion concentration in the condensate will be 10 ppb. As the condensate cleanup system will be in operation during a condenser leak, any chloride ions in dhe condensate will be removed.
om 9 e l
Attachment 1-49 2266A Question No. 281.4 Subsection 10.4.8.1-e discusses the blowdown demineralizer (10.4.8.1) systems removal of impurities from the blowdown and references Subsection 10.4.11. Subsection 10.4.11 states this system has been deleted from WNP-3. If this is correct, the referencing paragraph should also be deleted from the FSAR.
Response
The Steam Generator Blowdown Demineralizer System was deleted from the original design. This decision is now being recon-sidered. We will supply additional information concerning this decision by January, 1983 i l l e -
Attachment 1-50 2266A Question No. 311.1 As per Regulatory Guide 1.70, the site area map should include (2.1.1.2) the site boundary lines and if they are the same as the plant property lines, this should be stated.
Response
The site area map, FSAR Figure 2.1-1, provides the site boundary lines as well as the plant property lines. The plant property is indicated by the shaded area on Figure 2.1-1 and entitled
" Fee Owned Property". The site boundary lines are defined on Figure 2.1-1 by the combination of the " Fee Owned Property" and the " Exclusion Zone Easement". The required exclusion boundary (4300 ft radius) is also shown on the figure.
I 1 l 1 I
Attachment 1-51 2266A Question No. 311.2 As per Regulatory Guide 1.70, an estimate should be provided of (2.1.2.2) the time required to evacuate all personnel from the exclusion area.
Response
Evacuation time estimates are provided in Section 12.5 of the WNP-3 Emergency Preparedness Plan. Table 12-5 (attached) pro-vides evacuation times for the two, five, and 10-mile zones. Analysis of the one mile zone around WNP-3 indicates evacuation times within 10 minutes of that calculated for the two mile zone. The exclusion area (0.8 miles) can be approximated con-servatively by the one mile zone data found on the attached table. l l
TOTAL AREAS WITHIN 5 MILES AREAS WITHIN 10 MILES DESCRIPTION WITHIN 1 2 MILES
- I H M N TOTAL H M N I l l TOTAL PERMANENT POPULATION 109 4,098 152 1,617 5,867 6,207 1.100 20 7,838 15,165 PERMANENT POPULATION VEHICLES 55 2,049 76 -
809 2,934 3,104 550 3,919 10 7,583 es f TRAtJSIENT POPULATION 1,829 1,160 361 257 1,260 3,038 2,173 1,580 1,07S 2,096 6,924 m g TRANSIENT POPULATION VEHICLES 1,165 780 181 129 880 1,970 1,437 790 538 1,298 4,063 GENERAL POPULATION 1,938 5,258 513 257 2,877 8,905 8,380 2,600 1.095 9,934 25,706 TOTAL VEHICLES 1,220 2,829 257 129 1,689 4,904 4,541 1,340 548 5,217 11,646 NOTIFICATION TIME MINUTES 30 30 30 30 30 30 30 30 30 30 30 m i PERMANENT POPULATION EVAC. 1:05 1:10 1:00 - 1:10 1:10 2:00 1:20 y l Tit,'E NORMAL CONDITIONS pf,,,,) 1:35 2:00
,I GE ,ERAL POPULATION EVAC, rp 1:10 1:15 1:10 1.05 1:15 1:15 2:10 1:25 1:20
- TIME NORMAL CONDITIONS 1:35 2:10
( /
- oS) _
FERMANENT POPULATION EVAC. 1:15 1:30 1:05 - 1:45 1:45 3:05 1:50 2:25 TIME ADVERSE CONDITIONS - 3:05 (i:os) AD ERSE O :20 1:35 1:30 1:25 2:50 2:50 3:25 2d5 2:10 I, ITIO S 2:55 3:25 (Z: /o) CONFIRMATION TIME MINUTES 60 60 60 60 60 60 60 60 60 60 60 SPECIAL POPULATION EVAC. TIME-NORMAL CONDITIONS 1:40 - - 3:30 3:30 1:40 3:30 i - - 3:30 SPECIAL POPULATION EVAC. TIME-ADVERSE CONDITIONS 2:50 - - 6:00 6:00 2:50 - - 6:00 6:00 HRS: MINS
- O '" #* #" # ' '
TABLE 12-5
SUMMARY
OF RESULT OF EVACUATION TIME ANALYSIS r, w w ar-s (me<p sey fiepaashes s Pla- ',
Attachment 1-52 2266A Question No. 311.3 You state that your analysis for aircraft hazards is forthcoming. (3.5.1.6) Provide a schedule for furnishing this information.
Response
The aircraft hazards analysis results will be supplied by December 30, 1982. l l l l l l
I Attachment 1-53 ) 2266A l Question No. 410.1 Provide or reference a discussion of the testing and inspection (3.4.1.2) to be performed to verify that the groundwater drainage system capability and reliability are met and the instrumentation and control necessary for proper operation of the system are adequate.
Response
A discussion of the test performed to verify the design adequacy of the groundwater drainage system (GWDS) and establish the basis for the frequency of inspection required is provided in Subsection 3.4.2. The in-service surveillance requirements of the groundwater drainage system is specified in Subsection 3.4.2.2. t l
Attachment 1-54 2266A Question No. 410.2 Table 3.5.1-1 has several columns that have "under investigation" (3.5.1.1) listed instead of the necessary data. Provide this data or a schedule for providing it.
Response
The missile evaluation for WNP-3 is currently being reviewed by the Supply System. This information will be available in January 1983. l w - - , - .--- __._7 - -- - ..,,,.-r-,._
Attachment 1-55 2266A Question No. 410.3 Provide or reference the following as specified in Regulatory (3.5.1.1) Guide 1.70: A tabulation showing the safety-related structures, systems, and components outside containment required for safe shutdown of the reactor under all conditions of plant operation should be pro-vided and, as a minimum, should include the following:
- 1. Locations of the structures, systems, or components.
- 2. Applicable seismic category and quality group classifica-tions (may be referenced from Section 3.2).
- 3. Sections in the SAR where descriptions of the items may be found.
- 4. Reference drawings or piping and instrumentation diagrams where applicable (may be referenced from other sections of the SAR).
- 5. Identification of missiles to be protected against, their source, and the bases for selection.
- 6. Missile protection provided.
The ability of the structures, systems, and components to with-stand the effects of selected internally generated missiles should be evaluated.
Response
The evaluation of WNP-3 for missiles outside containment is cur-rently being reviewed by the Supply System. This information will be available by January 1983. l l l
Attachment 1-56 2266A Question No. 410.4 Provide or reference the following as specified in Regulatory (3.5.1.2) Guide 1.70: A tabulation showing the safety-related structures, systems, and components inside containment required for safe shutdown of the reactor under all conditions of the plant operation, including operational transients and postulated accident conditions, should be provided and, as a minimum, should include the following:
- 1. Location of the structure, system, or component.
- 2. Identification of missiles to be protected against, their source, and the bases for selection.
- 3. Missile protection provided.
The ability of the structures, systems, and components to with-stand the effects of selected internally generated missiles should be evaluated.
Response
The evaluation of WNP-3 for missiles inside containment is cur-rently being reviewed by the Supply System. This information l will be available for January 1983. ( 1 l l
Attachment 1-57 2266A Question No. 410.5 Table 3.5.1-3 states the information for the dry cooling tower (3.5.1.4) enclosure will be provided later. Provide this information or a schedule for furnishing it.
Response
The information indicated as later in Table 3.5.1-3 is: Dry Cooling Tower Enclosure 28 day Structure Minimum Thickness Concrete Strength Wall 2'-0" 50000 psi Roof Steel Grating N/A FSAR Table 3.5.1-3 will be amended to reflect this information. l
l \- :
~
.- 1547W-2 WWF-3 $ g/g7. 'T yggg ifI TABLE 3.5.1-3 TORNADO MISSILE GNCRETE BARITFE 28-Day Structure (l) Minism Thichees (2) Concrete Strensch Shield Building Cylindrical Wall 3'-0" 5000 poi Dome 2'-6* 5000 poi l Reactor Auxiliary Buildins '
~
1 Wall 3'-6* 5000 poi Roof 2' M 5000 poi Fuel Handlins Buildins l Vall 3'-6* 5000 poi Reef Slab 2'-0" 5000 poi condensete Storate Task Enclosure Wall 3'-0" 4000 poi Emot Steel Grating N/A Dry Coolins Tower Enclosure Q 'f- 0" vall Q yaco PS* I nf
@ s+=t Gndi 3 Notes:
l 1) The systems and componente protected by these structures are idehtified in Table 3.2-1.
- 2) See Subsection 3.5.3 for barrier design details.
f. ( 3.5-13 _m-- - . - - - - - - ,. _ -m--- -- - - - - - - - = - -
Attachment 1-58 2266A - R.uestion No. 410.6 The externally generated missile protection analyses should take (3.5.1.4) into account the effect on ventilation openings in the various facility buildings housing essential shutdown equipment. Reference or provide a discussion addressing this subject.
Response
The ventilation openings in the Reactor Auxiliary Building and Dry Cooling Tower Control Building housing essential shutdown equipment are all provided with steel grating for protection from externally generated missiles. The type of steel grating used for missile protection of ventilation openings is identi-fied by that described in Subsection 3.5.3.1.2. The barrier design procedure using steel grating is discussed in Subsection 3.5.3. Section 9.4 also contains discussions of missile protec-tion for the ventilation openings. i
l l Attachment 1-59 2268A i Question No. 410.7 This section does not provide the detail required by Regulatory (3.5.2) Guide 1.70 which states that it should be demonstrated that safety-related structures, systems and components are adequately protected against very low probability missile strikes by physi-cal barriers or protective structures. According to the Stan-dard Review Plan (NUREG-0800) this should even include such elements as essential service water intakes, buried components, and access openings and penetrations in structure. Provide or reference this level of detail for this FSAR section.
Response
A conplete response to this question will be available by Decenter 1982. l 1
Attachment 1-60 2268A Question No. 410.8 Subsection 9.2.5 does not define the number of cells per cooling (9.2.5) tower train, nor does it contain Figures 9.2.5-2a through 9.2.5-2d. Confirm the date by which you intend to supply this information.
Response
Each cooling tower train is divided into 10 cells. Figures 9.2.5-2a through 9.2.5-2d will be provided by February 1983. The FSAR will be updated to reflect this response. l l
N
~ WNP-3 FSAR Q410.8 9.2 5 ULTI". ATE EAT SIR If 2- .,
9.2.5.1 Design Basis The Ultimate Beat Sink (CES) provides heat rejection from the Componen: Cooling Water System (CC;;5) to the atmosphere during safe plant shutdown or accident conditions. The description of the Co=ponen: Cocling Water Sys:en is given in S.:bsection 9.2.2. The UES operates in eenjun:: ion vi:h the CCW ha: Exchangers, which rejects hea: through :he Service Wa:er Sys:e= (s ee Subsee: ion 9.21) during no: mal opera: ion and refueling. Ihe L~d5 co: plies with regulatory positions defined in NBC Regulatory Guide 1.27 GDC 2, 4 and 44 and 3 P ASE 9-2. j The UES is capable of providing sufficient heat rejection during the most unfavorable anticipated meteorological conditions (as discussed in Subsection 2.3.2.1) for an indefinite period of time in order to: a) Permit safe shutdown and cooldown of the plant and maintain the plant in a safe shutdown condition. b) In the event of an accident, permit control of that accident safely. The UES is designed to ASE III Safety Class 3, seismic Category I, IPPSS Quality Class I requirements. The UBS is capable of withstanding the effects of natural phenomena as well as loss of offsite power coincident with a single
, failure. UES component design data is given in Table 9.2.5-3 The failure of any single mechanical or electrical component of the UES vill not result in a loss of the heat sink safety function.
9.2.5.2 System Description The preliminary design of the UES is comprised of two independen: 100 percent capacity trains. Each train has a cooling tower of an air-cooled heat exchanger tiype with CCWS fluid passing through the tube side and air over the extended surface of the tubes. Cbucrete structures provide seismic and missile protection as well as structural support for heat transfer equipment. l As shown on Figure 9 2.5-1 Train A and Train B towers are separated by a concre:e shear vall and respective electrical equipment rooms. Each cooling tower train is divided into s. T4ch cell is separa:ed by a concrete shear vall and includes tube bundles, heat retention doors or louvers and induced draf: fans. Tans are controlled to maintain appropriate CCk'S vater temperatures. Major component design data is presented in- Table 9.2.5-3. The UES is capable of operating under a range of heat loads during nor=al and emergency conditions. At end of life conditions and 101.5T dry bulb ambient temperature, each train has a design heat rejection capability of 180 x 106 Btu /hr with 11,000 sp= CCL'S flov entering at 1537 and leaving at 1207. This capability exceeds the max
- cue expected heat rejee: ion requirements as discussed below.
- 9.2-44
. . .. . . . . < ,. . . . .....:............~....-... .. .'
- 1617W-2 wur-3 13 FSAR During normal operation the UBS in conjunction with the CCRS heat exchanger, can reject the maximum normal heat loads while maintaining CCES tecperature at '
or below 957. Table 9.2.5-1 presents expected tower perf or=ance requirenants during normal conditions. Parametric performance expectations for a range of ambient te=peratures with and without heat rejection to the S'n'S under normal and energency conditions is presented in Iable 9.2.5-2. Daring energency operation the OES provides sufficient coeling te safety-related heat loads identified in Subsection 9.2.2. In the event of the lors of one train for any reason, the redundant train can reject the maximuz instantaneous beat load. { Non-essential heat loads can be manually aligned as reactor decay heat affords spare UES capacity. The appropriate controls as discussed in Section 7.3 f pe'rmit the operator to monitor and select these heat load alignments. 9.2.5.3 Safetv Evaluation The CES with one tower is designed to meet maximum heat load operating alone following an accident. Its long term heat rejection capacity is sufficient to mitigate a postulated LOCA and return the containment pressure and temperature to ambient conditions. '- . The resul f an analysis of the 30-day period following a design basis accide are found in Tables 9.2.5-4 and 9.2.5-5 and Figures 9.2.5-2a through 9.2.5 2d i is analysis has determined the ' total heat rejected, the sensible heat re e ed, the station auxiliary system heat rejected, and the decay haat release rom the reactors.
. As per Branch Technical Position ASB-9-2, the decay heat curves f or fission products and for heavy elements were obtained using the assumptions and uncertainties set forth in the October 1973 draf t proposed ANS standard,
" Decay Energy Release Rates Following A Shutdown Of Uraniun-Fueled 'Iberaal ~
Reactors" ( ANS-5), to establish the heat input due to decay of radioactive material. An equilibrium fuel cycle and an increase in the calculated heat inputs ware assumed as f ollove: a) For the time interval O to 103 seconds, 20 percent was added to the heat released by the fission products to account for uncertainties in their nuclear properties. b) For the time interval 103 to 107 seconds,10 percent was added to the heat released by the fission products to account for uncertainties in their nuclear properties. c) .For the time interval O to 107 seconds, the heat released by the heavy elements was calculated (using the best estimate of the production rate f or each unit) and 10 percent was added to account f or uncertainties in their nuclear properties. e . . i l
f Attachment 1-61 2268A ,
! Question No.
l 410.9 Subsection 9.2.6 does not contain a storage facility failure ,
- (9.2.6) analysis. Provide this analysis or a date by which.it will be supplied.
Response , The condensate storage facility failure analysis has been in-J cluded in the Auxiliary Feedwater System Availability Analysis L provided in FSAR Appendix 10.4.9A. i I i I l i 'l l i i i i I
Attachment 1-62 2268A Question No. i 410.10 Subsection 9.2.8 does not contain Figure 9.2.8-1; provide this (9.2.8) drawing or a date by which it will be supplied.
Response
The Plant Makeup Water System flow diagram was included in the FSAR as Figure 9.2.1-3. This figure number will be revised to Figure number 9.2.8-1. The FSAR will be updated to reflect this response. 1 I l 1 - --. -__ - __ __ __ _. __ _
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-63 2268A Question No.
410.11 Section 9.4 does not include piping and instrurisentation diagrams (9.4) for any of the systems discussed; provide these diagrams or a date by which they will be supplied.
Response
The piping and instrumentation diagrams for the MVAC Systems discussed in Section 9.4 are the air flow diagrams shown on Figures 9.4.1-1 through 9.4.7-1 and the respective instrument schematic-logic diagrams are shown in Section 7.3 on Figures 7.3-4 through 7.3-92. This FSAR will be amended to include the above response.
-' ~ ~
5 . WNP-3 . 128 tv-1 FSAR g g;g, ll 9.4 lfI AIR CONDITIONING, _MEATING,_ COOLING _ AND VENTILATION SYSTEM 5 I The f ollowing sections describe the heating, ventilating and air conditioning , (HVAC) systems serving the plant during normal and ener5ency operating conditions. Table 9.4.1-4 provides an FSAR location guide f or the various HVAC and related syscame. Systems are designed to provide a suitable environment f or equipment and personnel. Ventilation sones and air distribution are erranged so that the ventilation air is drawn from clean areas to. areas of potentially greater radioactive contamination to final filtration and w/4f ust omhaust. De ventilation systems $are shown on Figures 9.4.1-1 through g/ 9.4.7-14 Table 9.4.1-1 indicates cno space temperatures in the plant during ( ' normal plant operation. System's equipment classifications are listed 'in Table 3 2-1.
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9.4.1 CONTROL ROOM AREA VENTILATION SYSTIN The Control Boon HVAC System ie designed to control the environment of the Control Roon Envelope as defined in subsection 6.4.2.1. De control Room
, Envelope will be referred to as the Control Boon". Se Control Room Area!
Vontilation System is in compliance with R6gulatory Guide 1.78 and 1.95 , requirements. he Control Room amargency Air Cleaning 1 kits are designed and , constructed in accordance with the recommendations of Regulatory Guide 152, with exception as delineated in Table 6 5-2. i 9.4.1.1 De sian Ba ses , The Control Roos BVAC System is designed tot a) Exclude entry of airborne radioactivity into the Control Boos and i remove radioactive material from the Control Roos environment such that 1 L.L61.. J... L. L L. . i L.- r.. .....L 1. 1 LL1 8.... 1 5..le . Criterion 19 ( Appendix A of ,10CTR$0) limits. b) Hatatain the Control Room sabiest temperature to assure personnel ' c omfort during normal plant operation as shown in Table 9.4.1-1, equipment qualification temperatures are addressed in section 3.11. c) Permit personnel occupancy and proper functioning of instrumentation and controls during all normal and design basis accident conditions a ssuming a single active f ailure. , d) Withstand the effects of tornado and tornado missile. Se HVAC system, with the . exception of some ductwork located in office areas and eastgency living quarters, is designed to seisaic Category I requirements. j e) Provide Control Room personnel protection by detection and limit of the introduction of smoke and chlorine into the Qantrol Room. k f) Provide accessibility f or adjustment and periodie inspections and testing of the system components to assure continuous functional g reliability. a,dr% uspecfi+e imbmeEschemaNc-k J k AN"*'* are shown m Sec} 7 Son Rsus7.3^ & & T3^2-
, -64 2268A Question No.
410.12 Subsection 9.4.7 states that the CCWS Dry Cooling Towers (9.4.7) Electrical Equipment Room Ventilation System design is concep-tual only. Confirm the date by which you intend to supply this information.
Response
The CCWS Dry Cooling Towers Electrical Equipment Rooms Ventila-tion System design is not final at this time. Subsection 9.4.7 will be amended after completion of the design and will be sup-plied in January 1983.
Attachment 1-65 2268A Question No. 410.13 As per Regulatory Guide 1.70, Revision 3, provide a discussion of (10.4.7) the piping analysis, including forcing function, or results of test prograns performed to verify that the uncovering of the feedwater lines could not occur or that the uncovering would not result in unacceptable damage to the system.
. Response I
The information requested in Regulatory Guide 1.70, Revision 3 applies to steam generators utilizing a top feed design. WNP-3 utilizes an integral economizer type steam generator. There-fore, this item is not addressed. l
-66 2268A Question No.
410.14 Subsection 10.4.9.3 discusses the design to prevent water hammer (10.4.9.3) in the pipe routing of the auxiliary feedwater system to the steam generators. It further states that tests acceptable to the NRC will be performed to verify unacceptable water hammer will not occur. Describe the proposed tests, how they will be conducted and when they will be conducted.
Response
A description of the proposed tests to verify acceptable water hammer mitigation in pipe routing of the auxiliary feedwater system to the steam generators will be available after July of 1983. Results of testing at other utility projects are being closely followed to determine test requirements. At this time it is believed that any necessary testing would occur during pre-core hot functional.
Attachment 1-67 2268A Question No. 421.1 Table 7.3-20 is shown as "to be supplied later". Provide either (7.3) the table or a date by which it will be supplied. Response , Table 7.3-20 is part of the description of the Dry Cooling Tower Electrical Equipment Room HVAC System. The design of the system is not finalized at this time. Upon completion of the design the following steps will be taken to update Chapter 7: a) Complete Subsection 7.3.1.1.4.8, "CCWS Dry Cooling Towers Electrical Equipment Rooms HVAC"; The description of the system will be given including initiating circuits, logic, bypasses, interlocks, sequencing, redundancy, diversity and actuated devices.
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b) Add Subsection 7.3.1.3.19 - The Final System Drawings in-cluding Instrument Schematics and Control Logic Diagrams and Instrument Location Arrangement Drawings will be given. c) Complete Table 7.3-20, "CCWS Dry Cooling Tower Electrical Equipment Room HVAC Failure Mode and Effects Analysis". This information will be supplied in January 1983. I _ _
awq) FSAR a #la r Ik 5 , Th3 control logic coacciated with the Diesel Generstdr Area HVAC System and ) typical control wiring diagrams are provided in Subsection 7.... 313 j Bypasses l 1 { i Components of the Diesel Generator Areas HVAC System are monitored continuously for bypassed or inoperable conditions and their status is indicated on both a system and a component basis on the Bypassed Inoperabl Status Panels in the Control Room. System status is alarmed in the Main e Control Board (C3-1). Bypassed-Inoperable Status Panels and conformance to Regulato . Redundancy Each diesel generator area has its own air handling unit HV2A(3). Upon failure of the operating Diesel Generator Area HVAC Systet, an alarm is sounded in the Main Control Room. Sequencing The Diesel Generator Areas HVAC System is automatically loaded onto Load Blo V of its associated diesel generator following a loss of offsite power. See Section 8 3 for discussion of the diesel generator load sequencing. and are not utilized following LOOSP.The non-safety electrical reheat coil Supporting Systems The following systems support the Diesel Generator Areas HVAC Systaat
- 1) Emergency AC Diesel Generators 2)
- 3) Reactor Auxiliary Building Main Ventilation System 480V AC System
- 4) Vital AC System 7.3.1.1.4.8 CCWS Dry Cooling Towers Electrical Equipment Rooms HVAC u.,,,,
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175W-6 WHP-3 FSAR y 7.3.1.3.17 Safety-Related 125 Volt DC System 1 Refer t o Chapter 8. 7.3.1.3.18 Typical Control Wiring Diagrama (CWD) Typical DC's for systen equipment and/or components are:
- 1) Solenoid Valve Figure No.
- 2) Air Opera t ed Va lve 7.3 .
- 3) Motor Operated Valve 7.3-87 7
- 4) Pump or Fan ( >100 Horsepotter) 7.3-88
- 5) Pump or Fan (( 100 Borsepower) 7.3-89 7.3-90 y a ,.s.t.s.t9 p, ~ l 5p k D~ ' y s.
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Attachment 1-68 2268A Question No. l 421.2 Section 7.6.2 contains no analyses of instrumentation installed (7.6.2) to prevent or mitigate the consequences of cold water slug injections and overpressurization of low-pressure systems; reference or provide these analyses or show these analyses are not applicable to WNP-3.
Response
Instrumentation and equipment to protect NSSS systems from cold water slug injections and overpressurization of low-pressure systens are within the CESSAR-F scope as noted in Section 7.6.2.1.6 of the WNP-3 FSAR. Therefore, this issue has been reviewed on the CESSAR-F docket and accepted in NUREG-0852,
" Safety Evaluation Report Related to the Final Design of the Standard Nuclear Steam Supply Reference System CESSAR System 80".
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-69 2268A I
Question No. 430.1 The Utility Grid Description (Subsection 8.1.1) and the Offsite (80) Power System (Section 8.2) should be revised to reflect the can-cellation of WNP-5 and any consequential changes in the BPA grid structure.
Response
The Utility Grid Description (Subsection 8.1.1) and the Offsite Power System (Section 8.2) text and figures have been revised to reflect the termination of WNP-5 and provided in Amendment No.1 to the WNP-3 FSAR.
. -70 2268A Question No.
430.2 As per Regulatory Guide 1.70, Revision 3, specifically provide or (10.4.1) reference discussions of the following:
- 1) The anticipated inventory of radioactive contaminants in the main condensers during operation and during shutdown,
- 2) anticipated air leakage limits,
- 3) control functions that could influence operation of the pri-mary coolant or secondary systems,
- 4) protection of safety-related equipment from flooding result-ing from failure of the condenser,
- 5) a procedure to repair condensate leaks and
- 6) the length of time the condenser can operate with degraded conditions without affecting the condensate /feedwater quality for safe operation.
Response
The main condenser serves no safety function. During normal operation and shutdown, the main condenser does not contain radioactive contaminants. Radioactive contaminants can only be present through primary-to-secondary system leakage due to a steam generator tube leak. Therefore, the six items as required per Regulatory Guide 1.70, Revision 3, are not applicable to the normal operation of the main condenser, as described in Sub-section 10.4.1. The following are responses to each of the specific items:
- 1. "The anticipated inventory of radioactive contaminants in the main condensers during operation and during shutdown."
Subsection 10.4.1.3, " Safety Evaluation" indicates that "During startup, normal operation and shutdown, the main condenser contains no radioactive contaminants. Non-condensible gases will be monitored for radioactivity prior to being discharged to the atmosphere as discussed in Sub-section 10.4.2. A full discussion of the radiological as-pects of primary to secondary leakage, including anticipated operation concentrations of radioactive contaminants is described in Section 11.1."
Attachment 1-71 2268A Question No. 430.2 2. " Anticipated air leakage limits" (contd.) The Main Condenser Air Evacuation System, described in Sub-section 10.4.2, evacuates the main condenser and turbine steam space during plant startup and thereafter maintains the condenser design vacuum of 2.24/3.08/4.14 in. of Hg. absolute in LP/IP/HP condenser shell during normal plant ! operation. The system maintains this vacuum by continuously I removing non-condensible gases and air in leakage from the three condenser shells. The Condenser Air Evacuation Sys-tem, as well as the Main Condenser, do not perform a safety function.
- 3. " Control functions that could influence operation of the primary coolant or secondary systems."
There are no controls in the operation of the condenser that could influence the operation of the primary coolant sys-tem. Operation of the condenser in relation to secondary systems is discussed in Subsections 10.4.1.2 and 10.4.1.3.
- 4. " Protection of safety-related equipment from flooding re-sulting from failure of the condenser."
l No protection of safety-related equipment from flooding re-l sulting from failure of the condenser is required. There is no safety-related equipment located in the Turbine Building.
- 5. "A procedure to repair condensate leaks" Subsection 10.4.1.5 describes methods of detecting and lo-cating condenser tube leaks. The following is the procedure to be followed to repair condensate leaks: In case of in-leakage to the condenser, the affected condenser section is removed from service, drained of cooling water and the leak-ing tubes are located and plugged. The affected condenser section is returned to service when repairs are completed.
Subsection 10.4.1.2 will be amended to reflect this response.
- 6. "The length of time the condenser can operate with degraded conditions without affecting the condensate /feedwater quality for safe operations."
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Attachment 1-72 2268A Question No. 430.2 The main condenser is not required to maintain the integrity (contd.) of the Reactor Coolant Pressure Boundary, and the ability to shutdown the reactor and maintain it in a safe shutdown condi-tion. Condenser operation with degraded conditions, such as inleakage, loss of vacuum, etc. are discussed in Subsection 10.4.1.3. l
WNP- 3 13 29'J-2 FSAR Gw.shw 430 A. . l the reaction of Hydrazine with f ree oxygen. By maintaining the proper ich chemistry, concentrations of impurities will not occur. Thu s , no ha rm f ul ? impurities will form in any scratch, nondatectable weld imperf ection or dent. Corrosion is also controlled by the addition of chemicals and ion exchange in the condensate demineralizers. Stress corrosion cracking will not occur because of the low operating temperatures, low chloride level and the use of stress relieved tubes. The tube material is stainless steel which has the highest resistance to erosion. The main condenser hotwell serves as a reservoir which supplies the low pressure condensate pumps. Minimum hotwell storage capacity is sufficient for providing adequate volume for condensate and f eedwater system surge protection. The storage capacity is about 116,000 gallons of water, equivalent to five minutes of condensate flow at full load operation. The condenser hotwell level is maintained by makeup from the condensate system (Subsection 9.2.6). The main condenser is serviced by an automatic makeup and letdown system to sustain a normal level in the hotwell. At low hotwell level a control valve opens to admit condensate from the demineralized water storage tank. At high hotwell level another control valve opens to discharge excess condensate f rom the system directly to the demineralized water storage tank via the low pressure condensate pump. The main condenser is cooled by circulating water (Subsection 10.4.5) which flows in series through tubes in the three condenser shells. The use of divided water boxes on each shell pemits isolation of one-half of the total circulating water flow through each shell. Utis permits access to
, the isolated water box on each shell for repair and/or inspection while M one-half of the circulating water flows through the c9er water box.
l Equalizing pipes between hotwells are provided so that only one hotwell level l control is required. Crossover ducts are provided between condenser shells for equalizing the exhaust steam flow. 10.4.1.3 Saf ety Evaluation Du' ring s tart up, normal operation and shutdown, the main condenser contains no radioactive contaminant s. In the event of primary to seconda ry tube leakage , radioactive contaminants will be present in the shell side of the steam generator. ~his will eventually enter the condenser. Noncondensible gases i will be nonitored for radioactivity prior to being discharged to the
.itnosphere as discussed in Subsection 10.4.2. A full discussion of the ridiological aspects of primirv to secondary leakage , including anticipated operating concentrations o f radioactive contaminant s is described in Section 11.1.
The n sin condenser is not required to maintain the integrity of the 'leactor # j g g Coolant Pressure Bounda rv , and the ability to shutdown the reactor and maintain it in 1 safe shutdown condition. f L cue f .% b do ffu M M sc/, fiu. cpc/t/ CN/4asu GecM,h e'se t h -- L-- se n.u , d u . 4 cae og mhr M & M. *du ~e kca M ~lPt agd. CA 4/uld encl,emc n , u hw)sw.x wlm y.>s ee o~jlM- - 10.4-2
Attachment 1-73 2268A Question No. 440.1 Provide the post-LOCA design heat load for the shutdown cooling (5.4.7.1.3) heat exchangers. Also, Note 2 in Table 5.4.7-2 should be re-placed by specific information rather than a general reference to the applicant's SAR.
Response
The Supply System, in conjunction with Combustion Engineering and Ebasco Services Inc., is preparing this information for sub-mittal. The response should be complete by December 1982. i l
Attachment 1-74 2268A Question No. 451.1 As per Regulatory Guide 1.70, provide estimates of the weight (2.3.1.2) of the 100-year return period snowpack and weight of the 48-hour Probable Maximum Winter Precipitation for the site vicinity. Using these estimates, provide the weight of snow and ice on the roof of each safety-related structure.
Response
The weight of the 100-year return period snowpack is estimated to be 19.7 psf corresponding to a snow depth of 25.2 inches. The weight of the 48-hour Probable Maximum Winter Precipitation (PMWP) is estimated to be 147.1 psf corresponding to a water depth of 28.28 inches. These estimates together with the ef-f ects on the roof of the safety-related structures are presented in Subsection 2.4.2.3. l
1 2193W-4 WNP-3 ON/ 1 lo(7 3 TABLE 15-2 (Cont'd) 15.6.2 - Double-Ended Break of Letdown Line Outside Contairment Pre existing Event Isduced No Spike Spike Seike Kr-85m 2.4(1) same same Kr-85 5.2(-1) same same Kr-87 1.3(1) same same Kr-88 3.6(1) same same Ze-131m 3.2(0) same same Ze-133 3.4(2) same same Ze-135 5.7(1) same same Ze-138 8.7(0) same same I-131 5.3(0) 6.7(1) t-M D6C-, I-M I-132 1.1(0) 1.4(1) -@ I-133 5.7(0) 7.3(1) - l I-134 8.7(-1) 1.1(1) - l I-135 3.-9(0) 5.0(1) - 15.6.3 - Steam Generator Tube Zupture With Loss of Offsite Power (2 hour) Kr-85m 9.3(-1) same same Kr-85 3.5(-1) same same Kr-87 8.5(0) same same Kr-88 2.5(1) same same Ze-131m 2.2(1) same same Ze-133 2.2(2) same same Ze-135 3.9(1) same same Ze-138 5.9(0) .same same I-131 9.1(0) 1.2(2) l T@) M Tb I-132 1.8(0) 2.4(1) - I-133 9.9(0) 1.3(2) -Gt I-134 1.5(0) 1.9(1) - I-134 6.7(0) 8.6(1) -
& Dee. Ege <. h.J C w.-te.
F . 15.1-12
s e S 2193W-5 WNP-3 FSAR v TA3LE 15-2 (Cont'd) . 15.6.3 - Steam Generator Tube Rupture (8 hour) Pre-existing Even: Induced No Spika Spike Spike Kr-85s 2.3(1) same same l Kr-85 9.0(-1) same same Kr-87 9.7(1) same same Kr-88 2.5(1) sama same Ie-131m 2.5(0) same same Ie-133 2.6(2) same same Ie-135 4.4(1) same same
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Ia-138 6.7(0) same same , I-131 9.1(0) 1.2(2) l Sh DO IN t I-132 1.8(0) 2.4(2) -@ ~ I-133 1.0(1) 1.3(2) I-134 1 5(0) 1.9(1) - I-135 6.7(0) 8.6(1) - 15.6.5 - Loss of Coolant Accident . Curies 2 hr 8 hr 30 Day Nuclide Containment Atmosphere Release Release Release Kr-85m 2.95(7) 3.7785(3) 1.1210(4) 1.5202(4) , Kr-85 9.36(5) 1.3890(2) 6.4622(2) 2.8514(4) Kr-87 5.41(7) 5.0429(3) 8.0183(3) 8.1223(3) l Kr-88 7.73(7) 9.1264(3) 2.2101(4) 2.5322(4) Ze-131m 8.24(5) 1.2201(2) 5.6319(2) 1.2423(4) Ia-133 2.37(8) 3.4987(4) 1.5993(5) 1.9821(6) Ie-135m 4.78(7) 1.3879(3) 1.3941(3) 1.3941(3) Ze-135 4.24(7) 5.8591(3) 2.1623(4). 4.2706(4) Ze-137 2.09(8) - - - l Ie-138 1.89(8) 5.0640(3) 5.6779(3) 5.0782(3) 1-131 2.93(7) 1.7476(2) 4.5884(2) 6.1862(3) I-132 4.28(7) 2.5461(2) 6.6189(2) 4.4326(3) 1-133 5.90(-) 3.4607(2) 8.4428(2) 2.2769(3) I-134 6.38(7) 2.5614(2) 2.8299(2) 2.8320(2) I-135 5.50(7) 3.0921(2) 6.4080(2) 9.5351(2)
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TABLE 15-3 (Cont'd) 15 .6 . 2 Double-Ended Break of a Le cdown Line outside Containment No. Iodine Seike Whole Body Thyroid E A3 (2 hr) 2.7(-2) 8.4(0) LPZ (8 hr) 3.7(-4) 1.2(0) Fr ee,xistina Soike Whole Body Thyroid EAB (2 hr) 1.3(-1) 1.l(2) LPI (8 hr) 1.8(-2) 1.5(1) Accident Induced Spike EAs (*2.hr) T-kd) ( Mcer]4 4.s(t) LP7- (5 br) 15.6.3 3 1(-3) Steam Generator Tube Rupture With IAF t.o(l) l l No. Iodine Spike l h ole Body Thyroid E AR (2 hr) 1. 6 2(-2) . 4.0(0) LPI (8 hr) 2.50(-3) 5.7(-1) Pre-exis tina Soike Whole Body Thyroid EAB (2 hr) 6.7(-2) 5.2(1) LPZ (8 hr) 9.5(-3) - 7.3(0) o Accident Induced Spike D 6.th) EAPs (#2kr} 10(al) 15b) LP'A (t nr1 3.o(-5) 15.6.5 Loss of Coolant Accident (LOCL) Thyroid Whole Body EAB (2 hr) 2.5(2) 1.4(1) LP.~ (30 day) 1.7(2) 5.4(0) 15.6.5 30-Day Control Room Doses Following IDCA Thyroid Whole Body Skin 1.6 (0) 1.2(-1) 2.5(0) 7 L-15.1-15
Attachment 1-76 2268A Question No. 460.1 Supply information relating to the effluent radiation monitors (11.5.2.4.2) for steam generator blowdown flash tank vent and steam seal gland steam condenser ventilation which the FSAR indicates as later or provide a schedule for submittal of this information.
Response
The following is the information indicated as later in Sub-section 11.5.2.4.2: Steam Generator Blowdown Flash Tank Vent Radiation Monitor The steam generator blowdown flash tank vent radiation monitor provides plant operations personnel with an indication and record of contamination of the Secondary Steam System and the potential for release via the steam generator blowdown flash tank vent. This contamination could occur due to leakage of primary reactor coolant into the secondary coolant through a steam generator. This monitor is an ambient type monitor located next to the steam generator blowdown, flash tank vent line 68012-200 at EL. 423 ft. in the Turbine Building. The monitor is collimated with a lead shield to reduce the effect of background. The un-shielded portion of the detector has an unobstructed view of the vent line. The ambient monitor is described in Subsection 11.5.2.3. The measured activity level is automatically transmitted to the system computer where it is recorded and available for display. If the activity exceeds setpoints an annunciation is made through the system CRTs and event typer. Receipt of these alarms will alert the operators to the possibility of contamina-tion of the Secondary Steam System and indicate the need for additional sanpling and further action. l The alarm setpoints are selected above plant background to give ! the greatest sensitivity for possible contamination without causing frequent false alarms. These setpoints may be adjusted continuously over the entire range of the monitor. Steam Seal Gland Steam Condenser Exhaust Radiation Monitor The steam seal gland steam condenser exhaust radiation monitor provides plant operations personnel with an indication and record of contamination of the Secondary Steam System and the potential for release via the steam seal gland steam condenser vent. This contamination could occur due to leakage of primary reactor coolant into the secondary coolant through a steam generator.
Attachment 1-77 2268A Question No. 460.1 The monitor is an ambient type monitor located next to the vent (11.5.2.4.2) line 6AE10-022 on EL. 455 ft. of the Turbine Building. (contd.) The monitor is collimated with a lead shield to reduce the ef-fect of background. The unshielded portion of the detector has an unobstructed view of the vent line. The ambient monitor is described in Subsection 11.5.2.3. The measured activity level is automatically transmitted to the system computer where it is recorded and available for display. If the activity exceeds setpoints an annunciation is made through the system CRTs and event typer. Receipt of these alarms will alert the operators to the possibility of contamina-tion of the Secondary Steam System and indicate the need for additional sangling and further action. The alarm setpoints are selected above plant background to give the greatest sensitivity for possible contamination without causing frequent false alarms. These setpoints may be adjusted continuously over the entire range of the monitor. FSAR Subsection 11.5.2.4.2 will be amended to reflect the addi-tion of this information. Information pertaining to the sensitivity of these monitors, now missing from Table 11.5-1, will be available by February 1983. [
WMp-3 g t.$g7, l I FSAR 1633W-4 /fq r pre-established setpoints an annunciation is made through the system I L CRTs and event typer. If the activity exceeds the high radiation alara setpoint, or if the monitor fails, as determined by the local microprocessor, a contact closure is made at the local alcroprocessor which is used to automatically terminate the waste gas discharge. The receipt of these alaras will ' alert the operators to analyse additional gas samples to determine the reason for the alare. The records of the total quantity of radioactive asterial released is used in writing the reports required by Regulatory Guide 1.21. The alars setpoints are selected in consideration of the requirement to prevent activity concentrations at the plant boundary or beyond from exceeding 10CFR20 limits, and to support the release limits set in the plant technical specification. The setpoints may be adjusted continuously over the entire range of the monitor. The range of this monitor uns selected to span the expected range of radioactive gas concentrations expected in the weste gas. - l e) Steen Generator 31oudown y1 ash Tank Vent Radiation Monitor
" """ } t sea hsed i f) Auxiliary Condensate Flash Tank Radiation Monitor The auxiliary condensate flash tank radiation monitse provides plant operations personnel with an indication and record of cont ==f== tion of the Auxiliary Steam System and the potential for release via various vents.in the Auxiliary Steam and Condensate System. This contaminatio'n could occur due to in-leakage from the various radioactive systems that are serviced by the Auxiliary Steam System. med io line 6 Mif.-200 o*s
%e 355 4. level .
This monitor is an ambient type monitor located next to the auxiliary condensate flash tgnk i ,t Rasctor Auxiliary Buildi The monitor ' is colliasted with'"les o educe the effect of background. The - unshielded portion of t detector has an unobstructed view of the flash tank. The ambient monitor is described in Subsection 11.5.2.3. The asseured activity level is automatically transmitted to the system computer where it is recorded and available for display. If the l activity exceeds setpoints an' annunciation is made through the system CITs and event typer. Receipt of these alaras will alert the operators to the possibility of contamination of the Auxiliary Steam and Condensate System and indicate the need for additional sampling and further action. The alara setpoints are selected above plant background to give the greatest sensitivity for possible contamination without causing frequent false alaras. These setpoints may be adjusted continuously over the entire range of the monitor. Eakaus} g) Steam Seal Cland Steam Condenser "n. deme Radiation Monitor we se.. L s . J '2. 11.5-23
l y [ Insert 1 The steam generator blowdown flash tank vent radiation monitor provides plant cperations personnel with an indication and record of contamination of the S:condary Steam System and the potential for release via the steam generator blowdown flash tank vent.. This contamination could occur due to leakage ci primary reactor coolant into the secondary coolant through a steam generator. This son.itor is an ambient type monitor located next to the steam generator blowdown.. flash tank vent line 6BD12-200 at F.1 423 ft. in the Turbine Buildinit. The monitor is collinated with a lead shield to reduce the effect of backround. The unshielded portion of the detector has an unobstructed view of the vent line. The ambient monitor is described in Subsection 11.5.2.3. The measured activity level is automatically transmitted to the system computer where it is recorded and available for display. If the activity exceeds cctpoints an annunciation is made through the system CRTs and event typer. Receipt of these alarm will alert the operators to the possibility of contamination of the Secondary Steam System and indicate the need for additional sampling and further action. The alarm setpoints are selected above plant backround to give the greatest censitivity;for possible contamination without causing frequent falso alarms. These setpoints may be adjusted continuously over the entire range of the monitor. l e e l
3 In art 2 ' Tha steam seal gland steam condenser radiation monitor provides plant operations psrsonnel with an indication and record of contamination of the Secondary Steam Syctsm and the potential for release via the steam seal gland steam condenser vcat. This contamination co61d occur due to leakage of primary reactor coolant into the secondary coolant through a steam generator. Th3 monitor is an ambient type monitor located next to the vent line 6AE10-022 on El. 455ftof the Turbine Building. The monitor is collinated with a lead shield to reduce the effect of background. The un:hielded portion of the detector has an unobstructed vie rof the vent line. The ambient monitor is described in Subsection 11.5.2.3. t Th3 measured activity. level is automatically transmitted to the system computer where it is recorded and available for display. If the activity exceeds setpoints an annunciation is made through the system CRTs and event typer. Receipt of these cicrus will alert the operators to the possibility of contamination of the Secondary Steam System and indicate the need for additional sampling and further action.
- Th3 alarm setpoints are selected above plant background to give the greatest censitivity for possible contamination without causing frequent false alarms.
Th:se setpoints may be adjusted continuously over the entire range of the monitor. e 9 D
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Attachment 1-78 2268A
! Question No.
471.1 As per Regulatory Guide 1.70 indicate whether, and if so how, the (12.3.4) guidance provided by Regulatory Guide 1.97 has been followed concerning area radiation and airborne radioactivity monitoring instrumentation. Reference or provide this information.
Response
The guidance provided by Regulatory Guide 1.97 concerning area radiation and airborne radioactivity monitoring instrumentation will be addressed in the WNP-3 design. Tlie appropriate information will be supplied by February 1983. l 1 i t l
WNP-3 FSAR I 4 1647W-5 - iff f) To provide the capability to alarm, initiate isolation of the Control
\ Room Normal Ventilation System, and actuate emergency ventilation.
g) To provide Control Room Operators with information regarding airborne radioactivity at the two Control Rooms outside air intakes. h) To provide the capacity to alarm'and initiate isc.lation of the containment purge and/or Fuel Handling Building.
- 1) To provide post-accident monitoring of conditions inside the containment.
12'.3.4.2.2 Criteria for Location of Monitors Considerations for locating the airborne radiation monitoring system monitors are based on the following: a) Areas in which the airborne radioactivity can abruptly increase and in which personnel normally have access. b) Inside the containment for the purpose of monitoring unidentified leaks. c) In Control Room outside air intake ducts for post-accident habitability monitoring purposes. d) Ventilation exhausts from areas in.which spent fuel is handled. e) The consolidated ventilation exhaust from sections of the Reactor 4 Auxiliary Building. f) The consolidated ventilation exhausted from the plant. g) Monitors are located as close as possible to sample points to insure representative air samples. i h) In-plant airborne monitors have their sample points upstream of any REPA
- filters. .
'12.3.4.2.3 Descriptions of Airborne Radiation Monitors Airborne radiation moni' tors are used when the concentration of airborne radioactivity must be observed. A number of these monitors are formally part of the process and effluent radiation monitoring systems and are referenced' below but are described elsewhere in this FSAR. The rest of these monitors l are defin' e d as in-plant airborne radiation monitors and are described in this section and are listed in Table 12.3.4-2 T44 a<r&ar e ra ./>
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- 4 * **
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,, f, , , , .,
+* /*** ** Ne e... < 9 m 12.3.4.2.3.1.1 Auxiliary Building Airborne Radiation Monito i-oc.,,/ , ;, 4.,,, ,
The auxiliary building airborne radiation monitors provides a measurement and a record of the airborne activity present in the Reactor Auxiliary Building. These monitors draw samples and monitor them for radioactive particulates, and radioactive gases and collects samples of the halogen that are in the exhaust air duct. The samples are available for later laboratory analysis.
- 12.3-31
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,n.- -, - -
Attachment 1-80 2268A Question No. 471.3 Regulatory Guide 1.70 states that information on the auxiliary (12.3.4) and/or emergency power supply should be provided. This informa-tion has not been provided for the general area radiation moni-tors. Provide or reference this information.
Response
l The General Area Radiation Monitors are not safety-related and ! thus no emergency power supply is provided for them in the case of loss of offsite power. Protection of the memory of the microprocessors associated with the General Area Radiation Moni-tors is accomplished in the manner described in Subsection 11.5.2.1. FSAR Subsection 12.3.4.1.3.1 will be changed to reflect the information on the auxiliary and/or emergency power supply to General Area Radiation Monitors. l t i
unt.s G 47I3 - PsAa , 1647W-3 /4 12.3.4.1 3.1 donatal Area Indiation Monitora
~
b area radiation monitors provide an indication to plant personnel of the l radiation dose rate in selected areas of the plant both through the radiation < aenitoring system CETs and local displays. h usasureaansa made by these comitors are used primarily for personnali protection, with their use fer the surveillance of equipment being incid. ental. h re are 70 C.M. radiation detectors serving as area radiation detectors, each with as associated set of loc.a1 displays. The locations of thess' detectors and their displays are shown on Figures 12 3-14 thrassh 12.3-28e and cre listed in Tabla 12.3.4-1. N 70 radiation detectors are arranged in j fours, threes or pairs, each group'of which report to a single microproesseor. bre are a total of 23 microprocsasors for these detectors. ! g Except hysically for similar the multiple to thedetectors ambient and displays, monitor the area described monitors are in Subeestion 113231.
-- ^y -
e local displays associated with each radiation detector consists of an enalog meter, a rotating beacon light and a howler (audible alars). h cualog meter is in a well sounted box continuously displaying the seasured radiation does rate os a five decade scale. h rotating beacon is a well mounted unit which has an amber color and21s activated by the lower of the two radiation .alara set points, the radiation alert setpoint. h howler is a - wall nousted unit which produces a loud distinctive sound and is activated by the upper of the two radiation alata setpoints; the high radiation setpoint.
~
hoe displays are placed as follows, with respect to their associated radiation detector. The analog meter is placed along the access path to the cres of the detector at a point before the detector has been reached. m , rotating beacon is located in the vicinit;y of the radiation detector so that . t it is visible in the immediate area of the detector. The howler is located in the vicinity of the radiation detector so that its sound emanates from that cres.
~
The measured dose rate for each of the radiation detectors is automatically transmitted to the systsa's computer where it is recorded and available for display through the system's CI s in the Vain Control Room and in the Health j Physics office. h dose rate is also displayed locally through the analog teter. If the seasured dose rata exceeds pre-established setpoints, an canunciation is made through the system CRT, event logger and the particular radiation detector's beacon and howler as appropriate. De receipt of these alaras will alert than operator and plant personnel to the presence of an elevated radiation
- 1evel so that surveys, investigations and j scher actions as appropriata any be takan, The high radiation and radiation alert alarm setpoints control the howler and flashing light displays l
- -. respectively. N alara setpoints are selected primarily for personnel protection and denote significant levels or changes in the seasured close
! -. rata h. alert alars setpoints are selegted for each monitor on the basis of l individual conditions. Typically the high alara setpoint will be set at approximately 11/hr. l The two area radiation monitors (RI-R.'f0028A,5) provide the radiation c, instrumentation that will be used to meet the criticality accident monitoring requirements of 10CyR70, Section 70.24 for the storage area of the new fuel. 12.3-29 l e-y g4 --
--7_,---w
Insert 1 The General Area Radiation Monitors are not safety-related and thus'no emergency power supply is provided in the case of loss of'off-site power. Protection af the memory of the microprocessors associated with these monitors is accomplished ~ as described in Subsection 11.5.2.1.- e O O 1 ~ e h
Attachment 1-81 2268A Question No. 471.4 The Annual Whole Body Dose Table is not complete. Furnish this (12.4.3) information or provide a schedule for furnishing it.
. Response The missing information is given as follows:
ANNUAL WHOLE BODY DOSE (mrem / year) Maximum Point Maximum Point Restricted Area Exclusion Area Boundary Boundary 1670 ft. SE 0.8 Miles SE Contained Source 5.4 0.8 Submersion in cloud 3.2(-2) 2.6(-2) Inhalation 7.2(-2) 7.2(-2) Also, the information pertaining to construction of WNP-5 was deleted as no longer applicable. FSAR Subsection 12.4.3 will be amended to reflect this information. l i l
y -. ~ . . p 1649W-3
. WNP-3 FSAR I I l
d controls, described in Section 12.5, will be used to maintain occupational exposures to airborne radioactivity as low as is reasonably achievable. All individual exposures in excess of two MPC-hour / day or 10 MPC-hour / week will be assessed. If an individual's exposure exceeds 40 MPC+ours in any seven I consecutive days, an evaluation will be made and necessary actions taken to I assure against recurrence. Records will be kept of internal exposure assessments, evaluations and actions. 12.4.3 CUPATIONAL EXPOSURE 1 Estimated annual whole body doses at the restricted area boundary, ;:~ l onnassuts.i.asmoees and exclusion area boundary are given below. These doses are based on occupancy of 50 weeks per year, 40 hours per week, and include doses from contained sources, submersion in the airborne cloud, and inhalation of airborne material. a ANNUAL WHOLE BODY DOSE (area / year) - Maximum Point NP-5 Maximum Point Ra'stricted Area Co true on Exclusion Area Boundary Sit Boundary 1670 ft. SE 45f.W 0.8 Miles SE Contained Source 5.4 0.8 Submersion In ('..nd 3.2(-2) (1 r) C' .:- ,' .7.6 ( et) s Cloud . Inhalation -
'NWumb { '. -... ' Y.2 h2 ( ata ) fl " ' ' Y 'M e
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Attachment 1-82 2268A Question No. 471.5 The SRP (NUREG-0800) and Regulatory Guide 1.70 state that the (12.5.1) applicant should indicate whether, and if so how, the guidance of Regulatory Guides 8.2, 8.8, 8.10 and 1.8 has been followed aid where applicable, describe the specific alternative approaches used. Provide or reference a discussion of your speci)Ic conformance or non-conformance to the guidelines in these Regulatory Guides.
Response
FSAR Subsection 12.5.1.1 states that Regulatory Guides 8.2, 8.8, 8.10 and 1.8 (as modified by the position statement in Section 17.2) were utilized as guidance in establishing the WNP-3 Health Physics Program. Regulatory Guide 8.2, " Guide For Administrative Practices In Radiation Monitoring," provides rudimentary guidance in radia-tion monitoring practices for administrative and management personnel who may not be trained or experienced in radiation protection. The administrative and management personnel of the WNP-3 Health Physics Program are experienced radiation protec-tion professionals. While the general guidance of Regulatory Guide 8.2 is followed, it contributes little to the organization of the WNP-3 Radiation Protection Program. Regulatory Guides 8.8, "Information Relevant Tr, Ensuring That Occupational Radiation Exposure At Nuclear Power Station Will Be As Low As Is Reasonably Achievable," provides guidance on orga-nization, personnel and responsibilities to maintain occupa-tional radiation exposures ALARA. WNP-3 FSAR Subsection t 12.1.1.2 provides specific information on implementation of l Regulatory Guide 8.8 with regard to organization, personnel and
- responsibilities.
Regulatory Guide 8.10, " Operating Philosophy For Maintaining Occupational Radiation Exposures As low As Reasonably Achievable," like Regulatory Guide 8.8, states that the qualifi- ! cations of the Radiation Protection Manager (RPM) should be l those specified by Regulatory Guide 1.8 and that the RPM should be given sufficient authority to enforce safe plant operation. FSAR Subsection 12.5.1.2 states that the Health Physics / l Chemistry Manager is the Radiation Protection Manager of the plant. He reports directly to the Plant Manager who has the authority to control all plant activities (FSAR Subsection l 13.1.2.2.1). Conformance to Regulatory Guide 1.8 is discussed below.
Attachment 1-83 2268A Question No. 471.5 Regulatory Guide 1.8 " Personnel Selection And Training," (12.5.1) references ANSI N18.1-1971 for criteria for the selection and (contd.) training of personnel, except for the Radiation Protection Manager (RPM). ANSI N18.1-1971 has been superseded, first by 4 ANSI / ANS-3.1-1978 and more recently by ANSI /ANS-3.1-1981. The WNP-3 position on compliance with Regulatory Guide 1.8 with regard to selection and training of radiation protection personnel will be stated in FSAR Section 17.2. I
l l Attachment 1-84 2268A Question No. 471.6 Regulatory Guide 1.70 and the SRP (NUREG-0800) state that the (12.5.2) description of the health physics instrumentation should include the instruments sensitivity. You provided the type of radiation the instrument detects and not the instrument sensitivity in Table 12.5-1. Provide the requested information.
Response
Table 12.5-1 will be revised by June 1983 to add instrument sensitivity. l l l t
Attachment 1-85 2268A Question No. 471.7 Regulatory Guide 1.70 states that it should be indicated whether, (12.5.2) and if so how, the guidance provided by Regulatory Guides 8.3, 8.4, 8.8, 8.9, 8.12, 8.14, 8.15, and 1.97 has been followed. If this guidance has not been followed, the specific alternative methods used should be described. Provide or reference a dis-cussion of this information. Reponse Regulatory Guide 8.3, " Film Badge Performance Criteria," is not applicable to WNP-3 since WNP-3 will use thermoluminescent dosi-meters (TLD) for personnel monitoring. Conformance to Regulatory Guide 8.4, " Direct-Reading and Indirect-Reading Pocket Dosimeters," will be discussed in Sub-section 12.5.2.2.4 in a future amendment as indicated by the attached markup. Regulatory Guide 8.9, " Acceptable Concepts, Models, Equations, and Assumptions For A Bioassay Program," is not directly ap-plicable to the specification of health physics equipment, instrumentation or facilities. Regulatory Guide 8.9 is used for guidance in establishing the WNP-3 bioassay program as described in Subsection 12.5.3 as amended by FSAR Amendment #1. Regulatory Guide 8.12, " Criticality Accident Alarm Systems," is not applicable to Subsection 12.5.2 and is not included in the SRP (NUREG-0800) review criteria. Regulatory Guide 8.14, " Personnel Neutron Dosimeters," was used for guidance in evaluating neutron dosimeters for use at WNP-3. Subsection 12.5.2.2.4 will be revised to discuss conformance with Regulatory Guide 8.14 as indicated by the attached markup. Regulatory guide 8.15, " Acceptable Programs For Respiratory Pro-tection," has been superseded by incorporation of the require-ments for respiratory protection programs into 10CFR20.103. WNP-3 complies with 10CFR20.103 in the selection of respiratory protective equipment as stated in Subsection 12.5.2.3. Regulatory Guide 8.8, "Information Relevant To Ensuring That Occupational Radiation Exposures At Nuclear Power Stations Will Be As Low As Reasonably Achievable," was utilized as guidance in specifying WNP-3 health physics equipment, instrumentation, and
- facilities. Subsection 12.5.2 will be revised to more specifi-cally address conformance with Regulatory Guide 8.8 as indicated by the attached markup.
WNP-3 13 94W-3 PSAR hf4L5h#>t 47b7 Regulatory Guide 1.8 has been followed in the selection of personnel f or the t.F health physics organization and in the development of training programs for plant personnel. Supply System pre employment practices include screening to determine that plant employees are responsible, conscientious and qualified to perf orm their duties safely. 12.5.1.3 Personnel Qualification and Training The experience and qualifications of the supervisory personnel initially assigned to the Health Physics /01emis try Department are outlined in Chapter
- 13. General employee training, specific f ormal training and on-the-job training programs are described in Section 13.2-. The Heal th Physic s/ Chemis try Technicians are trained in health physics and chemistry to levels commensurate with the duties they are assigned. In addition to the above training ,
technician training may consist of industry sponsored schools, seminars and training at other Supply System nuclear plants. Other training programs are established based on the needs of the individual technician. 12.5.2 EQUIPMENT, INSTRUMENTATION AND FACILITIES 12.5.2.1 Health Physics Facilities TLL W&*1 UA& iss m f ot % ;jiQg m $er ,he k ,n pey w hg ,&nlh f,ff$lff). The WNP-3% health physics f acilities include a single health physics access control point for each unit with auxiliary control points that can be established during refueling and equipment outages. Main locker / change rooms are located in each Administration and Service Building ( ASB) with separate f acilities f or both male and f emale employees. Auxiliary change rooms, personnel and equipment decontamination facilities, laboratories, dosime try station, clothing storage area, lab instrument shops, a hot machine shop and health physics of fice/ work area are located in the ASB. The majorit y o f the health physics f acilities are located on the EL. 390 ft. in the vicinity of the access control point. Se e Figure s 1.2-21 and 1.2-22 f or location s o f health physics f acilities and personnel traf fic patterns through the main access con trol area. 12.5.2.1.1 Access Control and Auxiliary Control Points Health physics access con trol is designed and equipped to maintain positive control over access to controlled areas (see Subsection 12.5.3.1.4 for the definition of controlled areas) and to prevent the spread o f radioactive contamination to uncontrolled a reas. Contamination control provisions include l adequata area to separate the flow of personnel entering and exiting the controlled area. Personnel monitoring stations and f acilities f or storage of potentially contaminated small tools will be located in the access con tro 1 a rea. Receptacles f or protective clothing will be provided at exit points from boundaried areas within the controlled area. Auxiliary control points can be established at o ther locations within controlled areas f or the purpose o f controlling contamination as close to the sourc e a s possible. 1 0 12.5-3 i ~
WNP=3 1394W-5 FSAR 47),7 2. c) Portable radiation monitort ag instrumentation.
,Q*]
d) Portable air sampling equipment. e) Pe rsonnel f ri skers and miniscalers. f) Working files and records. g) Status log of radiation contaminated areas and areas with airborne radioactivity. 12.5.2.1.7 La bo ra to ries Laboratory f acilities are located on the 405 f t. levels of the ASB and the RAB. The f acilities include a shielded low background counting room, hot and cold laboratories f or radiochemical and chemical analyses, hot and cold sample preparation rooms and counting room where samples will be analyzed qualitatively and/or quantitatively. The laboratories are equipped to pe :f orm routine analyses required for personnel protection, surveys and related health physics functions. The counting room (low background) is equipped with necessary instrumentation to perform routine counting on all plant radioactivity samples (water, air, swipe survey, e tc. ). La boratory de tec tion equipment is described in Subsection 12.5.2.2. 12.5.2.1.8 Hot Shops and Decontamination Facility A separate complex is provided within the ASB for decontamination, re pa i r , service and calibration of contaminated equipment, instrumentation and tools. he decontamination facility , ho t 21: . '. r i elec tronic sho p)(, ho t machine shop, shielded calibration room and hot storage room are located on EL. 390 ft. (see Figure 1.2-21 f or detail s) . All facilities are within the controlled area. 12.5.2.1.9 Calibration Facilities Facilities f or calibration o f all plant health physics ins trument s are ! provided by the plant staf f, a Supply System central services organization or ! by a qualified contractor.4 C111bration equipment will include pulse r generators, radiation sources, and electronic test equipment. Ra dia tion sources used f or calibrations will be traceable to Ohtional Bureau o f Standard s ( NBS). Periodic checks and maintenance are perf ormed on all test i equipment at a prescribed interval to maintain reliability. 1 l Instrument calibration records are maintained to insure that routine calibrations are performed at the specified frequencie s. We calibration records are also used as a tracking mechanism on instrument maintenance to t determine factors causative to reduced reliability. Die ins trumen t maintenance f acilities include an instrument shop and a contaminated instrument shop. l
, W. : , WF cd%k~ juil. ka nos As,- , G et toca d su L kh rad:nk :. W cn%r Ha- wm us:d ne$,inbeoVe us.
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- 12. . - 5
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WNP '> 1394W-6 FSAR Q7f 7 Storage of calibrated instruments is provided in the access control area with additional instruments stored at prescribed locations. Storage locations are selected to provide ready access for normal plant operations and to facilitate access to an adequate instrument supply during emergency or of f normal situations. 12.5.2.1.10 Ilhole Body Counter Whole body counting equipment will be maintained at a nearby f acility. The unit will be housed in a low background area and operated by trained Health Physics / Chemistry personnel. 12.5.2.1.11 Storage A radioactive material storage area will be maintained on the 390 ft. elevation of the RAB. This area is adjacent to the hot chemistry laboratory (see Figure 1 2-22 for details). 12.5.2.2 Health Physics Instrumentation A summary of the quantities, types and ranges of WNP-3/5 health physics instrumentation is provided XinTable 12.5-1. geGaislo f.S(d/pg)w t.97/4Mwcre assd a 5 (nskt
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)
The laboratory radiation detection instrumentation, located and used in the counting room, access control and laboratories, includes the following: a) A Ce (L1)$t gamma W spectrometer system with conventional lead shielding and pqmulti-channel analyzer. b) A liquid scintillation counter. I k O 5'I* U*$~ s ! le.o b~* k '**' A A P c) Twof;r- y .1s.; pgpor'tional counters. ' d) End window G-M type counting system. e) A NaI scintillation spectrometer system. Laboratory instrumentation is calibrated in accordance with manufacturer's recommendations using prepared liquid and sealed sources which are traceable to the National Bureau of Standards. 12.5.2.2.2 Portable Radiation Detection Instrumentation The portable radiation detection instrumentation consists of the following: { a) Neutron dose equivalent rate meters. b) Alpha L : . m Sc.' [ikk.% or pnfor[lnsl coubtekt & ts. ki < r se h & her i~s Y+ 0% Es Y &*nCHI*l
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c) Low and high range gamma dose rate meters. d) RemoteprobeG-Mtypesurveymeters[fr.*514t5) Portable radiation detection instruments are calibrated every six months, af ter repair, or any time a field check indicates an out of calibration condition, using appropriate calibration sources traceable to the National Bureau of Standards. Pulse type instruments may be calibrated with an electronic pulser and response to radiation verified with a radiation source. Instruments used to monitor radiography operations shall have been calibrated within the past three months. Mobile airborne monitors are discussed in Subsection 12.3.4. 12.5.2.2.3 Portable Air Sampling Equipment Portable air sampling equipment includes high and low volume air samplers capable of accepting particulate filters and charcoal cartridges for grab samples of radioactive particulates and halogens. Air sampling instruments are calibrated periodically in accordance with an established schedule. 12.5.2.2.4 y.Guth.s IMaln) Personnel Monitoring Instruments, f. t((,/19) 4 t. H (t/11) we re a hIi3 eA
- sy Personnel dosimetry badges containig gyyu igescent dosimeters (TLD) provide the primary l w l'" *"f h of4 exposure incurTed by personnel during normal and accident conditions. Eg o isassignedaTLDbadge hich isTr::: person entering plant controlled areasi d with the w (v' O) Results of the badgenanY,'w'Iie t period of exposure are recorded on a docup,eg,gg kept as an of ficial record. Badges used will be capable of recording C g; -
over a range of at least 10 mrem to greater than 1,000 rem. Neutron exposures are assigned by determining the dose 7 testsO rate versus the time spent in any areas wge signif,1,ca. .t The total neutron doseNgr by use of neutron exposure is present state of the art TLIfaneutron each exposed individual will be added to his permanent record at least quarterly.Tk/trlora -s*d An TL 6 nsee waW .La so ~.de r s sa swLs %. ee.p.;iss..sm4 s .f . C- tolr T. N . In addition to the dosimeter badge, persons entering the restricted area may be required to wear other dosimetry assigned by the health physics staf f such as direct reading pocket ion chambers, integrating dose meters, alarming electronic dosimeters, extremity badges or finger rings. WeL Iar PocketAdosimeters are tested for calibration / response and leak rate as l I required in Regulatory Guide 8.4, February 26, 1973. TLD's and TLD readers are calibrated prior to initial use and periodically thereafter using i / radiation sources traceable to the National Bureau of Standards. A -+ 12.5.2.2.5 Emergency Radiation De 6..'ds I.91 soss isW c.s psid* <" h ssla sr,tection Ins trumenta tionq wg<~ssf redles4lm oU
. table instruments are available for immediate access and consists of the j following:
a) Wide-range ionization chamber beta gamma survey instruments or instruments capable of measuring radiation fields in the nominal range of 1 mR/hr to 1,000 R/hr. .C i 12.5-7 C
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1394W-8 WNP-3 FSAR Y7/. 7 ti) W6O Portable G-M type survey instruments for use in monitoring surface contamination and counting emergency air samples. c) Portable air samplers. d) Direct reading dosimeters with nominal ranges of 0-500 mR, 0-SR, and 0-100R. l q) t 4 (,Lj g klyi y p w deu-/*/* .b b% b *o d m .b/r g g instrumenY$ aTallable for emergency use are checked regularly for proper operation and exchanged for calibration every six months. 12.5.2.2.6 Installed Radiation Detection Instrumentation installed Area Radiation Monitors (ARM) and Airborne Activity Monitors (AAM) are described in Subsection 12.3.4, Ef fluent Liquid Radioactivity Monitors (ELM), Effluent Gas Monitors (ECM) and Process Radiation Monitors (PRM) are described in Section 11.5. 12.5.2.3 Protective Clothing and Equipment Personnel protective clothing and equipment are stored in the clean clothing storage area and in the vicinity of the access control and auxiliary control points. The types of equipment and clothing available include the following: glow s a) Anti-contamination clothing - coveralls, hoods, boots,Aglove liners, and lab coats. b) Plastic gloves and plastic shoe covers. c) Plastic waterproof suits and hoods. d) Continuous airflow two piece plastic suits. Respiratory protective equipment is provided and required for personnel when levels of airborne radioactive materials approach or exceed applicable limits or when a potential for this condition exists. Respiratory protection procedures are described in Subsection 12.5.3.3.4. The respiratory protection program is conducted within the requirements of 10CFR20.103 and exposure is limited to average concentrations less than the values specified in 10CFR20, Appendix B Table 1. Allowance is made for use of respiratory protective equipment, as prescribed in 10CFR20.103, in determining an individual's inhalation of airborne radioactive materials. The following types of equipment are used: a) Full f acepiece air purifying respirators with high efficiency filters. b) Full facepiece pressure demand air line respirators. c) Full f acepiece pressure demand self-contained breathing apparatus.
,4 12.5-3
Attachment 1-86 2268A Question No. 471.7 Regulatory Guide 1.97, "Instrunentation for Light-Water-Cooled (12.5.2) Nuclear Power Plants to Assess Plant and Environs Conditions (contd.) During and Following An Accident," was used for guidance in specifying emergency radiation detection instrumentation. The portable emergency radiation detection instrumentation conforms to the requirements of Regulatory Guide 1.97 except that high range portable beta does rate measuring instruments are not com-mercially available. Subsection 12.5.2 will be revised to more specifically address conformance with Regulatory Guide 1.97 as indicated by the attached markup.
Attachment 1-87 2268A Question No. 471.8 The SRP (NUREG-0800) requires information describing the (12.5.3) implementation of Regulatory Guides 1.8, 8.2, 8.7, 8.8, 8.9, 8.10, 8.26, 8.27, and 8.29. This information is not completely d iscu ssed. Provide this material, including a specific discus-sion of the implementation of Regulatory Guides 1.8, 8.9, 8.26, 8.27, and 8.29.
Response
Regulatory Guide 1.8, " Personnel Selection and Training" refers to ANSI N18.1-1971, " Selection and Training of Nuclear Power Plant Personnel" which has recently been superseded by ANSI 3.1-1981, " Standard for Selection, Qualification and Training of Personnel for Nuclear Power Plants". The WNP-3 position on com-pliance with Regulatory Guide 1.8 will be stated in FSAR Section 17.2. Regulatory Guide 8.2, " Guide for Administrative Practices in Radiation Monitoring" refers to ANSI N13.2-1969, " Guide for Administrative Practices in Radiation Monitoring". Subsections 12.5.3.1.2 and 12.5.3.3.1 both state that Regulatory Guide 8.2 was used in developing radiation protection procedures. In addition, Subsection 12.5.3.8.1 has been revised with FSAR amendment No.1 to state that Regulatory Guide 8.2 was used as guidance, in part, for developing Radiation Exposure Records System. Regulatory Guide 8.7, " Occupational Radiation Exposure Records Systems" refers to ANSI N13.6-1966 (R1972), "American National Standard Practice for Occupational Radiation Exposure Records Systems" for guidance. Subsection 12.5.3.8 provides specific ! information and states that Regulatory Guide 8.7 was used to develop the Radiation Exposure Records System. j Regulatory Guide 8.8, "Information Relevant to Ensuring that l Occupational Radiation Exposures at Nuclear Power Stations will be As Low As Is Reasonably Achievable" was used as guidance in developing the Radiation Work Permit as discussed in Subsection 12.5.3.1.2. In addition, operation, maintenance, repair, sur-veillance and refueling procedures are reviewed, as applicable, to ensure occupational exposures are ALARA. Subsection 12.5.3 will be amended to reflect this as indicated by the attached. l l Regulatory Guide 8.9, " Acceptable Concept, Models, Equations, and Assumptions for a Bioassay Program" refers to several ICRP publications for guidance on developing an acceptable bioassay p rog ram. Subsection 12.5.3.4.2 has been revised by FSAR amend-ment No.1 to include a discussion as to how Regulatory Guide 8.9 has been implemented. Both in vivo and in vitro bioassay programs have been developed using the methodologies recomended by the ICRP.
WNP-3
=
.
- 1394W-9 FSAR
. G471.%
#3 12.5.3 lh 7-Teg PROCEDURES 12.5.3.1 Exposure Control 12.5.3.1.1 Maintaining Exposures ALARA
/
The occupational radiation exposure of personnel at WNP-3/5 is maintained ALARA by a combination of monitoring, protection, access control, training and planning. Monitoring consists of personnel exposure monitoring to maintain control over total radiation doses received and radiation, contamination, and airborne radioactivity surveys to identify and control radiological hazards. Protection involves the use of temporary shielding, where necessary, to reduce radiation levels and the use of anti-contamination clothing and respiratory protective' equipment to protect personnel from actual and possible sources of contamination and airborne radioactivity. Access control is employed to minimize exposure of personnel by limiting access to controlled areas commensurate with both the need for entry and the degree of radiological hazard involved. Training is used to keep personnel aware of radiological hazards and on the alert for eethods of reducing their exposure. Operating, - saintenance, surveillance and refueling procedures will be reviewed as applicable, cer$o ro,,em e to~insure
+4 that prescribed
$e,gais actions hry C- iales conform vig*h ALARA
- 9. .P,cks%/ 1919,JconceptspYik go hl977 Planning is utilized when high exposures are expected. It consists of job probriefing, preparation and review of written procedures for work in controlled areas, and " cold" practice of work to be performed. Subsection
%,,,, 12.1.3 describes a process that was incorporated into the preparation and M revision of all plant procedures which provides a positive method of assuring health physics input and ALARA consideration into all radiation exposure related activities.
12.5.3.1.2 Personnel Control Procedures The WNP-3/5 Plant Administrative Procedures and Health Physics Procedures contain the administrative control procedures for entry into radiation and high radiation areas. The procedures limit the entry and time spent in radiation areas to the time necessary to perform routine operations, maintenance and surveillance activities. The Radiation Work l'ermit (RWP) is used as the primary tool to insure the control at WNP-3/5. Health Physics / Chemistry personnel or personnel who are provided direct coverage by Health Physics / Chemistry Technicians may be exempt from the issuance of a RMP on a case to case basis as approved by plant management. Provision will be included on an individual basis for exposure tracking by job category and I function. - The radiation wock permit is issued for a particular task or function, and is required -before entering a radiation area. This permit provides current data on radiation levels within the area of interest, any restrictions on allowable work time, protective clothing and respiratory protective requirements, information on special tools or equipment needed, special radiation safety and personnel monitoring requirements and any other special instructions e 12.5 WNP-3
, ,1393W-10 FSAR
. V 12.5.3.8.3 Radioactive Materials y][j) v.;
Records of radiation and contamination surveys upon receipt of shipments of radioactive materials are retained for the time specified in 10CFR20.401. Records on inventory and sealed source leak tests of the by product materials not produced by operation of WNP-3/5 are retained until the activities of the by product materials decay to less than the exempt quantities. Records of disposal of the by product materials not produced by operation of WNP-3/5 are retained for the time specified in 10CFR20.401. 12.5.3.8.4 Radioactive Effluent Releases and Solid Radioactive Material Shipments Records of batch releases of radioactive effluent and shipments of radioactive material for disposal are retained for the time specified in 10CFR20.401. Reports to the NRC pursuant to the radioactive effluent releases and the shipment of radioactive material shall be made in accordance with the requirements in Regulatory Guide 1.21, Revision 1, June 1974. 12.5.3.9 Health Physics Program Review and Audits The WNP-3/5 Health Physics Procedures are reviewed annually by members of the central and/or plant health physics staff to determine areas where improvements may be desirable. The Radiological Assessment and Audit Section annually audits portions of each plant's radiation protection program. The areas to be audited are selected so that the entire radiation protection 7 program of each plant will be audited over a period of five years. g.gg,
~
Health Physics Procedures will be reviewed as required by the Technical Specification. . 4m g p/-~ , ,, p (,uste g,70 h5h/ 9 U-f ! Evaluations e etermine where significant occupational radiation exposures are occur
- are conducted continuously as a part of the operational ALARA progra These evaluations are conducted jointly by the Health Physics / Chemistry Department and the Radiological Assessment and Audic Section. The Radiation Exposure Records (RER) computer programs permit l analysis of radiation exposures by craft, job, component or system. The ALARA i program uses the methodology of NUREC/CR-0446 to identify and evaluate
{ potential areas of exposure reduction. l O e
?.. .
12.5-22 ,
- - , , , _ _ - _ , . , - - . , , . , - . , - , . . , , , , , , ~ _ , ,__,y _ . , - . . . _ __ _, ._ - _ , , , ,, y . , 7,, , , , . . , , . . _ _ . __ _ . . . ,
M Attachment 1-88 2268A Question No. 471.8 Regulatory Guide 8.10, " Operating Philosophy for Maintaining (12.5.3) Occupational Radiation Exposures As Low As Reasonably (contd.) Achievable" gives guidance with respect to modifying procedures and giving training to ensure exposures are ALARA. Procedures are reviewed, as applicable, to ensure occupational exposures are ALARA. Subsection 12.5.3 will be amended to reflect this as indicated by the attached. Training is discussed in Section 13.2. Regulatory Guide 8.26, " Applications of Bioassay for Fission and Activation Products" refers to ANSI N343 which was used in developing the bioassay program for WNP-3/5. Subsection 12.5.3.4.2 has been revised with FSAR amendment No. 1 to include a discussion of how Regulatory Guide 8.26 is being implemented. Regulatory Guide 8.27, " Radiation Protection Training for Per-sonnel at Light-Water-Cooled Nuclear Power Plants" was used in the development of training and retraining programs consistent with the ALARA objective to meet the requirements of 10CFR19, except that persons permanently assigned to nuclear facilities will be provided refresher advanced radiological training as-sociated with their specialty every two years. A detailed dis-cussion with respect to the implementation of Regulatory Guide 8.27 can be found in Subsection 12.5.3.5 as amended by FSAR amendment No.1. Regulatory Guide 8.29, " Instructions Concerning Risks From Oc-cupational Radiation Exposure" has been incorporated into the general employee radiation protection training. Discussion concerning conformance with Regulatory Guide 8.29 can be found in Subsection 12.5.3.5 as amended by FSAR amendment No.1. l I
Attachment 1-89 2268A Question No. 480.1 Identify the locations in the containment where water may be (6.2.1.1.2 ) trapped and prevented from returning to the containment sump. The quantity of water involved should be specified. Discuss how the static head for recirculation pumps may be affected.
Response
The only dead volume in the Containment where water may be trapped and prevented from returning to the containment SIS recirculation sump, is the reactor cavity and incore guide tube chase (up to EL. 375 ft). The total volume of water that may be trapped is approximately 96,700 gallons. Effects of dead' volumes have been considered in the containment flood level ele-vation analysis. The available NPSH of the containment spray pumps is based on the static head generated from the containment flood level. Discussion of the NPSH calculation is presented,in Subsection 6.2.2. 1 l l l
)
i
Attachment 1-90 2268A Question No. 480.2 Reference or provide a discussion of the administrative controls (6.2.1.1.3) and/or electrical interlocks that would prevent the inadvertent operation of the containment heat removal system or other sys-tems that could result in pressures lower than the external de-sign pressure of the containment structure. Identify the worst single failure that could result in the inadvertent operation of the containment heat removal system.
Response
A discussion of the administrative controls and/or electrical interlocks that would prevent the inadvertent operation of the containment heat removal system or other systems that could re-sult in pressures lower than the external design pressure of the containment structure will be supplied by December 31, 1982. i l
Attachment 1-91 2268A Question No. 480.3 Provide the results of the confirmatory review of the (6.2.1.5) containment pressure analysis for emergency core cooling system capability studies. Resp onse Final design calculations for the WNP-3 containment backpressure analysis indicate that WNP-3 will not conform to the pressure curve identified on Figure 6.2.1-24 of CESSAR-F because contain-ment heat sinks were found to be considerably greater than as-sumed at the preliminary design stage. Therefore, the large break LOCA analyses results in Subsection 6.3.3.2 of CESSAR-F will be replaced with WNP-3 specific results in the WNP-3 FSAR. The results will be provided by June 1983. i i G 1
Attachment 1-92 2268A Question No. 480.4 1. Establish a procedure to perform an inspection of the con-(6.2.2.3) tainment sump area in particular, to identify any materials which have the potential for becoming debris capable of blocking the containment sump when required for recircula-tion of coolant water. Typically, these materials consist of: plastic bags, step-off pads, health physics instrumen-
- tation, welding equipment, scaffolding, metal chips and screws, portable inspection lights, unsecured wood, con-struction materials, and tools as well as other miscella-neous loose equipment. "As licensed" cleanliness should be assured prior to each startup.
This inspection shall be performed at the end of each shut-down before containment isolation.
- 2. Pipe breaks, drain flow and channeling of spray flow re-leased below or impinging on the containment water surface in the area of the sump can cause a variety of problems; for exagle, air entrainment, cavitation and vortex formation.
Describe any changes you plan to make to reduce vortical flow in the neighborhood of the sump. Ideally, flow should approach uniformly from all directions.
- 3. Evaluate the extent to which the containment sumps satisfy
( each of the positions of Regulatory Guide 1.82. The follow-ing additional guidance is provided for this evaluation:
- a. Provide the size of openings in the fine screens and compare this with the minimum dimensions in the pumps which take suction from the sump, the minimum dimension in any spray nozzles and in the fuel assemblies in the reactor core or any other line in the recirculation flow path whose size is comparable to or smaller than the sump screen mash size in order to show that no flow l
t blockage will occur at any point past the screen.
- b. Estimate the extent to which debris could block the trash rack or screens. If a blockage problem is identi-fled, describe the corrective actions you plan to take.
- c. For each type of thermal insulation used in the contain-ment, provide the following informdtion:
(i) type of material including composition and density, (ii) manuf acturer and brand name, (iii) method of attachment,
4 Attachment 1-93 2268A Question No. 480.4 (iv) location and quantity in containment of each type, (6.2.2.3) (contd.) (v) an estimate aof the tendency of each type to form particles small enough to pass through the fine screen in the suction lines.
- d. Estimate what the effect of these insulation particles i would be on the-operability and performance of all pumps used for recirculation cooling. Address effects on pump seals and bearings.
Response
The staff concerns regarding containment sump designs and their - effect on long -term cooling following a Loss of Coolant Accident (LOCA) will be addressed as part of Enclosure 4, item (10),
" Effects of Containment Coatings and Sump Debris on ECCS and Containment Spray Operation."
i
1 I ! Attachment 1-94 2268A 1 a Question No.
! 480.5 Provide an evaluation of your conformance to Branch Technical i (6.2.4) Position CSB 6-4. Identify and justify any deviations.
t } Response
- The Supply System is reviewing the evaluation of conformance to BTP-CSB 6-4 at this time. This process will be complete by i Novenber 1982.
i i 4 h l i i
^
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-95 2268A Question No.
480.6 Provide instrument lines containment penetration information in Table 6.2.4-1 and Figure 6.2-36 which is labeled as "later".
Response
The following is a new writeup for the present FSAR Subsection 6.2.3.1.2 " Design Criteria for Instrument Lines". Instrument lines penetrating the primary reactor containment and that are connected directly to the containment atmosphere meet the requirements of General Design Criterion (GDC) 56. There are no instrument lines which penetrate the primary reactor con-tainment and form part of the reactor coolant pressure boundary. Compliance with the requirements of GDC 56 is achieved through conformance with Regulatory Position C.l.a through e and C.2.a of Regulatory Guide 1.11. Those instrument lines penetrating the primary reactor contain-ment and connected to the Containment Vacuum Relief System, meet the requirements for redundancy, independence and testability in that Train A instruments are connected to penetration 84 as shown on Figures 7.3-8 and 7.3-9. All instrument lines penetrating the primary reactor containment (Penetrations 84,86,87) are provided with a self-actuated excess flow check valve, located on the instrument line outside of the containment as close to the penetration as possible. The excess flow check valves are self-actuating and will close on excess flow due to loss of integrity of the instrument line outside of containment. The check valves are manual reset type and are provided with position indication for open/close and an alarm for closed position, in the control room. The valves are designed in accordance with ASME Section III Class II requirements and Quality Class I. The response time J instruments connected to those penetrations has been determineo to be acceptable with the excess flow check valves installed. Table 6.2.4-1 and Figures 6.2-36 through 6.2-36r provide details of the instrument line penetrations. The instrument line penetration information shown in Table 6.2.4-1 as later in Attachment 1 will be provided by August 1983. The instrument line penetration information shown on Figures 6.2-36 through 6.2-36r is given in Attachment 2.
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i i 1288W-3 U(W 4 FSAR T k 6.2.4 1 2 h4f d Design Criteria for Instrument Lines 7 - I ( Instrument .. criteris. lines penetrating the containment will be designed to R.G. 1 11 instrument line penetrations. Table 6.2.4-1 and Figures 6.2-36 through 6.2-36r 6.2.4.1 3 Design Requirements for Isolation Barriera All isolation account barriers transient, the accident are designed to perform their function taking into pc.,er sources. servere natural phenomena, and loss of normal Specifically, all isolation barriers are designed to seismic design temperature. Class 2 LOCA transient and environment, and containment Ca tegory 1, Safety minimum, to the containment structural acceptanceAlltest ESF pressure.All p systems are designed, as a minimum to the containment design pressure. All closed systems inside containment are designed, as a minimum, for an external pressure equal to the containment structural acceptance test pressure and have thernal relief protection through the steam generator relief valves . Detailed information on the design requirements is given or referenced Yn Subsection 6.2.4.2 3. 6.7.&,1 A Cant i. .%
!= usa non uependability Design Bases The Containment Isolation System meets the requirements of NUREG=0737 Task Action Plan II.E.4 2. The design criteria are as follows: .
a) The CIS complies with the requirements of Standard Review Flan 6 2.4, specifically that there be diversity in the parameters sensed for the initiation of containment isolation (refer to Subsection 7.3 1.1 2.3). b) Control systems for the automatic containment isolation actuation signal do not automatically reopen the containment isolation valves , upon reset of the CIAS but require deliberate operator action. l c) The containment setpoint pressure for nonessential systems lines is reduced to the minimum compatible with normal operating conditions. d) Containment vent and purge isolation systems are isolated upon a receipt of the high radiation signal. A detailed discussion of the specific design requirements is given or referenced in Subsection 6 2.4.2. 6.2.4.2 System Desig l Personne. air locks and the design of containment penetrations are discussed in Subsection 3 8.2. Material requirements for the C15 are discussed in Section 6.1. Guard pipes are discussed in Subsection 3 6.2.4. Table Containment 6.2.4-1 tabulates all the specific design information for the Isolation System (CIS) on a penetration-by penetration basis. The l discussion below provides design exceptions to the system's additional, general information and addresses any bases. ! k the design requirements for isolation barriers are met. Subsection 6.2.4.2.3 elaboratas hov 6.2-128 i
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10 Twseri 1 I directly to the containment atmosphere meets gnthe Criterion (GDC) 56. nnectedrequ reactor containment and form part of the reactor coolant pre . Regulatory Position C.1.a through e and C.2.s of Regul . . Those instrument lines penetrating the primary reactor containment and co nnected to the Containment Vacuum Relief System, meet the requirementsn for ancy,redu d 84 as shown in Figures 7.3-8 and 7.3-9. independence and testability n Ill 84,instrument 86, 87) lines penetrating the primary reacter containment are provided with a self-actuated excess flow check valve (Penetrations
, located on possible.
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Attachment 1-96 2268A Question No. 490.1 Chapter 4 states: "The initial fuel cycle for WNP-3 is not con-sistent with the extended fuel cycle described in Chapter 4 of CESSAR-F through Amendment 6. Although the fuel design para-meters in CESSAR-F may envelope the WNP-3 specific fuel design this has not been confirmed. WNP-3 specific evaluations are currently in process and if it is determined that Chapter 4 or portions thereof are not applicable, an amendment will be filed with WNP-3 specific "information". Provide a schedule for completion of these evaluations. Chapter 4.0 data is used in other chapters of the FSAR and/or CESSAR-F. Provide an evaluation of the effects of the WNP-3 fuel cycle data on the results in other chapters (e.g., accident analyses, instrumentation setpoints, etc.)
Response
The WNP-3 specific fuel cycle necessitates an evaluation of the fuel design parametcrs, and safety analysis results as presented in CESSAR-F. Section 4.3, which addresses the nuclear design, will change substantially. Other sections of Chapter 4, notably 4.2 and 4.4 will require some modification. All other areas of the NSSS scope impacted by the WNP-3 specific fuel cycle will be evaluated. We expect the specific fuel cycle design to be bounded by the envelope presented in the CESSAR-F safety analysis. The evaluation of the WNP-3 specific fuel design and amendments to Chapter 4 and all other sections of the FSAR, requiring re-visions due to the specific fuel cycle, will 5e provided by June l 1983. I i l
Attachment 1-97 2268A Question No. 620.1 Provide a Section 18 discussion regarding a detailed Control Room (18) Design Review per NUREG-0660, NUREG-0737 and SECY-82-ll1.
Response
In accordance with NUREG-0660 and 0737 Item 1.D.1 licensees and applicants for operating license are required to conduct a de-tailed Control Room Design Review in an effort to identify and correct human engineering design deficiencies. The Supply System shall submit for NRC review within two months prior to the start of the Control Room Review a progress plan that de-scribes how the Control Room Design Review will be accom-plished. The guidance provided in SECY-82-lli will be utilized in the development of the approach to Control Room Design Reviews. In addition to the above, acceptance criteria which is being developed in conjunction with the development of each subsection of Section 18.0 of the Standard Review Plan will be addressed by the Supply System upon finalization by the staff. For use as interim guidance the Supply System will utilize the guidelines for Control Room Design Review as set forth in NUREG-0700. l l l t
Attachment 1-98 2268A Question No. 630.1 This section should include a chart to show the schedule of, or (13.2.2) each part of, the reactor operator training program. The time scale should be relative to expected fuel loading and should also display the preoperational test period, expected time for examinations for licensed operators prior to criticality, and expected time for examinations for licensed operators after criticality. This section should delineate clearly the extent to which the training program has been accomplished at the approximate time of the subrittal of the FSAR.
Response
The extent of training accomplished at the approximate time of subnittal of the FSAR and a chart depicting the schedule for the Reactor Operator training ' program will be provided in a subse-quent amendment to the WNP-3 FSAR as shown by the attached pages. I 1 1
* " 1417W-5 WNP-3 FSAR - & llm (p}D.I f) Command responsibility and limits
[y Id b g) Administrative requirements for the particular SRO position. License Review Training (4 weeks) A comprehensive examination is given to the license candidates to determine their knowledge of the plant (written examination) and ability to safely operate (simulator examiution) . Based on the results of the examination, review training and/or individual tutoring any be provided. Instruction on plant design and operating problems at. similar plants will be provided. Training Program Evaluation The perf oraance of employees participating in the Cold License Training Program are monitored and evaluated throughout the program. Prequent examinations are given to license candidates in order to determine the effectiveness of the training and the knowledge of the trainees. Records will be maintained on an individual basis. . In the event the scheduled fuel loading date is substantially delayed, the cold license candidates will continue to maintain proficiency through participation in training similar in scope to the retraining program described in (Section 13.2.3.1). i i~ea. w . u q % .k ,L .0a . _ _ _ . - b Tk. +va.wiurg se Ate 4< li c. ase- orb wa;AJees is be.m4 goiJ N e e. 4veu.ua/j greys. L -h v . e4. g r o 9 t u e. W c. e4ta 1. li can c++Jc0 dal es w el re la ctumb . Tk sc k.eAuh b - M h v e 4. ye 3
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l l W* 1 i 7/82 ** ""UO NOW *ML MSTld, h 31 b .b ta s . I2/B6 fust. too ac .;.5 PREOPERATIONAL TESTING & ON.THE-JOB TRAINING ES" 5 pP rn
Attachment 1-99 2268A Question No. 630.2 Subsection 13.2.5 states that Regulatory Guide 1.101 is no longer (13.2.5) active. This was withdrawn at one time but has been reinstated as Revision 2 in October of 1981 and should be addressed in the FSAR. Subsection 13.2.5 also indicates that information regard-ing compliance with Regulatory Guides 8.2, 8.8, and 8.10 will be supplied later. Either supply this information or provide a schedule for its submittal.
Response
Subsection 13.2.5 will be revised in a subsequent FSAR amendment as indicated by the following: Regulatory Guide 8.2 dated February 1973, " Guide for Administra-tive Practices in Radiation Monitoring" was utilized in estab-lishing the Health Physics program (see Subsection 12.5.1.1). Regulatory Guide 8.8 dated June 1978, "Information Relevant to Ensuring that Occupational Radiation Exposures At Nuclear Power Stations Will Be As Low As Is Reasonably Achievable" was uti-lized in developing radiation protection training programs. Training programs address Provisions of the Guide. Regulatory Guide 1.101 dated October 1981, " Emergency Planning and Pre-paredness for Nuclear Power Reactors" establishes NUREG 0654/ FEMA-REP-1 as an acceptable method for compliance. WNP-3 will meet the intent of NUREG/ FEMA-REP-1 thereby complying with Regulatory Guide 1.101 (see Subsection 13.3). Regulatory Guide 8.10 dated May 1977, " Operating Philosophy for Maintaining Operational Radiation Exposure As Low As Reasonably Achievable" was used in establishing radiation protection training. The only exception taken to the guide is to provide training every two years instead of testing every year. (This was discussed with the NRC in conjunction with WNP-2 SER. The WNP-2 sumbittal was found acceptable.)
l
' 140 9W-12 WNP-3 a (,30 2 FSAR A^
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13.2.4.3 Senior Operator ,k ( . Senior operator candidates will be selected from the Licensed Reactor 40 Operators or from experienced and qualified plant staff. M' 4m Academic and Nuclear Plant Fundamentals (fs Senior Operator candidates wt11 receive an Academic / Nuclear Fundamentals N Program described in Table l',.2-1. g}( Senior Operators and Shif t Managers Duties (3 weeks) k Senior Operators and Shif t Managers will have instruction in subjects relating to their duties. b 13.2.4.4 Replacement Superintendents Managers and Supervisors (Same as N 13.2.2.3) ' k N. 13.2.4.5 Replacement Technical Department Sup* port Personnel (Same as
- 13. 2. 2. 4) - (
13.2.4.6 Replacement Maintenance Department Personnel (Same as*13.2.2.5) 13.2.5 APPLICABLE NRC DOCUMElff3 Qy The Supply System will comply with the below documents or take exception to j documents as identified below: a) 10CFR Part 50, " Licensing of Production and Utilization Facilities". No exceptions. 84h D q b) 10CFR Part 5 5, " Operators Licenses". No exceptions. . k 4'S c) 10CFR Part 19, " Notices, Instructions and Reports to Workers; W s Inspections". No exceptions. - 4'2 b 1 d) ' Regulatory Guide 1.8, " Personnel Qualification and Training", see FSAR Section 17.2. The Supply System will comply with the below documents or take exception to l documents as identified below: O dc M f l4 0 e) Regulatory Guide 1.1013 " Emergency Planning f or Nuclear Power Plants". m '- - ' ; r _.. ;;;in ?;;f rt:r7 ^_ir. e57ah43hh AAl/264 O(psy/pg44 febru " IT13 l
- -- f )- Regulatory Guide 8.2 ide f or Administrative Practices in Radiation i t k 1%E't Hh Mlonitoring"g lsyssks prqs36W1(yAs M.filfied/2 hr f4d' .%ecf1Pvs eshnk//5Jribf C./*/)
g) Regulatory Guice 8.8,g"Inf ormation Relevant to Ensuring That 47 occupational Radiation Exposures at !bclear Power Stations Will Be as bd Low s is Reasonably Achievable", Jtmou@- wd5 Kff//Ed /// /d,/d/f/>r (O ft2f1h4 f t"O YdCf7 evs '&'tri ypp pt f4??.$ . h) Regulatory Guide 8.10g" Operating P osophy or Maintaining vccupational Radiation Exposures As Low As Is Reasonably Achievable", pl == A LA nS a frlised fri e=sMh5hsw rddia&m ecken en 1'o 'the ' -frntn oy. The en /y exe ep' Hon 6Wde is to of ksky very yar ;,,'frovide* al,,,.,1y wer y hvo
. ,1409W-13 WNP-3 FSAR p/1
- 1) Regulatory Guide 8.13 " Instruction Concerning Prenatal Radiation Exposure". .
j) " Utility Staffing and Training for Nuclear Power", WASH-1130 Revised June, 1973. Contains advisory information and was considered in establishing staffing training. k) NRC Operator Licensing Guide, NUREG-0094, July,.1976. Contains information " intended to ass'st" i and was considered in establishing training programs. e 0 9 13.2-18
Attachment 1-100 2268A Question No. 640.1 State whether all tests at each given power test plateau will be (14.2.4) performed before increasing power to the next plateau (power level). -
Response
Add to FSAR Subsection 14.2.4.2.2-Startup Test Scheduling and Sequencing: Those tests which are deemed necessary to ensure the safe opera-tion of the plant will be conducted at all incremental power plateaus during Power Ascension Physics Testing (PAPT). These tests will be identified in the PAPT program procedure and will include such tests as verification of negative power coefficient of reactivity and steady state core performance including calorimetric NI calibrations. Certain other tests may be waived at intermediate power levels or rescheduled to a higher power level if WNP-3 is found to behave in an acceptable manner rela-tive to the lead CESSAR Standard Design plant (CESSAR-F, Chapter
- 14) and any test results obtained at lower power levels are found to be acceptable. Tests may be waived or rescheduled at the discretion of the Plant Manager.
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14.2.4.2.2 Startup Test Scheduling and Sequencing Scheduling and sequencing of testing during startup is performed under the direction of the Plant Manager. The test sequence and individual Startup Test Procedure identify hold points for data review and authorization to proceed and establishes the general plant condition f or each group of tests. 14.2.4.2.3 Startup Test Performance Be fore starting each test, the assigned test engineer will review the test procedure to assure that prerequisite activities or conditions have been satisfied as described in Subsection 14.2.4.3. The test will be stopped or curtailed if it cannot be performed safely or in accordance with the approved test procedure. Raquired test procedure i deviations or changes may be-effected in accordance with a " Test Change No tice" as described in Subsection 14.2.4.4.
- Should apparent deviations of test results from performance requirements or acceptance criteria be revealed, or should other anomalies develop, the plant will be placed in a saf e condition and relevant test data reviewed by the test engineer and Shif t Manager. If the apparent discrepancy or anomaly is substantiated, the situation will be reviewed by the Plant Operations Conmittee to ascertain if a plant safety question is involved. Control of any identified nonconformance or noncompliance will be in accordance with the plant administrative procedures.
Evaluation of the effect of the discrepancy or anomaly on plant safety will be
- performed at the appropriate level of review and appropriate corrective action will be taken before resumption of the test or test conditions at which the problem was revealed.
At the completion of an entire test procedure, the test engineer will assemble 4 all of the data and supporting information, nonconformance documentation and test results evaluations f or review by the Plant Operating Committee. Any data reduction or analysis required will be done as soon af ter the data is available as is practical so that the results of the analysis may be included in the completed tese package. Test records will be maintained as described in Subsection 14.2.6. 14.2.4.3 Control of Tes t Prerequisite s Conditions and, activities prerequisite to a given test will be identified in
~~
the~ applicable tes t procedure. Prior to commencement of the particular_ test . (khose tests which are deemed necessary to ensure the safe operation of the plant will be conducted at all incremental power plateaus during Power Ascension Physics Testing (PAPT). These tests will be identified in the PAPT program procedure and will include such tests as verification of negative power coefficient,of reactivity and steady state core perfonnance including calorimJtric NI cali-
, brations. Certain other tests may be waived at intennediate power levels or rescheduled to a higher power level if WNP-3 is found to behave in an acceptable manner relative to the lead CESSAR Standard Design plant (CESSAR-F, Chapter 14) and any test results obtained at lower power levels are found to be acceptable. Tests may be twaived or nescheduled at the discretion of the Plant Manager.
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Attachment 1-101 2268A Question No. 640.2 Clarify the description of the review process for preoperational (14.2.5) and startup test results. Include: (1) requirements for review and approval of the test data for each major test phase before beginning the next phase, and (2) a description of the require-ments for review and approval of the startup test data at each major power test plateau before raising power to the next test plateau during the power ascension phase.
Response
All test data and results will be reviewed by personnel formally judged to be qualified for such evaluation, as noted in Subsec-tion 14.2.5.1. Based on that review, all test results will be approved, as necessary, by the Startup Manager and, for post-core tests, the Plant Manager. This approval reflects a judg-ment that the test data and results adequately satisfy the test objectives and criteria specified in the test procedure. Instances of deficiency or incompleteness will be dispositioned formally, as noted in Subsection 14.2.5.2. In general, testing may proceed f rom SLT to Preoperational Phase to the Startup Phase as permitted by the procedure-specified test prerequi-s ites. However, prior to proceeding with fuel loading and major power test plateaus, a formal assessment may be required, as noted in Subsection 14.2.5.3, to consider the readiness of the plant equipment, procedures, and personnel for this next major phase. This readiness judgment will be made within the frame-work of intention to comply with all regulatory connitments, preparation to cope with unexpected plant performance, and avoidance of plant damage. l l
s , s Attachment 1-102 2268A Question No. - 640.3 Subsection 14.2.12 indicates that test descriptions for other (14.2.12) post-core hot functional and power ascension tests will be pro-vided later. Provide a schedule for submittal of this informa-tion or submit the test descriptions.
Response
Test descriptions for other post-core hot functional and power ascension tests noted in Subsection 14.2.12 will be developed i and approved thrde months prior to initiation'of fuel loading. x s O s*
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, -103 2268A Question No.
810.1 The Washington Nuclear Project 3 Emergency Preparedness Plan (13.3.2) indicates that additional information on some facilities in the design phase and one County and State emergency response plans will be provided later. Provide a schedule for submittal of this information.
Response
- 1. Additional information comparing Supply System facilities to NUREG-0696 wilt be available 0:tober 31,1983.
- 2. Washington State Plan - subhitted to FEMA in July. FEMA review expected Fall of 1982. Revised during Summer of 1983. The revised plan will be available September 30, 1983.
- 3. Grays Harbor County Plan - schedule same as Washington State.
- 4. Mason County Plan - to be submitted to FEMA by the end of 1982. The revised 'pla1 will be available September 30, 1983.
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Attachment 1-104 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 1) ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT Commission Memorandum and Order of May 23, 1980 defines the current staff re-quirements for qualification of this equipment. Additional guidance on this matter was provided in a subsequent NRR order dated November 26, 1980 (con-cerning record requirements), Supplements 2 and 3 dated September 30, 1980 and October 24, 1980, respectively to IE Bulletin No. 70-0I8, and a generic letter dated October 1,1980 to all holder of cps and OLs.
Response
The environmental qualification of safety-related and electrical equipment is discussed in Section 3.11. Tables 3.11-1 and 3.11-2 " Environmental Qualifica-tion of Safety-Related and Electrical Equipment" will be updated quarterly beginning December 1982 with a scheduled completion date of June 1984. Addi . tional information concerning documentation will be available by January 1983. I l 1 l l l l 1
Attachment 1-105 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 2) EMERGENCY RIEPAREDNESS Guidance on the preparation of emergency plans is presented in NUREG-0654 (FEMA-REP-1) " Criteria for Preparation and Evaluation of Radiological Emer-gency Response Plans and Preparedness in Support of Nuclear Power Plants".
The requirements for the emergency response facilities are included in NUREG-0596 " Functional CritEr ia for Emergency Response Facilities." Further guidance on emergency preparedness is provided in the revised Appendix E to 10CFR Part 50.
Response
The WNP-3 Emergency Preparedness Plan was written using NUREG-0654/ FEMA-REP-1 as guidance. Additional information with respect to the compliance with this document can be found in the WNP-3 Emergency Preparedness Plan, Appendix 8,
" cross reference to NUREG-0654". The WNP-3 Emergency Response Facilities were designed using NUREG-0696 as guidance. A detailed document comparing the de-sign of WNP-3 Emergency Response Facilities to NUREG-0696 will be available October 31, 1983.
Appendix E to 10CFR Part 50 refers to NUREG-0654 for guidance in developing Emergency Preparedness Plans. NUREG-0654 was used in developing the WNP-3 Emergency Preparedness Plan. l l l 1 l l l
Attachment 1-106 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 3) SAFETY-RELATED STRUCTURES, SYSTEMS AND COMPONENTS (0-LIST) CONTROL!.ED BY THE QA PROGRAM Staff requests for additional information regarding this issue have been sent to a nunber of OL applicants. A request from the Diablo Canyon review is pro-vided as Enclosure 5.
Response
The Supply System is in the process of developing the Q List. We expect to have finalized this list by February 1983. l
Attachment 1-107 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 4) FRACTURE PREVENTION OF CONTAINMENT PRESSURE BOUNDARY (GDC-51)
GDC-51 requires that under operating, maintenance, testing and postulated accident conditions, (1) the Ferritic materials of the containment pressure boundary behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The Ferritic materials of the contaiment pressure boundary which are assessed by the staff are those of conponents such as freestanding containment vessel, equipment hatches, personnel airlocks, primary containment drywell head, heads containment penetration sleeves, process pipes, and closure caps and flued heads and penetrating piping systems downstream of penetration process pipes extending to and including the system isolation valves. The acceptability of these materials within the context of GDC-51 is deter-mined in accordance with the fracture toughness criteria identified for Class 2 materials by the Suniner 1977 Addenda to ASME Code Section III.
Response
The FSAR has been reviewed and it has been determined that the ferritic mate-rials of the containment pressure boundary have been sufficiently identified in the FSAR to allow the NRC to perform their review. Refer to FSAR Subsec-tion 3.1.44 for discussion regarding GDC-51. CESSAR-F was not reviewed in this regard and is not considered necessary to be reviewed. { 1 l l l l l
Attachment 1-108 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 5) FIRE PROTECTION In accordance with Subsection 9.5.1., Branch Technical Position ASB 9.5-1, Position C.4.a(1) of NRC Standard Review Plan and Section III.G of new Appen-dix R to 10CFR Part 50, it is the staff's position that cabling for redundant safe shutdown systems should be separated by walls having a three hour fire rating or equivalent protection (see Section III.6.2 of Appendix R). That is, cabling required for or associated with the primary method of shutdown, should be physically separated by the equivalent of a three-hour rated fire barrier from cabling required for/or associated with the redundant or alternate method of shutdown. To assure that redundant shutdown cable systems and all other cable systems that are associated with the shutdown cable systems are sepa-rated from each other so that both are not subject to damage from a single fire hazard, we require the following information for each system needed to bring the plant to a safe shutdown.
Response
Safe Shutdown Steps and related equipment lists have been prepared for use in a safe shutdown analysis. An analysis of all essential equipment is beinp performed in order to insure that no single fire can prevent WPPSS Unit 3 from achieving a safe cold shutdown. In this regard a list of all equipment and power sources required to bring the plant to cold shutdown has been compiled. This list differentiates between equipment required to maintain hot standby and that required for cold shutdown only. To f acilitate this review a computer program was developed which lists all essential equipment that appear in a fire area. This report will be analyzed by fire area, in conjunction with the physical drawings, verify that the oper-ation of redundant essential components are not impaired by a single fire in any fire area. The results of this analysis will be presented by fire area and will include one acceptable means of separation as described in Appendix R to 10CFR50, Section III.G.2. This analysis will demonstrate that no single fire could prevent the plant from being brought safely to cold shutdown. Question No. ( 040.75 Provide a table that lists all equipment including instrumenta-tion and vital support system equipment required to achieve and maintain hot and/or cold shutdown. For each equipment listed:
- a. Differentiate between equipment required to achieve and maintain hot shutdown and equipment required to achieve and l maintain cold shutdown,
- b. Define each equipment's location by fire area,
- c. Define each equipment's redundant counterpart, 1
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Attachment 1-109 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 5) FIRE PROTECTION (contd.)
- d. Identify each equipment's essential cabling (instrumenta-tion, control, and power). For each cable identified: (1)
Describe the cable routing (by fire area) from source to termination, and (2) Identify each fire area location where the cables are separated by less than a wall having a three-hour fire rating from cables for any redundant shut-down system, and
- e. List any problem areas identified by Item 1.d.(2) above that will be corrected in accordance with Section III.6.3 of Appendix R (i.e., alternate or dedicated shutdown capability).
Response
A list of all equipment and their power sources required to achieve and maintain hot and/or cold shutdown will be provided l by May, 1983. This list will differentiate between equipment l required for hot standby and cold shutdown, define components location by fire area and list the redundant components of the essential systems. The listing will be developed assuming off-site power not available. i l In areas where the function of essential redundant components ( could be damaged by a single fire, the equipment and cables will i be either relocated or one means of insuring that one of the redundant trains is free of fire damage to meet the requirements ( of Appendix R to 10CFR so Section III.G.2 will be provided. The Supply System is taking exception to providing a listing of essential cabling for safe shutdown equipment. It is our posi-tion that all safety class electrical trays, conduits and pull boxes for redundant systems will be sufficiently separated to meet separation requirements of Appendix R. For areas where associated cables are separated by less than a barrier having a three-hour rating from cables required for/or associated with a redundant shutdown system, these cables will be protected at the power source by electrical devices (breakers, fuses). In the event of fire damage to these cables, they will become electri-cally isolated, thus preventing the propagation of damage to other cables in that raceway section. Instrumentation cables
- are self-protected by virture of low energy levels they convey.
l
-110 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 5) FIRE PROTECTION (contd.)
Question No. 040.76 Provide a table that lists Class 1E and Non-Class IE cables that are associated with the essential safe shutdown systems identi-fied in Item 1 above. For each cable listed: (*See note on page 3.)
- a. Define the cables' association to the safe shutdown system (common power source, common raceway, separation less than IEEE Standard-384 guidelines, cables for equipment whose spurious operation will adversely affect shutdown systems, etc.).
- b. Describe each associated cable routing (by fire area) from source to termination, and
- c. Identify each location where the associated cables are sepa-rated by less than a wall having a three-hour fire rating from cables required for/or associated with any redundant shutdown system.
Response
The Supply System is taking exception to providing a listing of Class IE and Non-Class IE cables that are associated with the essential safe shutdown systems. It is our position that these associated cables whether safety (Class IE) or non-safety (Non-Class IE) are both qualified as Class IE cables. For areas where associated cables are separated by less than a barrier having a three-hour rating from cables required for/or associated with a redundant shutdown system, these cables will be protected at the power source by electrical devices (breakers, fuses). In the event of fire damage to these cables, they will become electrically isolated, thus preventing the propagation of damcge to other cables in that raceway section. Instrumentation cables are self-protected by virtue of low energy levels they convey. Question No. 040.77 Provide one of the following for each of the circuits identified in Item 2.c above:
Attachment 1-111 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 5) FIRE PROTECTION (contd.)
Question No. 040.77 (a) The results of an analysis that demonstrates that failure (contd.) caused by open, ground, or hot short of cables will not affect it's associated shutdown system,
- Note *
(b) Identify each circuit requiring a solution in accordance with Section III.6.3 of Appendix R, or Identify each circuit meeting or that will be modified to meet the requirements of Section III.G.2 of Appendix R (i.e., three-hour wall, 20 feet of clear space with auto-matic fire suppression, or one-hour barrier with automatic fire suppression).
Response
The Supply System is taking exception to providing a listing of Class IE and Non-Class IE cables that are associated with the essential safe shutdown systems. It is our position that these associated cables whether safety (Class TE) or non-safety (Non-Class IE) are both qualified as Class IE cables. l l
- NOTE Option 3a is considered to be one method of meeting the require-
- ments of Section II.G.3 Appendix R. If Option 3a is selected the information requested in Items 2a and 2c above should be provided in general terms and the information requested by 2b l need not be provided.
l For areas where associated cables are separated by less than a I barrier having a three-hour rating from cables required for/or associated with a redundant shutdown system, these cables will I be protected at the power source by electrical devices (breakers, fuses). In the event of fire damage to these cables, they will become electrically isolated, thus preventing the propagation of damage to other cables in that raceway section. Instrumentation cables are self-protected by virtue of low energy levels they convey.
Attachment 1-112 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 5) FIRE PROTECTION (contd.)
Question No. 040.78 To assure compliance with GDC 19, we require the following in-formation be provided for the Control Room. If credit is to be taken for an alternate or dedicated shutdown method for other fire areas (as identified by Item 1.a or 3.b above) in accor-dance with Section III.G.3 of new Appendix R to 10CFR Part 50, the following information will also be required for each of these plant areas.
- a. A table that lists all equipment including instrumentation and vital support system equipment that are required by the primary method of achieving and maintaining hot and/or cold s hu tdown,
- b. A table that lists all equipment including instrumentation and vital support system equipment that are required by the alternate, dedicated, or remote method of achieving and maintaining hot and/or cold shutdown.
- c. Identify each alternate shutdown equipment listed in Item 4.b abcVe with essential cables (instrumentation, control, and power) that are located in the fire area containing the primary shutdown equipment. For each equipment listed pro-vide one of the following:
(1) Detailed electrical schematic drawings that show the essential cables that are duplicated elsewhere and are electrically isolated from the subject fire areas, or l (2) The results of an analysis that demonstrates that I failure (open, ground, or hot short) of each cable identified will not affect the capability to achieve and j maintain hot or cold shutdown.
- d. Provide a table that lists Class IC and Non-Class IE cables that are associated with the alternate, dedicated, or remote method of shutdown. For each cable so identified provide the results of an analysis that demonstrates that failure (open, ground, or hot short) of the associated cable will not adversely affect the alternate, dedicated, or remote method of shutdown.
i i
Attachment 1-113 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 5) FIRE PROTECTION (contd.)
Response
040.78 The Remote Shutdown Panel (RSP), local Diesel Generator (DG) (contd.) Control panels and Local Turbine Driven AFW Control panels lo-cated in the Reactor Auxiliary Building provides an alternative location from the Control Room, which allows the operator to:
- 1) Achieve prompt hot shutdown of the reactor; 2) Maintain the unit in a safe condition during hot shutdown; 3) Achieve cold shutdown of the reactor through the use of suitable procedures.
The Remote Shutdown Panel is fully described in FSAR Subsection 7.4.1.1. The RSP would be utilized as the alternate means of plant shutdown in the event that tFe Control Room becomes uninhabitable. The Supply System is taken exception to providing a listing of essential cabling for safe shutdown equipment. It is our posi-tion that all safety class electrical trays, conduits and pull boxes for redundant systems are sufficiently separated to meet the separation requirements of Appendix R. For areas where associated cables are separated by less than a barrier having a three-hour rating from cables required for/or associated with a redundant shutdown system, these cables will be protected at the power source by electrical devices (breakers, fuses). In the event of fire damage to these cables, they will become electri-cally isolated, thus preventing the propagation of damage to other cables in that raceway section. Instrumentation cables are self-protected by virtue of low energy levels they convey. Question No. 040.79 The Residual Heat Removal System is generally a low pressure system that interf aces with the high pressure primary coolant system. To preclude a LOCA through this interface, we require coinpliance with the recommendations of f' ranch Technical Position RSB 5-1. Thus, this interface mos. likely consists of two re-dundant and independent motor operated valves with diverse interlocks in accordance with Branch Technical Position ICSB 3. These two motor operated valves and their associated cable may be subject to a single fire hazard. It is our concern that this single fire could cause the two valves to open resulting in a fire-initiated LOCA through the subject high-low pressure system in terf ace. To assure that this interface and other high-low pressure interfaces are adequately protected from the effects of a single fire, we require the following information: l l l
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1 Attachment 1-114 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 5) FIRE PROTECTION (contd.)
Question No. 040.78 a. Identify each high-low pressure interface that uses redun-(contd.) dant electrically controlled devices (such as two series motor operated valves) to isolate or preclude rupture of any primary coolant boundary.
- b. Identify each device's essential cabling (power and control) and describe the cable routing (by fire area) from source to termination.
- c. Identify each locaton where the identified cables are sepa-rated by less than a wall having a three-hour fire rating from cables from the redundant device.
- d. For the areas identified in Item 5.c above (if any), provide the bases and justification as to the acceptability of the existing design or any proposed modifications.
Response
The assurance that high-low pressure interfaces are adequately protected from the effects of a single fire will be demonstrated in the safe shutdown analysis which will be completed by April 1983. 1
Attachment 1-115 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 6) EFFECTS OF MASONRY WALLS ON CLASS I STRUCTURES
- 1. Are there any concrete masonry walls being used in any of the Category I structures of your plant? If the answer is "No" to this question there is no need to answer the following questions.
- 2. Indicate the loads and load combinations to which the walls were designed to resist. If load factors other than one (1) have been employed, please indicate their magnitudes.
- 3. In addition to complying with the applicable requirements of the SRP Sections 3.5, 3.7 and 3.8, is there any other code, such as the " Uniform Building Code" or the " Building Code Requirements for Concrete Masonry Structures" (Proposed by the American Concrete Institute) which was or is being used to guide the design of these walls? Please identify and dis-cuss any exceptions or deviations from the SRP requirements or the aforenentioned codes.
- 4. Indicate the method that you used to calculate the dynamic forces in masonry walls due to earthquake, i.e., whether it is a code's method such as Uniform Building Code, or a dynamic analysis. Identify the code and its effective date if the code's method has been used. Indicate the input motion if a dynamic analysis has been performed.
- 5. How were the masonry walls and the piping / equipment supports attached to them designed? Provide enough numerical examples including details of reinforcement and attachments to illustrate the methods and procedures used to analyze and design the walls and the anchors needed for support-ing piping / equipment (as applicable).
- 6. Provide plan and elevation views of the plant structures showing the location of all masonry walls for your facility.
Response
A complete response to this request for information will be available by December 1982. t l
Attachment 1-116 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 7) INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (TMI Action Item II.F.2 in NUREG-0737)- Discussion of this item should address how coreand location thermocouple readouts rate of printout are provided (see Part in the control (4) of Attachment 1 to room including ).
item II.F.2
Response
Instrumentation for detection of inadequate core cooling are within the CESSAR-F scope. This issue is being reviewed on the CESSAR-F Docket as dis-cussed in NUREG-0852 " Safety Evaluation Report Related to the Final Design of the Standard Nuclear Steam Supply Reference System, CESSAR System 80". l l
Attachment 1-117 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 8) RRESERVICE AND INSERVICE INSPECTIONS Staff guidance in this review area has been sent to a number of pending OL app lican ts. A copy of that guidance is provided as Enclosure 8.
Response
The Preservice Inspection Program Plan will be submitted as a separate docu-ment by September 1983 and will be consistent with the requirements of ASME Section XI,1977 edition with addenda through Summer 1978. In accordance with 10CFR50, the WNP-3 Inservice Inspection Plan will be submitted six months prior to start of commercial operation.
Attachment 1-118 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 9) RIESERVICE INSPECTION AND TESTING OF SNUBBERS The staff has recently established requirements to ensure snubber operability which have been transmitted to pending OL applicants. A copy of those requirements is provided as enclosure 9 (Attachment I).
Response
The Supply System has reviewed the Preservice Examination and Pre-Operational Testing requirements discussed in the referenced Enclosure 9 (Attached). It is the intent of WNP-3 to fully comply with the requirements as stated in Enclosure 9. The Preservice Examination for snubbers on safety-related sys-tens will be detailed in the "WNP-3 Preservice Inspection Program Plan" which will be submitted for NRC staff review by August 1983. The Pre-Operational testing requirements will be included in the Pre-Operational and Startup Tests described in Chapter 14 of the FSAR by July 1983. l l l
Attachment 1-119 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 10) EFFECTS OF CONTAINMENT C0ATINGS AND SUMP DEBRIS ON ECCS AND CONTAINMENT SPRAY 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 Enclosure 10 (Attachment I).
Response
The Supply System is evaluating WNP-3 with respect to the Staff concerns men-tioned above. Further information will be available by November 1982. l
Attachment 1-120 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 11) SEISMIC QUALIFICATION A staff request for additional information in this review area has been sent to a number of pending OL applicants. A copy of that request is provided as Enclosure 11 (Attachment I).
Response
The Supply System will have further information available concerning Seismic Qualification by Decenber 1982. l l l l
Attachment 1-121 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4
- 12) SPECIAL LOW POWER TEST PROGRAM (TASK ACTION PLAN ITEM I.G.1)
The staff has established guidance on this matter for transmittal to all pend-ing and prospective OL applicants. A copy of that guidance is provided as enclosure 12 (Attachment 1).
Response
As noted in Subsection 14.2.12.6 of the WNP-3 FSAR, we propose to conduct the WNP-3 Low Power Testing in accordance with tests outlined in CESSAR-F Subsec-tion 14.2.12.4. That Subsection recommends natural circulation tests in the first-of-a-kind NSSS plant, in this instance Palo Verde. It is our intention to monitor the conduct and results of that testing prior to finalizing test requirements for Low Power Testing at WNP-3. As noted previously, we intend to have completed preparation of all postcore test procedures at least three months prior to fuel loading. i
Attachment 1-122 2268A REQUEST FOR ADDITIONAL INFORMATION - ENCLOSURE 4 13) INITIAL TEST PROGRAM DESCRIPTION (CHAPTER 14) which are comon to pending applications. Staff review rns of near term Sum r and the San Onofre 2 & 3 applications. typically ew of the expressed (i
Response
Staff concerns regarding startup testing as expressedainsedthe questions r i in its review of the Sumer and San Onofre 2 and 3 applications e re-will b vicwed as part of the further development of the WNP-3 startup test pro
,*,e .. 1415'J-9 WNP-3 FSAR M 0 8IM/6
((N IN TABLE 13 5-5 AT f2 / TYPICAL EMERGENCY PROCEDdRES seactor Trip /Turbina Trip Anticipated Transient Without Scrati Loss of Feedvater Loss of forced Reactor Coolant Flow Loss of Coolant Accident Steam Generator Tube Rupture Steamline Break ' Inadequate Core Cooling Mah-n 8/aekoat' - L 13.5-14 l i_. . _ _ _ . . - -_ _ -
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