ML20070N498

Forwards Response to Questions Generated During Acceptance Review of OL Application.Responses Scheduled to Be Provided for Review in Dec.Marked-up FSAR Pages Included for Cases Where Amend to FSAR Desirable Due to Responses

ML20070N498
Person / Time
Site:	Satsop
Issue date:	01/17/1983
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Knighton G Office of Nuclear Reactor Regulation
Shared Package
ML20070N502	List: ML20070N498 ML20070N526
References
GO3-83-46, NUDOCS 8301250525
Download: ML20070N498 (33)

Text

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Washington Public Power Supply System P.O. Box 968 3000GeorgeWashingtonWay Richland Washington 99352 (509)372-5000 January 17, 1983 G03-83-46 Docket No. 50-508 Director of Nuclear Reactor Regulation Attention: Mr. G. W. Knighton, Chief Licensing Branch No. 3 Division of Licensing US Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

RESPONSES TO NRC ACCEPTANCE REVIEW QUESTIONS

References:

a) Letter, DG Eisenhut to RL Ferguson, dated 8/20/82 b) Letter G03-82-830, GD Bouchey to HR Denton, dated 8/20/82 c) Letter G03-82-1085, GD Bouchey to JD Kerrigan, dated 11/22/82 Reference a) transmitted a set of questions generated during the NRC's acceptance review of the WNP-3 Operating License Application (Reference b). Reference c) represents the initial Supply System response to these questions and provided a schedule for those cases where our evaluations were not yet complete.

This letter transmits those responses scheduled to be provided for NRC review in December. In those cases where it is considered necessary or desirable to amend the FSAR due to our responses, we have provided marked up FSAR pages which show the changes which will be included in a subsequent amendme.it.

0301250525 830117 PDR ADOCK 05000508 A PDR Y

Mr. G. W. Knighton Page 2 January 17, 1983 G03-83-46 RESP 0f4SES T0 fiRC ACCEPTANCE REVIEW QUESTIONS If you require additional information or clarification, the Supply System point of contact for this matter is Mr. K. W. Cook, Licensing Project Manager (206/482-4428 ext. 5436).

Sincerely, G. D. Bouchey, Manager Nuclear Safety and Regulatory Programs AJM/ss Attachments: 1. NRC Question 100.2.

2. NRC Question 210.4 (3.9.3.4)

3. NRC Question 220.8 (3.8.4.8)

4. NRC Question 270.1 (3.11)

5. NRC Question 271.1 (3.10)

6. NRC Question 281.1 (6.1.1.2)

7. NRC Question 311.3 (3.5.1.6)

8. NRC Question 410.7 (3.5.2)

9. NP,C Question 440.1 (6.4.7.1.3)

10. NRC Question 480.2 (6.2.1.1) 11 Request for Additonal Information Enclosure 4 (Item 1)

Environmental Qualification of Safety Related Electrical Equipment

12. Request for Additional Information Enclosure 4 (Item 6)

Category I Masonry Walls

13. Request for Additional Information Enclosure 4 (Item 11)

Seismic Qualification cc: D. J. Chin - Ebasco NYO N. S. Reyonds - D&L E. F. Beckett - NPI J. A. Adams - NESCO D. Smithpeter - BPA A. V iett i - NRC Ebasco - Elma WNP-3 Files - Richland L _

r Attachment 1 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 CE SS AR-FSAR. This Safety Evaluation Report (SER) imposes requirements on applicants utilizing the CESSAR-FSAR and identifies open items. The' applicant should provide a plan for 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 implementation 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 Egnineering resolution. Although the latter items may not require specific licensee action at this time, licensee tracking is necessary to insure 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 understands that CE has submitted to the NRC all information necessary to close-out SER open items and confirmatory items and issue the final revision.to NUREG-0852. Accordingly, the Supply System considers that at this time, it would be premature to develop and implement a detailed plan to correlate interfaces between the WNP-3 FSAR and NUREG-0852.

t-

Attachment 1 (Cont'd)

QUESTION NO.

100.2

RESPONSE

The Supply System will continue to review NUREG-0852 and, through discussions with Combustion Engineering (C-E), assure that issues deferred to the referencing utility's docket &nd interface requirements identified are adequately addressed in the WNP-3 FSAR.

With respect to the open and confirmatory items in NUREG-0852, the Supply Sys-tem understands that only four of the itens will be likely to result in requirements for additional info,mation on the referencing utility's docket.

These are:

Open Item No. 1 - Environmental Qualification Open Item No. 3 - CPC Software Open Item No. 4 - ICC Instrumentation Confirmatory Item No. 3 - Fuel Performance Analyses All other open and confirmatory items should be resolved entirely within the CESSAR-F docket. After the NRC staff has reviewed C-E's submittals on the CESSAR-F docket and has revised NUREG-0852, the Supply System will update the WNP-3 FSAR as necessary to address the requirements for additional information.

Further, it should be recognized that Chapter 16 of CESSAR-F on Technical Specifications is currently undergoing NRC staff review. Upon completion of that review, it is expected that most of the technical specifications and license conditions imposed in NUREG-0852 will be resolved on the CESSAR-F docket. The remaining issues that are deferred to the referencing utility's docket will be addressed in the WNP-3 submittal of plant technical specifica-tions. The Supply System has committed (see FSAR Chapter 16) to provide the plant technical specifications no later than one year prior to fuel load.

The WNP-3 FSAR does not at this time include a Section 15.8 to discuss imple-mentation of fixes for anticipated transients without scram (ATWS) since the issue has not been resolved by the NRC. The interim fix of operatcr training and emergency procedures developments in NUREG-0852 is being pursued by the Supply System through participation in the CE Owners Group Emergency Proce-dures Guidelines Task. Comitments as to what procedures and/or training will be implemented must await completion of this task.

The second example in the question refers to an SER requirement that an ESFAS signal not override MSIS. The actual SER requirement is concerned with con-trol logic associated with an EFAS signal (Emergency Feedwater Actuation Signal) not an ESFAS signal (Engineered Safety Features Actuation System) as indicated in the question. The SER requirement is based on assuring

1 Attachment 1 (Cont'd)

QUESTION NO.

100.2

RESPONSE

availability of auxiliary (emergency) feedwater in the event of an EFAS signal. An MSIS (Main Steam Isolation Signal) does not result in closu're of the auxiliary (emergency) feedwater isolation valves in the WNP-3 design (See FSAR Section 10.4.3.2 and Figure 10.4-12). Therefore, no override of MSIS is

required to assure availability of auxiliary (emergency) feedwater and the SER requirements are met.

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Attachment 7 QUESTION NO.

311.3 (3.5.6)

You state that your analysis for aircraft hazards is forthcoming. Provide a schedule for furnishing this information RE SPC'4SE The aircraft hazards analysis scheduled for submittal in December 1982 is still under review by the Supply System. Upon completion of the review process and development of the impact evaluation, revised FSAR Sections 3.5.1.6 and 3.5.3 will be submitted for staff review.

Attachment 2 QUESTION NO.

210.4 (3.9.3.4)

The SRP (NUREG-0800) contains the following requirements:

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: (i) 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 supplier identified, (iv) specify whether the snubber was constructed to the rules of ASE 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-ralated components.

RESPONSE

A tabulation is provided below identifying:

i) safety-related systems which utilize safety related snubbers,

11) total number of safety-related snubbers utilized in each safety-related system, and 111) total number of safety-related snubbers utilized overall.

In addition, all safety-related snubbers specified by the A/E are provided for piping supports and are, a) of the mechanical type (see FSAR Subsection 3.9.3.4) and supplied by ITT Grinnel, b) constructed to the rules of ASME Code Section III, Subsection NF, and c) employed as ss ock arrestors.

Safety-related snun ars supplied by the NSSS Vendor for the Steam Generator Near Side are hydraulic, supplied by Paul-Monroe Hydraulics Manufacturer, ASME Code Section III, Subsection NF and are employed as shock arrestors. The Control Element Drive Mechanism (CEDM) snubbers are mechanical, supplied by Pacific Scientific, ASME Code Section III, Subsection NF and are dual purpose. The Control Element Drive Mechanism snubbers have been evaluated for fatigue strength. All other safety-related snubbers supplied by the MSSS vendor are within the System-80 scope and are addressed in CESSAR-F.

Attachment 2 (Cont'd)

QUESTION NO.

210.4

RESPONSE

System Total No. of Snubbers Utilized in System Reactor Coolant System / Steam 4*

Generator near side Fuel Pool Cooling 14 Main Steam 132 Feedwater 130 Steam Generator Blowdown 166 Control Element Drive Mechanism- 124 *

(CEDM)

Auxiliary Feedwater 3 Safety Injection 275 Chemical Volume Control System 101 Containment Spray 86 Diesel Generator 3 391fef Pilve Discharge 8 erv4n;cient of Radioactive 1 Piping in Chases MISC. Piping Combination 146 Structure l Total Number of Snubbers 1,203 Utilized

• NSSS Vendor

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I Attachment 3 QUESTION NO.

220.8-(33.8.4.8)

! Provide a Section 3.8.4.8 that discusses the effects of masonry walls on other structures in accordance with SRP 3.8.4 (NUREG-0800).

i

RESPONSE

The requested information on masonry walls was provided in response to an NRC Generic Letter dated April 21, 1980. A copy of that response is included as Attachment 12 to this letter. This information will be included in the FSAR as indicated in Attachment 12.

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4 Attachment 4 QUESTION NO.

270.1 1 (3.11.1)

, Tables 3.11-1 and 3.ll-2.are not complete. Provide the missing information or

! a schedule for providing it.

RESPONSE

! Tables 3.11-1 and 3.11-2 will be revised quarterly to incorporate information as it becomes available. An overall completion date is scheduled for June 1984.

i, Attached is the information for December 1982.

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LENCTu 0F TIME suv1 Row-QUALIFl= MNTAL SWO4ARY SEQD TO OPERATE EWVIROW- FUNC- CATION CONDI- 0F TEST SAFETT FUNCTION FOST ACCIDENT MENTAL T105AL TECM- T10W OF TEST DATAf CATE- NICAL POST DBA QUALIF. TESTS /

CATE- ANALYSES TEST RESULTS REPORT F.QU1F. CORT BASES NOGMAL DBA (FOST 2A HOURS) LOCA MSLB TIME TTFE MANUFACTURER MODEL 1ACAT10d COMY

$ R-30 l1 Rs H b tack closed Cleoed 5 5 DCA6 1 151- sers-verser Closed en on sec see TF186SSR R 303 h EL. 365.0 Except SEAS SIAS (S1-332) tor uet 188 Check Talve Test Open Closed Closed 10 10 R-10Al1 O 3D- Serg-Werner R8 H b see see E 303 e SC se on VR078SA EL. 362.5 Shell MSIS MSIS Side

• Seaple R-Serg-Werner RB H b Open Closed Closed 5 5 Cs6-2 CM- on see see

• • TSO44SAR R 303 e 52 Vent ce EL. 362.5 to Entes CIAS CIAS (CM-583)

MSR-10 H b opea Cleoed Closed 5 5 E 2CE- Berg-Werner RS BCP en en see see 1 VF0298AR WVD E 303 e EL. 365 Bleed- CIAS CIAS (Cu-506) ett 5 3 2 Cu- Berg 40ereer RAS C-5 t 1 see see TF030SSR A 352 (CM-505) EL. 362.5 Serg-Werner RS H b Open Closed Closed 2 C5- to se se VSO49SAR R 303 e Free- CIAS CIAS (CM-544) EL. 365.5

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LENGTH OF TIME ENVIRON-ENVIROP* FUNC- REQD TO OFERATE QUALIF1- MENTAL

• SUMMART

. CATION CDNDI- 0F TEST MENTAL TIONAL TECH- SAFETY FUNCTION ~ POST ACCIDENT POST DBA ' QUALIF. TESTS / TION OF TEST DATA /

CATE- CATE- NICAL ANALYSES TEST RESULTS REPORT

• 2QU1F. BASES NORMAL DBA (POST 24 HOURS) LOCA MSLS TIME NO. TTrE MANUFACTURER MODEL 1ACAT10N CORY GORY RAB D-5 a Lock A! 251- open VQO2753R A 141 3:

i (SI-H 8) EL. 335 .

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, 1 251- RAB S-5 e lack A 162 opea VQ046SA (SI-6M) EL. 335 231- RAS S-5 a lack A 139 open TQ04955

($1-667) EL. 335

NONE

.' Limit NAMCO EA-100- i

• Switch 31302 N0ht Limit RAM 00 EA-100-Switch 32302 Qa-3 Electro Anchor / E-6232 h Hydraulic Darling Actuator gy{

valve Tag pos.

> 5 eec 5 see I 2MS- Anchor / RAS S-6 b Open Closed Same I

Der 11ag A 746 M.S. os so E VD001SA Supply MSIS DBA 2 EL. A17.5 o

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"*a mS- me/ ras 9-6 b M.S. on se

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LENCTN OF TIME ENVISON-ENVIRON- FUNC- REQO 10 OrtRATE QUALIF1- WNTAL SIROGARY MENTAL TIONAL TECH- SAFETY FUNCTION POST ACCIDENT CATION CONDI- 0F TEST EC t. CATE- CATE- NICAL POST DBA QUALIF. TESTS / TION OF TEST DATA /

NO. TYPE MANUFACTUREE NODEL thCATION GORY CORY BASES NORMAL DBA (P0$7 24 HOURS) LOCA MSLB TIME ANALYSES TEST RESULTS REPORT 2MS- Anchor / RAB 3-6 b opes Cle we Same 5 sec 5 ese TD003SS Darling A 718 N.S. on as EL. 417.3 Supply NSIS DBA 2MS- Anchor / RAB D-6 b Opee Closed Sense $ see S sec T000485 Darling A 718 N.S. en as EL. 417.5 Supply NSIS DtA pggg A1995 Electre-PAIS959 Bydreolic M Actester Peel thneree qft-105 Limit NANDO EA180 Rev 1 Talve setteh Tag Nee. 1 2NS- Centrol R&B D-6 b Norest Operater If AF 13 13 i P012SA Compes- A 746 Clocad to opes Water eec see

." este . EL. 417.5 (PC) Talves Sys to U Atase- Noemally operaties E. pheric for het thte

" Diony standby vol. will 1 Valves and re- be le lease of operettoe decay for heat estended hot standby 2NS-7016ss Centrol RAB 3-6 b Norest Orerster if AF 13 13 g Ceepea- A 746 Closed to open Water sec see este EL. 417.5 (FC) Valves Sys to Atmoe- Massally operating pheric for het this Duer steadby wel, will

[ Valves and re- he te g lease of operation g decay or for a

" heat esteeded bot standby I

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e LENCTM OF TIME ENVIs0Nr ENVIRON- FUNC- REQO TO OPERATE QUALIFI- MENTAL

SUMMARY

MENTAL T10NAL TECN- SAFETY FUNCTION FOST ACCIDENT CATION * (X)NDI- OF TEST EQUIF. CATE- CATE- NICAL 895T DBA QUALIF. TESTS / TION OF TEST DATA /

NO. TYPE MANUFACTURER MODEL 1ACAT10N WY GORY BASES NORMAL DBA (FOST 24 HOURS) LOCA MSLS TIME ANALYSES TEST RESULTS REPORT tic MATRTE 33122 Not g Actuato SR60 Re-utred Lielt RAM (2 BA 740 EA-740 Switch 20100 Rev 2 3-ifay AS(D NF S31 AQS216 Solomote 664E 78/TR Velve Rev 0 3CC- Centro- ras a-3 h tised Open As dea 5 5 g 35115A metice A-137 to to see see EL. 335.0 moraal Pro-ehut- vide dove cootles 1 3CC- Centre- RAS B-3 b WTR B51285 metics A-186 EL. 335.0 w

• Acteater Aimes SA6-54 SAN 12-

~ 75 4 /CE 63

• 3CC- Centre- RAS S-3 b Open Closed en some se 11 11 Rev 1 351588 metico F 303 SIAS & DSA see sec 3CC- EL. rate of B516ss 362 water level decrosse elseal 3CC- Centre- RAS D-3 b Coettes closed on Some 11 11 1 B51384 metico F 301 water SIAS & AS DBA sec see EL. 362 to rate of

> mise, water-I 3CC- Centre- RAB B-3 b ogetp. level 11 11 E B5145& antics F 301 decrease see see I EL. 362 eignet n

g 3CC- Centre- RAB A-1 Open Closed on Some se 11 11

• B531sA antics F 101e SIAS & DBA sec see EL. 335 rate of water h 3CC- Centre- RAB A-1 level 11 11 g 353288 metics F 10lb decrease see see g EL. 335 elseet

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Attachment 5

, QUESTION NO.

271.1 (3.10)

Table 3.10-1 is not complete. Provide the missing inforrrpion or a schedule for providing it.

t

RESPONSE

Table 3.10-1 will be revised quarterly to incorporate information as it becomes available. An overall completion date is scheduled for February 1984. As of December 1982, no new information is available for inclusion into the FSAR.

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Attachment 6 QUESTION NO.

281.1 (6.1.1.2)

For all postulated design basis accidents involving release of water into the containment building, estimate the time-history of the pH of the aqueous phase in each drainage area of the bu'ilding. Identify and quantify all soluble acids and bases within the containment.

RESPONSE

In our November submittal, the Supply System indicated that a complete response to this question would be available by December 1982.

It has now become apparent, following preparation and review of an initial proposed response, that much more detailed work is required to provide an accurate response to this question than was originally thought.

We anticipate this effort to be complete and a response submitted for NRC review by March 1983.

1 1

Attachment 8 QUESTION N0.

410.7 (3.5.2)

This section does not provide the detail required by Regulatory Guide 1.70 which states that it should be demonstrated tnat safety-related structures, systems and components are adequately protected against very low probability 4 missile strikes by physical barriers or protective structures. According to the Standard 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

l l The plant Seismic Category 1 Structures that are designed to protect the housed safety-related systems and components from the externally generated missiles are identified in Table 3.5.1-3. The plant design does not utilize an essential service water intake. The Ultimate Heat Sink is provided by the dry cooling towers (see Section 9.2.5) housed within a protective structure as identified in Table 3.5.1-3.

FSAR Table 3.2-1, Equipment Classification, identifies the location and the method of tornado missile protection for safety-related systems and components. The missile protection considerations for buried components, access spenings, and penetrations in structures are given below.

Buried Piping i a) Component Cooling Water System piping connecting the RAB and Dry Cooling Tower structure, b) Diesel Oil supply lines connecting the RAB and Diesel Oil Storage Tank enclosures, and c) Auxiliary Feedwater System piping connecting the RAB and Condensate Water l Storage Tank enclosure.

All these buried pipes are covered with a minimum thickness of 1 1/2 in. to 2 in of lean concrete for cathodic protection, and further with a minimum depth i of 3 f t. 6 in. to 4 f t. 6 in. of Supply System Quality Class I compacted backfill. These covers adequately protect the buried piping against penetration by the design basis tornado missiles. The depth of earth l

penetration is calculated using the formula presented in the paper " Depth i Prediction for Earth-Penetrating Projectiles", by Wayne Young, Journal of Soil i

Mechanics and Forndations Division, Proceedings of the ASCE, Vol. 95, No, SM3, May 1969.

1

Attachment 8 (Cort'd)

QUESTION NO.

410.7 Buried Electrical Duct Runs The buried Class IE duct runs consist of those leading from the RAB to the Dry Cooling Towers and to the Condensate and Refueling Water Storage Tank i enclosures.

All these buried duct runs are encased in reinforced concrete with a minimum concrete cover of 5 .in. and further covered with a. minimum depthoof 4 ft. of

! Supply System Quality Class I compacted backfill. These covers have been-shown to be adequate against tornado missile penetration.

Access Openings All exterior access openings in the Seismic Category I. structures are provided with tornado resistant doors consisting of 17 doors in the RAB, 6 in the FHB, and 2 in the Condensate Water Storage Tank enclosure. With the exception of 4 doors located on the roof level of the RAB, all access doors are designed to withstand the effects of the tornado missile impact as discussed in Section 3.5.3. The doors located on the RAB roof are those which provide access to the elevator shafts and stairwells, and are designed only for the effects of the tornado wind and pressure drop.

Penetrations All ventilation openings in the Seismic Category I structures are provided r

with missile protection steel gratings as discussed in Section 3.5.3. The penetrations for the safety-related piping and electrical duct runs are located below grade and are the entry points to the buildings for the-safety-related buried components.

The penetrations above grade are primarily those exhaust or discharge pipes at' the RAB roof associated with main steam atmospheric dump valves, auxiliary

, feedwater pump turbine drive, and main steam safety valves. The diesel generator exhaust and diesel auxiliary vents penetrate the RAB walls. These penetrations are designed such that the missile impact will not impede the intended safety function of each system. Other piping penetrating the RAB walls above grade are the main steam and feedwater lines. The missile impact I

on these penetrations will not compromise the operability of the associated containment isolation valves.

WNP-3 FSAR Section 3.5.2 and Table 3.5.1-3 will be amended to reflect this response.

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, FSAR S 4/O 7 TABLE 3.5.1-3 L A TORNADO MISSILE 00NCRETE BARRIER 28-Day Structure (l) Minimum Thickness (2) Concre te Strength Shield Building Cylindrical Wall 3'-0" 5000 psi Dome 2'-6" 5000 psi Reactor Auxiliary Building Wall 3'-6" 5000 psi Roof 2'-0" 5000 psi Fuel Handling Building Wall 3'-6" 5000 psi Roof Slab 2'-0" 5000 ps1

.. 7 .

Condensate Storage Tank Enclosure

~~

Wall 3'-0" 4000 psi Roof Steel Grating N/A Dry Cooling Tower Enclosure Wall 2'-0" 5000 psi 1

Roof Steel Grating N/A Notes:

1) The systems and components protected by these structures are . identified in Table 3.2-1,

2) See Subsection 3.5.3 for barrier design details.

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FSAR Q AIO.7 3.5.2 STRUCTURES, SYSTEMS, AND COMPONENTS TO BE PROTECTED FROM Il EXTERNALLY GENERATED MISSILES

.5 Structures, systems, and components tnat are designed to withstand or to be protected from the effects ot externally generated missiles are those necessary to ensure:

a) Tne integrity of the reactor coolant pressure boundary.

b) The capability to shutdown the reactor and maintain it in a safe shutdown condition.

c) The capability to prevent accidents which could result in potential offsite exposures that are a significant fraction of the guideline values of 10CFR100. .

3.5.2.1 Protection Provided Safety-Related Structures, Systems And Components Against Impact of Externally Generated Missiles The plant seismic Category I Structures that are designed to protect the housed safety-related systems and components from the externally generated missiles are identified in Table 3.5.1-3. Table 3.5.1-2 identifies postulated tornado missiles. High trajectory and low trajectory turbine missiles are discussed in Subsection 3.5.1.3. ;NP-3 design does not utilize an essential service water intake system. Plant make-up water supply is non-safety and*

provided via the Ranney Well Make-up System (2.4.13) and therefore not A afforded missile protection. The Ultimate Heat Sink is provided by the dry i (..r cooling towers (see Subsection 9.2.5) housed within a protective structure as identified in Table 3.5.1-3.

Table 3.2-1 Equipment Classification, identifies the location and the method of tornado missile protection for safety-related systems and components. The 2 missile protection considerations for buried components, access openings, and penetrations in structures are given below.

j buried Piping i

h efecymrelated hnri Q ipisag lancl A .

a) Component Cooling Water System piping connecting the RAB and Dry Cooling Tower structure, ,

o) Diesel Oil supply lines connecting the RAB and Diesel Oil Storage Tank i enclosures, l c) Auxiliary Feedwater System piping connecting the RAB and Condensate Water Storage Tank enclosure, and

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-d-) Ci.w ical and Valnne Control Sycr e nipia c^nsec hsis die AAs~anc

- RefueLin g ater Storaze Ta %n(4estree c . A

3. 5- 1/. Ame : . - :ic. ^, (12/S2)

u96bd-2 Ep.3 C E Fsa g 4;g,7 All tnese buried pipes are covered witn a minimum thickness of .-: in. to 2 II/C=.

in. of lean concrete for cathodic protection, and furtner with a cir.imum deptn hjif of 3 ft. 6 in. to 4 ft. 6 in. of Supply System Quality Class I compacted ~

backfill. These covers adequately protect the biried piping against penetration by the design basis tornado missiles. Tne depth of earth penetration is calculated using the formula presented in the paper "Deptn Prediction f or Earth-Penetrating Projectiles", by Wayne Young, Journal of Soil Mechanics and Foundations Division. Proceedings of the ASCE, Vol. 95, No. SM3, May 1969.

Buried Electrical Duct Runs Ine buried Class IE duct runs consist of those leading from the RAB to the Dry Cooling Towers and to the Condensate and Refueling Water Storage Tank enclosures.

All these buried duct runs are encased in reinforced concrete with a minimum concrete cover of 5 in. and further covered with a minimum depth of 4 ft. of Supply System Quality Class I compacted backfill. These covers have been shown to be adequate against tornado missile penetration.

Access Openings

• All exterior access openings in the seismic Category 1 structures are provided witn tornado resistant doors consisting of 17 doors in the RAB, 6 in the FHB, and 2 in ene Condensate Water Storage Tank enclosure. With the exception of 4 ,_,

doors located on the root level of the RAB, all access doors are designed to p"" =

witnstand the effects of the tornado missile impact as discussed in Subsection Miit..

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3.5.3. The doors located on the RAB roof are those which provide access to tne elevator shafts and stairwells, and are designed only for tne effects of tne tornado wind and pressure drop.

4 Penetrations All ventilation openings in the seismic Category I structurer are provided witn missile protection steel gratings as discussed in Subsection 3.5.3. The penetrations for tne safety-related piping and electrical duct runs are located below grade and are the entry points to the buildings for the safety-related buried components.

Tnn penettations above grade are primarily those exnaust or discharge pipes at the R!d roof associated with main steam atmospneric dump valves, auxiliary feedwater pump turbine drive, and main steam safety valves. 'lue diesel generator exhaust and diesel auxiliary vents penetrate tne RAS walls. These peratrations are designed sucn that tne missile impact will not impede the iniended safety function ot eacn system. Other piping penetrating the RAB walls above the grade are the main steam and feedwater lines. Tne missile impact on these penetrations will not compromise the operability of tne associated containment valves.

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QUESTION N0.

440.1 (5.4.7.1.3)

Provide the post-LOCA design heat load for the shutdown cooling heat exchangers. Also, Note 2 in Table 5.4.7-2 should be replaced by specific information rather than a general referene.e te the applicant's SAR.

RESPONSE

Thepost-L0gAdesignheatloadfortheshutdowncoolingheatexchangeris 133.44 x 10 BTU /hr (for one SDCHX) based on the containment long term LOCA analysis and unfouled shutdown cooling heat exchangers. In addition to the above, notes 1, 2 and 3 will be deleted from the subject table since they are not applicable and new notes added for clarification ie:

(1) Maximum rates based on Containment Long Term LOCA Analysis and unfouled shutdown heat exchanger. Heat load is for one SDC Heat Exchanger.

(2) Total heat load for both SDCHX.

FSAR Subsection 5.4.7.1.3 and Table 5.4.7-2 have been amended (Amendment 2) to reflect the response to this question.

. 166' 9W-11 f /

-# l FSAR

((4m441 F $

OeptO.l Typical cooldown curves are shown in Figures 5.4.7-1 and 5.4.7-2 of this FSAR. .[ ' .;;n Replace CESSAR-F Subsection 5.4.7.1.2g with the following: l2 1..m.:~

~

g) The shutdown cooling heat exchangers are sized to remove decay heat 27-1/2 hours af ter shutdown based upon a refueling water temperature of 135F and a component cooling water temperature of 95F with an average reactor core burnup of two years.

5.4.7.1.3 Interf ace Requirements Refer to CESSAR-F Subsection 5.4.7.1.3 except for the following subsections.

2 Replace CESSAR-F Subsection 5.4.7.1.3 P.2.b with the f ollowing:

f b) Cooling water shall be supplied at the following temperatures and be able to remove the heat loads listed for the given conditions.

SHUTDOWN COOLING HEAT EXCHANGERS Cooling Water Design Heat Load Situation inle t Temperature (Million Btu / hour) l2 SDCHX' S Pos t-LOCA 65 - 12 0F 133.44(1) l2 Shutdown Cooling: """

3-1/2 hours af ter Shutdown 65 - 95F 247 27-1/2 hours af ter Shu,tdown 65 - 95F 87.7 l2 Replace CES SAR-F Subsection 5.4.7.1.3.P.2c with the f ollowing, c) For all conditions, cooling water shall be supplied as follows:

Required Value Per l2 Pa ramet er SDCHX Normal Allowable Delivery Pressure 80 psig Maximum Allowable Delivery Pressure 120 psig Required Flow rate 3500 gpm 1:aximum Allowable Flow rate 6,000 gpc l2 5.4.7.2 System De sign Re fer to CESSAR-F Subsection 5.4.7.2 except f or the f ollowing subsection.

2 Replace CES SAR-F Subsection 5.4.7.; .2.1, Shutdown Cooling Hea t Ex cha ng e r ,

first paragraph, with the f ollowing.

Note: (1) Design heat load f or post-LOCA condition is f or one SD Heat Exchanger basce on Contain:ent Long-!e r: LOCA Analysis and unfoule: 5 00 H: ,

! +

.4-E Amer '-  !. . 2. (12'E23 _,

IS36W-3 gp,3 # h

'saa a 490. I

,..q.c.,. TABLE 5.4.7-2 SHUTDOWN COOLING SYSTDi INTERFACE REQUIREMENTS FOR (DMPONENT CDOLING WATER Shutdown Shutdown Recirculation Cooling Cooling Following LOCA Mod e (3.5 hrs.) (27.5 hrs) (large Break)

Supply Temperature, F (Max) 95 95 120 Outlet Temperature , F 16 6.2 12 0.1 164.9 Flow per SDCHX, gpm 3500 3500 6,000(1) 2 _

Total Heat load, 10 r

247( ) 87.7( ) 133.44(1)

For both SDCHX 13tt

/

k (1) Maximum rate based on Containment Long Term LOCA Analysis and unf ouled shutdown heat exchanger. Heat load is f or one SDC heat exchanger. 2 (2) Total heat load f or both SDC heat exchangers.

5.4-17 Amendment 5 . 2, (12/E2)

Attachment 10 QUESTION NO.

480.2 (6.2.1.1.3)

Reference or provide 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 result in pressures lower than the external design pressure of the containment structure. Identify the worst single failure that could result in the inadvertent operation of the containment heat removal system.

RESPONSE

The Containment Spray System (CSS) is the safety-related system provided for containment heat removal. The system consists of two independent, separate and redundant trains.

The Containment Spray Actuation Signal (CSAS), as can be seen in CESSAR-F Table 7.2-5 " Failure Modes and Effects Analysis", cannot be spuriously or inadvertently actuated through any single component failure. A. single failure could result in actuation of only one containment spray pump or the associated containment spray isolation valve actuation relay; not both, thereby limiting either flow or flow path and resulting in no flow to the containment.

Testing of the CSS is arranged in a similar manner. The CS pumps and the associated CS isolation valves are actuated by separate test switches (located in different positions on the control board) and separate relays in such a way that both the pumps and valves cannot be simultaneously actuated.

In addition, the control board valve actuation switches are key-locked to their normal operating position, valves open to permit flow through the shutdown heat exchanger, except the containment isolation valve which is closed.

1

- Manual operation of any valve can only be accomplished via group key-locked switches except for the isolation valves which are actuated by individual key-locked switches located on the control board module, (see Figures 7.3-11 &

12) with the keys administratively controlled. Additionally, unlocking of the key switch produces an alarm in the control room, prior to any other action being taken.

6 Manual initiation of the CSAS can be accomplished from the control board, through the initiation of two of the four manual initiate pushbuttons located on CB-1.

i Attachment 11

, Request for Additional Information Enclosure 4 (Item 1)

! ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT 1

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.70-01B, and a generic-l letter dated October 1, 1980, to all holder of cps and L0s.

RESPONSE

l 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. Our response to Question No. 270.1 contains the information for Tables 3.11-1 and 3.11-2. Additional information concerning documentation will be available' by

January 1983.

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Attachment 12 Request for Additional Information Enclosure 4 (Item 6)

REQUEST FOR INFORMATI0fl ON THE USE OF CATEGORY I MASONRY WALLS

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 discuss any exceptions or deviations from the SRP requirements or the aforementioned 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 sup-porting 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 response to this request for information was submitted to the NRC via letter G03-81-854 dated March 25, 1981 (see Attachment A). Additional information regarding' masonry walls was provided to the staff by letter G03-82-1276, dated December 10, 1982 (Attachment B). As noted in Attachment 8, the initial re-sponse (Attachment A) should be modified by the following correction:

Change the first sentence of Attachment A, Item 5 to read, "The design cri-teria adopted for WNP-3 prohibits the use of concrete masonry walls for any safety-related piping system supports".

A description of the Concrete Block Walls (Masonry Walls) is given in Attach-ment C. FSAR Subsection 3.8.4.1.2 will be amended to reflect Attachment C.

ATTACHfEkT 12 Request for Additional Information Enclosure 4 (Item 6)

Attachment A i

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C ~l.4(, L S M z ,: ,

-r Washington Public Power Supply System A aclNT OPERATING AGENCY

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$ ....... m .o . ,

March 25,1981

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G03-81-854 Docket Nos. 50-508

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50-509 7 .

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. Mr. R. L. Tedesco, Assistant Director Licensing  !'N.83IO30$l Division of Licensing ' '

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 hr bMb 'd

Dear Mr. Tedesco:

CBNF9.3

Subject:

0 PROJECT NOS. 3 AND 5 -

INFORMATION REQUEST ON CATEGORY I HASONRY JALLS NUCLEAR UCtiNSih3

Reference:

Letter. Steven A. Varga (NRC) to Construction Permit and

( Operating License Applicants, Same Subject, dated April 21, 1980.

In order to satisfy your request ir. the above reference letter, the following response is provided concerning the design of Category I masonry walls for WNP-3/5. The numbers correspond to your question numbers.

1. Yes. Concrete masonry walls are being used in the Reactor Auxiliary Building which is a Category I structure. .

2. The loads ar.d load combinations used in the concrete masonry wall design are consistent with the Category I structure design as shown below:

Design Loads a) Dead load (0) b) OBE load (E) c) SSE load (E )

Load Combination ~s a) D+E Working Stress Design b) D + E' Ultimate Strength Design

• * =

m

,- ex Washington Public Power Supply System Mr. R. L. Tedesco Page Two l 3.- The concrete masonry walls have been designed primarily in '

accordance with Chapter 24, Masonry, of Unifonn Building Code (UBC) 1973 edition, following the Working Stress Design method specified therein. The wall design was further reviewed by the Ultimate Strength Design Method using ACI Building Code (ACI3/8-71)asaguide. Based on Ebasco's experience, the l concrete masonry block that meet the Supply System Quality t- Class I requirenents are not readily available commerically.

l Therefore, all concrete masonry blocks will be procured on i

the basis of the Supply System Quality Class G requirenents.

. however, with provisions for material certification that assures l adequate and consistent strengths as assumed in the design.

Additionally, the bonding mortar, concrete fill, and reinforcing steel to be used in the construction of concrete masonry walls will all be supplied in accordance with the Supply Systen Quality Class I requirenents.

Lf 4. The OBE and SSE loads in masonry walls located within the Reactor

( '-'-

Auniliary Building were determined by.using the maxima building accelerations at various floors of interest. The building accelerations have previously been established by the seismic dynamic analyses.

5. -The design criteria adopted for WNP-3/5 prohibits the use of concrete masonry walls for any piping systen supports. All masonry walls have been designed for the dead weight and the self-generated seismic loads only, and have not considered any piping support reactions. Typical details of the masonry wall reinforcing are shown in attached Figure 1.

6. The location of all masonry walls within the Reactor Auxiliary Building is shown in the following attached drawings:

1 - WPPS-3240 G-3108 Revision 2 Architectural Floor Plans, Sheet 1. EL 390.0' l .

l ff - WPPS-3240 G-3111 Revision 3 l

Architectural Floor Plans, Sheet 4, EL 335.0' l

. iii - WPPS-3240 G-3112 Revision 2 l

Architectural Floor Plans, Sheet 5, EL 362.5' and 373.5' l

D

. . g,, Washington Public Power Supply System

f. t.

h -

Mr. R. L. Tedesco Page Three iv - WPPS-3240 G-3113 Revision 3 Architectural Floor Plans Sheet 6, El 349.0' and 349.50' and 351.0' y - WPPS-3240 G-3122 Revision 1 Architectural Details and Sections, Sheet 4 vi - WPPS-3240 G-3125 Revision 0 Architectural Concrete Block Wall Details, Sheet 7 l Vii - WPPS-3240 G-3126. Revision 0 1

Architectural Concrete Block Wall Details Sheet 8 This completes our response to the above referenced letter, t

Very truly yours,

, { ,.-

~

i Program Director (Acting) WNP-3/5 Attachments cc: D. Smithpeter, BPA WNP-3/5 Files, Richland W. Woods. NUS l N.S. Reynolds, D & L Ebasco NY (Licensing)

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