ML20147G463

Submits Revised Notrump Agenda for 970312 Meeting & Proposed Agenda for Wc/T Meeting on 970313

ML20147G463
Person / Time
Site:	05200003
Issue date:	02/07/1997
From:	Novendstern E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Huffman B NRC
Shared Package
ML20147G451	List: ML20147G463 ML20147G474 ML20147G485 ML20147G514 ML20147G530 ML20147G539 ML20147G545 ML20147G555 ML20147G566 NSD-NRC-97-5033, Submits Formal Transmittal of Correspondence Previously Sent Informally Covering Period 970205-0316.Index of Encl Matl Attached
References
NUDOCS 9703280152
Download: ML20147G463 (132)

Text

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khhbk s d n- -em j g\s C O V E R ,

Brian A. Mc Intyre S H E E T l

4 To: Bill Huffman (NRC) )

cc: B. McIntyre (Informal NRC File;, Larry Hoch eiter (Fax), Bob Osterrieder,  !

Mike Young, Andy Gagnon, Dan Gamer, Bob Kemper, File 7.6 i

Subject:

NOTRUMP & WC/T l Date: February 7,1997 j Pages: Two, including this cover sheet.

7Q

COMMENTS

l Bill,

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1. Attrched is proposed agenda for our WC/T meeting on 3/13. Please give a copy to l Larnbrose and let me know if he has any comments. l l

l 2. Attached is the revised NOTRUMP agenda for the 3/12 meeting, which we discussed )

today.

3. bon 8 OCA S AR revision is in f' mal stages of publication and is scheduled to be shipped tomorrow.

4. The questions that we are working on from the original set are numbered:

Ic,d,e,i( 1 )

2e.f.g; 3; 7e,f 8a,b,f g.i 9a.c; 10a,b.c,d; 11; 12c,d,e.g h,i,j,k,1,m 13; 14; 15; 16; 17 Please pass on to Cliff for prioritization. Thanks.

From the desk of...

Earl H. Novendstern Manager, Advanced and VVER Plant Safety Analysis Westinghouse PO Box 355 Pittsburgh PA 15235 (412) 374 4790 Fax: (412) 374-4011

.,en i :.., i m ne_n o' l

9703200152 970321 I PDR ADOCK 05200003 i E PDR 1 1

1 AGENDA March 13,1997 Thursday,8:00 am l Westinghouse Rockville Office NOTRUMP MEETING

1. Introduction

2. SPES Results I

3. OSU Results

4. ACRS Meeting i

a. Executive Summary j

b. Proposed Agenda i

c. NRC Feedback on Approach 1

5. Documentation Closure

a. Report

b. RAls/Open items /DSER 6, Wrap-up l

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Merch 1 199' $ t#C NT.M 31' 1

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AGENDA March 12,1997 Wednesday,12:00 pm Westinghouse Rockville Office LONG TERM COOLING MEETING

1. Introduction

2. PIRT

3. WC/T Plant Model

4. Summary of Westinghouse Topical Report i

5. Recent Extended Time Calculation Results j

6. Summary ,

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7. ACRS Agenda e.,,o i. :.., . ': c ~ n,

9 FAX to DINO SCALETTI February 14, 1997 CC: Sharon or Dino, please make copies for: Bill Huffman Ted Quay Robin Nydes Chip Suggs Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEM #172 (M5.2.5-29) l In my quest to make sure we have provided NRC with everything needed to prepare an FSER, I am ,

researching open items from the smallest item number on. The relevant documentation related to l Open item #172 (MS.2.5-29) is attached. We provided the original comparison to STS with NSD-NRC-96-4833 on October 11, 1996. We then provided probability risk assessment information related to the differences from STS with NSD-NRC-97-4939 on January 14, 1997. This was l reiterated in the RAI responses provided by NSD-NRC-97-4972 of February 6,1997. This item I

(#172) was asked by a technical branch other than the Tech Spec branch. The letters identified above were in response to questions asked by the Tech Spec branch. Please help us provide the branch to branch coordination required to obtain proper review of this information. We believe that the letter identified above resolve the concerns of item #172. It seems a reasonable request that NRC acknowledge receipt of the information. We request that NRC provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N".

Thank you.

Jim Winters 412-374-5290

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AP600 Open item Traciumig System Database: Executive Sununary pese 2/14/97 Seledion: litem nol between 172 And 172 Sorted by item # +

Isem DSER Section/ Resp (W) NRC

• Tale /thvion No Bramh Question Type Detal Status Enguicer W Status letter No. / Duse 172 NRR/SPLH $25 MTG Of TECHSPEC/Suggs. C. Chud Action W ,

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M5 2.5-29 tREACTOR COOLANT PRESSURE BOUNDARY LEAKAGE) STS 3 415 states'that,'shoidd the contanment aar tooler corukasase now l f rase monitor hecume inoperable, a channel check should le performed on the contannient atmosphere radioactivny imwntaw usar per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Time APbOO

,TS 3.4 9 stases that a grab sample sloould be performed once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Westinghouse should prmade jushficasmus segarding the accegnatehty of the  !

ahemaec action Action. suberut T.S. 3 4 9 wnh June % rev. rha V28 l Chud - Wuh assesance of the Tech Specs in SSAR Rev. 9 [

l Action W - Need an explanason of Actum Tunes as they relate to STS, _ J I

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4 westinghouse Energy Systems Ba 355 Pittsburgh Pennsylvania 15230 0355 Electric Corporation NSD-NRC-9M833 DCP/NRC0616

/ Docket No.: STN 52-003

/ October 11,1996 j Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: T.R. QUAY

SUBJECT:

CLOSING THE LAST DSER OPEN ITEM FOR AP600 SSAR SECTION 16.1, TECHNICAL SPECIFICATIONS (TS) l Dear Mr. Quay. 1 i

This letter is written to close the last DSER open item for AP600 SSAR Section 16.1, Technical Specifications (TS). Westinghouse committed to provide written explanation of technical differences l between the AP600 TS and those presented in NUREG 1431, the Standard TS (STS). Attached are: l 1

1. A roadmap which identifies the sections comprising the STS versus those included in the i AP600 TS. For any TS that are included in the STS but not in the AP600 TS, an explanation is provided. For any TS that are included in the AP600 TS but not in the STS, those sections are shaded in the roadmap and explained. Explanations are also provided for other content differences between the STS and AP600 TS.

2. A description of general or overall changes whose explanations apply to multiple TS. l

3. A list of technical differences between the STS and AP600 TS. The TS and BASES are grouped by section and an explanation of each difference is provided.

4. A table of and explanation for those LCOs whose endpoint is defined as MODE 4 for the AP600, rather than MODE 5 or "Go to LCO 3.0.3" per the STS.

Discussions regarding ties between the AP600 PRA and the Technical Specifications will be provided in the response to RAI 630.10.

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.- . . . - _- -- .. - = . - . _ . - . .- -- ._--_ - -

J 4SD-NRC-%-4833 DCP/NRC0616 October 11,1996 i

This submittal closes Open item Tracking System (OITS) item 2353, which is the final open item for the AP600 Technical Specifications, if you have any questions regarding this transmittal, please contact Robin K. Nydes at (412) 374-4125.

I A [

Brian A. McIntyre, ger Advanced Plant Safety and Licensing ,

/nja l Attachment cc: W. Huffman, NRC A. Chu, NRC C. Grimes, NRC N. Liparulo, Westinghouse (w/o Attachments) 1

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Y westinghouse Energy Systems em 355 Electric Corporation Pittsburgh Pennsytverna 15230-0355 NSD-NRC-97-4939 DCP/NRC0705 Docket No.: STN-52-003 January 14, 1997 Doc'ument Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 .

ATTENTION: T.R. QUAY-

SUBJECT:

WESTINGHOUSE RESPONSE TO RAI 630.10

Dear Mr. Quay:

Enclosed are three copies of the Westinghouse response to RAI 630.10 regarding AP600 Technical Specification deviations from NUREG 1431 based on probability risk assessment. The NRC techmcal staff should review this response as part of their review of the AP600 Technical Specifications. "Ihis closes DSER open item tracking system item #3054. If there are any qu:stions regarding this transmittal, please contact Robin K. Nydes at (412) 374-4125.

Brian A. McIntyre, Manager Advanced Plant Safety and Licensing .

/jml enclosure cc: Angela Chu, NRC - (w/ enclosure) ,

W. C. Huffman, NRC - (w/cnclosure)

Nicholas Liparulo, Westinghouse - (w/o enclosure) l l

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, NRC REQUEST FOR ADDITIONAL INFORMATION

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Question 630.10. Provide a list of proposed AP600 Technical Specification requirements that deviate from NUREG-1431 based either totally or panially on probabilistic risk assessment (PRA) or PRA insights.

Response: The deviations from NUREG-1431 are explained in Reference 1. There are no AP600 Technical Specifications which deviate from NUREG-1431 with the PRA cs the basis.

However, selection of a standardized Completion Time or Surveillance Frequency consih;rt *'*"*hl t RA P results as described in Reference 2. Per NRC request, 1

[ attached is a list comparipg the NUREG-1431 Standardized Technical Specification i

'N(STS) completion times and surveillance frequencies to the AP600 TSs. Deviations from4TS-ti ifh are less restrictive than STS times are highlighted and any PRA I relationshi is given in the comment column.

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% 3L i SSAR Revision: NONE 1

References:

1. NSD-NRC-96-4833, Closing the Last DSER Open item for AP600 SSAR Section 16.1, Technical Specifications (TS),10/11/%. l NSD-NRC-96-4699, Westinghouse AP600 Technical Specifications Approach,5/3/96.  !

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i Westinghouse Er.ergy Systems sci 355 Electric Corporation Prftsburgn Pennsylvarta 15230 0355 j

NSD-NRC-97-4972 DCP/NRC0732 J Docket No.: STN-52-003 February 6,1997 Document Control Desk l U. S. Nuclear Regulator" Commisrion

Washington, DC 20555 TO: T.R. QUAY

SUBJECT:

RESPONSE TO RAls 630.11 THROUGH 630.14

REFERENCE:

LETTER FROM NRC TO WESTINGHOUSE (HUFFMAN TO LIPARULO),

.i

" REQUEST FOR ADDITIONAL INFORMATION ON WESTINGHOUSE AP600 TECHNICAL SPECIFICATIONS OPTIMIZATION METHODOLOGY", DATED DECEMBER 12, 1996.

• Enclosed for NRC review are the Westinghouse responses to the following Technical Specification RAls, provided by the above Reference.

630.11 Completion Time Anchor Point 630.12 Surveillance Frequency Baseline 630.13 Request for Response to RAI 630.10 3 630.14 Differences Between the Proposed Tech Specs Approach and Tech Specs Rev. 2 i This completes Westinghouse activity for Open item Tracking System items 4224 through 4227, a report for which is attached. Please advise as to the NRC status for these items, if you have any j questions regarding this transmittal, please contact Robin K. Nydes (412) 3744125.

J ff Brian A. McIntyre Manager 3

Advanced Plant Safety and Licensing

/jml enclosure attachment ~

, cc: W. Huffman, NRC (w/ enclosure / attachment) c A. Chu, NRC (w/ enclosure / attachment) -

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b Westinghouse W FAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION DATE: Th.g r7 f9p 7 NAME: L y ,,%

TO: LOCATION: ENERGY CENTER -

deu Idu6 owl EAST PHONE: FACSIMILE: PHONE: O m ee: c/f g -y y . 3 g p o COMPANY: Facsimile: win: 284 4887 U 5 Al/2.C outside: (412)374 4887 LOCATION:

o Cover + Pages 1+7 The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:

WIN: 264-5125 (Janice) or Outside: (412)374 5125.

COMMENTS

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8. Electric Power involve exclusively limited energy content cables (instrumentadon and control), these minimum distances are reduced to 3 inches and I inch respectively.

Within panels and control switchboards, the minimum horizontal separation between components or cables of different separation groups (both field-routed and vendor- '

supplied internal wiring) is 1 inch, and the minimum vertical separation distance is 6 inches.

The exceptions to the guidance in Regulatory Guide 1.75 are based on test results used to support exceptions to the separation gmdance for operating nuclear power plants. A cummary of test results from ten electrical separation test programs is documented in Reference 13.

These test programs support the AP600 exceptions.

n Non-Class IE circuits are electrically isolated from Class IE circuits, and Class IE circuits from different separation groups are electrically isolated by isolation devices, shielding and wiring techniques, physical separation (in accordance with Regulatory Guide 1.75 for circuits x in raceways), or an appropriate combination thereof.

} ,

When isolation devices are used to isolate Class IE circuits from non-Class IE circuits, the

_- y circuits within or from the Class IE equipment or devices are iden ified as Class IE and are treated as such. Beyond the isolation device (s) these circuits are identified as non-Class IE l and are separated from Class IE circuits in accordance with the above separation criteria.

i Power and control cables are installed in conduits or ventilated bottom trays (ladder-type).

j  ; Solid tray covers are used in outdoor locations and indoors where trays run in areas where  !

3 / falling debris is a problem. Instrumentation cables are routed in conduit or solid bottora cable 4

X [ tray with solid tray covers as required. The cables are derated for specific application in the  !

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location where they are installed as stated in subsection 8.3.1.3.3. The environmental design

/ of electrical equipment including Class IE cables under normal and abnormal operating Y conditions is discussed in Section 3.11.

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N Separate trays are provided for each voltage service level: 4.16 kV, low voltage power

\ i (480 Vac,120 Vac,125 Vde), high-level signal and control (120 Vac,125 Vde), and low ,

l $ level signal (instnunentation). 480%&powembla n=y LM widr420 Vactl25 Vdos'W~

l f an A c J e lm. Vertically stacked trays are arranged from top to bottom as stated in subsection 8.3.1.3.4. In general, a minimum of 12 inches vertical spacing is maintained between t ays of different service levels within the stack.

The electrical penetrations are in accordance with TFFF 317 (Reference 2). Class IE and non-Class IE electrical penetration assemblies are mairi.ained in a separate nozzle. 'Ihe physical separation of the Class IE electrical penetration assemblies are in accordance with Regulatory l Guide 1.75. The containment building penetratius are described in subsection 8.3.1.1.5.

I Raceways installed in v.ismic Category I structures have seismically designed supports or are shown not to affect safery-related equipment should they fail. Trays are not attached rigidly Revision: 8 June 19,1996 8.3-20 3 Westinghouse

8 e 1-I i INSERT 8.3-Y 4

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i A tray designed for a single class of cables shall contain only cables of the same class except that low 1 voltage power cables may be mixed with high level signal and control cables if their respective sizes i

do not differ greatly and if they have compatible operating temperatures. When this is done in trays, the power cable ampacity should be calculated as if all cables in the tray were power cable, unless i position and grouping are controlled.

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FAX to DINO SCALETTI Febru:.ry 18,1997 CC: Sharon or Dino, please make copies for: Bill Huffman Ted Quay Don Lindgren Chip Suggs Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEM #177 (M5.2.5-34)

In my quest to make sure we have provided NRC with everything needed to prepare an FSER. I am researching open items from the smallest item number on. The relevant documentation related to Open Item #177 (M5.2.5-34) is attached. We provided the original responses to RAls 410.16 I through 410.20 with ET-NRC-93-3840 on March 18, 1993. We then provided a revision to the SSAR describing our conformance with Position C.9 of the Reg Guide on December 20,1996. this information is consistent with the technical specifications. We believe that this information resolves the concerns of item #177. It seems a reasonable request that NRC acknowledge receipt of the information. We request that NRC provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N". Thank you.

m;+

Jim Winters 412-374-5290 9

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APtes Open item Tracking Systems Datshene: Executiv2Sununary Date: 2H8/97 Selection: litem nol between 177 And 177 Sorted by item #  ;

hem DSER Sectinn/ Tale /lksenpaum Resp (W) NRC f No Branth Quessum Type Iktal Staus Engurer Samus Status te,ict No / Dme 177 NRR/SPLH 5.2.5 MTG4M Landgsen,D Ckwed . Actum W [

MS 25-34 (REACTOR COOLANT PRESSURE BOUNDAR Y LEAKAGE) Add the responses to the folkming RAls to the SSAR. 410 16,410.17 410 18,410 19,410 20

,a :- . -. 2L Closed - SSAR Rev. 3 included the informatum from the RAI responses  !

' Actum W -Iksenhe the w.Jw. ._..= wah Ihitum C9 of R.e.g G_end.e. S_e_e re_s.ponse to RAI 41017 for confirmance informatum_. __..  ;

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.y Westinghouse Energy Systems sm 3ss Pmsburgh Pemsytvama 15230-0355 Electric Corporation l

ET NRC.93-3340 s

f NSRA APSI 93-0078 Docket No.: SIN-52-003 i i m' ' "If J -.,f.

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,f Document Control Desk t U.S. Nuclear Regulatory Commission  : 0t MAR l 81993 Washington, D.C. 20555 Brian A. Mc intyre NITENTION: R.W.BORCHARDT t l

SUBJECT:

WESTINGHOUSE RESPONSES TO NRC REQUESIS FOR ADDITIONAL l

INFORMAT10N ON THE AP600 )

l

Dear Mr. Borchardt:

4 i

Encised are three copies of the Westinghouse responses to NRC requests for additional information i on the AP600 from your letters of November 16,1992 and January 26,1993. This transmittal

]' completes the responses to the November 16,1992 letter. A listing of the NRC requests for additional information respondM to in this letter is contained in Attachment A. Attachment B is a complete listing of the questions associated with the November 16,1092 letter and the corresponding

Westinghouse letters that provided our response.

] If you have any questions on this material, please contact Mr. Brian A. Mclatyre at 412 374-4334.

1 1

i Nicholas J. Liparulo, ger Nuclear Safety & Regulatory Activities l

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4 cc: B. A. McIntyre Westinghouse  !

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, ET-NRC-93-3840 ATTACHMENT A

. AP600 RAI RESPONSES SUBMITTED MARCH 18,1993 RM m Issue f 410.016 \1 Reactor Coolant Leakage

410.017 't Reg. Guide 1.45, Position C.9

410.018 l} Reg Guide 1.45, Position C.8 l 410.019 / Reg. Guide 1.45, Position C.7 410.020 / l Reg. Guide 1.45, Position C.6

'41&o23 l First stage ADS hydrostatic loads 410.025 l Reg. Guide 1.52 l 410.027 l Equipment requiring protection from flooding 410.028 l Potential sources of flooding 410.030 l Maximun flood level

. 410.033 l Flood protection 410.034 l Flood protection 4

410.037 l PXS equipment location I 410.040 l Multi-door passageways leakage prevention i 410.043 l CCW layout 410.044 l Flood hazards 410.046 l Break protection from open cycle systems

410.047 l Water tight doors 410.048 l SFP cooling pumps & heat exchangers flood prot.

410.049 l Flood consequences 410.050 l Flooding protection ior remote shutdown panel

410.051 l Equipment requiring missile protection 410.052 l Turbine missiles l 410.053 l Secondary missiles 410.054 l Equipment protection 410.059 l Stored energy - nuts, bolts and studs I

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l NRC REQUEST FOR ADDITIONAL INFORMATION I

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Question 410.17 Position C.9 of RG 1.45 states that the technical specifications should address the availability of various types of j instruments for RCPB leakage to ensure adequate coverage at all times. Describe how the AP600 design will meet {

this regulatory position (Section 5.2.5).

Response: i l

SSAR Chapter 16, Technical Specification 3.4.9. defines the operability requirements for RCS leakage detection l instrumentation. In addition, instrumentation used to identify reactor coolant pressure boundary leakage is designed )

so that its operability may be determined at all ti ~a detector fail (sigWbrated. range or l self-monitored trouble detected), the strumentation system will alarm in the main control'ioBifrifstshe l specific leak detection monitor ut is questionable. The alarm prompts the operators to observe other sensors s  !

providing leak detection in ation. Technical Specification 3.4.9 allows leakage to be averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; therefore, operators ha sufficient time to determine if small leaks are from the reactor coolant system and to take corrective action in orderly manner. }

SSAR Revision: NO 410.17-1 W Westinghouse

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i 5, Reactor Coolant System cod Connected Systems  !

Reactor coolant pressure boundary leakage is classified as either identified or unidentified  !

leakage. Identified leakage includes:  !

Leakage from closed systems such as pump gasket or reactor vessel seal leaks that are captured and conducted to a sump or collecting tank Leakage into auxihary,systerns and secondary systems (intersystem leakaget(This I leakage is not considered to be part of the 10 gpm limit identified leakage in the  :

I bases of technical specification 3.4.8. His additional leakage must be considIred in I i the evaluation of the re'.idtors;oolanLmventory balanc i

Other leakage is unidentified leakage. l l

5.2.5.1 Collection and Monitoring of Identified Leakage Identified leakage other than intersystem leakage is collected in the reactor coolant drain tank. The reactor coolant drain tank is a closed tank located in the reactor cavity in the containment. The tank vent is piped to the gaseous radwaste system to prevent release of i radioactive gas to the containment atmosphere. The liquid level in the reactor coolant I drain tank and total flow pumped out of the reactor coolant drain tank are used to calculate the identified leakage rate. Rese parameters are available in the main control room. The reactor coolant drain tank, pumps, and sensors are part of the liquid radwaste system. The following sections outline the various sources of identified leakage other than intersystem leakage.  ;

I 5.2.5.1.1 Valve Stem Leakoff Collection Valve stem leakoff connections are not provided in the AP600.

f I 5.2.5.1.2 Reactor Head Seal ne reactor vessel flange and head flange are sealed by two concentric seals. Seal leakage is detected by two leak-off connections: one between the inner and outer seal, and one outside the outer seal. These lines are combined in a header before being routed to the reactor coolant drain tank. An isolation valve is installed in the common line. During normal plant operation, the leak-off valves are aligned so that leakage across the inner seal drains to the reactor coolant drain tank.

A surface-mounted resistance temperature detector installed on the bottom of the common reactor vessel seal leak pipe provides an indication and high temperature alarm signal in the main control room indicating the possibility of a reactor pressure vessel head seal leak.

The temperature detector and drain line downstream of the isolation valve are part of the liquid radwaste system.

l He reactor coolant pump closure flange is sealed with a welded canopy seal and does not I require leak-off collection provisions.

Revision: 10 W Wastinghouse ( h' 5.2-21 December 20,1996

4 FAX to DINO SCALETTI February 18,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don Lindgren Richard Orr Ed Cummins Bob Vijuk i Brian McIntyre l OPEN ITEMS FOR SSAR SECTION 3.8.3 l

This is a background package for the remaining open items for SSAR section 3.8.3. SSAR section 3.8.3 is of interest because by our joint NRC/W schedule, the FSER for this section should be turned into Projects by the end of Mar < h. There are 18 Open Items with NRC Status of Action W. Two (2) of these items (711 and 725) still require some Westinghouse action. Westinghouse believes the other sixteen (16) items were addressed in or prior to the January 16.1997 meeting with NRC.

Currently, our records show no additional outstanding Westinghouse action required for section 3.8.3, except items 711 and 725, and we request that NRC provide a definitive action for Westinghouse or provide direction to change the status of these items. We recommend " Action N" Thank you.

l l

Jim Winters l 412-374-5290 i l

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N AP600 0 pen Item Tracking System Database: Executive Summary Date: 2/18/97 Selectient: [nre st codej=' Action W' And [DSER Sectionl hie '3.83*' Sorted by item #

Item DSER Section/ Tule/th scriptam Resp (W) NRC No Branch Questum Type' Detail Status Engineer Staus Staus Uner No / Dae i

710 NHR/ECUB 383I1 DSER4)I Orr / Bechtel/ NRCSM Cksed Actim W Westmghouse stanM provale in the SSAR the connectum detals between "M" nniules, and between "M" naslules ami oshes types of numlules l Module trhavior study is m progress Design calculanons for unlules will be uplaied following completion of the trhanor study to include any changes I I

m metimmkdogy defined by the study. Additumal connectum detants will te developed dunng this updme and will be mcludcd in design daa to be audned dunng a meeting scheduled for September / October of 1996. Typical connectam detads wdi be th scnhed in SSAR. j Ckned in nretmg wah NRC I/16N7 - nunor SSAR change shown in draA revision ,

j 7tl NRR/ECGB 3 b 31-2 DSER4)I Orr /INI / NRCHM Actam W Actum W Westmghouse should demonstrate that the structure will not hft up dunng an SSE.

LafiofIof the CIS basemat from the conramnent vessel and NI basemat was included in the nuclear island basemat analyses Additional analyses of the l

CIS and NI tusemas respuise to seisnuc laats is m progress.. Dese analyses wdl demonstrate tha hftoff of one side of the CIS basemat is not segmficant. :

Result will be avadable at structural audit l 716 NRR/ECGH 38.32-S DSER4)I Orr / NRCSM Closed Actnm W Wesunghouse shouldJ usti the use of the ANSI /AISC N690 Standard and the ACI 349 Code for concrete filled steel M nulules. l Closed - his issue is addressed m the nuulule trhavior study and included in SSAR Rev. 7. Based on review by the NRC m a meeting on May 22,this issue is closed Meenng notes daed July I,1996 show this item as stdl open. Westmghouse to finalize all design cntena for structural nulules.

, Closed in meetmg with NRC I/16N7 _

717 NRR!ECGH 3833-1 DSER-OI Orr / Raz / NRCSM Chwed Actam W i

t4 Westmghouse shoukt address in the SSAR the entre constructum promss, from off-sne fahncaum to final on sne placement I

% Closed - NRC wdlieview sevision 7 of SSAR, subsectams 3 8 3 and 3 8 4. I p

See NRC letter dated 7/l5N6 - Address use of sectums 1.23, t 25, and 1.28 of AISC N690.

Action W - See NIsC letter of 12NN6.

Ckned in meetmg wnh NRC I/16N7 - mums SSAR change shown in draft revision 788 NRR/ECGB 38.3.3-2 DSER ol Orr / NRCSM Closed Actam W

,Westmgtnise should address the construction-induced stress followmg the cunng of the concrete.

Okwed - SSAR sutwectum 3 8 was revised to aidress stress in nulule due to concrete placenent.  !

NRC necemg notes 7/IN6 show this as Actum W - expand SSAR descripton of the metiumis for considenng the hydrustatic pressure due to constnatum m j the design j Chwed in meetmg wnh NRC t/16N7 - mmor SSAR change simwn in draft reusion j NRR/ECGB 38.33-3 DSER-OI Orr / NRCSM Resolved Actam W 719 Westmghouse shou:d consuler,in the design of the IRWST, the comtunation of the load inwn ADS actuatum and the SSE load In adJune, the thermal loadmg should be considered in the ernernal structural steel frame design.

~

.SSAR Revisum 7 subsection 3 8.3 3 l combines ADS and SSE loads Dermalloahng on seces atures is considered as shown in Table 3 8 4-I {

t Calculatums will he renewed dunng the structural nmxtule audit __, ,

383.4-3 DSER-OI Orr / NRCSM Resolved Actam W 722 NRR/ECGB

__ . . . . . ~ . . . ,

Westmghouse slumid demonstrate the adequacy of the design based on tir assumptam of a coinpmte sectum

-^ : : : . . : :: -.:: ; r.;:- =- ,

_ Resolved based on infonnanon m the nulule behavior stud _y.. _

Page- I Total Records: 18

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AP600 Open Item Tracking System Database: Executive Summary Date: 2/1837 Seletthise: inre st codej=' Action W' And [DSER Section] like '3.8.3* Sorted by hem #

leem DSER Sectionf Titic/Descnprion Resp (W) NRC No Hranch Questum Type Detad Status Engineer Status Status triter No. / Date 724 NRR/ECGM 3834-5 DSER-01 Our Ckwed Actim W

[Westmghouse should use a local 3D sohd model of the module geometry and maenals as the basis for developing equivalent isotropic sheli properties, nr fw justifying the equatums cunently used. _. . . _,.

' ~ ~

' Closed - Tlus issue was addressed in the nniule behavior study f NRC meeting noses 7/1/96 show this as Action W - to provide the analyus and design results to denumstrate and confinn the alequacy of the metint used l for design. thign calculmums are available for audit.  !

, Closed in meetmg with NRC I/16/97 I 725 NRR/ECGH 3834-6 DSER-OI Orr Actum W Actum W

[Westmghouse shmld acceptably address issues relating to the seisnuc nulehng of the containment internal structures. j

' ~ ~

[ Closed - Dis issue was addrewed in the numiule behavior study ')

NR_C meetmg notes 7/1/96 sluiw this as Actam W - design caletdatums to be audned by NRC , j 729 NRR/ECGB 383410 DSER-OI Orr Closed Actum W

,3 jWestinghouse should reuse the combined stress equations in Section 3A.31.3 of the SSAR to reflect reahstic actum of the walls if tuanial bendmg is b Pluimd. ~

M 'M - Tius issue was addressed in the nulule' behavior study )

NRC meetmg notes 7/1/96 show this as Actum W - to reemannne intera tuni equations desenhed in SSAR. }

, Closed in meetmg with NRC t/l6/97 _ . _ j 730 NRR/ECGH 3834-11 DSER-OI Orr / NRCSM Resolved Actum W

,Westmghouse should complete the design of the connection detals and provmb the design for staff resiew. l

~

Resolved - Sciected connectum detaats will' be available for review dunng the s'tructural nafule audit. )

731 NRR/ECGH 3834-12 DSERol Orr / NRCSM Resolved Adam W Westmghouse should compde design summary reposts usmg the format and artnbutes desentvd in Appendix C to Secaon 3 8 4 of the SRP, and sinnskt sutumt the reports for staff review.

Resolved - De design repor' will be available for review dunng the structural unlule audet. l

. 732 NRR/ECGH 383413 DSER OI Orr / NRCSM Resolved Actnm W De staff wdl perform a structural design audit of the contanment intemal structures. j Resolved - De structural module audit is planne'd for late 1996 _

}

2347 NRR/ECGH 383 MTG-OI Orr / NRCSM Closed Action W Westmghouse slumsid desenbe the design ss used for the structural nulule design in the SSAR. f i

IDis is part of the nulule behavior study in progress as well as the update to the hydrudynamic analyses See open item 3 8 f3 4 l0 (item # 729) and item # j 2348. l Closed in rnectmg with NRC t/16/97 l 38.3 MTG-OI Orr / NRCSM Audit N Action W 2348 NRR/ECGB ~

.Westinginsse shoukt revise Appendia ' 3F to address questio's n related to analysis met *als and ADS Icads for the structural module design u: - . .

n. . . ; .. . n 2 , _. _. .2 - rr . .. . . . . ,

! Appendix 3F lias been replaced by matenalin subsectum 3.83. Detaled questions have been aldressed in design calculmums wi.ich are available for audit j in late 1996.

Review of docume_ntation suit canpleted dunng auda _ ,  !

P. ige. 2 Total Records: 1B i

_ . . _ _ _ _ . _ - . _ .__._____________.m .______.=_--____.._-..____.__.m. _.______-_______.__.________________________.___._.__m._.-_-_ _ _ _ _ - _ _ _ _ . _ _ _ _ _ _ _- ____ __--_ ___

e *.

. s AP600 Open Item Tracking Systeni Database: Executive Sunissary Date: 2H8/97 Selection: [nre st code]=' Action W' And [DSER Section] like '3.8.3* Soned by item # .

Item DSER Sectionf Tule/Desenptimi . Resp (W) NRC No Branch Quesium Type Detal Staus Engmeer Samus Staus  % m, / Dme 2149 NRRiCGH 353 MTG4)I Orr / NRCSM Ched Actum W Westinghouse simwald complete analysas'of a 30 mch wall in the M-1 structurU nuntule and rde the analysis avadable for audit.'

4

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Analyses of 30" wall are being finahad and wdl le avantable for hinEl996 , -

Closed in meetmg wnh NRC I/16/97 - nunor SSAR change shown in draA revision ,

3037 NRR/ECGB 383 MTG4)I Orr Closed Actmin W NSD-NRC-96-4732 5/31/96

,Descnbe how concrete cratking is consulered in the thermal analysis and provnie justificmion for the adaquacy of the niettubs used. ' ^ ]

~l

"~

l Closed - Respmse provided in item I of lener NSD NRC-%d732, dmed Mahk 1996 3247 NRR/ECUB 3834 RAl4)I Orr Cksed Actum W RAI 230 98 Aptd 5.1996 letter: Westmghouse should complete the new design of structural nuxtules (using stear stiuh) and sulmut the & sign for staff review. .. *

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Ched - The struttural nmulule design wnh shear studs and other changes is descnhaiin SS' A R sutsectum 3 8 3 I Rev. 7 ~ ~ ~~l P

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Page: 3 Total Records: 18 L

. . . _ . , _ .__m-_ .__. _ -___._____.______._....____.__._m _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ _ . ______m __m._____...___m______-_ _ _.=._ _ ___.m._ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ ___ m_.____

3. _ _ .

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Westinghouse FAX COVER SHEET e

i i RECIPIENT INFORMATION SENDER INFORMATION 1

i

! DATE: 2-#8 'l 7 NAME: 0 'A %c,_

l TO: LOCATION: ENERGY CENTE'R -

& sash EAST I

PHONE: FACSIMILE: PHONE: Omce: qiz- pq -y m

! COMPANY: Facsimile: win: 284-4887 i L S )J R c. outside: (412)374-4887 !

l LOCATION:

4 i

Cover + Pages 1+ g

The following pages are being sent from the Westinghouse Energy Center, East Tower,  !

! Monroeville, PA. If any problems occur during this transmission, please call:

i WIN: 284 5125 (Janice) or Outside: (412)374 5125.

i COMMENTS:

i Lc-l Ae a im sc.o ev6 co r eda dd e tm ~ r e bh

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6. Engineered Safety F::ctures the capacity of the recombiners. Consequently, the containment hydrogen concentration will exceed the flammability limits. This massive hydrogen prcJuction is postulated to occur as the result of a degraded core or core melt accident (severe accident scenario) in which up to 100 percent of the zirconium fuel cladding reacts with steam to produce hydrogen.

I The hydrogen ignition subsystem consists of 5860 hydrogen igmters strategically distributed throughout the containment Since the igniters are incorporated in the design to address a

• I low-probability severe accident, the hydrogen ignition system is not Class IE. Although not I class IE, the igniter coverage, distribution and power supply has been designed to minimize I the potential loss of igniter protection globally for containment and locally for individual I companments. The igniters have been divided into two power groups. Power to each group I will be normally provided by offsite power, however should offsite power be unavailable, then I each of the power groups is powered by one of the onsite non-essential diesels and finally I should the diesels fail to provide power then approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of igniter operation is I supported by th : non-Class IE batteries for each group. Assignment of igniters to each group i is based on pre viding coverage for each companment.or area by at least one igniter from each I group.

The locations of the igniters are based on evaluation of hydrogen transpon in the containment and the hydrogen combustion characteristics. Locations include companmented areas in the containment and various locations throughout the free volume, including the upper dome.

For enclosed areas of the containment at least two igniters are installed. The separation between igniter locations is selected to prevent the velocity of a flame front initiated by one igniter from becoming significant before being extinguished by a similar flame front I propagating from another igniter. The number of hydrogen igniters and their locations are I selected considering the behavior of hydrogen in the containment during severe accidents.

I ne likely hydrogen transpon paths in the containment and hydrogen bum physics are the two I imponant aspects influencing the choice of igniter location.

I I The primary objective of installing an igniter system is to promote hydrogen burning at a low I concentration and, to the extent possible, to bum hydrogen more or less continuously so that I the hydrogen concentration does not build up in the containment. To achieve this goal, I igniters are placed in the major regions of the containment where hydrogen may be released, I through which it may flow, or where it may accumulate. ne criteria utilized in the evaluation I is provided in Table 6.2.4-6. De location of igniters throughout containment is provided in I Figures 6.2.4-5 through 6.2.4-12. He location of igniters is also summarized in Table 6.2.4-7 l identifying subcompanment/ regions and which igniters by power group provide protection.

I ne locations identified are considered ap roximations & 2.5 feet) with the final locations governed by the installation _ details. De igniter locations identitiEi mc c6Hs &

I I roximations & 2.5 feet) with the final locations governed by the installation details.

He igniter assembly is designed to maintain the surface temperature within a range of 1600 to 1700*F in the anticipated containment environment following a loss of coolant accident.

A spray shield is provided to protect the igniter from falling water drops (resulting from Revision: 11

[ W85fingh00$8 -6.2-43 Draft,1997

Y

6. Engineered S:fety Features l

l Table 6.2.4-6 l _

l IGNITER LOCATION 'RITERI l

ax.v I

• A sufficient number of igniters hM [placed in the major transport paths (including dominant natural I circulation pathways) of hydrogen so that hydrogen can be burned continuously close to the release point.

l This prevents hydrogen from preferentially accumulating in a certain region of the containment.

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• Igniters (minimum of 2) showetIE located in/ major regions or compartments where hydrogen may be I released, through which it may flow, or where it may accumulate.

I I

• 11 is preferable to ignite a hydrogen-air mixture at the bottom so that upward flame to agation can be I promoted at lean hydrogen concentrations. Igniters within each subcompartment/ .uld ' ocated in the I vicinity of, and above, the highest tential release location within the subcompartment. 'Gre.__

l l

• r)AJ GLAA

• In compartments with relative , small openings in the diling, the potential may ex' t for the hydrogen-air I

mixture to rise and to collect tear the ceiling. Therfore, one or more igniters " miaced

- near the I ceiling of such compartments. Igniter coverage.hi %cprovided within the upper 1[%  ;

I height subcompartments or 1 ft,from the ceiling whichever is less. In cases where the highest potential I " Sonsidered.

I release point is low in the corhpartment, both this and the previous criteria d~TaAa_,

I

• To the extent possible, igniters should-Be placed away from walls and other large surfaces so that a flame I front created by ignition at the bottom of a compartment can travel unimpeded up to the top.

l l

• A sufficient number of igniters installed in long, narrow compartments (corridors) so that the I flame fronts created by the igniters need to travel only a limited distance before they merge. This limits l l the potential forj significant flame acceleration.

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• Igniter coverge M %tovided to contol combustion in - areas where oxygen rich air may enter into an inerted region with combustible hydrogen levels during a dent scenario.

l

• Igniters se @ located above the flood level, if possible. Those which may be flooded.shetrt8Ta've I redundant fuses to protect the power supply.

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• In locations where the potential hydrogen release location can be defined, i.e. above the IRWST spargers, I at IRWST vents, etc igniter coverage shoghHa@rovided as close to the source as feasible.

I "

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• Provisions for installation, maintenance, and testing n 2g ^ considered.

-b Revision: 11 o wnow:.i nii o:is97 Draft 6.2 200 W W65tiligh0US8

f h Westinghouse FAX COVER SHEET D

! RECIPIENT INFORMATION SENDER INFORMATION 1

DATE: I -it -9 7 NAME: G4 Heu;,

i ace <.A3,v.ty j B.E 3 y m '

TO: - LOCATION: ENERGY CENTER -

EAST l PHONE: FACSIMILE: PHONE: Office: 9i 2 - m -ym  !

! COMPANY: Facsimile: win: 284-4887

! Osync outside: (412)374-4887 LOCATION:

k 1

l 1

1 Cover + Pages 1 + JAF W i <

i The following pages are being sent from the Westinghoua Enorgy Center, East Tower, l

) Monroeville, PA. If any problems occur during this transmission, please call: J i

) WIN: 284 5125 (Janice) or Outside: (412)3Y4 5125.

a  !

COMMENTS:

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j A telecon was held on Wednesday. February 12,1997 between Westinghouse and NRC Probabilistic Safety Assessment Branch to discuss NRC questions on an AP600 PRA sensitivity study. One of the questions the NRC asked during the telecon related to the failure rate Westinghouse used for squib j valve failure to operate. Westinghouse accepted an telecon action item to provide to NRC the Sandia 2

data which was used to develop the AP600 PRA squib valve failure rate. .

J Attached is a copy of the Sandia data for squib valve failure to operate. This information is being

provided to NRC in response to the 2/12/97 telecon action item.

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l t CNCLASSIFIED Sandla National Laboratorier.  ;

t. . .m o, . :: .:..o - .. m s 4:see i i 1

date: Fe:;r ary 19.1996 i

I to: T.m Sveter, West.cghouse

%J}$a.M' ,

I from: C J. DeCaru 6116 subject: Explosive Valve Reliability information

! Jere Har'an asked me to send you information on the reliability of our explosively i actuated valves.

s Explosively actuated valves that cut tubes or punch membranee have assessed f fatture rates that range from .0002 to .0006. No failures have been observed in i these valves. Tnere are cifferences in assessed failure rates because of different l quantit.ss of test data, not Decause failures have occurred.

l i I am enclosing the data assessment sheets for our standardized mini valves.

l

^* that have standardized internal features. The assessed These are cutY-'"decause failure rate E 00Q there have been no failures in over 3600 post- l development destructive tests combined over the several valves in the family.

1 4

Feel tree to call me at 510 294 2561 if you have questions.

ejd:8116 Copy to:

MS 1452 J. G. Harlan.1552 li MS 9202 R. L Bierbaum,8116 i

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i i l i UNCLASSIFIED D strib,t.On 2 l

i i ST AND AR0lZED MINI V ALVES I

WEAPON SYSTEM Several Weapon Systems I i l

13 valves; MC3006, MC3205 MC3206. l

} Comoonents MC3294/MC3784, MC3295. MC3297/

MC3785 Side B. MC3298, MC3425, ll i

MC3427/MC3427A. MC3428/MC3428A.

MC3570/MC4232, MC3604 MC4241 l

(' Gas Transfer System

• - 4 i MAJOR ASSEMBLY i FAILURE EVENT General Failure of the standerdized mini valve to i

' property cut one of two tubes and transfer '

i

' gas, given the proper input to the MC3004/MC2949A/MC3479,MC3753 l

actuators.

C -%, q 2

AS SE SSED FAILURE M 0.0002 gESSMENT DAI,E sunuary 996 Date RFI IABILITY ENGINEER C. J. DeCarii, 8116 Date REVIEWEB '

R. S. Tilley, 8116 Rollebility Assessment Date: i Cumulative data for thirteen mint veNes is summarized below. The sampling rate I for produc6on acceptance D testing of all mint-valves was 5%. This was adopted in Octocer 1981 because of the success history of the MC3006, MC3205 and MC3206 valves and because so many mini valves were to be produced. Prior to l l

this time, mini vanes were tested at a much higher rate to accumulate a data base.

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- UNCLASSIFIED  ;

D stnbut!on 3 i CUVULATIVE VALVE TEST DATA FOR THIP EEN MINI VALVES

$ Data Sotrce No. Tested No. Fal'ed Comme ~s

' A. Development 1048 0 See Table 1 B. Production D tests 2703 0 See Table 1 j i

C. Surveillance i C1 NMLT/SLT 1362 0 See Table 1 l C2 NMFT/SFT 607 0 Total 5720 0 See Table 1 l

Total w/o development 4672 The 0.0002 assessment is a 50% upper binomial confidence limit based on zero i failures in 4672 post development tests.

s' CNCI ASSIFIED C stnDut.on TABLE 1

SUMMARY

OF MINI VALVE D TEST DATA (JANU ARY 1996)

Product Acceptance Stockpile Evaivation Devefoo No. Lot No. t ab. Fha-*

MC No

'43 547 '07 161 100 3006 354 318 45 355 86 3205 194 210 33 223 86 3206 ..

97 161 71 31 59 3294/

3784 78 37 41 116 44 31 72 3295 3297/ 66 227 14 3t 72 3785 415 35 174 36 31 36 3296 65 279 421 48 4 Vafves 29 76 34 76 46 3570 -

59 4232 46 13 4241 56 30 1048 2703 1342 607 TOTALS GRAND TOTAL WITHOUT DEVELOPMENT TES1s s 4472

~

• Includes 4 valves: MC3415. MC3427/A, MC3428/A MC3604 5

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! PRORC I

4 SCEnWRC i

i r= Tr=nte ENERGY DYNAMICS OlVISION 7.e w. ... s, .,

a.. . m ,,ve re ,,e ,; sok- FW-fGod ....;';;;;*:

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Con'peny: , NY2A '

Attn: A k '- OA 8AJ _.

From: L 1 Ja dis 2 , __

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The message is intendes for tne une of the indiv6 duel or erety to which it is edereened and may con-tem information tnet is prnnleged, conheental and esempt trem dieciosure under applicomte law. If the reader of me maneege is not tne etenced rocialent, or me empesyee er egent responsete for deswer-ing the message to be etenced rocceent, you are hereby 6 that any diesemawtaen. sistreution or copying of this communscation is stnewy prehitWtas if you have rece6ved thle communcotion e error, pieene nosty we e ' vnelsetely by telephone, and retum the enginal message to us et the eneve addrene ve regular postel service Thank you.

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! PB CIFIC*

!, SCIE nTIFIC 1

l Energy Dynamics Division i

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! TECHNICAL MEMORANDUM i l

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f 5 THE RELIABILTTY OF FYROTECHNICALLY l ACTUATED VALVES l

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1 14 FEBRUARY 19% l i

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Prepwed by:

Y M A

! Jolin Greendade, Senior Stas Engineer i

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THE RELIABILITY OF FYROTECHNLCALLY ACTUATED VALVE 8 l

j .o INTRODUCTION l

! Pacific Scientfic/ Energy Dynamics Division (PS/EDD) has prepared hs 1

4' technical mernorandum in response to an informal requert for informaton reistive I

!, to the general reliability of pyrotechnically actuated valves, one of this company s product lines. Such valves, in both Normally Open (N 0) and Normally C!osed

! (N.C) sonfiguranons, have been widely used for rnany years in military and

. aerospace applicanons where rapid and positave volving of a liquid or gaseous I working medium is required in a one time control event, such as fuel shut-eff, i

fire suppressant deployment or 3as sampling. The characteristics which have made pyro actussed valves so suitable for such applicamens, in fact the only l reasonable choice in many cases, are their small sins and weight (compared with I

all other competitive approashes),6eir ability a valve very high pressure f working media (in some cases as high as 10,000 psig), their exsrernely rapid

! acaisson time (typically <5 maass) and their relasive lack of complexity. N latter aanbves no doubt contribens e their high opernaonal reliability which i typically is well in essess of 0.999, even aner long panods of dormancy.

I j PS/EDD produced a nusnbar of high reliability pyro valves for the DOE. for use s

in nuclear weapons,

(

j All pyre actuated devices (FADS), including valves, by their very nasure ars "one shoe devises. Cansequendy, he reliability of a PAD cannot be established empiribh by eenduseng a lar0s nwnbar of operational tests repeatedly on the same ladividual imit, as is done with electrisal and electronic components for insmes. Instead, the predicted reliability of FADS must be derived omns data from teses of similar assemblies, to results of stress analyses and recorded failure rean data relasive a similar individual components of the device.

In order a illussrums the methods used m anive at the predicend reliability levels of a pyre monished valve, a relisbility analysis is presented in the following 8

l i secton relatave to a typical PS/EDD 2 way pyro actussed valve, somewhat more comptes than most, as it might be used m a ground-based appiscation such as a

nuclear reactor coolmg system An estimate is denved of its operanonal relia'o ility, as well u its mission rehability he la

ter takes into account an j assumed pened of dormaney.

i 2.0 THE RELIABILTT1 ANALYSIS OFA TYFICAL PS/EDD PYRO i

ACTt'ATED VALVE The Assumed Seensrie l 2.1 The valvt selectsd as an emarnple for analysis is PS/EDD's 51-5575 2 Fuel Valve, shown in Figure 1, which was qualified for use in die TSSAM program.

In this 2 way valve, which performs both N 0 and N C fbncmons, the mlet and by pass ports are in contact with the fuel and are interconnected, prior to actuation During astuation of the valve, a nipple covering the outlet port is sheared away and the inlet and oudet ports are than interconnessed by way of a transverse bore through the piston. The pisaan blocks off the by-pass port as it ,

l completas its stroke.

1 In te following sub-escnons, a reliability analysis of this device is presented which is based on the assumptions below, which are believed to be consistent with a ground based applicanon at a nuclear energy facility i

~

Darmant (Nan-Onaramenan Pened to years a)

Yahm Acmanen Thns 5 msess b)  ;

Valve Funenam* T6 20 minutes c)

('During this period the assuased valve would handle pressunaed flow-down wi6eet less of structural or saaling integrity. A fwiesion nime of 20 mmutes may be longer than required but will provide a conservasive esdmass of reliability.)

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! 2.2 The Reliability Medel The valve can be considered as an assembly of mochanical" components, both

]

I functional and structural, a set of interface seals and a matans self conwned i s assembly which provides the required actuanon gas pressure, namely, the pyro

! carmdge Smcs all of the componenu of the overall assembly must fun =6en j

l correctly and/or mamtasn their structural integnty during the overall life of the

!' device, the componenu are considered senes dependent This permiu the i

j following very simple Reliability Model.

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! COAPONENT REL-  % R, R, i

! Valve Assembh (Inal. Cariridge) Operamenal Ral R, = (R,)(%) --(1) i 2.3 ReRability Amatrais i

i 2.3.1 Omaradenal RaMabiptr (L) 2.3.1.1 ha Pus CartrMas RaMahiEtr (L)

As approach often need fbr prediating he operanomal reliability of pyro cartridges is bened en a relaniendup besween the number of tests conducted wineet failers (N) 6e Confidenes Level (CL) and the Raliability (R). His relasionship, which is derived from 6s binomial 6eorem (Raf.1) is esproued by the equasien

, , Ier (1-31 ....g3 Larim 10

- . _ __ _ _ . _ . . . _ . . _ - _ _ _ _ = _ _ .. . _ _ _ _ _ _ . . _ _ _ . . _ . _ _ . - _ _ . _ _ _ _ _ . _ _ . _ . -

t*

l i ,

i which gives:

i

! p . p a so/= . . . 1 33 a

~

i 1

Through the years, g.lgag 25,000 sundg pyro cartedges have been successfully j

fired by PS/EDD alone, therefore 25,000 would not be an unreasonable value for l

I N in Equanon 3 1 hen, at a CL of 90%, which is also a reasonable level, the I operanonal reliability of the pressure carindge (Re), as given by Equanon 3, would be i

4 . ,u.u em n.see. .,,,,se t

2.3.1.2 The "" mat =hmer ma The $2-5815 2 valve incorporases nina (9) 0 nns seals. Failure rate data for 0, ,

l nnes is given in NHLD-91 (Ed. 2). That document gives a generalized failure 1

resa (1) of a sneo raaures/W heure for " MIL" type 0 rings subjected m a 4 l In the currendy ensumed scenario the "Oround Mobile" (G.M.) environment.

i, valve would be subjected to a mors benign " Ground Fized" (GP) environment.

i In the absence of spesi6s A dass for the W environment it is aerumed to be half i

that of the GM environment i.e., A., = 0.5 A , or, L.,_= 0 5 (Useo) - 3.2941 faslaras/10' As stated in 1.1, he assumed Functies Time (t) could be 20 f

minutes, i.e., .3333 hours0.0386 days <br />0.926 hours <br />0.00551 weeks <br />0.00127 months <br />. As sesed in WIL-STD-756 (Rd. 3), if the failors rate l

e and operusional time are known the reliability is given by the equadon

m. -u . . . . m l

This equasion, which is for a single component can be modined for (n) like

! components se follows 1

3 e e se . . . .(5)

J i

i i

I . _

il .

4 5*

l i

I Substitueng for n, t and L is (5) gives the sd operauonal reliability i

i a, . , + > . m.s. u un ne. . .,,,,,o j

2313 De Mechanical Ps.rts Rahahihty (Ro) l In the 51-5875 2 Valve there are ame (9) mechanical" parts, a)) of which must retaan 6eir structural intagnty during the operaton of the valve and three of i

! them must perform certain funceons. Thus:

a I, comannants (9) Funcipefis (5) l Body i

Adapter i

Plug

{ .

j Lee Plug 1

l Inlet Fitens By Pass Fitting Outlet Fining - Sheer the Cosure Nipple Piseen

/ s Release the Initial Lock

\ ,

r Provide Metal / Mesal Seal r Provide Finallock SealPivs  ;- Most shear Open .

We can eensider the overall opersaional reliability of the medesical components

(%)

• a,, - (a )(4) . . . . (s)

Whers %is the sirucevrai rei.

R., is the festional rel.

Then, if we assume that eash esmpement has been designed with a structwal Safety Facent af et least 1.5 (vari 5ed by seress analysis), esperience has shown that is sawstural reliability will be se least .999999. Therefore-Rw. = (.999999)' = .999991

\L

4 s*

i l.'

For each of the five listed functions we will conservatively assume a funenonal )

I reliability of .999990. Thus-

$  % = ( 999990)' = 999950 The mechanical pans reliability follows.

j Rg = ( 999991)( 999950) = 999941 t

f 2.3.1 A Calculaban of Ey I

Substituting from 2.312,2.3.1.2 and 2.3.! 3 in Equanon (1) we obism the valve's operanonal reliability:

Ry = (.999908)( 999990)(.999941) = .999339 l

2J.2 Massion Rallahdary (R-)

According so HCL-STD-756 (Raf. 3) the " Mission Reliability" is given by the equanon:

R, = (R,)%) . . . .. (7) l Where Roy is the probah;!ity of the unit fusstening as required after being l

dorment for a spesified pened of tims. NPRD-91 (Raf. 2) gives the fbliowing definition for 'dsrment"

" Dormant Component or equipment is normassed to a system in the normal operesional configuraman and esperiences non-operational and/or periodic operational seresses and savironmental stresses. The system may be in a dormant stans for prolonged periods before being veed in a mission."

A value Ser R,, for the entire assembly can be derived fhun Equenon 4. given failere rase (A) data for similar equipment under dorment conditions. De closest available dass in NPRD 91 (Ref. 2) is for a ' Valve. Hydraulic" at a " MIL

• quality level, which has a (A)D of .001s anilurean o' hours. For the assumed

, 13

.. .. - .- . . . - . . - . . - . . . .. - - . ~ . _- .. - - _ _ _ - . - . .

i 1* i d

.i dormancy period of 10 years, i e., 87.648 hours0.0075 days <br />0.18 hours <br />0.00107 weeks <br />2.46564e-4 months <br /> (which meludes 2 extra days for i

leap years) we obtmn:

1

p,,. ,<enei.e....... . ,,,,e43 A

k

! Then from Equinon 7, the esemated Mission Reliability for the valve is.

4 t

}

g - (a)(y . (.sssesel(.ssessal = .ssstat -.

1 Jh' b . .

. */

/ '/u i

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d l.

1

)

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4 I -

a

1 I

1

a s

.h j .' l

,l RUZE ENCES:

1 1 Technical Memorandum Pyremechanical Reliability" - N Butterfield, June 1931, f

)

Un numbered. Martin Mariens T M.

2 NPRD 91 *Non-Electronic Pars Rehability Data 1991* Rehabihty Analysis Center,

' Rome. NY. Document generaad under centract to Acme Laborazary Onms AFB,

. NY.

3. MIL STD be Relisbelity Modeling ud Prediction.

4 l

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ea 'm I i

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i FIG.1 THE PS/EDD Fart No. 51-5875-2 FUEL VALVZ _

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February 18,1997 K 2 K ?? f *

< s l

.k FEB l 91997

Subject:

Informal Transmittal of Information on s

- Revision to WCAP-14407 Chapter 2 Tables U Brian A. Mc Intyre I To: Ed Throm Fax: 301-415-3577 (8 pages) ,

cc: Jim Gresham )

Brian McIntyre  !

Mike Loftus l l

This transmittal provides to NRC a draft of a sample revision of WCAP 14407, Section 2, Tables 2-3 and 2-4 which provide the link between the PIRT phenomena and the ,

containment pressure Evaluation Model. A revision to Tables 2-3 and 2 4 is an action l from a telecon between Westinghouse and NRC on January 17,1997. l The sample phenomena chosen cover a range of methods used to bound phenomena, and are Density of Break Source (1D), Mixing and Stratification in 1 Containment Volume (2A), Gas Compliance in Containment Volume (2C), and Evapo. ration on Steel Shell (7N). The following are being addressed .

- Consistency in terminology and numbering with December 19,1996 PIRT:  !

- Minor updates for consistency; i

- "Page format" rather than the more cumbersome " table format." l This table is being provided early to aid in the review process. It is anticipated that a draft of revised tables 2 3 and 2-3 could be available in time for the March 6,1997  ;

i

. meeting.

l I

O '

h ;> n tlSt J'oel Woodcoc l l

1 4

-a -

2-10 Table 2-3 FIRT Application to Evaluation Model: Inside Containment-LOCA-All Phases and MSLB P8 f.'I' Fe9 T "'".".s"e'le-C AF.00 SCs .e wkh Reeped t. M.demog Meeh.d Use .i V.ed.e6.e R seles la eM. Ume.sedney te gp( M.d.a. rhe e s.e.s the se M. des. Tem sese, mer e s.6 dee.d to senC pey.n C r": ww WCOTHtC ted s.e he. del Ibeled I W.susn. A Muis- H G emmenswease en the AM woes .astysed NTD NaC-95450 FNede .i - -; lsT enciendee .w .ad WOOTittC h.s twee Covereeng : _ en Dewaded W A)TittC .se . wahd

( C_

C. ,

g.eeesssig , wtsh _WLX77HBC E.rtsmuse a GFntBC o marm== ner.a e=ewde

, . g see .re r h-*sded hi see.sn 7.isdeerd wMh the LST _L

- of - "

G.s.se pe.eldes toege d.e.6.ee of gewesedng equ.ses.e s _ ,_ __;g behee4.e erses semh .tr. W Maessmusa Tert er.IspedAr esen

.ad behuse presesse amed en e.s$snse.n wesh Feirtasesee 2 G7ntIC 0% estessee hessakhay Tere.er.t he.n e deersehe, g.museg ,

E.rt. ewe 3 G7nec Usee's M.m I deervees h.ee e. isertee uses.us go.ees NTD NRC-954e62 eras me,-t mA me.

ooTwic oceeg. noview.

t=d nep.n WCAP 9082 v.ud.ere

_oOTnic W e.e. .,.r eferren. enseg=.I ares wth

.se.s. a s e.s.r e, u s r.s , n e tsr ie-e WcAr ion se,.s.= i .-red de g .ee.-d som w r.es. W w nnen 6-. see m. nee i.e m Iv A so ded

,.s.n.w. en se* s s e.e. wee ee eie. d _e.hde.ed w tsi

]O h ded pee.ameese p ins.,ed 6 WCAP-94382. 8.e ameegent .yee.eeng ded, e.eeeby s.

deserienn esseres seems g.oeveeng , erees seducteg he.t sean.e.t pasan wee, ,

vooed whe. rcs . - ~

IST m r e.r, d be.y.nry stelewig on=*. sed pw.=rwe sw m ades m edshe em .how. g d v.s e.,e.o.de egeoesnes. weeh 550 m.de LST _ 5.r t.DCA B me.dri M.delmg of bu.y.ary

.ad e wannw.e es meer=ht.

Sa b.we f.e hae IV A Soneded C risw rieed t. neenaag wahaa asi su swee.n 9. st =d.-a is *e ae a Urra .ad i=== _WGOTHIC no. del h.e ses e, .e aus. - 2.,,. .e d.ed , s .eg .ee. v.nd ed w rege.ns med mahdeg t.ang4ersa ll.)CA to dreven by eene. .; she tsf lA ser.sehree.m mesmees. she .,,.ee t.nese pore.s. .f she a , , e disi eseees e.cge Sw APeOS

.e,.ees. sed .eisT M$LS te e.es t adaed due to w h e*-=, we omere.wd pm awese

. , saw-e gaud

.gmmes. whh SSS n.de IST sn. del tendeteg of bas.y.nry d emee.inaw ao newpe.ba,

. = .- .

' # f -,

a g&\ h WM N}

Af tQlA C (_

\ m _A _ _ - ._.

Containment Phenomena Identification and Ranking Table SePeember 1996 m.\3006w 2.wpf.ItW90996 i

_ . - . . . . - _ _ _ _ _ _ _ _ _ . - _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _____________._m_ _ _ _ _ _ _ _ _ . . - _ _ _ _

_ r 3

l '4 4\ lib Cb YCrkRf' ,

i

Table 2-?? Summary Bases and Report Cross Reference for PIRT Phenomena l

! Phenomena - Density of Break Source (item ID in PIRT) l

i Ranking - High for all phases j t

AP600 BC's or Phenomena Models - Buoyancy forces are included in the lumped parameter 3

junction governing equations ,

j- i j Test Bases - LST internal buoyant flows. ,

l Report Submitted to NRC:  ;

1 l

o WCAP-14326, Separate Effects Tests i o WCAP-14382, LST validation results  ;

o' WCAP-14407 Section 9 Mixing within containment 5

Report

Conclusions:

if J o Lumped modeling overmixes noncondensibles above operating deck, thereby reducing

heat removal from vessel when PCS is dominant i o Distributed parameter modeling shows good agreement with 550 node LST model.

Modeling of buoyancy and entrainment is acceptable.

i i Applicability of LST with respect to Phenomena:

i 4

o Steam injection point elevation, direction, and momentum affects tests performed.

! o LST had prototypical buoyancy dnving forces and covered the range of Froude

! numbers for LOCA and MSLB.

s

- Validation of Modeling Method and/or WGOTHIC

1

! o WGOTHIC has been validated with the LST.

} o Effect of density on break source is evaluated in WCAP-14407, Section 9. -l 3 l

. Use of Validation Results in this Evaluation Model:  ;

i i

o Buoyant plume rising from the SG companment is shown to be a limiting scenario for j pressure (WCAP-14404, Section 9).

How Uncenainty is Handled: i o Bounded by selection of limiting scenario with respect to circulation and stratification effects relative to break density.

l I

I

4 N )

r l Table 2-?? l 3

Phenomena - Mixing / Stratification in Containment Volume (item 2A in PIRT)

Ranking - High for all phases j

i AP600 BC's or Phenomena Models - Mixing within the containment upper regions and 4 l

mixing between the upper and lower portions of the containment, as it is influenced by I circulation and stratification, is examined outside of the evaluation model (WCAP-14407, ,

Section 9) and methods are developed to bound the potential effects.

[

i Test Bases - Large Scale Tests

Report Submitted to NRC

i o WCAP-13566, "AP6001/8th Large Scale Passive Containment Cooling System Heat .

Transfer Baseline Data Report" j o WCAP-14135, " Final Data Report for PCS Large Scale Tests, Phase 2 and Phase 3" j o- WCAP-14382 shows code influence on mixing / stratification using measured and I nominal inputs 2

o WCAP-14407 Section 9, Mixing evaluation 1

, Report

Conclusions:

l o Blowdown is the same as standard plants.

1 o Post blowdown LOCA is driven by buoyant plume and LST covers range for AP600.

i o. MSLB is well mixed above the operating deck due to high velocity jet.

o Distributed parameter modeling shows good agreement with 550 node LST model.

j o Buoyancy and entrainment effects on condensation to internal sinks are bounded.

j o Effects of circulation on steam distribution were ranged to select a bounding scenario.

I j Applicability of LST with respect to Phenomena:  ;

i j

j o Upper and lower regions of containment represented in the LST,

- stratification data from above deck region is applicable

- lack of SG flow path in LST prevents its use for studying circulation effects i 1 l- Validation of Modeling Method and/or WGOTHIC

o WGOUilC model has been validated with the LST (WCAP-14382)

Use of Validation Results in this Evaluation Model:

. o Evaluation model bounds effects of mixing / stratification as discussed in Section 9.

O is' How Uncenainty is Handled:

o Bounded 4

i U i

, l l

l i

4

0 V

1 Table 2-??

Phenomena - Gas Compliance in Containment Volume (item 2C in PIRT)

Ranking - High for all phases AP600 BC's or Phenomena Models - Gas constituents in the governing equations.  ;

l-Test Bases - All tests analyzed with WGOTHIC l

l Report Submitted to NRC:

1 o NTD-NRC-95-4563 Enclosure 1 - GOTHIC Qualification Report provides large database of tests with air, hydrogen, and helium Enclosure 2 - GOTHIC Technical Manual describes governing equations  :.

Enclosure 3 - GOTHIC User's Manual describes how to invoke various gases , ,

o NTD-NRC-95-4462, EPRI Report RA-93-10, GOTHIC Design Review Final Report ,

o WCAP-14382 validates WGOTHIC with separate effects, integral tests with steam and air Report

Conclusions:

o Effects of multi-component compressible gases are correctly included in governing

. equations.

Applicability of LST with respect to Phenomena:

l o LST includes air and steam in an enclosed volume.

Validation of Modeling Method and/or WGOTHIC:

o WGOTHIC has been validated with the LST (WCAP 14382).

- 1 Use of Validation Results in this Evaluation Model: j o Governing equations in WGOTHIC are a valid representation'of compressible, multi-component gas behavior.

o Maximum Technical Specification pressure used in conjunction with low estimate of i ' containment volume.

1 l

i How Uncertainty is Handled: i

o Bounded i

l i i

o i l

y i Table 2-?'

Phenomena - Evaporation on Steel Shell (item 7N in PIRT)

Ranking - High for all phases AP600 BC's or Phenomena Models - Empirical correlation for the Sherwood number which is derived by dimensional analysis using the heat and mass transfer analogy and Colburn j factors. Application of a correction for mass transfer rate gives the AP600 forced convection l mass transfer correlation.

l Test Bases - Gilliland and Sherwood evaporation tests and Westinghouse STC Dat plate l evaporation tests.

Report Submitted to NRC:

o NTD-NRC-95-4397, " Supporting Information for the Use of Forced Convection in the l AP600 PCS Annulus" o WCAP-14326, Separates Effects Tests gives correlation (sections 2.0,2.1), entrance effect used for separate effect test (section 2.2), and correlation validation with tests (sections 3.6, 3.7 and 4.2) l Report

Conclusions:

)

o AP600 shown to operate in forced convection dominant regime.

o Correlation is biased 6.4% conservative with reasonable scatter over the range, o Once the outer shell heats up to at least 2F above ambient, the AP600 annulus i operates in forced convection.

Applicability of LST with respect to Phenomena:

o LST includes tests with and without fan on, covering the annulus from mixed convection through forced convection regimes, o WCAP-14382: Predictions of total evaporation (page 8-3)and wall heat Gux (page 8-6) validate models in an integral setting.

Validation of Modeling Method and/or WGOTHIC:

o WCAP-14382 summarizes WGOTHIC separate effects validation results (sections 3.2.1, and 4.4)

Use of Validation Results in this Evaluation Model:

o Forced convection correlation, modined for mixed convection effects to allow transient startup is appropriate for AP600.

o A conservative bias of 0.83 times the nominal correlation is used.

Q I

M~ l r

i How Uncertainty is Handled:

o Bounded 1 i

1 I

j l

I l

I i

i i

i l

I J

l 1

l i

i i

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l l

l

Westinghouse FAX COVER SHEET D

! l l 1 1

6 RECIPIENT INFORMATION SENDER INFORMATION OATE: h[kf~7 4

NAME: h))ja j{JMn j TO: LOCATION: ENERGY C' ENTER -

{ //7 67 YO/') EAST PHONE: FACSIMILE: PHONE: Omce:

l COMPANY: Facsimile: win: 284-4887 i l outside: (412)374-4887

LOCATION

Cover + Pages 1+b b57 m 067.

1 2 The following pages are being sent from the Westinghouse Energy Center, East Tower, i Monroeville, PA. If any problems occur during this transmission, please call:

(

j WIN: 284-5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

TOM ' ff b (Af11DN$

1 0fC 0Cd Y' 0A 7 C Av10}h f k, $ 0d5

/

k Nllfk Wh

/dbn

i

~ Changes Previously Approved by NRC )

to be Incorporated in SSAR Revision 11 l

l i

j l

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l l

l e

7. Instrumentation and Controls Operability, Availability, and Testing The diverse actuation system is designed to provide protection under all plant operating conditions in which the reactor vessel head is in place. The automatic actuation processors, I

in each of the two redundant automatic subsystems of the diverse actuation system, are I

provided with the capability for channel calibration and testing while the plant is operating.

I To prevent inadvertent DAS actuations during online calibration, testing activities or I, maintenance, the normal activation function is bypassed. Testing of the diverse actuation system is performed on a periodic basis.

Equipment Qualification and Quality Standards The diverse actuation system is capable of functioning during and after normal and abnormal events and conditions that include:

• Excessive temperature Ambient vibration Radio frequency and electromagnetic interference 3 nvjudig ochloWd MiCPs j g ne diverse actuation system equipmentAis designed and qualified in accordance with the industry standards listed in subsection 7.1.4.1.8. He adequacy of the hardware and software '

is demonstrate.1 through the verification and validation progediscussed in subsection 7.1.1 15. His program provides for commercial dedication of commercial off-the-shelf hardware and software. As the diverse actuation system performs many of the protection functions associated within the ATWS systems used in existing plants, the diverse actuation system is designed to meet the quality guidelines established by Generic Letter 85-06,

" Quality Assurance Guidelines for ATWS Equipment that is not Safety-Related."

7.7,1.12 Signal Selector The plant control system for the AP600 derives some of its control inputs from signals that are also used in the protection and safety monitoring system. He advantages of this design are:

• Re nonsafety-related plant systems are controlled from the same measurements which provide protection. This permits the control system to function in a manner which maintains margin between operating conditions and safety limits, and reduces the likelihood of spurious trips.

Reducing the number of redundant measurements for any single process variable reduces the overall plant complexity at critical pressure boundary penetrations. This leads to a reduction in separation requirements within the containment, as well as to a decrease in plant cost and maintenance requirements.

To obtain these advantages, measures are taken to provide the independence of the protection and control systems. The criteria for these measures are contained in the Standard IEEE Revision: 10 3 W85tingh00S8 7.7-19 December 20,1996

l

7. InstrumentDtion and Controls i

Opciration procedures prohibit testing two divisions at the same time. There are no built-in interlocks to prevent simultaneous testing of two integrated protection cabinets. However, the '

l use of bypasses by the tester provides that the protection and safety monitoring system cannot be placed in an unsafe condition if the procedure prohibiting simultaneous testing is violated.  ?

For example, testing two divisions results in two bypasses, which causes the voting logic to  ! 4 revert to a one.out-of-two coincidence for the remaining two unbypassed divisions.

Attempting to test three or four divisions at the same time causes a plant trip. The operational . ,

procedure restricting simultaneous testing of two or more divisions is for operability reasons {

i to avoid unnecessary trips. *

l

• in addition to periodic tests, the system performs error detection and data link testing as part ,

of its normal operation. Where practical, the on-line error detecting features are designed to t I automatically place the channel in which the error was detected into a trip or bypass state (either by direct bypass or reconfiguration). When a channel is automatically placed into a trip state, the operator has the option to subsequently place that channel in a bypass state. If the automatic configuration of the channel is not practical, the on.line error detecting feature causes alarm annunciation to the operator.

7.1.2.13 Safety.Related Display Instrumentation Safety-related display instrumentation provides the operator with information to determine the effect of automatic and manual actions taken following reactor trip due to a Condition 11,10.-

or IV event as defined in Chapter 15. His instmmentation also provides for operator display of the information necessary to meet Regulatory Guide 1.97. A description of the equipment used to provide this function is provided in subsectinn 7.1.2.6. A description of the data provided to the operator by this instrumentation is provided in Section 7.5.

7.1.2.14 Auxiliary Supporting Systems ne safety-related system equipment is supported by the supply of uninterruptable electrical energy. His electrical power is supplied by the Class IE de and UPS system discussed in Chaper 8.

e 7.1.2.15 Verdiestion and Validation Adequacy of the hardware and software is demonstrated for the protection and safety

[f

. monitoring system through a verification and validation (V&V) program. Details on the verification and validation program are provided in WCAP 13383 (Reference 4). De software development process which is documented in this document is consistent with the following standards:

ANSI /IEEE ANS-7-4.3.2 (1993); " Application Criteria for Programmable Digital Computer Systems in Safety Systems for Programmable Digital Computer Systems in Safety Systems of Nuclear Power Generating Stations" O

Revision: 5 February 29,1996 7.1 26 3 Westkigh00$8

2

== -

7. Irutrument: tim cid Cectrols IEC 880-1986; " Software for Computers in the Safety Systems for Nuclear Power Generating Stations" IEEE 8281983; "IEEE Standard for Software Configuration Mana'gement Plans"

=

EE 829-1983;"!EEE Standard for Software Test Documentation" l

.._EE 830-1984; "lEEE Standard for Software Requirements Specifications" IEEE 10121986f"IEEE Standard for Software Verificatio 1666 IC% -199h '1666 6 aide h S/+we CmEyou WNp d (A2D ,'

WCAP-13383 also provides for the use of commercial off the-shelf hardware and software through a commercial grade dedication process.

7.1.3 Plant Control System ne plant control system is a nonnfety-related system that provides control and coordination of the plant during stanup, ascent to power, power operation, and shutdown conditions. De plant control system integrates the automatic and manual control of the reactor, reactor coolant, and various reactor support processes for required normal and off-normal conditions.

The plant control system also provides control of the nonsafety-related decay heat remov,al systems during shutdown. De plant control system accomplishes these functions through use of the following:

i

• Rod control

• Pressurizer pressure and level control

• Steam generator water level control Steam dump (turbine bypass) control

• Rapid power reduction ne plant control system provides automatic regulation of reactor and other key system parameters in response to changes in operating limits (load changes). The plant control system acts to maximize margins to plant sat 'imits and maximize the plant transient performance ne plant control system als" ies the capability for manual control of plant systems and equipment. Redundant wimi logic is used in some applications to increase single-failure tolerance.

De plant control system includes the equipment from the process sensor input circuitry through to the modulating and nonmodulating control outputs as well as the digital signals to other plant systems. Modu!ating control devices include valve positioners, pump speed controllers, and the control rrs d equipment. Nonmodulating devices include motor starters for motor-operated valves and pumps, breakers for heaters, and solenoids for actuation of air-operated valves. De control cabmets contain the process sensor inputs and the modulating and nonmodulating outputs, he plant control system also includes equipment to monitor and control the control rods.

k Revision: 5 y Westingh00Sg 7.1 27 February 29,1996

= art

7. Instrumentatirn cod Cutrols

(

7.1.4.2.22 Conformance to the Requirements for Identification of Redundant Safety System Equipanent (Paragraph 4.22 of IEEE 2791971)

Distinctive markings are applied to redundant divisions of the protection and safety monitoring system. j l

The color coded nameplates described below provide identification of equipment. associated with protective functions and their divisions associations.

Division Color Coding i

Division A BROWN with WHITE lettering Division B GREEN with BLACK lettering Division C BLUE with WHm! lettering Division D YELLOW with BLACK lettering Non-cabinet mounted protective equipment and components have an identification tag or nameplate. Small electrical components such as relays, have nameplates on the enclosure that houses them. .

l 7.1.5 AP600 Protective Functions Protective functions are those necessary to achieve the system responses assumed in the safety analyses, and those needed to shut down the plant safely. The protective functions are (

l grouped into two classes, reactor trip and engineered safety features actuation.

Reactor trip is discussed in Section 7.2. Engineered safety features actuation is discussed in Section 7.3.

7.1.6 Combined License Information SCD' pescbve Grdim s Rh-LR,g M 4u M :=.=..; m d idd ..y,~c.e a cm saw

.yyaeww 'n C3Ccdtdok6 of bd 7 4 s c m-a df>nt w th

% meWodobg p ese4fd in Eafereno 8.

%b. eJ L.c- ~l' V '; A A** N n~

L Revision: 5 February 29,1996 7.1-48 T Westinghouse

i

7. Instrumentatios : d controts

~

7.1.7 References

1. IEEE 6031991, "lEEE Cnteria for Safety Systems for Nuclear Power Generator Stations."

2. IEEE 796-1983. "!EEE Microcomputer System Bus."

3. WCAP-13382 (P), WCAP-13391 (NP) " AP600 Instrumentation and Control Hardware l Description."

4. WCAP-13383 (P), WCAP-13392 (NP) "AP600 Instrumentation and Control Hardware and Software Design, Venfication, and Validation Process Report."

5. IEEE 2791971. "!EEE Criteria for Protection Systems for Nuclear Power Generating Stations."

6. IEEE 3841981, "lEEE Criteria for Independence or Class IE Equipment and Circuits."

7. WCAP-8897 (P), WCAP-8898 (NP), ' Bypass Logic for the Westinghouse Integrated Protection System."

6. WGP 1%05(P1WtAP-l%Ob(nPL" cashg%nr.,e 2eyu tedodao$y & Pokch Syses, M600. "

L Revision: 5 y Westinghoust 7.1-49 February 29,1996

l c i

1. Introduction and General Description of Plant I

1 Table 1.6-1 (Sheet 11 of 15)

MATERIAL REFERENCED l SSAR l Section Westinghouse Topical Number Report Number Title 4

I 6.2 WCAP 14382 WGOTHIC Code Description and Validation

I WCAP 8077 (P) Ice Condenser Containment Pressure Transient Analysis l l WCAP-8078 McAods l WCAP-8264-P-A (P) Wesdnghouse Mass and Energy Release Data for l WCAP 8312-A Containment Design l WCAP 10325 (P) Westinghouse LOCA Mass and Energy Release Model l for Containment Design March 1979 Version l WCAP 8822 (P) Mass and Energy Releases Following A Steam Line I WCAP 8860 Rupture l WCAP 7907 P-A (P) LOFTRAN Code Descripdon l WCAP-7907 A l WCAP-12945-P (P) Code Qualification Document for Best Estimate Analysis .

l WCAP 14407 (P) WGOTHIC Applicadon to AP600 l WCAP-14408 1 1 6.3 WCAP-8966 Evaluation of Mispositioned ECCS Valves l 7,1 WCAP-13382 (P) AP600 Instrumentation and Control Hardware l WCAP 13391 Descripdon l WCAP-13383 (P) AP600 Instrumentadon and Control Hardware and l WCAP 13392 Software Design, Verificauon, and Validadon l Process Report l WCAP 8897 (P) Bypass Logic for the Westinghouse Integrated Protection I WCAP-8898 System 5

l [7.2 WCAP-13594 (P) FMEA of Advanced Passive Plant Protection System l WCAP 13662 l 8.3 WCAP-13856 AP600 Implementation of the Regulatory Treatment of

, 1 Nonsafery-Related Systems Process l 10.2 WCAP-il525 Probabilistic Evaluadon of Reduction in Turbine Valve l Test Frequency WC AP- 14(o05 ( r ) W shn hg n x Se @ d E ION b W@E -l%Db (99) Pretc.Mcm QWs - APiOO l (P) Denotes Document is Propnetary Revision: 7 April 30,1996 1.6-12 T Westhghouse

7

1. Introduction and General Description of Plant Table 1.8 2 (Sheet 3 of 4)

SUMMARY

OF AP600 STANDARD PLANT COMBINED LICENSE INFORMATION ITEMS Item No. Subject subsection 6.4-2 local Toxic Gu Services and Monitoring 6.4.7 6.4-3 Procedures for Training for Control Room Hat /tability 64.7 I 6.6-1 Inspection Programs 6.69.1 662 Construcuon Activities 5.etp.e.4 w btWcw, 6cErcW. M E D#%

4

% g .I 6 ,6.9.2 a . (,,

" 8.2 1 Offsite ElectricrJ Power 8.2.4 i 8.31 Onsite Electrical Power 8.3.3 9.1 1 Fuel Storage and Handling 9.1.6 9.51 Offsite Communications Interfaces 9.5.2.5.1 l 9.5-2 Emergency Response Facility Communications 9.5.2.5.2 9.5-3 Security Communications 9.5.2.5.3 9.5-4 Cathodic and Environmental Protection for Fuel Oil Tanks 9.5.4.7 10.1 1 Erosion-Corrosion Monitoring 10.1.3

. 10.2 1 Turbine Maintenance and Inspection 10.2.6

{

10.4-1 Circulating Water Supply 10.4.12.1 10.4 2 Condensate Feedwater and Auxillary Steam System Chemistry Control 10.4.12.2 10.4-3 Potable Water 10.4.12.3 l 11.2 1 Liquid Radwaste Processing by Mobile Equipment i1.2.4.1 11.2 2 Cost Benefit Analysis of Population Doses from Liquid Effluents 11.2.4.2 a

11.2 3 Identification of Ion Exchange and Adsorbent Media for Liquid Radwane i1.2.4.3 l

l 11.2 4 Dilution and Control of Bonc Acid Discharge 11.2.4.4 l 11.3 1 Cost Benefit Analysis of Population Doses fmm Gaseous Effluents 11.3.4.1

!!.3 2 Identificanoc of Adsorbent Media for Gaseous Radwaste 11.3.4.2 1 11.4 1 Solid Wasse Management System Pmcess Concel Program i 1.4.6 l 1

11.5 1 Plant Offsite Dose Calculation Manual (ODCM) 11.5.7 i

12.1 1 ALARA and Operanonal Policies 12.1.3  !

12.2 1 Additional Contained Radiation Sources 12.2.3 12.3 1 Administrative Controls, Criteria and Methods for Radiological Fww 12.3.5

0.1 Revision

9 l August 9,1996 1.8 12 y Wggthghoggg 1

1 7. Instrumentation and Controts core cooling monitor. The incore instrutnent assemblies house both fixed incore detectors and O

I core exit thermocouples. The incore instrumentation system is described in subsection 4 A.6.1.

7.1.2 General Protection Subsystem Configuration o

The protection and safety monitoring system is illustrated in Figure 7.1-2. The functions of the protection and safety monitormg system have been decomposed into physically and electrically separate microprocessor based subsystems. Each subsystem is located on an independent computer bus to prevent propagation of failures and to enhance availability. In most cases, each subsystem is implemented in a separate card chassis. Subsystem independence is maintained through the use of the following:

Separate de power sources with output protection to prevent interaction between subsystems upon failure of a subsystem.

Separate input or output circuitry to maintain independence at the subsystem interfaces.

Deadman signals: A device, circuit, or function that forces a predefined operating condition upon the cessation of a normally dynamic input parameter to improve the reliability of hard-wired data that crosses the subsystem interface.

Optical coupling or resistor buffering between two subsystems or between a subsystern and an input / output (1/0) module.

4 - vXA9 -19080 WCAP 13382 (Reference 3) provides a description of the h ware elements which comprise

( i the protection and safety onitoring system configuration. e&CMct 8 70Vid25M discophen eF % sa ar6& care and epccthew.

7.1.2.1 Functional Components The type and number of boards used to implement the functions of a microproccoor based subsystem are purposely limited to aid serviceability and to restrict the number of spares. In addition, the basic function of a particular board remains fixed among subsysterr.s to facilitate the development and maintenance of the subsystem software. IEEE 796 (aeference 2) bus cards are typically used to provide functions as listed below.

Functional Processor The functional processor performs the major computations required to achieve the specific function of the microprocessor based subsystem. Tasks performed by the functional processor include movement of data between subsystem memories or I/O registers for the purpose of input or output, on-line compensation of the analog inputs, conversion of input data to engineering units, and diagnostic testing. A functional processor is included in each subsystem.

O Revision: 5 February 29,1996 7.1-6 T WesMgh00S8

7, Instrumentation and Controls __

7.1.7 References

1. IEEE 603-1991, "IEEE Criteria for Safety Systems for Nuclear Power Generator Stations."

2. IEEE 796-1983, "IEEE Microcomputer System Bus."

3. WCAP-13382 (P), WCAP-13391 (NP), "AP600 Instrumentation and Control Hardware Description."

l 4. WCAP-13383, Revision 1 (NP), "AP600 Instrumentation and Control Hardware and Software Design, Verification, and Validation Process Report."

5. IEEE 279-1971, "IEEE Criteria for Protection Systems for Nuclear Power Generating Stations."

6. IEEE 3841981, "IEEE Criteria for Independence or Class lE Equipment and Circuits."

7. WCAP-8897 (P), WCAP-8898 (NP), " Bypass Logic for the Westinghouse Integrated Protection System."

8. WCAP-190eo(P), WCAP-)%91(uP3, " APfoCB lf6VuffmMW G-d

( Gwd ScRwore Archekdcre ord Cxcab '

vscopW ", % m C lE) v Resision: 10 3 b i' $ iGUSS s 7.1-49 December 20,1996 8

J

1. Introduction and General Description of Plant Table 1.6-1 (Sheet 1I of 15)

MATERIAL REFERENCED SSAR Section Westinghouse Topical i Number Report Number Title 1 6.2 WCAP 14382 .W_ GOTHIC Code Description and Validation l WCAP-8077 (P) Ice Condenser Containment Pressure Transient Analysis I WCAP-8078 Methods I WCAP-8264-P-A (P) Westinghouse Mass and Energy Release Data for l WCAP-8312-A Containment Design l WCAP-10325 (P) Westinghouse LOCA Mass and Energy Release Model I for Containment Design March 1979 Version l WCAP 8822 (P) Mass and Energy Releases Following A Steam Line l WCAP-8860 Rupture l WCAP 7907 P-A (P) LOFTRAN Code Desenption l WCAP-7907-A I WCAP-12945-P (P) Code Qualification Document for Best Estimate Analysis -

l WCAP-14407 (P) .W_ GOTHIC Application to AP600 ,

I WCAP-14408 i l 6.3 WCAP 8966 Evaluation of Mispositioned ECCS Valves l 7.1 WCAP 13382 (P) AP600 Instrumentation and Control Hardware l WCAP-13391 Description l WCAP 13383 (P) AP600 Instrumentation and Control Hardware and l WCAP-13392 Software Design, Verification, and Validation l Process Report I WCAP-8897 (P) Bypass Logic for the Westinghouse Integrated Protection l WCAP 8898 System l 7.2 WCAP-13594 (P) FMEA of Advanced Passive Plant Protection System l WCAP 13662 l 8.3 WCAP 13856 AP600 Implementation of the Regulatory Treatment of I Nonsafety Related Systems Process l 10.2 WCAP il525 Probabilistic Evaluation of Reduction in Turbine Valve l Test Frequency d - N KC(P) AP(46 InWWdah and bhkcI bhd k N MCtI fw&McktW4 Cpsahh k'sf'h l (P) Denotes Document is Proprietary Revision: 7 April 30,1996 1.6-12 3 WOSthgh0058

l Changes Reflecting Resolution of NRC PAM/ ERG Comments 1

1 i

j J

7. tastrunnestation and Controis Table 7.5-5 Summary of Type B Variables Function variable Type / Category Moaltered Reactivity Control Neutron fhtx Bt Control rod position B3 Bonc acid concentration B3 Reactor Coolant System Integnty RCS pressure B1 RCS wide range That 81 RCS wule range Tcoid Bt Contamment water Icvel Bt Coomamsat pressure Bi Reactor Coolant Inventory Control Pressuruer level Bt Pressunser reference leg temperature Bt Pressuruer pressure B1 Reactor vessel - hot leg water level B3 Reactor Core Cooling Cars eut temperanus BI ,

RCS subcoohag 81 RCS wule range Tw B2 RCS wide range Tcond B2 RCS pressure B2 Reactor vessel - hot leg waser level B2 Heat Sink Mainisaance IRWST water level 81 PRHR flow Bt PRHR outlet tempersare B1 PCS sacrage tank waner level BI Passive contatamesa coohng waser flow B1 IRWST to RNS secoon valve stasms Bt Conwament Envuoament Costnament pressare B1 Remoealy operased contauuneet isolance valve B1 W Sb~sbu5 i

Revisies: 8 7.5 29 Juss 19,1996 T WeltkWhelas ,

7. !astrumentados and Controls Table 7.5 7 (Sheet 1 of 4)

Summary of Type D VariaWs System variable TpCatepn Reacuvity Control System Reactor tnp breaker stams D2 l Control rod posioon D3 l Pressunzer Level and Pressure Control Pmsurizer safery valve status D2 Pressunzer level D2 RCS pressure D2 Pressunac; pnasure D2 Reference leg temperamre D2 RCS Imops RCS wWe range Tw D2 RCS wide range Teog D2 RCP breaker sensus D2 ,

W- y Prestm and Level Contml Steam generanc; PORY stems D2 Steam gemensor PORY block valve stams D2 Steam generaser tafory valve stams D2 Main feedwaser'isolanon valve stams D2 Steam generosor level (wide range) D2 Steam gensrasar level (aanow range) D2 Steam generusar Wwdows isolanos valve D2 stams Rc P amrin3 unter b,perabm D2 Revision: 8 y@ 7.5 31 June 19,1996

7. Instmmentation and Controis Table 7 5-7 (Sheet 2 of a)

Surnmary of Type D Variables System Variable Type /Catepry Secondary Pressure and Level Contml Steam line pressuru D2 Main feedwater pump stuus D2 Main feedwatzr control valve status D2 Main steam line isolation valve status D2 Main steam line isolanon bypass valve status D2 Startup Feedwater Startup feedwater flow D2 Startup feedwater control valve status D2 Startup feedwater isolation valve stams D2 Main to startup feedwater crossover valve D2 stanas Safeguards Coeuunment pressure D2 Accumulator level D2

, Core makeup tank level D2

\

• IRWSTW}in e t.to/a $s(MOV) valve status on D3 in IRWST': f . Je e.bo

. on status

' valve isolatk o n D2 (22 EO - Sq,vi h ADS first stage, second nage and third stage D2 valve stams ADS fourth stage valve stams (MOV) D2 ADS fourth stage valve status (non-MOV) D2 PRHR heat exchanger inlet isolance valve D3 stanas PRHR heat exchangg"CDnb.rol -e -'"-- D2 valve status Ramesor vessel head vent valve stanas D2 t dis chorse isolaban CMTjg .r_ ' valve ssanas D2 i

i CMT inlet naam valve stanas D2 Accumulasor Isolabs m valve stanas D3 4

?$ .b Revision: 8 June 19,19M 7.5 32 y WestkWhouS8

ll

7. Instnumentados and Controls

]

)

, Table 7 5 7 (Sheet 4 of 4)

{ Summary of Type D Variables

, System Variabl* TWCamry Containment Cooling Containment temperature D2

, water .sc rie4 MOlabm j

, PCSktorage tan 1( 9 alve status D2 (MOV) 1 1

Mer-PCSjstorage tr

'aak,b :olablon

c_.c.. 2; valve status D2 (non-MOV)

Passive contmament cooling water flow D2 PCS storage tank water level D2 j HVAC System Stams MCR remrn air isolation damper stams D2 i MCR toilet nhane isolation damper stams D2

! MCR supply air isolation damper stams D2 MCR air delivery isolation valve stams D2 MCR air storage bottle pressure D2 MCR supply air radiation level D2 Main Steam Turbine stop valve stams D2 Turtnne control valve position D2 Coedenser senem dump valve stams D2 Revision: 8 June 19,1996 7.5-34 36

9. Instrumentadon and Controis Table 7.5 9 (Sheet 2 of 4)

Summary of Type F Variables Variable Type / Category Startup feedwater control valve status p3 Main feed **ater flow F3 Steam generator level (WR) p3 Steam flow F3 i

Main steam line isolanon valve status F3 j Main feedwater pump status F3 Stanup feedwater pump stams F3 ,

l Condenser steam dump valve status F3 l Condensate storage tank level 7/ F3 q )

Pressunzer spray 6 colo les to alve stanas F3 ,1 Aux.iliary spray line isolation valve stams F3 Makeup flow F3 Mdeup pump 5:atus p3 j Letdown flow F3 Cnev faliy cx>aler 90~p L reder skA>.s N \

CnJen ze e- backpeuo rc.

f=3 Ac.eumulahear ved volve s+alus F3 I

- i Revision: 8 3 WSElligh0MB0 7.5 37 June 19,1996

7. Instnimentation and Controts e 1 l

Table 7 5 9 (Sheet 3 of 4)

Summary of Type F Variables Variable Type / Category

, Bonc acid unk level p3 Bonc acid flow p3 ,

Makeupg_blencl t

x; .__ _x bde valve status p3 Makeup flow control valve status F3 RNS flow F3 IRWST to RNS suction valve status F3 RNS discharge to IRWST valve stams F3 CCS swge tank level F3 CCS flow F3 CCS pump status p3 CCS flow to RNS valve stams F3 CCS flow to RCPs valve status F3 i CCS heat exchanger intet temperature F3 CCS heat exchanger outlet temperamre F3 Diesel generator status F3 I i

8 Conuinment fan cooler stams F3 Clulled water pump stams p3 Clulled wasar valve staeus p3 Containment temperamme F3 Mais comeet room supply air isolatio. damper stams F3 uni. < e reen air inauso. damp., stan. F3 f Main consol room supply air redi- F3 sernce wa now F3 beesel g eneroker looet F3

-- %6 ,7 b diesel-fackect busa F3 Pow supply to c),ael-6ked hutes P3

- RNS pump s haloa F3 Revision: 8 June 19,1996 7.5 38 ,

T Westhghoust

7 Instrumentation and Controis Table ' 5-9 (Sheet 4 of 4) i i

Summary of Type F Variables I I

Variable Type / Category Service water pump status p3 Service water pump discharge valve status F3 Service water pump discharge temperature F3 Instrument air header pressure F3 i

Spent fuel pool pump flow F3 l i

Spent fuel pool temperature p3 Spent fuel pool water level F3 e

Main to startup feedwater crossover valve status F3 n ,,,_ m em o, . : - -- . . . g g. ; . 7 p3 I MWN g O bbt- pets,5 /.Soloben VO t/c. Sbbu.T

• O Revision: 8 7.5 39 June 19,1996 3 W8821ghouS4

1 l

l 1

Changes for Auxiliary Spray and CVS Letdown '

l I

I I

1 1

7. Instrumentation and Controls I

Condition I results from a coincidence of two of the four divisions of reactor loop average

! temperature (T,,,) below the Low 2 setpoint coincident with the P-.1 pernussive (reactor tnp).

{ This blocks the opening of the steam dump valves. This signal also becomes an input to the 1

steam dump interlock selector switch for unblocking the steam dump valves used for plant cooldown. "Diis function may be manually blocked when the pressunzer pressure is below

{ the P-11 setpoint. The block is automatically removed when the pressunzer pressure is above the P 1I setpoint.

l Condition 2 consists of two controls. Either one of these controls can be used to manually initiate a steam dump block.

The functional logic relating to the steam dump block is illustrated in Figure 7.2-1, sheet 10.

j 7.3.1.2.17 Control Room Isolados and Air Supply Initiados l l Signals to initiate isolation of the main control room and to initiate the air supply are i generated from either of the following conditions:

1 ,

l I l 1. High control room air supply radioactivity level  !

! 2. Loss of ac power sources -

! Condition 1 is the occunence one of two control room air supply radioactivity monitors j detecting a radioactivity level above the High-2 serpoint.

l Condition 2 results from the loss of all ac power sources. A preset time delay is provided to l pemut the restoration of ac power from the offsite sources or from tim onsite diesel generators

' before inicanon. De loss of all ac power is detected by undervoltage sensors that are connected to the input of each of the four Class IE banery chargers. Two sensors are

! connected to each of the four banery charger inputs. De loss of ac power signal is based on

! the detection of an undervoltage condinon by each of the two sensors connected to two of the

four banery chargers. De two out-of four logic is based on an undervoltage to the battery j chargers for divisions A or C coincidset with an undervoltage to the banery chargers for l divisions B or D.

l

The fhmenonal logic relating to consol room isolation and air supply initianon is illustrated l in Pigsse 7.2- sbest 13.

A u x W car s reg cm d.

i 7.3.1.2.18,\ tendevu Partilensise une Isolas6es owilieu3 Sproa cmc 1

)

! A signal to isolaan letdown purificance linajs s genermed upon the coincidence of

) .

I pressuruer level below Low-l setpost in any two of four (visions. His helps to

! maintain reactor coolant synes inventory. Dis funcoon can be manually blocked when the

! pressuruer water level is below the P-12 seapotat. His funcoon is automancally unblocked j when the pressuruar water level is above the F 12 seapoet. The funcoonal logic relating to j this is illustrated in Figure 7.21, shast 12.

i

. Revision: 10 i December 20,1996 7.3 16 TM _

i

7. Instrumentation and Controis i

7.3.1.2.19 Containment Air Filtration System Isolation

! A signal to isolate the containment air filtration system is generated upon the coincidence of containment radioactivity above the High-1 setpoint in any two of four divisions. His limits '

activity release to the environment. The functional logic relating to this is illustrated in

, Figure 7.21, sheet 13.

7.3.1.2.20 Normal Residual Heat Removal System Isolation A signal for isolating the normal residual heat removal system lines is generated upon the

~

coincidence of containment radioactivity above the High-2 setpoint in any two of four l divisions. This signal also isolates the chemical and volume control system as discussed in I subsection 7.3.1.2.15. His limits activity release to the environment. He functional logic I relating to this is illustrated in Figure,7.2-1, sheet 13.

7.3.1.2.21 Spent Fuel Pool isolados A signal for isolating the spent fuel pool lines is generated upon the coincidence of spent fuel pool level below the Low setpoint in any one of two divisions. This helps to maintain the water inventory in the spent fuel pool due to line leakage. The functional logic relating to this.,

is illustrated in Figure 7.2-1, sheet 13.

cabo rusEKT 7 3. I. 2 21 7.3.1.3 Blocka, Permissives, and Interlocks for Engineered Safety Features Actuados ne interlocks used for engineered safety features actuation are designated as "P xx" permissives and are listed in Table 7.3-2.

7.3.1.4 Bypasses of Engineered Safety Features Actuations The channels used in engineered safety features actuation that can be manually bypassed are indicated in Table 7.3-1. A description of this bypass capability is provided in subsection 7.1.2.10. He actuation logic is not bypassed for test. During tests, the actuation logic is fully tested by blocking the actuation logic output before it results in component actuations.

7.3.1.5 Design Basis for Englaeered Safety Features Actuados The following subsections provide the design bases information for engineered safety features actuation, including the information required by Section 3 of IEEE 279-1971. Engineered safety features are initiated by the protection and safety monitoring system. Those design bases relating to the equipment that initiates and accomplishes engineered safety features are given in subsection 7.1.4.1. The design bases presented here concern the variables monitored for engineered safety features actuation and the minimum performance requirements in generating the actuation signals.

Revision: 7

{ Westingh00sg 7.3-17 April 30,1996

i p

i i

s

[ INSERT 7.3.1.1.22) 7.3.I.2.22 Chemical and Volume Control System Letdown Isolation  !

i A signal to isolate the letdown valves of the chemical and volume control system is generated upon .

the coincidence of low loop 1 and loop 2 hot leg levels. This helps to maintain reactor system ,

inventory. The functional logic relating to this is illustrated in Figure 7.2-1 sheet 16. These letdown .l valves are also closed by the containment isolation function as described in subsection 7.3.1.2.1. i I

\

i l

l i

l l

7 Instrutnentation and Controis Table ? 31 iSheet 7 of 8)

ENGINEERED SAFETY FEATURES ACTUATION SIGNALS No. of Channels / Actuation Actuation Signal Switches Logic Permissives and Interlocks

16. Main Control Room Isolation and Air Supply Initiation (Figure 7.21. Sheet 13)

a. High-2 control room 2 1/2 None supply air radiauon

b. Undervoltage to Class IE 2/ charger 2/2 per charger None battery chargers and 2/4 chargers8 A u d lrez e 3 $ preg orld 17.] Purification Line Isolation (Figure 7.2-1. Sheet 12)

I a. Low-l pressurizer level 4 2/4-BYP" Manual block permitted below P.12.

Automaucally unblocked above P 12.

18. CLatainment Air Filtration System Isolation (Figure 7.2-1. Sheet 13)

a. High-1 containment 4 2/4-BYP' None radioacuvity

19. Normal Residual Heat Removal System Isolation (Figure 7.21. Sheet 13)

a. High 2 containment 4 2/4-BYP' Nct:s radioacuvity

20. Spent Fool Fool faaladaa (Figure 7.21. Sheet 13)

a. Low spent fuel poollevel 2 1/2 None

21. Opes In Containament Refoe!!ag Water Storage Tank (IRWST) Injectice 1.ine Valves (Figure 7.21. Sheet 16)

a. Automanc reactor coolant (See items 3d and 3e) system depressurtzanos (fourth stage)

b. Coinculset loop I and I per loop 2/2 None loop 2 low hot leg level (after deley)

c. Manualiamance 4 switches 2/4 switches) None

22. Open IRWST Cameht Radrement6ss Valves la Seeiss wtth Check Valves (Figure 7 21. Sheet 15) -

a. Extended undervoltage to 2/chargw 1/2 per chargw None Class IE banery chargers and 2/4 chargers Revision: 10 December 20,1996 g 7.3 29

7. Instrumentation and Controls

~

Table 7.3-1 (Sheet 8 of 8)

ENGLNEERED SAFETY FEATURES ACTUATION SIGNALS No. of Channels / Actuatloa Actuation Signal Switches Logic Permissives and Interlocks

23. Open All IRWST Contal= ment Recirculation Valves (Figure 7.21. Sheet 16)

b. Safeguards actuation signal (See items la through le)

(automatic or manual) coincident with Low IRWST tevel 4 2/4 BYP' None (Low 3 serpoint)

c. Manual initiation 4 switches 2/4 switches ' None Notes:

1. 2/4-BYP indicates automatic bypass logic. The logic is 2 out of 4 with no bypasses; 2 out of 3 with one bypass; I out of 2 with two bypasses; and, automatically actuated with three or four bypasses

2. Any two channels from either tank not in same division. ..

3. Two switches must be actuated simultaneously.

4. Also, closes power-operated relief block valve of respective steam generator.

! 5. The two-out-of-four logic is based on undervoltage to the banery chargers for divisions A or C coincident with I an undervoltage to the battery chargers for divisions B or D.

6. Any two channels from either loop not in same division.

7. Any two channels from either line not in same division.

M* O W 'tol o d % lu ~e. [o ,[,,l f ,/ p f,Q)_ Ze,j,Q (Sp,y.] yy e g [j '93 2. /Vo n c_

l e ve }

Revision: 7 April 30,1996 7.3-30 3 W95tingh0084

7. Instnamentation and Controls l

Table 7.3-2 (Sheet 3 of 3)

LNTERLOCKS FOR ENGINEERED SAFETY FEATURES ACTUATION SYSTEM Designation Derivatloa Function P-12 Pressurizer level below setpoint (a) Permits manual block of core makeup tank actuation on low pressurizer level to allow mid-loop operation (b) Permits manual block of reactor coolant pump trip on low pressuriar level to allow i mid loop operation , f 40xsItttry Sprou '

arb (c) Permits manual block ohurification line l isolation on low pressuriar level to allow mid-loop operation P72 Pressuriur level above setpoint (a) Prevents manual block of core makeup tank actuation on low pressurizer level ,

(b) Prevents manual block of reactor coolant pump trip on low pressuriur level auxiftar SJ,my cwd (c) Prevents manual block of/ purification me  ;

isolation on low pressurizer level (d) Provides confirmatory open signal to the core makeup tank cold leg balance lines Revision: 7 y@ 7.3 33 April 30,1996

i '

EIE II E I .

Zli L P .Sf,d -

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( Revision 5 3 WSEthgh0088 February 29,1996 7.1-59

6 i

7. Instrumentation and Controls Eii jj 9 , A DL A(i a

Condition 3 results from a coincidence of two ef the four divisions of narrow range steam generator water level above the High-2 setooint for either steam generator.

De functional logic relating to the tripping of the turbine is illustrated in Figure 7.21,

sheet 14.

I 7.3.1.2.9 In Containment Refueling Water Storage Tank Containment Recirculation 4

. Signals to align the in-containment refueling water storage tank containment recirculation I isolation valves are generated from the following conditions:

i

1. Automatic or manual safeguards actuation (subsection 7.3.1.1) in coincidence with low 2

in-containment refueling water storage tank water level i 2. Manual initiation I 3. Extended loss of ac power sources here are four parallel containment recirculation paths provided to permit the recirculation of, j the water provided by the in-containment refueling water storage tank. Two of these paths

are provided with two isolation valves in series while the remairting two paths are provided l with a single isolation valve in series with a check valve.

i j Conditions I and 2 result in the opening of all isolation valves in all four parallel paths.

Condition 3 results in the opening of the two isolation valves that are in series with the check j valves.

i Condition I results from the coincidence of two of the four divisions of in containment i refueling water storage tank water level below the Low-3 setpoint, coincident with an j automatic or manual safeguards actuation.

j

! Condition 2 consists of two sets of two momentary controls. Manual actuation of both I controls of either of the two control sets initiates recirculation in all four parallel paths. A

. two control simultaneous actuation prevents inadvertent actuanon.

i Condition 3 results from the loss of all ac power for a period of time that approaches the

24-hour Class IE de battery capability to activate the in containment refueling water storage i tank containment recirculation isolation valves. De timed output holds on restoration of ac

power and is manually reset after the batteries are recharged. De loss of all ac power is i i detected by undervoltage sensors that are connected to the 'mput of each of the four Class IE

( l banery chargers. Two sensors are connected to each of the four banery charger inputs. De l loss of ac power signalis based on the detection of an undervoltage condition by either of the I two sensors connected to rwo of the four banery chargers.

i C / A.) s E A T Z3. /. 2,9.2

No interiocks or permissive signals apply directly to the activanon of the in-containment i refueling water storage tank containment recirculation isolation valves. However, automatic

(

Revision: 7

. _ b mm. Staful CafL30J996

(INSERT 7.3.1.2.9]

The safeguards actuation signal, which is part of condition 1 is latched-in upon its occurrence. A deliberate operator action is required to reset this latch. This feature is provided so that the actuation signal to the recirculation isolation valves is not cleared by the reset of the safeguards actuation signal as discussed in subsection 7.3.1.1.

9 6

i' i

4

7. Imtrumentation and Controls Condition I results from a coipctdence of two of the four divisions of reactor loop average temperature (T,y) below the Low 2 setpoint cotncident with the P-4 permissive (reactor tnp).

This blocks the opening of the steam dump valves. His signal also becomes an input to the steam dump interlock selector switch for unblocking the steam dump valves used for plant cooldown. This function may be manually blocked when the pressunzer pressure is below the P 11 serpoint. The block is automaucal}Y removed when the pressunzer pressure is above the P-ll serpoint.

Condition 2 consists of two controls. Either one of these controls can be used to manually '

initiate a steam dump block.

The functional logic relating to the steam dump block is illustrated in Figure 7.2-1, sheet 10.

7.3.1.2.17 Control Room Isolation and Air Supply Indtiatica Signals to initiate isolation of the main control room and to initiate the air supply are generated from either of the foDowing conditions:

1. High control room air supply radioactivity level

2. Loss of ac power sources ,

3. Movw ed s m tids ovs Condition 1 is the occunence one of two control room air supply radioactivity moniten detecting a radioactivity level above the High-2 serpoint.

Cond! don 2 results from the loss of all ac power sources. A pttset time delay is provided to permit the restoration of ac power from the offsite sources or P ' the onsite diesel generators before initiation. The loss of all ac power is detected b; Arvoltage sensors that are connected to the input of each of the four Class IE banery chargers. Two sensors are connected to each of the four banery charger inputs. The loss of ac power signal is based on the detection of an undervoltage condroon by e.ach of the two sensors connected to two of the four banery chargers. The twt >out-of-four logic is based on an undervoltage to the banery chargers for divisions A or C coincident with an undervoltage to the banery chargers for divisions B or D.

CI A> S E RT 7 3. l. 2. r7 7 The funcnonal logic relating to control room isolation and air supply initianon is illustrated in Figure 7.21, sheet 13.

7.3.1.2.18 Latdows PartScation Line Isolat6em A signal to isolass the letdows punficance line is generssed upon the coincidence of l I pressurtzer level below the Low-1 seepoint in any two of four divisions. This helps to mamtain reactor coolant syssess inventory. This funcuon can be manually blocked when the pressunzer water level is below the P-12 seapotat. This funcnon is automarnity unblocked when the pressurtzer waser level is above the P 12 W_=_1he funcnonal logic reta:ing to this is illustrated in Figum 7.21, sheet 12.

(

Revision:le Decessber 20,1996 7.3-16 T Wesitighouse

, [ INSERT 7.3.1.2.17]

Condition 3 consists of two momentary controls. Manual actuation of either of the two controls will result in control room isolation and air supply initiation.

9 e

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7. Instrumentation and Controls j

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j Table 7 3-1 (Sheet 7 of 8) i i

ENGINEERED SAFETY FEATURES ACTUATION SIGNALS l 5 No. of I j Channetal Actuation j Actuation Signal Switches Logic Permissives and Interlocks l

) . 16. Main Control Room Isolation and Air Supply laitiation (Figure 7.21, Sheet 13)

a. High 2 contros room 2 1/2 None Supply aar radiauon i b. Undervoltage to Class 1E 2/ charger 2/2 per charger 1 None 4 battery chargers and 2/4 C. M ar imhabios 2$wftNes

17. Purtfication Llae Isolation (Figure 7.21. Sheet 12) t 3 }e(gg Nog I a. Low l pressurizer level 4 2/4-BYP". Manual block permitted below P 12.

Automancally unblocked above P 12. l l

j 18. Comenta===t Air Filtretloe System Isolation (Figure 7.21. Sheet 13) l a. High-l containment 4 2/4-BYP' None  !

j radioacovity .

19. Norsial Reekleal Heat Removal System Isolation (Figure 7.21. Sheet 13)

a. High-2 containment 4 2/4-BYP' None l radioacuvity a

i 20. Spent Feet Pool isolation (Figure 7.21, Sheet 13)

a. Low spent fuel pool level 2 1/2 None l 21. Opes la Costainment Refueling Water Storage Tank (IRWST) tajection 1.ine Valves l (Figure 7.21. Sheet 16) i a. Automauc reactor coolant (See items 3d and 3e) i system depresswizanos (fourdi stage)

b. Coincident loop I and I per loop 2/2 None loop 2 low het leg level (after 4:4)

c. Manealinnianos 4 switW 2/4 switches8 None

( 22. Open IRWFT Contaisseest Recircolaslen Valves la Seetes wish Check Valves (Figwe 7.21. Sheet 15)

a. Estended undervoltage to 2/ charger, 1/2 per charger Noes Class IE banery chsym and 2/4 chargers i

Revision: 10 o.c.imber2o,1996 W wasmpouse 7.3-29

7. Instrumentation and Controls Table 7.3-3 SYSTEM LEVEL MANUAL INPUT TO THE ENGINEERED SAFETY FEATURES ACTUATION SYSTEM To F1gure 7.21 l Manual Control Divisloes Sheet Manual safeguards actuation #1 ABCD 2 & 11 Manual safeguards actuation #2 ABCD 2 & 11 Manual passive residual heat removal actuadon #1 AB 8 Manual passive residual heat removal actuation #2 AB 8 Manual steam line isolation #1 B D 9

Manual steam line isolation #2 B D 9

Steam /feedwater isoladon and safeguards block control #1 B 9 Steam /feedwater isolation and safeguards block control #2 D 9 Manual feedwater isola: ion #1 B D 10 Manual feedwater isoladon #2 B D 10 Manual steam dump interlock selector #1 B 10 Manual steam dump interlock selector #2 D 10 Pressurizer pressure safeguards block control #1 A 11

Pressurizer pressure safeguards block control #2 B 11 .

l Pressurizer pressure safeguards block control #3 C 11 i Pressurizer pressure safeguards block control N D 11

Manual core makeup tank actuation #1 ABCD 12

Manual core makeup tank actuadon #2 ABCD 12 Core makeup tank .actuadon block control #1 A 12

Core makeup tank actuauon block control #2 B 12

Core makeup tank actuation block control #3 C 12 Core makeup tank actuation block control M D 12 Manual containment cooling actuation #1 & #2 AB 13

] Manual containment cooling actuation #3 & N AB 13

Manual containtnent isolation actuanon #1 ABCD 13 ABCD Manual containment isolation actuation #2 13 Manual depressurization system stages I,2. and 3 actuanon #1 & #2 ABCD 15 Manual depressurizanon system stages 1,2. and 3 actuanon #3 & N ABCD 15 Manual depressuruanos sysaem stage 4 actuation #1 & #2 ABCD 15 Manual depressuruance system stage 4 actuanoe #3 & N ABCD 15 Manual IRWST acnianos #1 & F2 ABCD 16 Manual IRWST acennos #3 & N ABCD 16 Manual containment recuculanon actuanoe #1 & #2 ABCD 16 Manual coneman==e reeveulation actuanoe #3 & N ABCD 16 q Manoel condrol room J5elmhos ord oien,ff J y,e,h.6,o#/ A8C,h j3

( Marao l cedrol rose 15elden occi QIir Supply in, Me #2 A8O )3

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Revision: 7 April 30,1996 7.3-34 3 WOEthighouS8

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One remote shutdown workstation is provided. ,Ihe remote shutdown workstation contams

controls for the safety-related equipr tent required to establish and maintam safe shutdown.

Additionally, control of nonsafety g ged components is available, allowing operation and control when ac power is available.7 w sumo. i.;dec.r. -ci=;in

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The remote shutdown workstation is provided for use only following an evacuation of the

• 1. 1. p i 8 main control room. No actions are anticipated from the remote shutdown workstation dunng j il(f f /f'fh normal, routine shutdown, refueling, or maintenance operations.

I

l Q t 18'l$o W The remote shutdown workstation has sufficient communication circuits to allow the operator ll

• to effectively establish safe shutdown conditions. As detailed in subsection 9.5.2, m p j.cg g /g communication is available between the following stations:

j towf $ht - Main control room j (Mo ft 5 Mew - Remote shutdown workstation j g,g.g , = Onsite technical support center

'a Diesel generator local control station i Sussec.he w 18,17.3

) g'jO /MU Operator control capability at 6: mmote shutdown workstation is nonnally disabled, and operator control functions are normally performed from workstations located inside the main

, ),gg,y g control room; however, operator cos:rol capability can be transferred from the main control room workstations to the remote workstation if the control room requires evacuation. This j

! % (%te operator control transfer capability can not be disabled by any single active failure coincident g with the loss of offsite power.

! ayW/o_ bog The control transfer function is implemented by multiple transfer switches. Each individual

~

transfer switch is associated with only a single safety-related or single nonsafety-related  !

j1 dis j A6Mfj division. These switches are located behind an unlocked access panel. Entry into this access panel will result in alarms at the main control room and remote shutdown workstation. The l

! d d CeOMi, access panel is located within a fire zone which is separate from the main control room.  !

)

! Actuation of these transfer switches results in additional alarms at the main control room and i remote shutdown workstation, the activation of operator control capability from the remote workstation and thp deactivation of operator control gility from the main control room workstations. The Z ,

' .- f - - - '

_ operator displays located in the main control room and on the remote shutdown workstation are not affected by this control transfer function.

7.4.3.1.2 Controls at Other Locations In addition to the controls and indicators provided at the remote shutdown workstation, the

- following controls are provided outside the main control room:

• Reactor trip capability at the reactor trip switchgear

• Turbine trip capability at the turbine Revision: 5

- N ME W W astt1Ef10u88

7. Instrumentation and Controls

• Stan/stop controls for the diesel generators. located at each diesel generator local control panel 7.4.3.1.3 Design Bases Information According to GDC 19. the capability of establishing a shutdown condition and mamtammg the station in a safe status in that mode is an essential function. He controls and indications necessary for this function are identified in subsection 7.4.2. To provide the availability of the remote shutdown workstation after control room evacuation, the following de ,a features are provided:

• The remote shutdown workstation conforms with the guidelines provided by ANSI 58.6 1983 (Reference 1).

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• Re remote shutdown workstation achieves and maintains safe shutdown conditions from full power conditions and maintains safe shutdown conditions thereafter.

• He remote shutdown workstation achieves safe shutdown when offsite power is available and when offsite power is not available.

• Re remote shutdown workstation operates safety-related systems, independent from the I main control room.

• The remote shutdown workstation is designed for a single failure. When a random event, such as a fire, or an allowable technical specification maintenance results in one safety-related division being unavailable, a single failure in a redundant division is not postulated. When a random event oJur than Qt causes a main control room evacuation, I a coincident single failure in the systems controlled from the remote shutdown panel is considered.

• Access to the remote shutdown workstation is under admmistrative control.

I 7.4.3.2 Analysis The analysis of the systems required for safe shutdown is provided in subsection 7.4.1. The l

following discussion is limited to the remote shutdown workstation. l Conformance to NRC General Design Criteria i General Derita Criteries # - The remote shutdown workstation provides adequate controls and indications located outside the main control roorn to establish and maintain the reactor Revision: 5 yg 7.4 13 Febniary 29,1996

f l 1

4

7. Insy6mentauohnd Controls i .

and the reactor coolant system in a safe shutdown condition in the event that the main control room must be evacuated.

Conformance to NRC Regulatory Guides Regulatory Guide 1.22 - The remote shutdown workstation is tested periodically dunng l station operation.

i l Regulatory Guide 1.29 ihe remote shmdown workstation is designed 1..- J J.. N

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Conformance to IEEE 2791971 bck f 6/Mf O f The remote shutdown workstation and the cofg design %gy features j control capability from the main control room to the remote shutdop'w'orkstation "*d-"-conforms to applicable portions of IEEE 279-1971. The seneroPcircuits- x - .;r weriesenesen are designed so that a single fail does not prevent maintaining safe shutdown. i This is accomplished by redundant W the systems required for safe shutdown, trsing l independent safety-related power divisions.

I G*Mf*Megts & 1 l

To prevent interaction between the redundant systems, the redundant control channels are wired independently and are separated from ech other. Nonsafety related circuits available i for (but not required for) safe shutdown are electrically isolated from safety-related circuits. l 7.4.4 Combined License taformation This section has no requirement for information to be provided in support of the Combined License application.  ;

7.4.5 References

1. ANSI 58.61983, " Criteria for Remote Shutdown for Light Water Reactors."

Revision: 5 February 29,1996 7.4 14 T Westkighouse

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7. Instrumentation and Controls De protection and safety monitoring system provides signal conditioning, communications, and display functions for Category I variables and for Category 2 variables that are energized from the Class IE de uninterruptible power supply system. ne plant control system and the data display and processing system provides signal conditioning, communications and display

, functions for Category 3 variables and for Category 2 variables that are energized from the non-Class IE de uninterruptible power system. The data display and processing system also provides an altemate display of the variables which are displayed by the protection and safety monitoring system. Electrical separation of the data display and processing system and the protection and safety monitoring system is maintained through the use of isolation devices in the data links connecting the two systems, as discussed in subsection 7.1.2.11. The portion of the protection and safety monitoring system which is dedicated to providing the safety-related display function is referred to as the qualified data processing cabinets. Rese cabinets are discussed in subsection 7.1.2.6 and are illustrated in Figure 7.1-8.

The qualified data processing cabinets are dividLi into two separate electrical divisions. Each of the two electrical divisions is connected to a Class lE de uninterruptible power system with 1

sufficient battery capacity to provide necessary electrical power for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If all ac power sources are lost for a period of time that exceeds 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the power supply system will be energized from ac power sources which are brought to the site fmm other locations.

See Section 8.3. 7_3 g % g h/ g,g.

/ n c /.%

i Instrumentation associated with primary variables that are energized from the Class IE de uninterruptible power supply system are powered from one of the two electrical divisions with

72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> battery capacity. Instrumentation associated with other variables that are energized 2

from the Class IE de uninterruptible power supply system are powered from one of four electrical divisions with 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery capacity. If a variable exists only to provide a backup to a primary variable, it may be powered by an electrical division with a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery capacity. In such cases, provisions are provided to enable this variable to be powered by an alternate source if it is needed to resolve a discrepancy between two primary variables in the event that all ac power sources are lost for a period in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1 Class IE position indication signals for valves and electrical breakers may be powered by an i I electrical division with 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery capacity. His is necessary to make full use of all four l Class IE electrical divisions to enhance fire separation criteria. De power associated with J l the actuation signal for each of these valves or electrical breakers is provided by an electrical I division with 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery capacity, so there is no need to provide position indication I beyond this period. The operator will verify that the valves or electncal breakers have I achieved the proper position for long-term stable plant operation before position indication is I lost. Once the position indication is lost, there is no need for further monitoring since the I operator does not have any remote capability for changing the position of these components.

Electrically operated valves, which have the electrical power removed to meet the single failure criterion, are provided with rMnMaat valve position sensors. Each of the two position sensors is powered from a different non-Class IE power source.

Revision: 8 June 19,1996 7.5-12 3 W0gthgh0088

4 a

7. Instrumentation and Controis

]

I l 7.3.1.2.19 Containment Air Filtration System Isolation A signal to isolate the containment air filtration system is generated upon the coincidence of l containment radioactivity above the High 1 setpoint in any two of four divisions. His limits I activity release to the environment. The functional logic relating to this is illustrated in Figure 7.2-1, sheet 13.

l 7.3.1.2.20 Normal Resideal Heat Removal System Isolation

A signal for isolating the normal residual heat removal system lines is generated upon the

, coincidence of containment radioactivity above the High-2 setpoint in any two of fout

I divisions. This signal also isolates the chemical and volume control system as discussed in

. I subsection 7.3.1.2.15. This limits activity release to the environment. He functional logic

. relating to this is illustrated in Figure 7.2-1, sheet 13.

7.3.1.2.21 Spent Fuel Pool Isolation dwo d.k r e.e.  ;

, A signal for isolating the spent fuel Ilines is g rated upon the coincidence of spent fuel I pool level below the 1.ow setpoint iri aapene of divisions. His helps to maintain the water inventory in the spent fuel pool due to line lerAage. The functional logic relating to this' l

is illustrated in Figure 7.2-1, sheet 13.

1 7.3.1J Blocka, Permissives, and Interlocks for Engineered Safety Features Actuatios '

The interlocks used for engineered safety features actuation are designated as "P-xx" permissives and are listed in Table 7.3-2.

7.3.1.4 Bypasses of Engineered Safety Features Actuations j De channels used in engineered safety feamres actuation that can be manually bypassed are i

indicated in Table 73-1. A description of this bypass capability is provided in j subsection 7.1.2.10. De actuation logic is not bypassed for test. Durmg tests, the actuation

logic is fully testod by blocking the actuation logic output before it results in component actuations.

7.3.1.5 Design Basis for Engineered Safety Features Actuation The following subsections provide the design bases information for engineered safety features actuation, including the information required by Section 3 of IEEE 279-1971. Engineered safety features are initiated by the protection and safety monitoring system. Dose design bases relating to the equipment that initiates and accomplishes engmeered safety features are given in subsection 7.1.4.1. The design bases presented here concern the variables monitored for engineered safety features actuation and the minimum performance requirements in generating the actuation signals.

Revision: 7 E _Westhgh00$g 7.3-17 April 30,1996

4.

4 S i

i 4

7. Instrumentation and Controis -

l 1

j Table 7.31 (Sheet 7 of 8) f ENGINEERED SAFETY FEATURES ACTUATION SIGNALS No. of I Chaamels/ Actuation j Actuation Signal Switches Logic Pe 1 missives and Interlocks

16. Main Control Roosa Isolation and Air Supply Initiation (Figure 7.21. Sheet 13) l i a. High 2 control room 2 1/2 None ,

j supply air radiation j l b. Undervoltage to Class IE 2/ charger 2/2 per charger None

battery chargers and 2/4 chargers8 f 17. Purification Line Isolation (Figure 7.21 Sheet 12)

I a. Low-l pressurizer level 4 2/4 BYP' Manual block permitted below P-12. l Automatically unblocked above P 12.

l 13. Costalasseet Air FBtration Sysseen Isolation (Figwe 7.21. Sheet 13)

a. High l containment 4 2/4-BYP' None -

l radioactivity 1 l

2

19. Norunal Residual Heat Roanovel System Isolastes (Figwe 7.21. Sheet 13)

a. High 2 containment 4 2/4-BYP' None

! radioactivity 1

20. Spent Feel Post Isolastem (Figwe 7.21. Shast 13) l a. l.ow spent fuel pool level # 44- 2 None i 3

21. Opes la-Costaloonset Reheellag Water Storage Task (IRWST) Indecties Llos Valves (Figwe 7.21. Sheet 16)

a. Automaec reactor coolant (See items 3d and 3e) sysism ' -h (fourth sage)

b. Coincuisat loop 1 and i per loop 2/2 None loop 2 low het les level (afier deisy)

c. Manualimmianos 4 switches 2/4 switches8 None

22. Open IRwsT casamimment Reserceimansa vet,es la seress wie check valves (Figwe 7 21. Sheet 15)

a. Extended undervoltage to 2/ charger 1/2 per charger None Class IE banery chargers and 2/4 Chargel Revision: 10 Decesatser 20,1996 W itineinshnuma _9fds)

i

] l 1

i a l

'. 1 Mi ,

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" =T" j kCVl %C. fOSAoLO 5 3 donne.ls anel 0 7 TT 1

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me o== = e== =

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==rr.r r ".

a ggum.m.= e . is amen seem es as

'h h 1 TF O ~

""gye.7 5 "g% ,e"# gm Figms 7.21 (Sheet 13 of 20)

F e d Diagramms Caetahumant & Other Pra**

Revision: 7

/ April 30,1996 7.2-51

1 i

7. Instrinnentation and Controls

]

i l

l Table 7.51 (Sheet 7 of 12)

Post Accident Monitoring System i

Ranger Typer Que*h Neanbar af QDPS Veneh6e Steses Casagery leserieneese Power leeceases Renarts l Eeviremessesai Ce Rote N Sepply (Note 2)

Chund wsw puesp Oe/Off F3 N= Noes IW NonlE No stamm j

(%Insd water velve Opse/ P3 Nees Nees thake Nom.tE No stana Ciceed Speos fuel pass c3000 r3 Nees None I/pesop Neo4E No posep now 5pm Spentthespool 50- D2,F3 Mad Neen i Nee.tE No ineparames 250*F 5 pees thel pool 08006 Dt F3 Mad Yes X (E Yes waan kval (N etc. y)

C'MT to reactor Opsel D2 Hast No thake Neo-LE Ne veenal valve Clamed senes CMT anses notessee Opent D2 Herst Yes IJeake IE Yes .

l valve stana Cleend (Ness 7) t'bfT level 48006 D2.P2 Hersh Yes I/ test IE Yes DtWrf is sencear Oyse/ D2 Herat Nees thake Neo!E Ne veenes vake steen Closed (Nee MOV)

UtWST es Open/ D3 Name None lhake NeoLE No remser vesset Oceed valve sisees (MOV)

ADS: iksi. Open/ D2 Hast Yee I/ rake LE Yes l nacond and Closed (Neen 7) thed eenge wake sines ReYision: 8 June 19,1996 7.5-20 3 Westhghoust

-- _~ . - _ - .. = __ __ _ __ - _ , , - - -

- 1 l

7 I 4, il 1

7. Instrumentados and contrais Table 7.5-1 (Sheet 12 of 12)

, Post. Accident Monkoring Systern 4

', Rasent Typet W (

Number of QoPs I Varishie stanu Cassgery Power E*'ir*=='='as Insrummemes indsmens Be j summis a sr.d sapesy es 2)

Pos

• 10'I. E3 Name Nous i No

'Non lE samphag manue  !@

radianos mMr 1

Mass muse las IF I. C2.E2 Mad Naas IAase Non-lE No re- level IE io=

1 Todiascal supput IF I. D Name Hans  !

I Noe.lE No casear radamen 80' a

, nur M- . N/A E3 Ness Name N/A Non-lE No $as yarde Parnassus I Halir. bF f I 1. Taunt now - - a ensamed base en een of huast now dersomL 1

l 2. The saris soformaman is avaaminis a ens enceinscal supymit esser via me unamiser tum6 Informance aveamble om aus gusheed dna m ryssa j

as also avnGable a she rumme shundows sortsumaa. I

3. Notes pus: 177to 105 P W IFU m 17 pCWus ledess: 1782, gp pg,,

d. The ammha of sonnemaan ngered snar muhis penas aumenons a two. A tad damsel as available estegh tangenry - to rumalve mdemmmune suhigory if ammamary (Sea mesannes 7.5 ok

, $. Nabis put IF Ien 10'3

- ,Fn . ,F ,o.s 8

ladsmas: IFII e 17 pCWus

6. Dayes af sabemahng a cnicahund from prussirmar puummere and RCS het hg ammpaumme.

l l 7. This aussunnus a met seemed sAsr 24 heart j

f 1

i 1

J e

Revision: 8 3@ 7.5-25 June 19,1996

l 9

DSER Open Item Tracking System Report for Chapter 7 Items not Statused " Resolved" by NRC '

9 e

l

Al%00 Opea item Tracking Syst;m Dat: base: Exec tive S;menary pain 2/18M7 Selection: lnrc si codejo'Rewiw# And lDSER Section] hke'78* Sorted by item 8 Item DSL R Sectam/ 1 stic/lkwnption Resp (W) NRC No Heanth Questnm 1)pc Iktail Status Engmeer Status Status I. citer Nu. I Date 1018 NRR/IllCB 7 4 4-8 dst.R4)I ITAAC/Deutsch, K Action N Action N NSI ANRC-%-4737

~ " ~ ~

%;cstmghouse should descnbc in the SSAR, CDhl, and 1 FAAC the digital system design process Westmghouse should provide a detaded descntpeton of the digital sy stem design process in the SSAR and CDM math a correspamdeng IT A AC.

.. = ===n_2 z- .

m. =-==-:

Action W . WCAP-13383, which descnbes the digital system design process is being updated The certified design matenal and IT AACs udt be modified The SSAR has been modified to reference the design process and to mdicate the soRmase design standards the design process donitwms so lhis informatum is peovided in Revisnm 3 of the SSAR, Subsection 71.2 I5. The WCAP and IT AAC resishms must be completed beftwe tes item can be closed out NRC has requested a presentasion when all elements are compicted WCAP-13383 rev due 5/30N6 rha 5/7/96 WCAP-t3383 m eepro 6/14 for &l7 teicase rin &l436 Closed - Response provided by NSD-NRC-96-4737 Per an i 1/21 W/NRC telecon, the NRC thinks the 1&C ITAAC is deficent arul requested that we *fis" the iI AAC or justify /csplam deviatkwis from the SRP I4.3 5 to NRC satisfaction NRC to provide specific comments on the IIAAC. ekn 12/2 ___ _

1039 NR101tlCH 7 I 7-l DSLR4){ ITAAC/lkutsch Action N Action N NSIFNRC-%-4737 Westinghouse should descnbe a commercia! grade item dedication program for digital systems Westinghouse has not addressed the commercial grade acm dedication program that is necessary to ensure sufficient quahty m the design el safety-related and nonsafety-related IAC systems using commercial of-the-shelf equipment The design, verificatum, and ialatatami process for COIS  ;

soRware and hardware should be clearly documented for design certificaten - - - -

= = ..r . - . ._ = .:..: __

Action W - WCAP-33383 is bems updated to mclude a commercial grade item dedication process lhe SSAR has been namhfacd to reference this process. This information is provided m Revision 3 of the $$AR, Subsection 71215. The WCAP resision must be ciwnpleted before this item can be closed out WCAP in repro 6/14 for 6/17 assuance rkn 6/14 Closed - Response provided by NSD-NRC-96-4737.

Same as item 1038 rkn 12/2 ~ ~~~ ~

DStR4)I ll AW/lic:.tscii, K ' Actum A-^ Action N 104i NRR/IllCB 726-1 1he staff has not yet completed sts cvstuatum of the software architecture design

. because WCAP I4080 was submitted in July 1994, the staff has not completed ets revicw of the document and as umtmuing its cvaluathwi of the soRware archnecture based on both the proposed design and the associated design process. The results from this esatuainm udt be presented m the final 5ER for AP600 Closed - Westmghouse has completed necessary submittals to support stalT review Per Il/2I W/NRC telecon, when the NRC agrees with the design process through their resice of the Ii AACs, this stem udl be clo ed ekn 12/2 ITAAC/Deutsch Atthm N Actum N N I D-NRC-95-4464 1943 NRR/IllCD 728-1 DSI R4)I

~

W'est'Ang~housishould provide ' alasc ~iss' ion 'conccrnmg the quahticatkm of digital equipment to the elettrtwnagnetet ens mmment Westinghouse has not addressed the issue of electromagrictic environmental quahfication and has not comnutted to the appropnate stamfarJs

- - - _ _ = _ m: _

Closed - List of standards revsewed by NRC durmg meeting on May I5-16. Standards incorporated mto Resiuon 3 of the SSAR, Fulnccthm 71416

+

Per an iI/2I W/NRC telecon, the techmcal issues are resolved When NRC agrees math deugn process thru f I AAC res seu, this stem udt be cimed 4

1%ge i 1otal Records. 14

_____.__m _ _ _ _ _ _ _ _ _ _ _ ..__ _ _ _._ . _ . _ ._ . _ .. - . _

r AP600 Open it:en Tracking Syntem Datbese: Ezecctiva Semunary Dat:: 2/I8/97 Selection: lnre sa codelo' Resolved' And lDSER Sectionj hite '7** Sorted by item 8 hem Inst R Sccimm/ i ngefne3t,,,,,o, Resp (W) NRC ,

No Branch Questam 1)pe _ Detael Staeus Lagmeer Status Status I etter No /

l)sec

'4044 NRR/HICH 725-2 DSLR4M ITAAC/Deutsch. li Closed Actam W . NI D-NRC-95-4464 I

~ ~ ~

We'stmghouse should provide informaniim concernericEirosunentalipuolificEmm of PMEchec~nts addressing k$d sciispermeau'c'rlses ab'ove she rouen anubient emperienced by the compinsents durmg operation ,

le is deseralde to have addoseonal snargin bank into the design he components should, therefore, be quahlied by testmg to higher eeniperatures  !

shan specified an the SSAn lot a given room emrmunent Westinghuine should address this concern in the SSAR. Westinghuine should also e provide mold enverunenent esguipsument apushfecation in alue CDM wish she m.% " ;, ITAAC

=:= :: ;== v = = . = . . . _ . = - - - z.:

Closed - Technical infornissoas asseeded to by NRC during usiecting on May 15-16 Additional technical meurmatum regarihng the equernient design margen to loss ofIIVAC has been incorporseed into Revision 3 of the SSAR, Subsection 714 I 8 rkn 12/2  ;

Westinglioisse needs to decule approach to close this item An 12/6 Action N - NRC stdl has the acleon to evalessac the Wessenghouse proposal on procedural fix ofinstrument oserheatmg after 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> permal (6/21 I I

meetag wish W/SPLB/HICB) Based on 11/21 W/NRC telecon, dois anproach is reasonable, see quahficatum program in SSAR Section ) I1.

Action W - NRC resquested W provide proposed COL stem for qisahfication margen and instr.unent setpoint data ur document en the CDM and

• m.,_, ' ITAAC (W is " . options, did not conuma 80 enhet approach) An 12/2 Westinghouse does not consider there to be an applicalde COL action to identify. Technical inlormatum related to design margm agamst a kiss ol j HVAC was provided in SSAR 7.1416 and is considered tecluiscally resolved, as waws previously agreed to by NRC. His item is umsedereJ closed since there is no Wessinghouse action reepsered at this time to address this ince (sence she NRC relaecs this conunent to the PMS II AAC, the responsside engineer is changed to ITAAC) An 1/14/97.

Action W -(freen NRC on 1/28/97) Submit revised CMD & ITAAC to include COL action so include additional design margm to acawnodate a loss of the morneal IIVAC. Provide an alarm if useernal cabinet terrqueraswes reach an excessive value jw=1/28 his was previously closed and as stdl considered closed, nicaneg there is no Westinghouse action identdied or resguired o close this nene. all necessary submittels have been estade. For background. SSAR Section 7.1.4 5 6 was revised on Feb 1996 to address ihn Specifically, there is a ,

senernce which reads,"he cabumets contameng the digital eqinpmen; are provided with temperature senuws which provide an alarm if intemal cabinet temperasures reach an excessive ,alue? His is closed An I/3087.

?

Per telecon wish Hulbert li today, the action is for Westinghuine to include this alarm in the I T AAC rLn 2/18/97

~ _ _

[

DSLR4M iTAAC/Lmdgren/lkuts Action N Actum N 1949 NRR/HICB 7.55-1 __ ._.

j TVestenglioisse shoistd describe the design feasures of she incare instrumentateon system j

in its response to Q492.5 deced Duly 25.1994, Westinghouse stones shal infornietion on the enipksyment of fined mcore detectors en cimjunctum

• wish an online power desersbution neonieorung system will be provided to she NRC to support the fmal SI R.

- - = . = _ - . - = - . . -

j Closed - The technical informasson was accepted by the l&C Branch of NRC diaring the meetuig on May 15-16 This technical informatam has heen incorperseed into Revesion 3 of the SSAR, Subsection 4 4 61.

Open for ITAAC based on fem froen NRC I/21/97. An [

t for Chapter 7 shis seem is resolved (NRC/R$8 to conununicate any concerns unh quahficaemm of thernmicouples and inwrwnent undant capabshty outside the scope of Chapter 7) An 12/2 _. _

b i

l'ege 2 1 otal Records' I4

- . - . . - . - - . - - _ - ~ _ - - - - - - - . .. - ... ~_. - . . - - - . . - . - - . - - . . - . . . - -.

APtes Open It en Tracking Sy:tems Database: Executive Samannary pate: 2/18/97 i Selection: lnrc st codelo* Resolved' And (DSI:R Sectionl like '7*' Sorted by item 8 leem DSI R Sectum/ I stic/lkscription RCSP (W) NRC No Branch Questoon type Detsel Stasus Engineer Status Statu5 l.ener.No / - . ~ .De.se. -

1952 NRR/IIK H 7621 D$l R4M Schulz. T. Closed Actum N

~ ~~

Westeghouse'shoenid pro' vide additEnal deisgli Atmeis3f the accunielAEwlatEnTalve inecriticis~M to safety to confirm th5s'aliidesign meets the relev ant sceguerensents of the SRP, mcluding ILLE 279. _

Closed - Additacnal technical enfwmatiost has been mcorporated into Revision 3 of the SSAR, Subsection 7 6 2.1. I igure 7.2-1 eas alw naalified to include additional technical detail Action NRC - Per 11/21 telecon, NRC to revicw technical enformation already provided smce this operasur is nimsafety, nut unportant to safety, has sepersee power, possteve 3 position indicatums, and power removed at power (ciesessent with Tech Specsl and hmit switch alarms rkn 12/2 lechnical informasson prowided NRC to advise to sesolution status. rkn III4N7 Per fan, NRC considers this syen for interlocks concern (f SLR open ince 7 6.2-1I skn 1/2IN7 1053 NRR/lilCH 763-1 DSER-OI Schulz, L Closed Actum N Westeghouse should provide additional design detaels of the IRWS) discharge valve interlocks important to safety to cimfirm that the desegn . . - .

neerts the relevant reaguirennents of the SRP,includmg ILLE 279.

- ~-

= - - - _ . - - -  ;

Closed - Addnional technicalinformation has been incorporated into Revision 3 of the SSAR, Subsection 7 6 2.2. Figure 7.2-1 =as also modified  ;

r so include additional technscal dessel Action NRC- See 1052 An 12/2 j Tectnical information has been provided NRC to ad,ese regarding resolutum sta us rkn Ill4N7. e Per fan, NRC considers this open for interlocks concern (iSER open item 7 6 2-1) rkn 1/2 tN7 NRR/lllCH 7.72-1 DSLR4)1 ITAAC/DeloscJrank Action N Actum N N I D-NRC-95-4464 1055

~

(WestM should prudaUs~tiIm~a~l infiAnasson concerneng the design'of the DAS _ [ j

~^ ~

- ~

Closed - Technical informasson accepted by NRC during essecting on May 15-16 This addetidal te'clmecleI detail has lYen encmporated into  ;

Revision 3 of the $SAR, subsection 7 71 11.

NRC action to review 11 AAC Per 11/28 telecon, this seem is now subject to DAS 1TAAC cannment resolutum/completam rkn 12/2 t ITAAC/lkutsch Acthm N Actum N NND-NRC-4875  ;

2023 N RR/IIR H 7 DSER4M50

~~ ~ ~~~

' ~ ~~

2fNoC-~ Eliidustry S't'andards for Dignal Syseems ~~

i While the SSAR references ILLE standads 279. 384,603 and 796 for the deugn of AP600 IAC systems, the stall es umccrned that there es no '

reference to digital microprocessor-related standards Specifically they are concerned about the latL of standards relased to muliertener archiaccture,conimunications protocols, and hardware / software design 1he staff wants Westm6 ouse h to make an espixit commitment to endustry hardware and software related standards No detaded docuenentainon of the process and no phase d 1TAAt' for verification nf the design [

cTion W -leem 1037 closes all histisinIsentence ofitem. Remaining action to addressWAtarted'documentatum of the process arm no phased j i

11 AAC for verificasion of the desaget SSAR Ch 7 I commets to a VAV program, meeteng Standards, etc., such that NRC capectations are snet. When the fi A AC for 1%fS is cumplete, this stem will be closed rkn 5/7N6 I

~

Closed -ITAAC subented by NSD-NRC-%-4875 of II/7N6 l

Per i1/2_1 telecon for DSI R Ch 7 NRC = ants to descuss ITAAC approach with Westmghou c.

i t

r

[

l'<ge 1 Iotal Records: 14 l

AP600 Open It2m Tracking Sy; tem Database: Esecttin Summary page: 2H8/97 i Selecties: lnrc si codelukesolved* And (DSER Sectionj like '7'* Sorted by item #

leem DSI R Section/ I nic!Dewraptum Resp (W) NRC >

l ' No Heanch Question T)pe Detast Staeus Engmcer Status Status I citer No / Dase 2025 NRR/IllCB 7 DSLRol50 $$AltREV/Mdict Confrm-W Actam N

~~ ~ ~ '

~~

29 l'avtronmental Qualification of Dh5 bi ~uipment aniScnN 1he DSB R indicases that the DAS equipment must be designed and qualified to the environment m which n needs to perform lhe Westinghouse position is that the DAS equipment will be designed to function the environment en which et needs to perform llowever, the DAS equipmens udl not be sisbrected to a full-blown 10 Cl R 50 49 / IEEE 323 qualificatum program ~-

I

=_u====. -

~

=d - SSAR Chapter 7 sectum 7 7.I II revised to address Close Per an i1/21 telecon, NRC thunks the DAS sensors and actuated devices (c g , PRilR solenoid valvel should bc quahtied to a higher (PMSI standard but Westinghouse does not agree.

Hy 12/6 fax, W proposed SSAR change to clarify qualificatum, NRC to review approach En 12/6 Completed in SSAR Rev 10 rka 1/14/97 Whenpst I checked and it didni get into SSAR Rev to it Wil.l. get into Rev i1. See NSD-NRC-97-4947 rLa 1/30N7 ~ ~ '

2272 NRR/SRXH 762 M1G-Of Dedh ~ Closed Actum N j

__ APRIL 19,1995 (HSil) DISCUSSION ITLMS i

15 Availahdny of Safeguards -Interlocks (SSAR Sectum 7 6 2)

Sectien 7 6 2 of the SSAR discusses the inactlock systems to verify the availabdity of safeguard functams,i e , to ensure openmg of the isidation  ;

valves of the accumeslasors,IRWST, and PRHRilXs.11 ese valves are motor +perased, normally open valves, and are controlled froen the main control room and remose shutdown wod station a SSAR Section 7 6 2 stases that, as a result of the confumatory safeguard open signal (which wdi automatscally open the isolasion valves, r overridmg bypass leasures to allow the isolation valves to be closed),isolatum of an accumulator with the itCS as pressure (or isolatitui of the IRWST gravny injection ime when the tank is seguired to be operable, or isolation of the PRilRilX inlet Ime when the PRilRllX is requued to be '

operabic)is acceptable What are the design reliabdity of these inscriocks to ensure these isolation valves =ill be open upon the confirmatory safeguard open signals? Is this practice acceptable for current operstm8 reactor to allow accumulator isolated at pressure?

~

~

E5 sed - At he Reactor Ustem Branch Meeting on 4/25/95, WestuMe Efe'rre'ditE use ofid'nTtal e inteslocks on the accumulators and~ i IRWST as those currently used on the accumulators at current plants (powet locked-out) CMIs and PRilR interlocks are not powcr lothed out but instead redundant ontrollers are provided for each valve along with three-way redundant valve positions. Westmghouse aho referred to the 3 Revision 2 SSAR 6 3 for the design detads. Revision 3 of the SSAR, Sectam 7 6 includes the interlock in'ormation Based on It/2I telecon, NRC doesn't think the SSAR Section 7 6 is sellicient and has been asked to prowide specific comments rkn 12/6 Closed since these is no Westmghouse action at this tune. NRC to advise regardmg status. rkn III4N7_

i TECllSPI C/Schulz Closed Actum N 2273 NRR/SRXB 762 MTG-OI

~~~

~

APRIL 19,19951S11)IEU SS'lO5'INMS

15. Avadabsiny of Safeguards - Interlocks (SSAR Section 7 6 2) '

b. SSAR Section 7 6 2 also states that the maximinui permissible tune that j

an accumulaser valve (or IRWST shscharge valve, or PRHRilX inlet valve, respectively) is closed when the reactor is at pressure as specified in the TS Where are they specified? - ~

Action W - Section 3 5 I of the lech Sg;ecs specifies the ma$imism permissble alv'e tunes the revised Tech Specs wdi be submnted June 1996, as which tune this item can be closed Closed - wnh issuance of the Tech Specs an SSAR Res. 9.

Action NRC - Per 11/21 telecon, NRC to rew new Tech Specs to ensure this is resolved / closed rkn 12/4 "

Westinghouse actum is ctanplete for this item NRC lo advisc on status rkn iII4N7. _ _

Page 4 lotal Records: 14 6

_ ~ . . . . . . . . _ . . - . . . . - - _ . _ ___. - . . - - _ _ _ _ _ - - - . _ - _ - - - _ . - _--._.___---_ .-.--.__. ___._- _ -_-_ - -_____.___ _ ___._._---_._____--__ - _._._---_.--.-- _ _ _- - - --____-_____ - _-_.__.-_-__ - -

APHO Opea Itm Tracking System Database: Executive Summary Date: 2/18/97 Selecties: lnre st mdej<>' Resolved

• And lDSER Section] 1Ae '7 Sorted by item #

leem Ing R sectum/ g ,,gefg ye,c,,,,,, Rc5P (W) NRC No Hranch Type Iktail Status Engineer Status Status Irtter No / thu Questusa .

4257 NRMIK'N 7 mig-COM SS ARREV/Ikunch, Ke Contnn-W Confrm-W

_ k ^ .' _ _ .'" "'"

~

l

'W5:nghouse has confmned this is an appropriate reference and transmitted the SSAR markups to ERfiEiiiicm'isidd kic^n$cWnwked '

changes are included in the neat SSAR rev ska Ill5N7.

Refer to NSIANRC-97-4947 for changes to be included in SSAR Rev II. rkn 1/3_tN7, . _ _ _ -

t i

?

I l=,Fe $ Iotal Records. 14 I

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ .________.m_m ___ _ _ _ _ _ _ _ _ __ _ . _ . _ _ _ _ _ _ . _ _.

h i

FAX to DINO SCALETTI February 20,1997 CC: Sharon or Dino, please make copies for: Ted Quay Bill Huffman

. Diane Jackson i Tom Kenyon Joe Sebrosky Robin Nydes Cindy Haag Don Lindgren Bob Tupper Bruce Rarig Brian McIntyre Ed Cummins Bob Vijuk

RC is requested to please acknowledge receipt of information related to each of the following Open Items. These are a subset of the items with " Action W" in "NRC Status" for which I have personally checked, since the first of the year, that we have submitted what we believe is the resolving l

information. Unlike those on the other list I will send you today, I have not prepared a background package for each of these, However, the reviewer in each case should have a submittal from 1 Westinghouse as identified in OITS for the item. Recognizing that reviewing for completeness of the j response in each case constitutes an NRC action, we recommend that receipt acknowledgement be accompanied by direction to change their "NRC Status" to " Action N". If these are truly " Action W", please provide a description of the action Westinghouse is expected to take. We know of no action required. This is the sixth weekly request of this type.

5, 21, 142, 157, 164, 172, 173, 177, 182, 184, 262, 300, 305, 308, 333, 405, 457, 458, 628, 681, 698, 706, 710, 716, 717, 718, 719, 722, 724, 729, 730, 731, 732, 801, 802, 805, 807, 809, 972, 973, 1009, 1037, 1038, 1039, 1040, 1041, 1043, 1045, 1052, 1053, 1055, 1101, 1102, 1195, 1197, i 1210, 1225, 1226, 1227, 1228, 1231, 1232, 1317, 1354, 1356, 1360, 1361, 1365, 1366, 1367, 1368,  !

1369, 1370, 1371, 1392, 13 % , 1458, 1461, !697, 1698, 1699, 1700, 1701, 1702, 1703, 1704, 1707, 1716, 1727, 1730, 1731, 1742, 1745, 1747, 1740, 1753, 1760, 1809, 1810, 1811, 1885, 1888, 1996, 1999, 2018, 2019, 2023, 2024, 2025, 2034, 2040, 2043, 2044, 2045, 2051, 2199, 7200, 2201, 2202. J 2272, 2273, 2347, 2348, 2349, 2442, 2457, 2515, 2676, 2683, 2684, 2686, 2691, 2698, 2939, 2942, 2945, 2958, 2959, 2960, 2961, 2 % 2, 2963, 2964, 2965, 2966, 2 % 7, 2 % 8, 2 % 9, 2970, 2971, 2972, 2973, 2974, 2975, 2976, 2977, 2978, 2979, 2981, 2982, 2983, 2984, 2985, 2986, 3057, 3098, 3122, ,

3126. 3127, 3128, 3197, 3247, 3264, 3265, 3266, 3267, 3268, 3269, 3270, 3271, 3372, 3398, 3399, )

3400, 3401, 3402, 3427, 3439, 3468, 3469, 3470, 3471, 3472, 3473, 3505, 3517, 3895, 3944, 3945, )

3946, 3947, 3948, 3949, 3950, 3951, 3952, 3953, 3954, 3955, 3956, 3957, 3958, 4123, 4124, 4125, j 4126, 4127, 4128, 4129, 4130, 4131, 4132, 4133, 4134, 4135, 4136, 4137, 4138, 4139, 4140, 4141, 1 4142,4143,4144,4151,4224,4225,4226, and 4227.

Thanks Jim Winters  !

412-374-5290 l I

, i

e FAX to DINO SCALETTI February 20.1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay

Don'Lindgren 4 Richard Orr Ed Cummins l Bob Vijuk Brian McIntyre l OPEN ITEMS FOR SSAR SECTION 3.8.4 l- This is a background package for the remaining'open items for SSAR section 3.8.4. SS R section 3.8.4 is of interest because by our joint NRC/W schedule, the FSER for this section should be turned

, into Projects by the end of March. There are 8 Open items with NRC Status of Action W. Four (4) of these items (745, 750, 751 and 755) still require some Westinghouse action. Westinghouse believes the other four (4) items were addressed in or prior to the January 16,1997 meeting with NRC. We are still waiting for the NRC letter report on the January 16,1997 meeting. Currently, our records show no additional outstanding Westinghouse action required for section 3.8.4, except items 745,750,751 and 755, and we request that NRC provide a definitive action for Westinghouse or provide direction to change the status of these items. We recommend " Action N". Thank you.

Jim Winters 412-374-5290 tc

AP600 Open Item Tracking System Database: Executive Sununary Date: 2/20N7 Selection: lnrc st cuJe]=' Action W' And lDSER Section) hke'3.8.4* Sorted by llem #

Item DSER Sectsonf . Title /Desenptum Resp (W) NRC No Branch Questum Typc Detal Status Engurer Samus Staus letter No. / Deee 740 NRR/ECGH 384.1-3 DSER4)1 Orr / NRCSM Ckned Actum W s osmgtunne slumkl provide in the SSAR a descriptm and design detals of nuxtules located in the austhary lukhng, and sluiukt indicane the difference j ximeen these nuxtules and those locased msade the contanmein.

Additumal infonnmum as provided in RAI respmse has been incorpurned in SSAR Rev 3 for the structural naulules in the fuel handhng area and for the fmned ihmrs I;sther discussion wdl be added in SSAR Revision once nuxtule telmior study is completed (see DSER Ols on sectun 3.8.3).

Okmed in nectmg with NRC III6f97 nunor SSAR change sim>wn in draft revision 745 NRR/ECGB 38431 DSER-OI Orr / Prasad Actum W Actum W N1DNRC-95-44M Westingkmise should accepaldy address the issues regardmg the consideranon of hve lomt m the seisnue nulet SS AR subsection 3 8 4 3 2 3 has been added and 3.7.2.3 I has been revised.

Meetmg 6f15/95 - NRC mdl provale copy of staff positum on cominnmum of hve lomt with SSE. Westmgkure shoukt address this positim NRC posnion provuled in letter of July I 8.1996 ,

NRC Status : Actum W - The staff reviewed the Wesungluiuse letter respmse wah draft maiup. The resemse is unacceptable. ( D Westmglumme di es not include hve ked in the global scismsc rnudel (West has calc. to show msignificance - SSAR reviwm is needed) (2) For kx al effects, the Wesunglunne u*e of 25% of hve lomt for SSE cases (rmher than 100% in all cases)is unacceptable. (3) fig ghibal effects. Westmglunase does not include snow kmd as l defined in the staff position (NRC letter dated July 18.1996)- SSARy_ z is needed (12/16/96) 750 NRR!ECGH 3844-2 DSER4)I Orr/ANSALDO Actum W Actum W Westingluurse shoukt provide for stad review the final design calculatum for the shickt lxnktmg and the passive contamnent coolmg water storage tank f N Meskatology was presented to NRC in meeting on March 2. The final design calculation for the shickl tunkhng and the passive cuntanment coulmg weer Q

% storage tank were reviewed during the meeung in June 1995. Westmghouse slumaid address comments identified in meetmg runes. i h Comments from meetmg have been addressed and were discussed dunng meeting on Marth 7.1996-Closed: remanmng issues are tracked under new R Al 230100 transnutted by letter of Arnt 5.1996 NRC to review design calculations dunng meetmg in Denmber NRC Status: Actum W - Westinghouse mdl pide the design of the nng beam mcludmg torsiemal nuwnera and axial tension (I2/16/96) 3844-3 DSER-OI Orr/ANSALDO Actum W Actum W 751 NRR/ECGH Westmgknsse stundd adJ COL Action Item 3 8.4 4-1 to the SSAR.

Design calculmums for shield txukhng nx>f were reviewed by NRC staff dunng nretmg in June. Tank defontiarniris dunng tank filling are small and snomtonng of tank deflectums and compenson against predictions is not meanmgful.

NRC Status. Action W - Post-constructum testing is necessary to confirm adequacy of the FCS tank. I12/16/96)

Orr / NRCSM Ckned Actum W 754 NRRKCGH 3844-6 DSER4)I Westinghouse shoukt provide analysis prucedures and desagn details of the spent fuel ptxd includmg fuel racks the fuel transfer canal and the new fuel storage area The spent fuel pool and transfer canal are part of module 'M20/M21. Additional details wdl be provided in response to 013 8 4 5-l . Analysis a detals of the fuel racks are covered in SSAR Chapeer 9. .

i NRC to review design calculmums dunng meeting in January,1997.

SSAR 3 8 4. Rev II includes seference to Sectum 9.1 for the new and spent fuel racks SSAR subsectum 9 I 6 requires the G)L apphcani to perform j confirmatory analyses. Dese include reconcibanon of loads on the sann1. ures.

Page. 1 Total Rectmis: 8 I

v-( -t.

AP600 Open hem Tracking System Database: Executive Summary Date: 2/20/97 Selectiost: lnre st ccJej=' Action W' And [DSER Section] like '3.8.4 Sorted by item #

item DSER Sectum/ Tale /Descnptum Resp (W) NRC No Branch Questum Type Detal Status Engmeer Status Staus W No / thne 755 NRR/ECUB 38447 DSER4)I Orr / McDermott Action W Actum W Westmghouse's hst of components provided in the June 30.1994 respone to Q220 83 should include txth the IRWST (as part of containment insemal a structures), and the att baffle (as part of the shield taniding).  !

The descnption of the open item is misleading This open item covers the design reptuts for the nuclear island tecm.a. ausiliary tuskhng, contanment l internal structure and the shield buildmg Iksign reports for the nuclear island basemat, auuhary Innteng.contamment internal sinacnare and the shoekt j tunkimg wdl be available in meetmgs in December,1996 and January 1997.

I NRC Status: Action W - The staff reviewed the summary reports shield tnnkhng utm;f structure, ba,emat, and aunhary tuukimg in the Deumber 1996 nretmg. The contamment intemal structures summary repon will be reviewed in January 1997 audn meetmg. 02/16%

Actum W - meeting 1/l&97 - deternune kwals on air baffle due to flo_w and vones sheddmg and denumstrase adequacy of air baffle.

757 NRR/ECGB 3845-1 DSER-Ol Orr / NRCSM Ckiel Actam W Westinghouse should molude in Appendia 3A ol'the SSAR.a desenpuon of cntena teed for the different cunfigursums and applicanonsif there are i differenas an the detads of these modules __ l See 013 8 4.1-3. l Chud in meetmg wah NRC l/l&97 - nunor SSAR change shown m drah revision j 753 NRIUECGB 3845-2 DSER-OI Orr / Ritz / NRCSM Ckud ' Actum W Wesunghouse should provide requiremeras in the SSAR for erudatar construction m the auuhary tuuleng. l Additumal information to be added to SSAR as discussed un Dewmber,1994 neenng Le See NRC letter dmed 7/15/96

• rnore informanun needed on qualmy control

• Ckned un meenng wah NRC t/l687 - nunor SSAR change shown in draA revision h4 Page: 2 Total Rectwds: 8

o j

' s c,

FAX to DINO SCALETTI  !

February 20,1997 CC: Sharon or Dino, please make copies for: Ted Quay Bill Huffman  :

Diane Jackson Tom Kenyon  ;

Joe Sebrosky Cindy Haag Don Lindgren Robin Nydes ,

i Brian McIntyre Ed Cummins Bob Vijuk This is a reminder list of the Open Items where we have recently provided background documentation )

showing the difference between "W Status" and "NRC Status" In all cases, we believe the next action is with NRC and await your definitization of a Westinghouse action or your direction to change

{

the "NRC Status" to something other than " Action W" Note that we have received no information  ;

from NRC on items on this list for over a week. I Open Item Number Westinghouse Submittal Request for Status Change 21(RAI 471.24) 2/14/97 2/14/97 142 (M3. Il-9) 2/29/% 2/3/97 157 (M5.2.5-13) 1/9/97 2/12/97 164 (MS.2.5-20) 1/10/97 2/12/97 172 (M5.2.5-29) 1/14/97 2/14/97 j I

173 (MS.2.5-30) 1/14/97 2/17/97 177 (M5.2.5-34) 12/20/ % 2/18/97 405 7/8/% 2/11/97

! 681 (DSER 3.8.2.4-3) 2/11/97 2/17/97 706 (DSER 3.8.2.4-28) 2/11/97 2/17/97 l

710 (DSER 3.8.3.1-1) 1/16/97 2/18/97 716 (DSER 3.8.3.2-5) 1/16/97 2/18/97 717 (DSER 3.8.3.3-1) 1/16/97 2/18/97 718 (DSER 3.8.3.3-2)  !/16/97 2/18/97 j 719 (DSER 3.8.3.3-3) 1/16/97 2/18/97 l 1 of 3 1

L

. . .- -_ . - - . - - . . . __ - . ~ . _ - - . . . . . . - - . _ .__ _ .- - ~ _ -

i 4-Open item Number Westinghouse Submittal Request for Status Change _

1 722 (DSER 3.8.3.4-3) 1/16/97 2/18/97 I 724 (DSER 3.8.3.4-5) 1/16/97 2/18/97 729 (DSER 3.8.3.4-10) 1/16/97 2/18/97 I 730 (DSER 3.8.3.4-11) 1/16/97 2/18/97 731 (DSER 3.8.3.4-12) 1/16/97 2/18/97 l 732 (DSER 3.8.3.4-13) 1/16/97 2/18/97 740 (DSER 3.8.4.1-3) 1/16/97 2/20/97 1

754 (DSER 3.8.4.4-6) 1/16/97 2/20/97 l 757 (DSER 3.8.4.5-1) 1/16/97 2/20/97 758 (DSER 3.8.4.5-2) 1/16/97 2/20/97 j 1210 (DSER 12.4.2-2) 4/30/% 2/6/97 1227 7/8/% 2/11/97 1228 7/8/% 2/11/97 1231 7/8/% 2/11/97 I

1232 7/8/% 2/11/97 1354 (DSER 18.8.1.3-1) 1/30/97 2/14/97 l 1356 (DSER 18.8.1.3-3) 1/30/97 2/14/97 1360 (DSER 18.8.1.3-7) 12/19/ % 2/14/97 1 1361 (DSER 18.8.1.3-8) 12/19/67 2/14/97 1365 (DSER 18.9.3-3) 12/3/% ~2/14/97 1366 (DSER 18.9.3-4) 12/3/96 2/14/97 1367 (DSER 18.9.3-5) 1/10/97 2/14/97 1368 (DSER 18.9.3-6) 12/3/96 2/14/97 1369 (DSER 18.9.3-7) 12/3/% 2/14/97 l 1370 (DSER 18.9.3-8) 12/3/% 2/14/97 l 1371 (DSER 18.9.3-9) 12/3/% 2/14/97 1392 (DSER 18.11.3.4-1) 1/30/97 2/14/97 i 13% (DSER 18.13.3-1) 11/7/% 2/14/97 l 1809 (DSER-CN 3.10-2) 9/5/96 2/13/97 2 of 3 l

l I

s.

l 1

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Open Item Number Westinghouse Submittal Request for Status Change 1810 (DSER-CN 3.10-3) 9/5/96 2/13/97 1811 (DSER-CN 3.10-4) 9'5/96 2/13/97 1888(DSER-COL 3.8.2,4-1) 2/11/97 2/17/97 2034 7/8/96 2/11/97 2043 1/10/97 2/14/97 2044 1/30/97 2/14/97 2347 1/16/97 2/18/97 2348 1/16/97 2/18/97 2349 1/16/97 2/18/97 3057 5/30/96 2/18/97 3247 (RAI 230.98) 4/30/96 2/18/97 3264 (RAI 220.95) 12/9/% 2/13/97 3265 (RAI 220.%) 12/9/% 2/13/97 I 3266 (RAI 220.97) 12/9/% 2/13/97 3267 (RAI 220.98) 12/9/% 2/13/97 3268 (RAI 220.99) 12/9/% 2/13/97 3269 (RAI 220.100) 2/11/97 2/13/97 &

2/17/97 3270 (RAI 220.101) 2/11/97 2/13/97 &

2/17/97 3271 (RAI 220.102) 2/11/97 2/13/97 &

2/17/97  !

4617 2/14/97 2/14/97 Note that the status was changed for Items 4, 21, 30, 37, 123, 134, 135, 137, 138, 139, 140, 141, 144,158,586, %9,970,971,1300, and 1301 so they have been removed from the table.

Thanks for your help.

3 of 3

-- -- . - . .- - . _ ~ ~ . - - . . . .. - - .. ___- - .

3 V

FAX to DJNO-SeAtETTI. j j February 20,1997 CC: Sharon or Dino, please make copies for: Ted Qua

/ Bi uffman Diane Jackson J

Tom Kenyon

Joe Sebrosky  !

Cindy Haag Don Lindgren

{ Robin Nydes  ;

. Brian McIntyre  :

! Ed Cummins

- Bob Vijuk i

This is a reminder list of the Open Items where we have recently provided background documentation

! showing the difference between "W Status" and "NRC Status". In all cases, we believe the next

action is with NRC and await your definitization of a Westinghouse action or your direction to change j i the "NRC Status" to something other than " Action W". Note that we have received no information  !

j from NRC on items on this list for over a week.

Open Item Number Westinghouse Submittal Request for Status Change 21(RAI 471.24) 2/14/97 2/14/97 j I 142 (M3.11-9) 2/29/% 2/3/97 157 (M5.2.5-13) 1/9/97 2/12/97 164 (M5.2.5-20) 1/10/97 2/12/97 172 (MS.2.5-29) 1/14/97 2/14/97 173 (M5.2.5-30) 1/14/97 2/17/97 177 (MS.2.5-34) 12/20/ % 2/18/97 405 7/8/% 2/11/97 681 (DSER 3.8.2.4-3) 2/11/97 2/17/97 706 (DSER 3.8.2.4-28) 2/11/97 2/17/97 710 (DSER 3.8.3.1-1) 1/16/97 2/18/97 716 (DSER 3.8.3.2-5) 1/16/97 2/18/97 717 (DSER 3.8.3.3-1) 1/16/97 2/18/97 718 (DSER 3.8.3.3-2) 1/16/97 2/18/97 719 (DSER 3.8.3.3-3) 1/16/97 2/18/97 1 of 3

i V I Open item Number Westinghouse Submittal Request for Status Change 722 (DSER 3.8.3.4-3) 1/16/97 2/18/97 724 (DSER 3.8.3.4-5) 1/16/97 2/18/97

, 729 (DSER 3.8.3.4-10) 1/16/97 2/18/97 730 (DSER 3.8.3.4-11) 1/16/97 2/18/97 I 1

731 (DSER 3.8.3.4-12) 1/16/97 2/18/97 d

732 (DSER 3.8.3.4-13)  !/16/97 2/18/97 j 740 (DSER 3.8.4.1-3) 1/16/97 2/20/97 i l 754 (DSER 3.8.4.4-6) 1/16/97 2/20/97 757 (DSER 3.8.4.5-1) 1/16/97 2/20/97 )

758 (DSER 3.8.4.5-2) 1/16/97 2/20/97 1210 (DSER 12.4.2-2) 4/30/96 2/6/97 j 1227 7/8/% 2/11/97 )

1228 7/8/% 2/11/97 1231 7/8/% 2/11/97 1232 7/8/% 2/11/97 1354 (DSER 18.8.1.3-1) 1/30/97 2/14/97

]

1356 (DSER 18.8.1.3-3) 1/30/97 2/14/97  !

i 1360 (DSER 18.8.1.3-7) 12/19/ % 2/14/97 1 j 1361 (DSER 18.8.1.3-8) 12/19/67 2/14/97

, 1365 (DSER 18.9.3-3) 12/3/% 2/14/97 l 1366 (DSER 18.9.3-4) 12/3/% 2/14/97 j 1367 (DSER 18.9.3-5) 1/10/97 2/14/97

. 1368 (DSER 18.9.3-6) 12/3/% 2/14/97

1369 (DSER 18.9.3-7) 12/3/% 2/14/97 1370 (DSER 18.9.3 8) 12/3/% 2/14/97

. 1371 (DSER 18.9.3-9) 12/3/% 2/14/97 1392 (DSER 18.11.3.4-1) 1/30/97 2/14/97 13% (DSER 18.13.3-1) 11/7/% -2/14/97 1

4 1809 (DSER-CN 3.10-2) 9/5/% 2/13/97 2 of 3

a b

i Open Item Number Westinghouse Submittal Request for l Status Change l 1810 (DSER-CN 3.10-3) 9/5/% 2/13/97 1811 (DSER-CN 3.10-4) 9/5/% 2/13/97

- '1888(DSER-COL 3.8.2.41) 2/11/97 2/17/97 2034 7/8/% 2/11/97 2043 1/10/97 2/14/97 4

2044 1/30/97 2/14/97

[ 2347 1/16/97 2/18/97 4

1 48 1/16/97 2/18/97 2349 1/16/97 2/18/97 3057 5/30/% 2/18/97 l 3247 (RAI 230.98) 4/30/% 2/18/97 j 1

] 3264 (RAI 220.95) 12/9/% 2/13/97 3265 (RAI 220.%) 12/9/% 2/13/97 j 3266 (RAI 220.97) 12/9/% 2/13/97 j 3267 (RAI 220.98) 12/9/% 2/13/97  ;

)

l 3268 (RAI 220.99) 12/9/% 2/13/97

)

3269 (RAI 220.100) 2/11/97 2/13/97 &

2/17/97 )

1 3270 (RAI 220.101) 2/11/97 2/13/97 &

2/17/97 3271 (RAI 220.102) 2/11/97 2/13/97 &

2/17/97 I 4617 2/14/97 2/14/97 .

l I

l i

- Note that the status was changed for Items 4,21,30,37,123,134,135,137,138,139,140,141, I 144,158, 586,969,970,971,1300, and 1301 so they have been removed from the table. l Thanks for your help.

l I

3 of 3 l

J

\

8 i

FAX to DINO SCALETTI  ;

i February 20,1997 CC: Sharon or Dino, please make copies for: Diane Jackson Ted Quay Don Lindgren Richard Orr Ed Curnmins Bob Vijuk Brian McIntyre OPEN ITEMS FOR SSAR SECTION 3.8.5 This is a background package for the remaining open items for SSAR section 3.8.5 for your information. SSAR section 3.8.5 is of interest because by our joint NRC/W schedule, the FSER for this section should be turned into Projects by the end of March. There are 5 Open Items with NRC Status of Action W. All 5 of these items still require some Westinghouse action. Thank you.

Jim Winters 412-374-5290 l

/ h2.

AP600 Open Item Tracking System Database: Executive Summary Date: 2/20/97 Selection: lnrc st codel=' Action W' And [DSER Section)like '3.8.5* Sc ted by itern C Ikw DSER Section/ Resp (W) NRC Title /Desenprmn Type Detail Staus Engineer Status Status I ettet No / Date No Branch Questum NR RECCH 385-8 DSER Of Orr Actkm W Actum W 766 Westinghouse should povmie the walklaum pacLage of INITEC"s in-lumse computer codes for review and should venfy the alequacy of the piwi proceswd results obtamed Inun these cales Chised - Vahdavon pacLage is available for review in proposed basemas meetmg m December 1996 Cmipanums of sesults of computer code versus hand cataculation are included in revised skxumentaion Actum W - Westinginusse will povkle technicalinformatum and Gnal design calculmions for the staff to review in meetmg in Decenter.

NRC Status- Action W - Westinghouse has not provided (m enghsh) the AR M A computer code vahdakm and ven6catmn package (12/16/96) 767 NRRSCGH 385-9 DSER Ol (hT Acamm W Actum W Westinghouse should perfonn akhtional review of the basemas analyds, arnt should use dmphfied analydebawd on ACI U6 parduse4 to verify the dedgn adespiacy.

Westmghouse has gerformed akhtmnal review. Simplified analyws have tren performed to venfy deugn adequacy Action W - Westingimmse w di pmskje technical informaion and final denga calculmions for the staff to revtew an December meeting NRC Staus: Actum W - Simple analyus was reviewed and accepted flowever, the December 1996 deugn audat identified a lxk of conustency in the analysis and design of the entire structural syacm (12/16/96) 7'e NRRECGH 385-10 DSER OI Orr / Hechtel / NRCBM Actum W Action W Westinghuse should perform akhtsonal analyses im constnaction loals.

l Analyses have been performed for construction loats as descnbed in SSAR subsection 3 8 5 4.2. The sesults are used in remforcement deugn As discussed in June,1996 nretmg. Westinghusse will consider effect of kmg tenn wttlement NRC requessed thz the acceptable willenent dunng constsucikm shoukt he akhessed in the SSAR.

NRC Staus: Actum W - The SSAR 3 8 5 dran is incomplete; umwe mfunnamn is needed Westmghouw dies not consaler the effects of walenent dunng construction and dewarenng dunng construction. (12/16N6) N Reu4ution of this stem will also resolve Item 547 (DSER Ol 2.5 4 8 1) 769 NRRNCGH 3 RS-11 DSER4)I Orr / NRCBM Actam W Actum W N

Westmgkuse snould perform akhtsonal analyses to evalume the effects of (1 ) incal soft spots of soil foundation. (2) unt sgmngs to the fuushimum mai design with sum-uniform stiffnesses, and (3) s mi stiffness casresponding to other soil cakhthms used in the design Atkhtsonal analyses have been gerformed as desenbed in SSAR subsectma 3 8.5 4 2.

Action W - Subject was discuswd dunng meeting on July 11.1996 Westinghuse wdl mclude in the SS AR site interface cntena related to the haal vanatuhty of soil stiffness telow the foundation The allowable vanabiley will he mcluded in the design of the lusemat Sod varashdary has been addressed m the documentaion available for review in uretang in December. A draR SSAR resasson is being prepared NRC Status: Actum W - he SSAR 2.5 4 drah is incomplete. nuwe infurinatumi is needed on the geotechnical program A cross reference is needed m SSAR 3 8 5. (12/16/96) 772 NRRECGH 385-14 DSER4)1 On/HPC Action W Acthw W Westinghouse should comrmt in the SSAR to use coated reinforcing hars for the dedgn of the NI fn=wtarion

.The seistmc Category I stnaccures below gra** are psutected against fkushng by waerstops and a waierproofing system ne =merpnnfing syuem as provided by the introductum of a cementituuss crystallirse waserproofing additave to the nailed soil retention wall shoscrete or to the shotcrete apphed to the rtw k surface. T1us waierproofing system plus concrete cover to the reinforcement is consulered to prowkle adequate prutettum N RC Stasus:Actum W - Westinghouse needs to ensure a consistent waterproof syuem, specifically the interface treween the numbnal and the shutcrete sidewall. This information needs to te in SSAR 3 41.1.1 A csoss-reference is needed in S3AR 3 8 5 6 (12/16N6)

Page: 1 Total Records: 5 A.

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. 1 FAX to DINO SCALETTI i j February 20,1997 j r

CC: Sharon or Dino, please make copies for: Bill Huffman )

Ted Quay Don Lindgren Ron Vijuk Terry Schulz Mike Corletti Ed Cummins Bob Vijuk i Brian McIntyre OPEN ITEM #182 (M5.4.11-5) la my quest to make sure we have provided NRC with everything needed to prepare an FSER, I am researching open items from the smallest item number on. The relevant documentation related to Open Item #182 (M5.4.11-5) is attached. We provided the original fax of a markup on January 10, 1997 (over a month ago). We intend to include this SSAR change in Revision 11 to be issued next week. We believe that this information resolves the concerns of item #182. It seems a reasonable request that NRC acknowledge receipt of the information. We request that NRC provide a defmitive j action for Westinghouse or provide direction to change the status of this item. We recommend j

" Action N". Thank you. j l

I 1

Jim Winters 412-374-5290 i

/ I

i

AP600 0 pen Itene Tracking Systene Dath: Executive f -- y Date: 2/2697 i Selection: litem nol between 182 And 182 Sorted by item #

1 4

Isem Rese (W) NRC

, DSER Secton/ Tule/Descripske No Branch Type Detail Status Eagumeer Status Status letter No / Da_c __

4 Questka

$ 182 NRR/SPLB 5 4 Il MTG 4)I Corletti M Chud Actam W

• Mi411-5 (PRESSURIZER ' $RLIEF DISCHARGE) Secima 5 4 Il.3 states that the IRWST is sized based on the heat load and sacam volume f4diosing i an actuation of the ADS tAes this include sacam, water,and noncundensaNe gases from all thee ADS stages? Provide the analysis Ckud - See Section 6.3 for a dascussion of the IRWST dunas accuation of the automatic depressunzanon syssern DISCUSSED AT If25/QS MEETING

j. BETWEEN WESTINGlK)USE AND NRC itANT SYSTEMS BRANCH Acekm N - Need so resiew folkming Westinghanase prmidmg of specific SSAR seference Specific seference provided by fan markup of SSAR on Janumy 10,1997.

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FAX COVER SHEET

[DW) Westinghouse l

RECIPIENT INFORMATION ~ SENDER INFORMATION DATE

3ve/ (0, /99 7 NAME: y,,, gg,g TO: LOCATION: ENERGY CENTER -

b,Lt ku /~AvvW EAST PHONE: FACSIMILE: PHONE:

Office:c// 2- 17</-52 5c COMPANY: Facsimile: win: 284 4887 l

() S A[dC. '

outside: (412)374 4887 LOCATION:

a Cover + Pages 1+2 The following pages are being sent from the Westinghouse Energy Center, East Tower,

, Monroeville, PA. If any problems occur during this transmission, please call:

, WIN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

8, oc.

1;,g1 Sa on O SA7)s F/ 7He= kk(/2 scut 3Ycwl i2f2/9C r < l*btt[ E.) 0 pen b CP1 I97.Now nAr rpecof.c 51AL s u o re e n ao 11 b. 3, 2. 2, 3 b,-t aeg,u,,,(

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5. xcactor Coolant System and Connected Syst:ms e

lifetime (thermal, weight, and pressure) are applied, and stresses are compared to allowable values. Subsection 3.9.3 discusses the modeling and analysis methods.

5.4.10.4 Tests and Inspections Nondestmetive examinations are performed according to the procedures of the ASME Code,Section V, except as modified by the ASME Code,Section III, Subsection 57.

5.4.11 Pressurtur Relief Discharge i ne AP600 does not have a pressurizer relief discharge system. He AP600 has neither power operated pressurizer relief valves nor a pressurizer relief discharge tank. Some of the  ;

- functions provided by the pressurizer relief discharge system in previous nuclear power  ;

p! ants are provided by portions of other systems in the AP600.

l i

ne safety valves connected to the top of the pressurizer provide for overpressure protection of the reactor coolant system. First- second , and third-stage automatic depressurization system *.alves provide for depressurization of the reactor coolant system and venting of noncudensable gases in the pressurizer following an accident. Rese functions are  ;

discussed in subsections 5.2.2,5.4.12, and in Section 6.3. The AP600 does not have power 4 operated relief valves connected to the pressurizer. l ne discharge of the safety valves is directed through a rupture disk to containment '

atmosphere.

ne discharge of the first , second , and third-stage automatic depressurization system valves is directed to the in-containment refueling water storage tank. For the automatic depressurization system valves, the following discussion considers only the gas venting function. Only the first stage automatic depressurization valves are used to vent non-condensible gases following an accident. De sizing considerations and design basis for the in-containment refueling water storage tank for the depressurization function are discussed bd '"* W Section 6.3. He provisions to minimize the differential pressure between the contamment atmosphere and the interior of the in-containment refueling water storage tank are also discussed in S+ E3 s 6 r,shn 4 3, 2.

He safety valve on the normal residual heat removal system, which provides low temperature overpressure protection, discharges into the in<ontainment refueling water storage tank. See subsection 5.4.7 for a discussion of the connections to and location of the safety valve in the normal residual heat removal system.

5.4.11.1 Design Bases ne containment has the apability to absorb the pressure increase and heat load resulting from the discharge of the safety valves to containment atmosphere. He in-containment refueling water storage tank has the capability to absorb the pressure increase and heat load from the discharge, including the water seal, steam and gases, from a first stage automatic Revision: 5 y Westinghouse 5.4-63 February 29,1996 d"

9

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5. Reactor Cootnt System and Connected Systems

.- l I

depressurization system valve when ured to vent noncondensable gases from the press I

following an accident. De venting of noncondensable gases from the pressurizer foi ,

an accident is not a safety related function.

5.4.11.2 System Description Each safety valve discharge is dirceted to a rupture disk at the end of the discharge piping A small pipe is connected to the discharge piping to drain away condensed steam !eaking past the safety valve. He discharge is directed away from any safety related equipment, structures, or supports that could be damaged to the extent that emergency plant shutdown is prevented by such a discharge.

He discharge from each of two groups of automatic depressurization system valves is connected to a separate sparger below the water level in the in-containment refueling water storage tank. The piping and instmmentation diagram for the connection between the automatic depressurization system valves and the in-containment refueling water storage I

tank is shown in Figure 6.31. De in containment refueling water storage tank is a stainless steel lined compartment integrated into the containment interior structure. He discharge of water, steam, and gases from the first-stage automatic depressurization system valves when used to vent noncondensable gases does not result in pressure in excess of the in-containment refueling water storage tank design pressure. Additionally, vents on the top of the tank protect the tank from overpressureg4 d,ua.J ~ mh4 W 2.

I Overflow provisions prevent overfilling of the tank. He overflow is directed into the refueling cavity. De in-containment refueling water storage tank does not have a cover gas and does not require a connection to the waste gas processing system. De normal residual I

heat removal system provides nonsafety-related cooling of the in-containment refueling I water storage tank.

5.4.11.3 Safety Evaluation De design of the control for the reactor coolant system and the volume of the pressurizer is such that a discharge from the safety valves is not expected. De containment design pressure, which is based on loss of coolant accident considerations, is greatly in excess of the pressure that would result from the discharge of a pressurizer safety valve. De heat load resulting from a discharge of a pressurizer safety valve is considerably less than the capacity of the passive containment cooling system or the fan coolers. See Section 6.2.

Venting of noncondensable gases, including entrained steam and water from the loop seals in the lines to the automatic depressurizations system valves, from the pressurizer into spargers below the water line in the in-contamment refueling water storage tank does not result in a significant increase in the pressure or water temperature De in-containment refueling water storage tank is not susceptible to vacuum conditions resulting from the cooling of hot water in the tanku ne in-containment refueling water storage tank has capacity in excess of that required r venting of noncondensable gases from the pressurizer following an accident. e l m ,d>J m d ini b d M .

Revision: 5 February 29,1996 5.4-64 T Westinghouse

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- .. .. - . -.. . . - . - . . . - - . - - _ . - - ~ _ _ . . -

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I

• l FAX to DINO SCALETTI

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Febmary 20,1997  !

CC: Sharon or Dino, please make copies for: Bill Huffman )

Ted Quay  ;

Don Lindgren  !

Ron Vijuk Terry Schulz Mike Corletti Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEM #184 (M5.4.11-7)  ;

In my ques: to make sure we have provided NRC with everything needed to prepare an FSER, I am j researching open items from the smallest item number on. The relevant documentation related to Open Item #184 (MS.4.ll-7) is attached. We provided the original fax of a markup on January 13,  !

1997 (over a month ago). We intend to include this SSAR change in Revision 11 to be issued next l week. We believe that this information resolves the concerns of item #184. It seems a reasonable ,

request that NRC acknowledge receipt of the information. We request that NRC provide a definitive i action for Westinghouse or provide direction to change the status of this item. We recommend ,

" Action N" Thank you.

F l

s\.  ;

Jim Winters  ;

412-374-5290 i

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f l

I

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AP600 Open item Tracking Systeen Database: . Executive Sununary Date: 2/2Mr1 Selection: litem nol between 184 And 184 Sorted by item #

leem DSER Sectum/ Tale /IWupam RCSP IWI' NRC No- Hranch Queque Type Detm3 5tatus Engmeer Starus - Staus ' %m f gw 154 NRR/SPLM 54.I1 MTG4M. Corletti.M . Ched Action W M5 4 Il-7 (PRESSURIZER RELIEl DISCI 11 ARGE) Where is the instnnnentamm for the ADS valve dewitarge imes thwussed?

Closed - See Sectum 6.3 for instrumentatum requirements for the autosnatic depressimization systein DISCUSSED AT I/15N5 MEE11NG HEIVEEN WESTINGlK)USE AND NRC PLANT SYSTEMS HRANCll Actum N - Need to seview folloming Westmghouse provahng of speranc SSAR reference Actum W - Gise NRC euphcit. specific reference to the part of Section 6 3 the answers the questam.

Action N - FAX with inaLup of approprime changes to Chaper 5 prowled on 1/13N7.

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P. ige. I Total Records: 1

WlWestinghouse

_ FAX COVER SHEET

'o l RECIPIENT INFORMATION SENDER INFORMATION DATE: _ .3AauM4 13, /99 7 NAME:

TO:

LWm 1.OCATION: ENERGY CENTER -  !

Oeu. Murma EAST i

PHONE: FACSIMILE: PHONE:

office: t/I2 -5/V- 5 29o COMPANY: Facsimile: win: 284 4887 !

O S AJetc.

outside: (412)374 4887 LOCATION:

Cover + Pages 1+g The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:

WIN: 284 5125 (JanIce) or Outside: (412)374 5125.

COMMENTS:

8,u.

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". 5. Reactsr Coolant System and Connected Systems _ _ _ _ _ .

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Leakage from other flanges is discussed in subsection 5.2.5.3, Collection and Monitoring of Unidentified Leakage, #l

( J %&s 'Ocmum% J "g, yA ess *g# p..,4 o,W 5.2.5.1.3 Pressurizer Safety Relief Valves , j 4[,,

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Temperature is sensed downstream of each pres rizer safety relief valve by a resistance j temperature detector on the discharge piping t,ya..r.;r. Of dm ..p;u.e d;a. High temperature indications (alarms in the main control roem) identify a duction of coolant inventory as a result of seat leakage through : px....;- r af.;y va es se detectors are part of the reactor coolant system. 'Ihis leakage is rained to the reactor coolant drain tank during normal plant operation and vented to c tainment atmosphere uring accident conditions,9M mp=m ic dix. This identified le age is measured by th change in level of the reactor coolant drain tank.

0" E N

, 4 gs 3 ..fq 4 5.2.5.1.4 Reactor Coolant Pump Drain wd,sw.y 44. ,

Leakage from the reactor coolant pump drain is directed to the reactor coolant drain tank.

His identified leakage is measured by 1e change in level in the reactor coolant drain tank.

5.2.5.1.5 Other Leakage Sources In the course of plant operation, various minor leaks of the reactor coolant pressure boundary may be detected by operating personnel. If these leaks can be subsequently observed, quantified, and routed to the containment sump, this leakage will be considered identified leakage.

5.2.5.2 Intersystem Leakage Detection Substantial intersystem leakage from the reactor coolant pressure boundary to other systems is not expected. However, possible leakage points across passive barriers or valves and their detection methods are considered.. Auxiliary systems connected to the reactor coolant pressure boundary incorporate design and administrative provisions that limit leakage. Leakage is detected by increasing auxiliary system level, temperature, flow, or pressure, by lifting the relief valves or increasing the values of monitored radiation in the auxiliary system.

De normal residual heat removal system and the chemical and volume control system, which are connected to the reactor coolant system, have potential for leakage past closed valves. For additional information on the control of reactor coolant leakage into these systems, see subsections 5.4.7 and 9.3.6 and the intersystem LOCA discussion in subsection 1.9.5.1.

Revision: 10 December 20,1996 5.2 22 3 Westinghouse 4

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5. Reactor Coolant System and Connected Systems 1

5.4.11.4 Instnunentation Requirements

,5,2,53 He instrumentation for the safety valve discharge pipe containment, and in-containment refueling water storage tank are discussed in subsecti 5.4.9 and in Sections 6.2 and 6.3, respectively. Separate instrumentation for the monitoring of the discharge of noncondensable gases in not required.

5.4.11.5 Inspection and Testing Requirements Sections 6.2 and 6.3 discuss the requirements for inspection and testing of the containment and in-containment refueling water storage tank, including operational testing of the spargers. Separate testing is not required for the noncondensable gas venting function.

5.4.12 Reactor Coolant System High P.>lnt Vents ne requirements for high point vents are provided for the AP600 by the reactor vessel head vent valves and the automatic depressurization system valves. The primary function of the reactor vessel head vent is for use during plant startup to properly fill the reactor coolant system and vessel head.

Both reactor vessel head vent valves and the automatic depressurization system valves may be activated and controlled from the main control room.

He AP600 does not require use of a reactor vessel head vent to provide safety-related core cooling following a postulated accident.

He reactor vessel head vent valves (Figure 5.4-8) can remove noncondensable gases or steam from the reactor vessel head to mitigate a possible condition of inadequate core cooling or impaired natural circulation through the steam generators resulting from the accumulation of noncondensable gases in the reactor coolant system. De design of the i

reactor vessel head vent system is in accordance with the requirements of 10 CFR 50.34 (f)(2)(vi).

De first stage valves of the automatic depressurization synem are attached to the pressurizer and pmvide the capability of removing noncondensable gases from the I

' pressurizer steam space following an accident. Venting of noncondensable gases from the I

pressurizer steam space is not required to provide safety-related core cooling following a I

postulated accident. Gas accumulations are removed by remote manual operation of the first stage automatic depressurization system valves.

De discharge of the automatic depressurization system valves is directed to the in-containment refueling water storage tank. Subsection 5.4.6 and Section 6.3 discuss the automatic depressurization system valves and discharge system.

He passive residual heat removal heat exchanger piping and the core makeup tank inlet piping in the passive core cooling system include high point vents that provide the capability of removing noncondensable gases that could interfere with heat exchanger or core makeup tank operation. Dese gases are normally expected to accumulate when the Revision: 5 y Westingh0080 h(q '

5.4-65 February 29,1996

.