ML20151J608

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Summary of 970709 Meeting W/Westinghouse in Rockville,Md Re Discuss Design Certification Issues for AP600.List of Attendees & Handouts Encl
ML20151J608
Person / Time
Site: 05200003
Issue date: 07/31/1997
From: Joseph Sebrosky
NRC (Affiliation Not Assigned)
To:
NRC (Affiliation Not Assigned)
References
NUDOCS 9708050196
Download: ML20151J608 (53)
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Text

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8 k UNITED STATES

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j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 2066H001 July 31,1997 APPLICANT: Westinghouse Electric Corporation

-FACILITY: AP600 ,

SLBJECT:

SUMMARY

OF JULY 9, 1997, SENIOR MANAGEMENT MEETING WITH WESTINGHOUSE TO DISCUSS DESIGN CERTIFICATION ISSUES FOR THE AP600 The subject meeting was held on July 9,1997, in the Rockville, Maryland, office of the Nuclear Regulatory Commission (NRC) between representatives of i Westinghouse and the NRC staff. Attachment 1 is a list of meeting a uendees. I Attachments 2 and 3 are the handouts provided by the staff and Westinghouse i respectively during the meeting. j Prior to this meeting, a meeting was held between Westinghouse and the staff on July 8, 1997, to focus the discussion for the senior managers. The agenda of Attachment 2 was developed during the July 8, 1997 meeting.

The staff provided a high level status of the AP600 review (Attachment 2).

The staff stressed that the safety evaluation report dates were estimates and the staff was in the process of developing a detailed schedule for the i completion of the review.

The staff then discussed the status of the top 27 technical issues (Attach-  :

ment 2). Although the staff provided a handout that gave the status of all of these issues, only the following issues were discussed with the senior managers: regulatory treatment of non-safety related. systems (issue 2);

passive thermal-hydraulic performance reliability (issue 21); inspections, tests, analyses and acceptance criteria (issue 3); prevention and mitigation of severe accidents (issue 8); post-72 hour support actions (issue 9); WG0THIC code documentation and qualification (part of issue 17); and fire protection

-(issue 12).

For issues 2 and 21, Westinghouse had informed the staff in previous meetings that they were in the process of developing administrative controls for some systems. The staff believes that Westinghouse's approach could significantly simplify the review for both issues.

For issue 3, Westinghouse had recently submitted considerable information that the staff was in the process of evaluating. The staff committed to develop a schedule for the completion of the review in this area as part of the effort l to develop an overall schedule.

i For issue 8, Westinghouse committed to provide the staff with a schedule for t

responding to the Commission's staff requirements memcrandum of June 30, 1997.

Westinghouse also committed to provide the staff with a schedule for respond- -,); ;

. ing to the staff's position letter concerning issue 9.

l NRC FILE F r COPY 9708050196 970731 PDR ADOCK 05200003 h PDR l

July 31,1997 The WG0THIC code was then discussed. The staff stated that they may not be I able to determine if WG0THIC can adequately predict containment performance l because of undefined uncertainties and poor quality of reports. Westinghouse I replied that it did not agree with the staff's assessment and that it had a high degree of confidence in both WG0THIC, the containment design and that the quality of the more recent reports had been improved. Westinghouse stated  !

that it would put people in the Rockville office, if necessary, to expedite i the review of the code. Because of the continuing problems in this area, the staff stated that it would evaluate options to determine the most efficient 1 approach for performing the review of the code. l Westinghouse then presented the status of the fire protection issue (Attach-ment 3). The two issues involved with fire protection are: (1) the post-fire shutdown condition, and (2) the :;hutdown capability of the plant. Westing- l house stated that its legal interpretation of SECY-94-084 is that the accept-  !

ability of safe shutdown (420 'F or below) apply to fire protection systems  !

and that safe shutdown is also the safest mode for fire-related shutdowns for the AP600. For the second issue, shutdown capability of the plant, Westing-house believes that its passive safe shutdown systems are those necessary and ,

sufficient to comply with fire regulations and that alternate dedicated I systems are not necessary to comply.

The staff did not agree with Westinghouse's interpretation of SECY-94-084. I The staff agreed to review past SECYs and staff requirements memorandums l pertaining to evolutionary and advanced light-water reactor for guidance on fire protection issues. As part of the review, the staff agreed to assess the intent of the documents and applicability to the AP600. Westinghouse agreed to work with the staff to schedule meetings to identify resolution paths for the issues. Westinghouse also agreed to assess the possibility of providing the staff with an evaluation of vulnerabilities of the AP600 design at safe shutdown and at cold shutdown. A draft of this meeting summary was provided to Westinghouse to allow them the opportunity to comment on the summary prior to issuance. ,

l Original signed by Dino C. Scaletti FOR Joseph M. Sebrosky, Project Manager  !

Standardization Project Directorate l Division of Reactor Program Management Office of Nuclear Reactor Regulation ,

i Docket No.52-003 I Attachments: As stated I

cc w/atts: See next page  !

DISTRIBUTION
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DATE 07/ V 97 07/3t/97 l OfflCIAL RECORD COPY

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JMoore, 0-15 B18 JNWilson l SNewberry, 0-8 E2 l Glainas, 0-7 D26 CBerlinger, 0-8 H7 EThrom, 0-8 H7 BPalla, 0-8 H7 HWalker, 0-8 D1 KWest, 0-8 D1 Econnell, 0-8 D1 JLyons, 0-8 D1 Alevin, 0-8 E23 JFlack, 0-10:E4 l l HBrammer, 0-7 HIS

.TCheng, 0-7 H15 NDudley, 0-5 E6  !

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Westinghouse Electric Corporation Docket No.52-003 cc: Mr. Nicholas J. Liparulo, Manager Mr. Frank A. Ross Nuclear Safety and Regulatory Analysis U.S. Department of Energy, NE-42 Nuclear and Advanced Technolcgy Division Office of LWR Safety and Technology i Westinghouse Electric Corporation 19901 Germantown Road P.O. Box 355 Germantown, MD 20874 Pittsburgh, PA 15230  ;

Mr. Ronald Simard, Director Mr. B. A. McIntyre Advanced Reactor Program Advanced Plant Safety & Licensing Nuclear Energy Institute Westinghouse Electric Corporation 1776 Eye Street, N.W.

I Energy Systems Business Unit Suite 300 i Box 355 Washington, DC 20006-3706 Pittsburgh, PA 15230 Ms. Lynn Connor Ms. Cindy L. Haag Doc-Search Associates l Advanced Plant Safety & Licensing Post Office Box 34 i Westinghouse Electric Corporation Cabin John, MD 20818 Energy Systems Business Unit Box 355 Dr. Craig D. Sawyer, Manager Pittsburgh, PA 15230 Advanced Reactor Programs GE Nuclear Energy l Mr. M. D. Beaumont 175 Curtner Avenue, MC-754 ,

l Nuclear and Advanced Technology Division San Jose, CA 95125 l Westinghouse Electric Corporation l One Montrose Metro Mr. Robert H. Buchholz l 11921 Rockville Pike GE Nuclear Energy l Suite 350 175 Curtner Avenue, MC-781 l Rockville, MD 20852 San Jose, CA 95125 Mr. Sterling Franks Barton Z. Cowan, Esq.

! U.S. Department of Energy Eckert Seamans Cherin & Mellott NE-50 600 Grant Street 42nd floor l 19901 Germantown Road Pittsburgh, PA 15219

! Germantown, MD 20874 Mr. Ed Rodwell, Manager l Mr. S. M. Modro PWR Design Certification l

Nuclear Systems Analysis Technologies Electric Power Research Institute ,

l Lockheed Idaho Technologies Company 3412 Hillview Avenue i l Post Office Box 1625 Palo Alto, CA 94303 l Idaho Falls, ID 83415 Mr. Charles Thompson, Nuclear Engineer AP600 Certification NE-50 19901 Germantown Road Germantown, MD 20874 f

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WESTINGHOUSE AP600 SENIOR MANAGEMENT MEETING ATTENDEES JULY 9, 1997 1

! l fLA_ML ORGANIZATION HOWARD BRUSCHI WESTINGHOUSE >

l BOB VIJUK WESTINGHOUSE I ED CUMMINS WESTINGHOUSE JIM WINTERS WESTINGHOUSE I

BRIAN MCINTYRE WESTINGHOUSE .

BART C0WAN WESTINGHOUSE  !

CHARLES THOMPSON DOE  ;

R. PATRICK MCDONALD ARC '

ED RODWELL EPRI >

l DAVID STELLFOX MCGRAW HILL B0B MAIERS PENNSYLVANIA-BUREAU OF RADIA- l

, TION PROTECTION l SAM COLLINS NRR l TIM MARTIN NRR/ADT i SCOTT NEWBERRY NRR/DSSA  !

GUS LAINAS* NRR/DE i

CARL H. BERLINGER* NRR/DSSA/SCSB l EDWARD THROM* NRR/DSSA/SCSB i

B0B PALLA* NRR/DSSA/SCSB L.B. MARSH

  • NRR/DSSA/SPLB HAROLD WALKER
  • NRR/DSSA/SPLB K.S. WEST
  • NRR/DSSA/SPLB E.A. CONNELL* NRR/DSSA/SPLB l JIM LYONS* NRR/DSSA/SRXB l ALAN LEVIN
  • NRR/DSSA/SRXB l JOHN FLACK
  • NRR/DSSA/SPSB

, ]

l H.L. BRAMMER* NRR/DE/ECGB THOMAS CHENG* NRR/DE/ECGB NOEL DVDLEY NRR/0ED0 MARYLEE SLOSSON NRR/DRPM l SEYMOUR H. WEISS NRR/DRPM l TED QUAY NRR/DRPM/PDST l JERRY WILSON NRR/DRPM/PDST i TOM KENYON NRR/DRPM/PDST BILL HUFFMAN NRR/DRPM/PDST r DIANE JACKSON NRR/DRPM/PDST j J0E SEBROSKY NRR/DRPM/PDST

  • PART TIME j 4

Attachment 1

} Attachment 1 l

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NRC HANDOUTS PROVIDED DURING JULY 9,1997, SENIOR MANAGEMENT MEETING l

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l Attachment 2

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l l AP600 SENIOR MANAGEMENT MEETING July 9,1997 I. Introduction NRC/W II. Status of NRC review NRC III. Status of Top 27 Technical Issues NRC/W IV. Fire Protection W V. Conclusion NRC/W J

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. Estimated 2

Chotr Title SER Input' Comments 1 Introduction and General 9/97 Level of Detail (KI #1) & Post-72 Discussion Hr Actions (KI #9) issues open.

2 Site Envelope Characteristics 8/97 Soil / Structure / Seismic Interaction issue (Kl #5) open. Most SERs i' complete, with open items.

3 Design of Structures, 8/97 Basemat issue (KI #6) open.

Components, Equipment, Final audits - 8/97.

and Systems Most SERs complete, with open items.

4 Reactor 6/97 SER complete, no open items 5 Reactor Coolant System and 8/97 Shutdown / Low Power (KI #22) &

Connected Systems Operations Containment Bypass (Kl #24) issues open. Some SERs complete, with open items.

' Based on timely receipt of Westinghouse's final submittals. Dates subject to change based on reviewer availability.

2 Unresolved key technical issues (KI) listed by number established in 12/96 letter.

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. Estimated . '

Chotr Title SER input' Comments2 1

6 Engineered Safety Features 9/97 Mitigation of SA (KI #8) & H 2  !

Mitigation Systems (Kl #11) issues open. Some SERs  ;

complete, with open items '

7 Instrumentation and Controls 9/97 SER complete, with open items 8 Electric Power Systems 8/97 SER complete, with open items 9 Auxiliary Systems 10/97 Fire Protection (KI #12) & SFP Cooling Sys (KI #13) issues open.

10 Steam and Power Conversion 10/97 Overspeed Protection issue .

System (KI #14) open.

11 Radioactive Waste Management 8/97 i

12 Radiation Protection 7/97

' Based on timely receipt of Westinghouse's final submittals. Dates subject to change based on reviewer availability. i

'Umesolved key technical issues (KI) listed by number established in 12/96 letter.

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. Estimated  ;

2 Chotr .Titig SER inout' Comments 13 Conduct of Operations 9/97 Safeguards issue (KI #15) open. I 14 initial Test Program 9/97 ITP issues (Kl #16) open. i ITAAC TBD ITAAC issue (KI #3) open.

i 15 Accident Analyses 9/97 DBA Radiological Consequences  !

(KI #7) & Accident Analyses (KI

  1. 18) issues open.

1 16 Technical Specifications 10/97 Tech Spec issues (KI #26) open.

17 Quality Assurance 10/97 QA inspection ~9/97 18 Human Factors Engineering 8/97 i

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' Based on timely receipt of Westinghouse's final submittals. Dates subject to change based on reviewer f availability.

  • 2 Unresolved 'wy technical issues (KI) listed by number established in 12/96 letter.

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i Estimated Chotr Title SER Input' . Comments 2  ;

19 Severe Accidents & PRA -

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- Level 1 PRA 7/97 RTNSS (KI #2), Integrated PRA Insights (Kl #20), T-H Uncertainty (KI #21), Adverse Systems Interaction (KI #25) issues open.

- Level 2 & 3 PRA 9/97 In-Vessel Retention issue (KI #23) open.

- Shutdown PRA TBD 20 Generic issues 10/97 Some SERs complete,  :

with open items 21 Testing & Code Development Code Documentation & Qual.

issues (KI #17) open.

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- LOFTRAN 7/97 SER complete, with open item

- NOTRUMP 8/97 I

- WCOBRA/ TRAC LBLOCA 9/97

- WCOBRA/ TRAC LTC 7/97

., - WGOTHIC 11/97

' Based on timely receipt of Westinghouse's final submittals. Dates subject to change based on reviewer availability.

2 Unresolved key technical issues (KI) listed by number established in 12/96 letter.  ;

Key Technical Issues - AP600 -

1. Content of the SSAR - Level of Detail and Adequacy of Figures
2. Regulatory Treatment of Non-Safety Related Systems (RTNSS)
3. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) ,
4. Leak-Before-Break Design Criteria For FW Piping System
5. Soil / Structure / Seismic interactions
6. Site-Soil Variability (Basemat)
7. DBA Radiological Consequences  ;
8. Prevention and Mitigation of Severe Accidents ,
9. Post-72 Hour Support Actions
10. Containment Isolation
11. Systems Reliability of Hydrogen Mitigation Systems
12. Fire Protection Program
13. Spent Fuel Pool Cooling System ,
14. Overspeed Protection
15. Proposed AP600 Security Plan
16. Initial Test Program t
17. Code Documentation and Qualification (V&V of Codes)
18. Chapter 15 Accident Analysis  ;
19. Westinghouse *s Proposed LCO 3.0.3
20. Integrated Use of PRA Insights
21. Passive System Thermal-Hydraulic Performance Reliability
22. Shutdown and Low Power Operations
23. Extemal Cooling of the Reactor Pressure Vessel / Severe Accidents
24. Containment Bypass /SGTR
25. Adverse Systems interactions
26. Technical Specifications Review
27. Quality Classification of Systems

' identified in 12/6/96 letter.

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1. Content of the SSAR - Level of Detail and Adequacy of Figures
  • Westinghouse provided a written response and updated SSAR in Revisions 13 and -l 14.
  • Westinghouse to submit additional revision based on telephone conference callin June in Revision 15.
  • Staff is reviewing current changes and draft markup revisions for Revision 15.

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2. Regulatory Treatment of Non-Safety Related Systems (RTNSS)~
  • Schedule impact issue The RTNSS process, defined in SECY-94-084, involves a two step process which involves, (1) identification of systems subject to RTNSS and, (2) defining the appropriate regulatory oversight for the RTNSS identified systems.

The remaining effort in these areas has a potential for schedular impact and will need continued management attention .

Passive System Thermal-Hydraulic Performance Reliability (Key issue 21)

The status of this review is discussed in the follow-on slide.

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2. Regulatory Treatment of Non-Safety Related Systems (RTNSS) (cont.)

General RTNSS Status -

Based on meetings between Westinghouse and NRC on 4/3/97 and 5/6/97, the staff has been informed that Westinghouse is developing administrative availability controls .

t on DAS, the standby diesel-generators, RNS and some additional systems (such as  !

hydrogen igniters and post-72 hour equipment). The staff issued a letter on 6/9/97 providing conditions under which it would find such an approach acceptable.

i The staff believes that such an approach could significantly simplify the review in this area and is waiting for Westinghouse's submittal.

3. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) e Schedule impact issue i

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  • Westinghouse has recently submitted the following documents:
- Revision 3 to the Certified Design Material (5/19/97)

- Response to the majority of the RAls (6/16/97)

- Revision 13 to the SSAR which supports changes to thi CDM (6/13/97)

- A table that cross-references important design parameters to their treatment in Tier 1 (6/20/97) 2

3. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) (cont.)
  • Major Changes to the Certified Design Material and SSAR i

- Additional systems added to address staff's concern

- Piping ITAAC added that is significantly different than the evolutionary plants' '

i - Radiation Protection approach that is significantly different than the evolutionary i i plants'  !

Changes to the SSAR made to support the ITAAC including a revision to SSAR .

Chapter 14.3 and changes made to individual SSAR. chapters to support numbers given in the ITAAC  :

  • An internal meeting with'the task group leaders was held on 7/2/97 to discuss the ITAAC. The staff is in the process of developing a review schedule. There maybe some conflicts with some groups between writing the draft SERs for their respective areas and reviewing the ITAAC at the same time.

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4. Leak-Before-Break Design Criteria For FW Piping System ,
  • This issue is resolved.
5. Soil / Structure / Seismic Interactions
  • Westinghouse revising RAI responses as discussed in a telephone conference call in June.
  • Westinghouse completed its re-analysis for seismic due to post 72-hour actions.

Audit of calculations expected in July or August.

6. Site-Soil Variability (Basemat)
  • The staff is reviewing Westinghouse's proposal for construction sequence and geotechnica! investigation provided by Westinghouse.
  • The resolution of the basemat is linked to the shallow soil site issue, 4  ;
7. DBA Radiological Consequences e The staff is awaiting RAI responses on aerosol removal in containment-(10/96) and i- EQ in containment (8/96).
  • Westinghouse submitted information on 4/17/97 on Westinghouse input assumptions for calculation of aerosol removal coefficients in containment. .

e The staff informed Westinghouse that use of NUREG/CR-5055 dispersion methodology was unacceptable and recommended that Westinghouse use the newly-developed ARCON96 code during a June 1997 meeting. Westinghouse is evaluating its options.

8. Prevention and Mitigation of Severe Accidents 1
  • Potential schedule impact issue
  • Westinghouse needs to develop proposal to address June 30,1997 SRM.

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9. Post-72 Hour Support Actions
  • -Schedule impact issue The passive safety' systems are designed with sufficient capability to mitigate all design basis events for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without operator actions and without non-safety-related onsite or offsite power. For long-term safety (post-72 hours), the AP600 design includes safety-related connections for use with transportable equipment and supplies to provide the extended support actions for safety-related functions.

In SECY-96-128, the staff stated that local communities struggling with disaster response should not be given the additional burden of providing for nuclear power safety. The staff recommended the Commission approve the position that the site be capable of sustaining all design basis events with onsite equipment and supplies for the long term. After 7 days, replenishment of consumables such as diesel fuel oil from offsite suppliers can be credited. On 1/15/97, the Commission issued an SRM approving the staff's position.

Status

  • Westinghouse has revised the AP600 SSAR to include the Post-72 hour design changes and completed most of the supporting analyses (seismic assessment and control room dose analyses).
  • The staff is still reviewing the Post-72 hour design changes and is reviewing recent supporting analyses provided by Westinghouse. The staff has issued several comment letters and continues to assess the Westinghouse responses.

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9. Post-72 Hour Support Actions (cont.)

Issue

  • Westinghouse has indicated that the Post-72 hour equipment will be analyzed to remain functional following safe shutdown earthquake (SSE) loads but does not need to be classified as Seismic Category I per GDC-2. In addition, Westinghouse does not believe that the Post-72 hour equipment needs to be designed to withstand tornado loadings or missiles per GDC-2.

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i e The staff has developed a position letter on the seismic, tornado wind, and missile analyses criteria for Post-72 hour equipment. This letter (just issued on 7/7/97) ,

i states that the Post-72 hour equipment should be analyzed using the same methods as used for Seismic Category 11 SSCs. The Post-72 hour equipment should also be

designed to withstand severe Category 5 hurricanes (wind speeds of greater than 155 mph) including the effects of sustained winds, maximum wind gusts, and associated wind-borne missiles.
  • Other Post-72 hour related issues still under review by the staff:

- Seismic analyses of the Post-72 hour design changes to the PCCS tank.

- Main Control Room Dose Rate Calculational Methodology.

- Acceptability of Spent Fuel Pool boiling as the safety related way of heat removal.  !

- Acceptability of the Main Control Room Habitability System Human Factors Environment.

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10. Containment isolation
  • This issue is technically resolved based on Westinghouse letter dated 4/22/97
11. Systems Reliability of Hydrogen Mitigation Systems
  • A meeting between Westinghouse and the staff was held on 5/20/97 to review .l remaining open items on use of PARS for AP600 DBA hydrogen control.

Westinghouse has tne following remaining areas to address for the use of PARS: j

- Demonstrating a well mixed environment  !

- Concerns about mixing above 135 foot elevation j

- Concerns about mixing below 135 foot elevation i l

- Environmental Qualifications

- Technical Specifications

, - Battelle Testing

- Compliance with regulations

- Debris clogging of PARS l l

- Location of PARS relative to pipe breaks  ;

A e Westinghouse is preparing a major revision to SSAR section 6.2.4 to address the issues above. Westinghouse stated that a markup of the revision should be sent to i the staff by week of 7/7/97.

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11. Systems Reliability of Hydrogen Mitigation Systems (cont.)
  • A position letter on technical specifications controls for hydrogen igniters was issued ,

on 4/3/97 which Westinghouse responded to in a 6/24/97 letter. Westinghouse does  ;

not believe hydrogen igniters meet the inclusion criteria for technical specifications, however, Westinghouse did propose short term availability controls. These controls are similar to what the staff believes Westinghouse will propose for RTNSS controls.

The staff is currently evaluating Westinghouse's position.

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i i 12. Fire Protection Program i

  • Potential schedule impact issue
  • Major issues:

- Post-fire endstate .

- Shutdown capability.. i

  • Except for two major issues, Westinghouse preparing responses and design changes  ;

in response to staff position papers on issued in May and June.  ;

13. Spent Fuel Pool Cooling System
  • Staff review in progress. Final acceptability of the SFP is linked to resolution of dose calculations for control room habitability. <

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'j4. Overspeed Protection e Westinghouse informed the staff that they will revise its design with both an electrical trip and mechanical trip. This is expected in mid-July and SSAR Revision 15.

15. Proposed AP600 Security Plan f e The revised Security Report was received on 2/28/97.

e The staff identified key concerns with Westinghouse's proposal during a May 13, 1997 meeting.

e A meeting /telecon will be held in mid-July to discuss Westinghouse's proposed resolutions to the issues.

16. Initial Test Program e Westinghouse responded to the majority of the staff's comments on 5/9/97. Staff's response provided in 6/25/97 letter e Approximately 25 issues remain open. While continued d.alogue should be able to resolve most of these is. sues, there are some issues (e.g., first plant only testing) that-may require senior management involvement.

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17. Code Documentation and Qualification (V&V of Codes) ,

NOTRUMP

  • Westinghouse submitted remaining outstanding RAI responses on the NOTRUMP Final Validation Report on 6/17/97.
  • Staff received final revision of NOTRUMP validation report on 7/7/97.

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  • Staff should have draft SER completed in August.

LOFTRAN  !

  • Most open items for LOFTRAN are resolved. l
  • On 5/9/97, the staff provided Westinghouse with PRHR heat transfer data from j several ROSA tests for use in making (blind) predictions of ROSA test data using the LOFTRAN heat transfer correlation. The staff is waiting for the results of the ROSA  :

data analyses to complete its LOFTRAN assessment.

  • A draft of the LOFTRAN SER has been completed. i WCOBRA/ TRAC - LBLOCA
  • Westinghouse has replied to all the staff's RAls on this ap31ication of the code.
  • The staff is reviewing the RAI responses.

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17. Code Documentation and Qualification (V&V of Codes) (cont.)

WCOBRA/ TRAC - Lona Term Coolina  ;

  • Westinghouse has replied to all the staff's RAls .on this application of the code. -

, o The staff is completing its technical review of the applicatiori of WCOBRA/ TRAC to AP600 long term cooling analyses.

WGOTHlC e Schedule impact issue e Westinghouse submitted new information on May 23,1997. Two-dimensional model  :

used in addition to WGOTHIC is needed to remain below one-half of design pressure  ;

e.fter 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.  !

e Based on reports today (all but one " final"), the staff may not be able to determine if WGOTHIC can adequately predict containment performance because of undefined uncertainties and poor quality of the reports. The small margin forces the staff to define uncertainties that were Previously unimportant but may be an impact of this code.

e The WGOTHIC review is behind schedule, based upon the November 1996 schedule.

Submittals are approximately 3-4 months late. Draft SER expected in November (four months later than Nov. 96 schedule).

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18. Chapter 15 Accident Analysis e Westinghouse completed all Chapter 15 documentation with revision 13 of SSAR
  • which was received by the staff on 6/13/97 (except Revision 1.of WCAP-14601 on i accident evaluation models which is due by 7/14/97). ,

o The staff is in the process of reviewing this material.

e The staff has issued some additional RAls based on its review of the revised Chapter 15 analyses to date.

19. Westinghouse's Proposed LCO 3.0.3 e Westinghouse has agreed to incorporate the standard LCO 3.0.3 wording into the AP600 technical specifications.

e This issue is technically resolved.

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20. Integrated Use of PRA Insights .

I e Westinghouse must use insights from the sensitivity, uncertainty, and importance analyses in an integrated fashion, in conjunction with assumptions from the entire PRA, to identify design certification and operational requirements (such as ITAAC, RAP, TSs, administrative controls, procedures) as well as COL and interface require-ments.

Status  !

e The staff is in the process of writing the draft SER for the majority of the level 1 PRA. Any open items from this SER will be forwarded to Westinghouse for resolutio.n in accordance with current staff practice.

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  • There is a potential schedule impact concerning the shutdown portion of the PRA, and the draft SER for this portion of the PRA maybe delayed.

e Potential changes to the level 1 insights due to the staff's review will also be forwarded to Westinghouse for resolution.

e Staff is reviewing RAI responses on SMA. Final audit needed.

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21. Passive System Thermal-Hydraulic Performance Reliability
  • Schedule impact issue Westinghouse has stated that the AP600 can respond in an acceptable manner to risk-significant PRA accident sequences by using only passive safety systems, and that, as a result, no regulatory oversight of active, non-safety-related systems is required. To 1 support this statement, Westinghouse has used a DBA ana:ysis code (NOTRUMP) to perform sensitivity studies on risk-significant, low margin, accident sequences from the .

focused PRA. The sensitivity studies use conservative, bounding inputs and assumptions, and demonstrate that there are large margins to core damage. The analyzed sequences have been selected using the PRA thermal-hydraulic computer code (MAAP4) to " screen" sequences from the focused PRA. The margins approach is undertaken in lieu of attempting to quantify thermal-hydraulic uncertainties in the PRA, related to passive system performance.

i Status

. t e The MAAP4 benchmarking report was received April 18,1997.

  • T-H uncertamty report was received by the staff on 6/24/97.
  • Staff review in progress. Issues may be contingent on the administrative availability t controls Westinghouse is planning to propose to the staff.

t j 15

e

! 22. Shutdown and Low Power Operations j i

  • Westinghouse submitted a major portion to the shutdown evaluation report on 4/18/97.
  • On 6/6/97, Westinghouse submitted the remaining informatiun related to the shutdown evaluation report.
  • The staff is prepanng RAls for submittal to Westinghouse.
i
23. External Cooling of the Reactor Pressure Vessel / Severe Accidents  ;
  • The status of the reports associated with the IVR issue follows:-

IVR main report:

Westinghouse has responded to the majority of the RAls in this area. The staff review confirms the effectiveness of external reactor vessel cooling for the " final bounding state." The staff continues to have some concerns with the IVR report including:

- " stratified intermediate states" that could pose a greater threat to the reactor pressure vessel (RPV) than the " final bounding state" evaluated in the report

- damage to the structural integrity of the RPV therma insulation (by hydrodynamic loads) that could degrade ex-vessel heat removal capability 16 .

-..~.i

23. External Cooling of the Reactor Pressure Vessel / Severe Accidents (cont.) .

1 IVR main report (cont.):

Residual concerns / uncertainties to be addressed by reliance on results of ex-vessel calculations.

Reports on IVSE: ,

The staff is evaluating Westinghouse's 6/13/97 submittal that provided responses to the staff's concerns, a revision to the reports, and resolution of the peer review i comments. ,

Exvessel Phenomena:

Westinghouse is in the process of responding to staff RAls concerning core concrete interaction. The staff is evaluating Westinghouse's recent submittals and RAI responses concerning ex-vessel steam explosions.

24. Containment Bypass /SGTR
  • Westinghouse submitted a revised SGTR analysis on 3/24/37.
  • Staff resolution of this issue is based on satisfactory assessment of the NOTRUMP -

MAAP benchmarking report which is still under review.  ;

17

_________.____._.__._____.___._________._________________________m __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

25. Adverse Systems Interactions e Westinghouse issued final ASI report to incorporate staff comments on 5/9/97 t

e Issues related to the focused PRA evaluation (e.g., SGTR end RCP common mode failure) are still under review by the staff.

i

26. Technical Specifications Review e Position letter on optimized technical specifications issued on 3/27/97.  !

e Westinghouse letter dated 6/6/97 provided Westinghouse response to staff position -

letter. The staff has reviewed this response and has additional comments which are being prepared for issuance to Westinghouse.  ;

i e Numerous additional technical branch issues are still in the process of being resolved between the staff and Westinghouse.  !

I i

27 Quality Classification of Systems-i e Pending SSAR (Revision 14) changes, this issue is resolved. .

18 i

. ~r Status of Top 27 issues i

Staff and Westinghouse Action Required: 9 issues _

Staff Action Required: 6 issues i

Westinghouse Action Required: 4 Issues Technically Resolved: 4 issues

Schedule Impe ~' !ssues
5 issues 4

i I

f 19

- - . _ _ _ . . - . . - . . _ . - - _ - - - - .-.-..__--.___.--.---..,_.a.-_--_-- --.---.-.-_--___--.__u- - - - - _ _ - - - - _ - _ . - _ _ - _ - - - - _ _ _ _ _ - - - - - - - - . _ _ . _ - - - - - _ _ - _ _ _ _ - - - _ - - . - _ - _ - -

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WESTINGHOUSE HANDOUTS PROVIDED

!, DURING JULY 9,1997, SENIOR i

MANAGEMENT MEETING l

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Attachment 3

Safe Shutdown Following a Fire .

k NRC Senior Management Meeting July 9,1997 J. W. Winters t

a

_ . . _ _ _ _ _ _ . _ . _ _ _ _ . ____.m____.____.____. _ . _ _____ _ _ ___ _ _ _ _ ._ _ _ _ _ __ _

i Outline Westinghouse position

" Legal" basis I

Technical basis for using " safe" shutdown AP600 is different ,

- Safe shutdown is safest plant mode after a fire Challenges to achieving " cold" shutdown following a fire Conclusion Recommendation i 2em44r ppa 7597 i

.  ?

Westinghouse Position k

AP600 can use " safe" shutdown rather than " cold" l shutdown for compliance with fire protection i regulations.

Using " safe" shutdown, alternate / dedicated shutdown

-system rules need not apply to AP600.

t

f I

k

.Mm44H ppt 7347 f

t Definitions .

" Safe" shutdown for AP600 is safe, stable, suberitical, J

passive, long term operation with reactor coolant temperatures less than 420 oF.  !

" Cold" shutdown for AP600 is safe, stable, subcritical, active, long-term operation with reactor coolant temperature less than 200 oF. t

" Cold" shutdown for active plants is the same as for AP600.

AP600 can achieve long term safe shutdown with safety related systems.

Active plants cannot maintain " hot" or " warm" shutdown long term with safety related systems.

3.~,, ,.., 1

\

~

AP600 Safe Shutdown AP600 provides improved safe shutdown for fires.

. More reliable safety-related systems i Simplified passive features

- One time alignment of valves, mostly fail safe

- None/few support systems required

- Diversit9 in safety systems ,

Full pressure RHR, no switching of systems

No operator actions, no control actions l

. Simplified fire hazard analysis contributes to safety

. Safe shutdown conditions prevent subsequent accidents Operates at low percentage of design pressure Reactor coolant not circulated outside containment 3.m m1 t

" Legal" Basis i

SECY-94-084 said that " safe" shutdown could be used  ;

rather than " cold" shutdown for passive plants. It used Appendix R as an example of why this position was required.

NRC staff in a July 24,1996 letter endorsed the details of .

this position.

- Using the " safe shutdown" approach, AP600 meets the conditional logic of Appendix R so that alternate /

dedicated systems are not required. This approach is consistent with ABWR.

. , . . ~ ,

i 1

Technical Basis for Using " Safe" Shutdown 1 AP600 is different.

NRC agrees that " safe" shutdown is a safe mode. .

" Safe" shutdown is a safe mode after any fire.

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_ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.___m_ .- _ __ _

. .i l

AP600 is Different Overall design and licensing approach is to achieve simple,  ;

passive safety.

" Safe" shutdown as an end state for safety analysis was considered.by NRC as necessary consequence of passive design. j We believe that time and design effort have made " safe"  !

shutdown the safest plant mode.  ;

i i

t be4 alt ppe ?H7 I

Safe Shutdown Mode Plant is safe, suberitical, and stable.

Reactor coolant is < 420 oF and < 380 psi in about 25 l hours.

Cooling is provided by safety related equipment. ,

i Reactor coolant boundary is intact and as it was before fire.

Reactor coolant boundary is entirely within containment.

Containment is closed.

i i

Conditions are indefinite.

.,..~,..,

Safe Shutdown, RCS Temp.

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Passive Systems Characteristics

- Use of passive processes One time alignment of valves Most valves are fail safe. j No continuously operating components

- No need for maintenance during safe shutdown.

l

- Essentially no support systems required.

Only need de power to actuate few fail-as-is valves Capable of long term operation j.

- Automatic, full RCS pressure operation  :

No need to switch systems, greater pressure margin.

- Passive systems have internal diversity For more probable events 3maaN ppe 7597 i

_ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ . _ _.____________________m_ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ . _ -- - - - - - . - - - - - -- .- ..-. .,._.. .

4 Passive Systems Characteristics

- AP600 automatically achieves safe shutdown.

Decay heat

- PRHR HX actuates on low SG level '

  • Transfers decay heat from RCS to containment.

- PCS actuates on high containment pressure (< 8 psig)

  • Transfers heat from containment to atm.

RCS makeup /boration~

- CMTs actuate on low RCS temp, Pzr pres, or Pzr level.

  • Recirculation provides makeup and boration.

All these valves are fail-safe.

  • PRHR HX is located inside containment Reduces chance of RC leak outside containment

_ _ , _ _ _ _ ___ _ _ _ _ _ __ __ _ _ _ _ _ _ _______ _ _______l

Simplified AP600 Decay Heat Removal

. t i

Standard 2-Loop AP600 Ultimate Heat Sink Ultimate Heat Sink

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Safe shutdown is safest plant mode

- AP600 safe shutdown conditions Reduces RCS pressure to < 12% of design

- Current plants operate at > 20% of RHRS design '

Chance of subsequent pressure boundary failure insignificant Passive systems can mitigate pipe failure

- AP600 safe shutdown systems are more reliable.

Simple passive features  !

Full pressure PRHR actuates automatically '

No switching of systems Multiple failure capability / diversity Essentially no support. systems required.

l

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f

. ~

Challenges to Achieve " Cold" Shutdown Following a Fire Additional sprinkler coverage

Physical separation of system heat exchange equipment Physical separation of I&C equipmen~t and cable Physical separation of power cable and MCC i

Physical separation of power feeder busses Required versus optional operator actions following fire Physical separation of non-safety plant monitor bus Additional containment penetrations Additional reactor coolant boundary penetr'ations  ;

Jemeen pps ?r87

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F Review of current arrangement of cold shutdown equipment I

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

Conclusion NRC allowed use of " safe" shutdown.

NRC provided no direction to the contrary for fire .

protection.

r NRC has identified no new issues to question the safety of ,

AP600 after a fire.

AP600 is based on a passive, safety related design and licensing approach.

" Safe" shutdown is an acceptable basis for implementation  :

of fire protection regulations.

.ww ,,. . mr D

_ . _ _ _ _ _ _ _ . . _ _____..___._..__-_______._____...___-...._..-_____-_____.____-______._._____m_

. _ _ _ . _ _ . _ = _ _ _ _ ____a ___.__ _ _ ___ -__w+ ._ r _ -m-- w i-_m. - -

Recommendations l

Westinghouse recommends the following criteria be I applied to AP600:

l " Safe" shutdown can be used rather than cold shutdown for design basis accidents per SECY-94-084.

" Safe" shutdown can be used as a basis for compliance .

with fire protection regulation. t In addition, where cold shutdown capability is jeopardized by a fire outside containment in a single fire area, fixed fire suppression shall be provided to protect equipment necessary to achieve cold shutdown.

______ .. _.=___.__. ._._.___._.- __. __ _ __.____._.___________ __ _ __.. _____ _ ____ _ _ __ _ _ ___

._ _ __

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