ML20210Q756

Summary of 970808 Meeting W/Westinghouse Electric Corp in Monroeville,Pa Re Fire Protection Issues for AP600 Design. List of Attendees,Agenda & Slides Presented by Westinghouse Encl

ML20210Q756
Person / Time
Site:	05200003
Issue date:	08/22/1997
From:	Kenyon T NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9709020132
Download: ML20210Q756 (45)

Text

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I p'f " % h. g_gg UMTED STATES en S NUCLEAR REGULATORY COMMISSION

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f WASHINGTON, D.C. 30446 4 001

.h# August 22, 1997 APPLICANT: Westinghouse Electric Corporation FACILITY: AP600

SUMMARY

OF AUGUST 8, 1997, MEETING TO DISCUSS FIRE PROTECTION ISSUES ON THE AP600 On August 8,1997, representatives of the U.S. Nuclear Regulatory Commission (NRC) and Westinghouse met at the Westinghouse offices in Monroeville, Pennsylvania,-to discuss fire protection issues for the AP600 design.

Attachment 1 is a list of attendees. Attachment 2 are-the-agenda and slides presented by Westinghouse.

Westinghouse first described the layout of the AP600, using the 3D electronic presentation package and the model of the plant. Westinghouse then made presentations on smoke control, loss-of-offsite-power (LOOP event, fire probabilistic risk assessment (PRA), and risk significance o)f shutdown modes, as follows: "

Smoke Control The participants discussed a LOOP event coupled with a fire in the main control room. The staff expressed its concerns that the stairwells need to be pressurized for smoke control to allow ingress and egress-to the fire area to extinguish the fire and to prevent the spread of smoke from one fire area to another. Westinghouse indicated in their submittals that they comply with the philosophy of the NFPA standards; however, the staff expressed its concerns with technical compliance of the AP600 with NFPA 92A and NFPA 204M. Westing--

house agreed to reevaluate compliance of the AP600 with the NFPA standards.

The results of this evaluation will be provided in a revised response to the June 3, 1997, NRC letter..

Fire PRA Westinghouse is assessing the fire PRA [which was based on Revision 1 of the standard safety analysis report (SSAR ] to determine whether the fire PRA is (D)

- still bounded conservatively as a resu)lt of the changes to the design (through Revision 14). Westinghouse stated that its preliminary evaluation shows that l no PRA revision is necessary as a result of design changes. Westinghouse gi  ;

committed to include in the PRA a statement that the design changes made since the fire PRA was developed do not significantly affect the results.

Safe Shutdown Vulnerability Assessment Westinghouse then discussed its plans to perform a qualitative analysis to address the relative risk between keeping the AP600 at safe versus cold shutdown. The vendor intends to evaluate the most probable challenges to NBC FILE CENTER CDPV 9709020132 970822

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I August 22, 1997

. long-term decay heat removal in both -states. . Westinghouse indicated that it ,

will consider spurious events that would attack the integrity of the safe or cold shutdown mode. The staff indicated that Westinghouse should look at mission times in terms of the technical specifications.

The participants then discussed Westinghouse's analysis of _a LOOP event coupled with a fire. Westinghouse stated that the results of their realistic analysis show that the pressurizer safety valve would not lift during such an event. Westinghouse stated that they believed that their evaluation of the vulnerability of the AP600 to fires should be based on realistic assumptions because such evaluations assume multiple failures and the use of non-safety-related systems to mitigate the consequences of a fire. The staff stated that they would consider the appropriateness of using such fire analysis, and agreed to get back to the vendor with its conclusions.

Independent of the fire analysis, the staff expressed its concern with the constant opening and closing of the pressurizer. relief valves for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> during a design basis accident (Chapter 15) LOOP event. Westinghouse stated that, although they may not be leak tight, the pressurizer relief valves will

-close when the event ends.

Other Methods to Achieve Cold Shutdown Westinghouse then discussed some conceptual design changes to the AP600 that i would make the normal residual heat removal (RHR) system more reliable in the event:of_a' fire. Westinghouse considered the existing interfaces to the reactor _ coolant system, and concludea enhanced reliability of the normal RHR system was:the best way to improve the " fire survivability" of the path to cold shutdown, When asked about installing a dedicated shutdown system, Westinghouse stated that such a system would introduce significant concerns (additional containment penetrations and high/ low pressure interface con-cerns)',:and was not an attractive option.- Westinghouse stated that they were also considering other options discussed at previous meetings,-such as incorporating additional sprinklers ~in the turbine building:to protect non-

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safety-related (defense-in-depth) shutdown equipment. In response to a question-from Westinghouse, the staff stated that the NRC_ fire protection

_ guidance do not require a plant to go to cold ~ shutdown conditions after a fire; but they require a plant.to have the capability to go there.

The. staff acknowledged-that none of the proposed changes to the AP600 dis-cussed at this meeting would guarantee the ability to go to cold shutdown; however, these change would enhance such capability for the-AP600 to go to _

l

_ cold shutdown,:and is, therefore, a step in the right direction to addressing the-staff.'s concerns.

Review of Past Documentation-The staff briefly discussed the preliminary results of its ongoing review of previous documentation concerning fire protection and safe shutdown. The staff is currently reviewing past Commission papers, staff requirements

August 22, 1997 memorandum, related communications with the Advisory Committee on Reactor Safeguards (ACRS), and transcripts of key meetings with the Commission and ACRS to determine the " regulatory footprint" that was established for the, passive LWRs concerning fire protection and safe shutdown conditions.

Preliminary results indicate that:

1. The staff agreed that safe shutdown could be defined.as 420*F, provided

. Westinghouse did analysis to demonstrate that it was an acceptable long-term state. ~ Westinghouse indicated that it provided that evaluation in its Shutdown Evaluation Report, WCAP-14837,

2. The staff believes that the active, non-safety-related shutdown systems are the "Ist line of defense to reduce challenges to the passive systems ,

in the event of transients or plant upsets."

3. The staff stated that there was nothing unique about the AP600 design that' would lead them to conclude that the enhanced fire protection criteria should not be applied to the design.

4. In its technical specification discussions with Westinghouse,-the staff recently concluded that cold shutdown is still the most desirable end-state after an unanalyzed event. The staff stated that hot shutdown was only acceptable when-the condition was analyzed; for unanalyzed events,  :

the appropriate end-state is cold shutdown.

5. The staff concluded that all passive plants must have a reliable means of maintaining decay heat' removal capability during all phases _ of shutdown activities, including refueling and maintenance (such as during mid-loop operation).

The staff indicated that these conclusions were preliminary and that it was about one-third of the way through its review of the documentation. Westing-house stated that if no clear position is found regarding-the need to go to cold or safe shutdown after a fire,. it would elevate discussion to senior

-management should the NRC not- accept ssfe shutdown as the final end state.

Remainina Issues The staff indicated that Westinghouse was headed in the right direction to resolve its fire protection concerns, but that it was-still completing its review.- The following action items remain to be completed:

'1. The staff indicated that it still may have some questions on the recent submittals, regarding such issues as smoke control and fire barriers.

'2. -The staff needed to caucus on the appropriateness of realistic versus -

conservative assumptions in Westinghouse's analysis of a LOOP during a fire event, and'would try to provide Westinghouse its conclusions by the third week of August.

August 22, 1997

3. Westinghouse should tell the staff how the non-safety-grade shutdown control systems are affected by fire.

4. The staff will try to complete its review of past documentation b.y the third week cf August.

5. Westinghouse should also identify to the staff where the AP600 does not comply with the Commission papers and Commission guidance during the next fire protection meeting.

6. Westinghouse should discuss the results of its safe shutdown vulnerabil-ity assessment during the next fire protection meeting.

7. Westinghouse should review the definitions of " associated circuits" in IEEE standards and Generic Letter 86-10 to determine whether the AP600 design includes such circuits.

The next fire protection meeting was tentatively setup for the last week in August 1997. A draft of this meeting summary was provided to Westinghouse to allow them the opportunity to comment on the summary prior to issuance.

original signed by:

1 -

Thomas J. Kenyon, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003 Attachments: As stated cc w/atts: See next page DISTRIBUTION:

Docket File PDST R/F SCollins/FMiraglia, 0-12 G18 PUBLIC RZimmerman, 0-12 G18~ SWeiss TQuay TKenyon BHuffman DTJackson JSebrosky WDean, 0-5 E23 ACRS (11) JMoore, 0-15 B18 Econnell, 0-8 D1 RSwest, 0-8 D1 LBMarsh, 0-8 01 DOCUMENT NAME: A:SA GRD. SUM n .w . ..., . :a. . ,4i. .i. i. in. 6..: c . co,, iinoui .it nm.nis.ncio.or. r . co , with .ti.cnm.nis.ncio.ur. N - No copy 0FFICE PM:PDS7){HIPM D:PDST:DRPM l NAME TJKenyohlsg TRQuay MA)

DATE 08/11/97 08/tt/97' W 0FFICIAL RECORD COPY l

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N Westinghouse Electric Corporation Docket No.52-003 cc: Mr. Nicholas J. Liparulo, Manager Mr. Frenk A. Ross Nuclear Safety and Regulatory Analysis U.S. Department of Energy, NE-42 Nuclear and Advanced Technology Division Office of LWR Safety and Technology Westinghouse Electric Corporation 19901 Germantown Road

_ P.O. Box 355 Germantown, MD 20874 Pittsburgh, PA 15230 Mr. Russ Bell Mr. B. A. McIntyre Senior Project Manager, Programs Advanced Plant Safety & Licensing Nuclear Energy Institute Westinghouse Electric Corporation 1776 Eye Street, N.W.

Energy Systems Business Unit Suite 300 Box 355 Washington, DC 20006-3706 Pittsburgh, PA 15230 Ms. Lynn Connor Ms. Cindy L. Haag Doc-Search Associates Advanced Plant Safety & Licensing Post Office Box 34 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 Mr. M. D. Beaumont 175 Curtner Avenue, MC-754 Nuclear and Advanced Technology Division San Jose, CA 95125 Westinghouse Electric Corporation One Montrose Metro Mr. Robert 'l. Buchholz .

11921 Rockville Pike GE Nuclear Energy #

Suite 350 175 Curtner Avenue, MC-781 Rockville, MD 20852 San Jose, CA 95125 Mr. Sterling Franks Barton 2. Cowan Esq.

U.S. Department of Energy Eckert Seamans Cherin & Mellott NE-50 600 Grant Street 42nd Floor 19901 Germantown Road Pittsburgh, PA 15219 Germantown, MD 20874 Mr. Ed Rodwell, Manager Mr. S. M. Modro PWR Design Certification Nuclear Systems Analysis Technologies Electric Power Research Instit.ute Lockheed Idaho Technologies Company 3412 Hillview Avenue Post Office Box 1625 Palo Alto, CA 94303 Idaho Falls, ID 83415 Mr. Charles Thompson, Nuclear Engineer AP600 Certification NE-50 19901 Germantown Road Germantown, MD 20874

NRC/W5STINGHOUSE AP600 FIRE PROTECTION MEETING MEETING PARTICIPANTS AUGUST 8, 1997 HAME ORGANIZATION R. NYDES W AP600 SR PROJ ENGR E. CONNELL NRC/NRR/SPLB K. S. WEST NRC/NRR/SPLB T. KENYON NRC/NRR/PDST T. QUAY NRC/NRR/PDST D. HUTCHINGS WESTINGHOUSE L. B. MARSH NRC/NRR/SPLB F. T. JOHNSON WESTINGHOUSE C. HAAG WESTINGHOUSE E. CUMMINS WESTINGHOUSE T. SCHULTZ WESTINGHOUSE R. VIJUK WESTINGHOUSE

B. MCINTYRE WESTINGHOUSE M. CORLETTI WESTINGHOUSE S. SANCAKTAR WESTINGHOUSE Attachment 1

4

- Discussion Topics for August 8, 1997 AP600 Fire Protection Meeting hmg Erejsenter Discussion Tonic 10:00 - 11:00 Tom Johnson AP600 Layout and Model 11:00 - 11:30 Don Hutchings Smoke Control 11:30 - 12:00 Tom Kenyon Status of Review of Staff Positions on Fire Protection and Safe Shutdown 12:00 - 1:00 LUNCH - cafeteria 1:00 - 1:30 Cindy Haag Fire PRA 1:30 - 2:00 Mike Corletti Approach to Determining Risk Significance of Shutdown Modes 2:00 - 2:30 Terry Schulz Discussion-of Loss'of Offsite Power and No Pressurizer Relief Valve Lift 2:30 - 3:00 Ed Cummins Methods to Achieve Cold Shutdown 3:00 - 3:30 Ed Connell Remaining Items to Complete the FSER Attachment 2 l

AP600 Stairwell Smoke Control  :.,/ ,

Building Fire Area \ Fire Zone Stairwell Method of Smoke Control-1200 1202 AF 01 S01 VBS / Self-closing doors ggg) 1201 AF 01 1204 AF 02-S02 S03 VBS / Self-closing doors Self-closing ' doors (Not used, leads to PCS tank) 1205 AF 01 SO4 VAS / Self-closing doors 1202 AF 05 S05 VBS / Self-closing doors

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d PRA INTERNAL FIRE ANALYSIS E-BACKGROUND

- PRA Chapte~r 57, internal fire analysis

- At-power operations submitted June 1996

- Shutdown operations submitted Sept.1996

. Fire analysis based on information in SSAR Chapter 9.5 & Appendix 9A, Revision 1

- SSAR information related to fire areas and fire protection was updated in June 1996 (Rev.11) and again in 1997

. June 27,1997 NRC position letter

- Noted discrepancies exist between AP600 fire protection configuration described in SSAR Appendix 9A and the configuration assessed in the PRA

- Stated position is Westinghouse revise PRA internal fire analysis and the SSAR, as appropriate, to reflect the actual fire protection design of the AP600

- AP600 PRA produces conservative estimate of fire core damage frequency because of the bounding nature of the analysis.

Examples of conservatism include:

- Assume fire in area with nonsafety-related equipment causes all nonsafety-related equipment to be unavailable.

- Assume' one hot short causes spurious ADS, in reality, two hot shorts would be needed.

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ACTION

- Assess changes made between SSAR Rev.1 and Rev.14, and the effect on the PRA internal fire analysis conclusions and insights.

Would changes cause conclusions and insights to change?

Would results change? If yes, increase or decrease; minor or significant effect?

DISCREPANCIES BETWEEN SSAR & PRA

- Fire Areas Many areas were renumbered Some areas combined Primarily electrical and I&C equipment rooms within same division One new zone identified within containment (area 1000 AF 01) zone 1250 AF 12555, VES air storage / operating deck staging area

- Automatic Fire Suppression Equipment I Automatic fire suppression equipment considered in PRA fire analysis to help mitigate the propagation of fires from one area to another Capability of auto fire suppression should no longer be credited for 6 fire areas 2

Pil PRA INTERNAL FIRE ANALYSIS "e*

ASSESSMENT

. Determine effect of changes on Fire ignition Frequencies

. Determine if any new potential for fire propagation (or removal of propagation)

. Evaluate effect of changes on at-power and shutdown operations results, conclusions, and insights.

OUTCOME

. Preliminary results of assessment indicate little to no change to internal fire core damage frequency.

. AP600 PRA internal fire analysis still produces a conservative estimate of the plant core damage frequency. -

- Conclusions and insights remain unchanged.

3

"s o o PRA INTERNAL FIRE ANALYSIS

SUMMARY

& CONCLUSION

. Discrepancies exist between SSAR fire protection configuration and the configuration assessed in the PRA

= Assessment of differences show little to no change to the conservative internal fire core damage frequency.

- PRA internal fire analysis conclusions and insights remain unchanged.

. It is not necessary to revise the PRA internal fire analysis.

1 1

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AP600 SAFE SHUTDOWN CONDITIONS

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August 8,1997 Terry L. Schulz

swmmmmmu-SAFE SHUTDOWN FC'R FIRE

. Fire Analysis Should Be Based On Expected l l Performance  !

f . Performed plant analysis for Safe Shutdown l I

. Initiating event is loss of offsite power

. Used Chapter 15 transient analysis code (LOFTRAN) l

. Used some best estimate assumptions / inputs

. Results show No operation of Pressurizer safety valve

- RCS to < 420 F and < 500 psig in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Containment increases to ~ 22 psig and 220 F  :

. AP600 Achieves " Comfortable" Safe Shutdown Conditions  :

TLS 8/8/97

ASSUMPTIONS / INPUTS

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- Initial Core. Power 100 % 102% >  !

- Decay Heat Margin 0 sig m a 2 sigma

- Initial Pzr Level 828 ft3 1142.ft2  :

- Turbine Stop Valve 0.3 sec 0 sec Close Time

- Heat Losses from expected none Pzr Steam Space -

- CVS Makeup Start 10 % 37 %

Stop Levels 20 % 52 %

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. AP600 Safe Shutdown '

e Vulnerability Assessment -

Assess the Acceptability of Safe Shutdown as a Safe End State After a Fire Evaluate Relative Vulnerabilities

- AP600 in Safe Shutdown

-AP600 in Cold Shutdown Primarily a Qualitative Assessment Augmented by Risk Based Evaluations l

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• Vulnerability Assessment 5 e Examine Two Shutdown Conditions -

AP6C O " Safe" Shutdown e PRHR Operating I

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,e Vulnerability Assessment

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Relative Vulnerability of Both End States

-Loss of Operating Decay Heat Removal System

-No Calculation of Core Damage or Severe Releases Mission Time - 7 Days Transition Risk to End States Not Calculated

-Ignores Cold Shutdown Transitional Risk

--Qualitatin Assessment Provided .

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Vulnerability Assessment E o Evaluation of the Most Probable Challenges to the Loss of Decay Heat Removal

-Loss of Offsite Power

-Mechanical Failures (Pumps, Valves) -

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-Loss of RCS Integrity ,

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Reliability of Support System 6

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• Cold Shutdown Vulnerabilities 'E l" Evaluate Two Cold Shutdown Cases Two Trains of RNS / CCW / SWS Operating

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-Most Relevant Comparison

-Assesses the Benefit of Providing Segregation to Auxiliary Systems Westinghouse 7 i

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Presentation to the US NRC August 8,1997 W. E. Cummins, Project Manager AP600 FOAKE

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. In Senior Management meeting, Westinghouse offered to assess methods to improve " fire survivability" of path to cold shutdown

. Context of assessment is that Safe Shutdown end state will be found to be acceptable by NRC

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Shutdown .

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em" Option 1 Evaluation .

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. Relatively minor change to plant design cross connect CCS and FPS through new norma!ly closed valves in Annex Building provide a dump path from CCS in Annex Building

. Fire System designed to survive any fire Cooldown is not needed until long after the fire

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(72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />)

. Relies on RNS (some low probability RNS vulnerabilities?

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Concepts of Improving ,., y., j 4

Survivability j

Option 2 .

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. Obtain a cooldown method which does not rely on RNS '

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. If RNS impacted, then other fire areas (Turbine Building) are available ~

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Concepts of Improving '

.1 Survivability .

Proposal  :

. Use Startup Feedwater and PORV to " approach cold shutdown"

. Cool Steam Generator Secondary by some method to achieve cold shutdown l bleed and feed (?)

recirculate using blowdown (?) .

. Use Natural Circulation Cooling of Reactor Coolant

. Appears feasible - design and calculations not done m

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Option 2 Evaluation 1

Not as simple / straight forward as Option 1

. Assumes plant damage isolated to RNS

! . Does not rely on RNS

. Does clapend on Turbine Building Fire Area (Startup Feedwater, Blowdown)

. Design details not developed (conceptual)

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. Perception ofimprovement? -

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) . Fire Survivability?

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