ML20203K505

Summary of 971125 Meeting W/Wog in Rockville,Maryland Re in Baffle Barrel Bolting (B3) Program,Status of Current WOG Programs,Review of Submittals,Multiflex 3 Analysis,Dynamic Analysis & Modeling Approach & Risk Informed Assessment

ML20203K505
Person / Time
Issue date:	12/12/1997
From:	Craig C NRC (Affiliation Not Assigned)
To:	Essig T NRC (Affiliation Not Assigned)
References
PROJECT-694 NUDOCS 9712220404
Download: ML20203K505 (74)

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NUCLEAR REGULATORY COMMISSION WAsHlWGToN, D.c. thetHuet December 12. 1997 MEMORANDUM FOR: Thomas H. Essig, Acting Chief Generic lasues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Ructor Regulation FROM:

Claudia M. Craig, Senior Project Manage

.AAd) k.d Generic lesues and Environmental Projects Branch Division of Reactor Pr., gram Management Office of Nuclear '<esctor Regulation

SUMMARY

OF WESTINGHOUSE OWNERS GROUP (WOG)

MEETING ON BAFFLE BARREL BOLTING (B') PROGRAM The subject meeting was held at the NRC offices in Rockville, Maryland on November 25,1997, between members of the WOG, Westinghouse, and NRR staff. The purpose of the meeting was to continue discussions on the WOG B' program, specifically the status of current WOG programs, the review of submittals, the Multiflex 3.0 analysis, the dynamic analysis and modeling approach, the risk informed assetsment, WOG/NRC interactions, and the schedule for review and approvals. Attachment 1 is a list of participants. Attachment 2 is a copy of the presentation material.

The WOG provided an overview of the history of the issue and provided the B' program strategy. Basically, the WOG B' program provides the justification to support a baffle design that allows o certain number of degraded bolts to exist. The WOG discussed the WCAP that was submitted to justify increasing the LOCA break opening time. The WOG requested the WCAP be reviewed and approved by March 1998 to support other activities associated with the program and the lead plant inspections scheduled for Fall 1998. The WOG discussed the Multiflex 3.0 analysis and the differences between this revision and the approved Multiflex 1.0 version. Pa t of the justification for the revision is based on published papers that were discussed at the meetirJ. The WOG will provide copies of the papers to support staff review of the revised analysis. The WOG plans on submitting the Muhiflex 3.0 analysis to the staff for -

review and approvalin January 1998. The WOG requested the Multiflex 3,0 WCAP also be reviewed and approved by September 1998 to support the fall outages. The WOG discussed its risk informed assessment to be implemented to estimate the risk associated with the possibility of degraded botts in the battle barrel region. The WOG plans on presenting more details and the results of the risk assessment at a meeting in February 1998. At the present time, the WOG does not plan on submitting any risk assessment documentation for NRC review i

and approval. The staff stated that it may need more than presentation material to fully

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understand the risk assessment, and will explore whether the WOG may need to submit additional risk assessment information to support any decision making activity that follows.

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2 December 12, 1997 T. E ssig The WOG requested general NRC foodback on the approach of the B8 program by January 1998. The staff stated that at the present time, it had not identified any significant comments on the program, however, the staff has ordy received one submittal and discussed the program in three mootings. The WOG indicated that it recognized that it is proceeding with some risk in light of the unroviewed status of some of those initiatives, The WOG stressed the importance of receiving timely NrtC review and approval to support the Fall 1998 outages during which the inspections will be performed.

- The WOG and staff discussed the topics for the next mooting. The tepics will include the,

Multiflex 3.0 submittal (to be provided to the NRC in January 1998) and the details and results of the risk assessment. Tentative dates for the next meeting in mid-February 1998 were discussed.

Project No. 694 Attachments: As stated oc w/stts: See next page P

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O December 12. 1997 T. Essig 2-i The WOG requested general NRC feedback on the approach of the B8 program by January i

1998. The staff stated that at the present time, it had not idertified any significant comments on the program, however, the staff has only received one submittal and discussed the program in three meetings. The WOG Indicated that it recognized that it is proceeding with some risk in light of the unreviewed status of some of these initiatives. The WOG stressed the importance of receiving timely NRC review and approval to support the Fall 1998 outages during which the inspections will be perfonned.

The WOG and staff discussed the topics for the next meeting. The topics willinclude the Multiflex 3.0 submittal (to be provided to the NRC in January 1998) and the details and results of the risk assessment. Tentative dates for the next meeting in mid-February 1998 were discussed.

Project No. 694 Attachments: As stated cc w/atts: See next page DISTRIBUTION:

See attached page DOCUMENT NAME: 11.25_97. MIN n

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OFFICE PGEB SC:PGEB BC:EMEB S

NAME CCrahd MCangk RWesh JFlack M/

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DISTRIBUTION: w/ attachments: Summary of November 25,1997, meeting with WOG dated December 12,1997 Hard Copy Central Files i

PUGUC Project File PGEB R/F l

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WOG/NRC MEETING BAFFLE BARREL BOLTING PROGRAM NOVEMBER 25,1997 MEETING PARTICIPANTS hlama Organization Michael McNeil NRC/RES/DET C.I. Grimes NRC/NRR/PDLR 4

P.T. Kuo NRC/NRR/PDLR Keith Wichman NRC/NRR/EMCB Charles A. Tomes WPSC Lee Rochino RG&E Francis Grubelich NRC/NRR/EMEB J. Rajan NRC/NRR/EMEB R.E. Schwirlan Westinghouse EMT James A. Barsic Westinghouse NSD Barry Sloane Westinghouse NSD Albert Jones Southem Nuclear Farley Adel El-Bassioni NRC/NRR/SPSB Kenneth Balkey Westinghouse NSD Louise Lund NRC/RES/EMMEB C.E. Carpenter NRC/NRR/EMCB David Forsyth Westinghouse NSD Ralph Shell TVA Bob Dorsum B&WOG Karl Jacobs NYPA George Wrobel RG&E Kamal Manoly NRC/NRR/EMEB Samuel Lee NRC/.NRR/SPSB Sarah Malik NRC/NRR/SPSB Steven DITommaso WOG Project ONice Roger Newton WEPCo Claudia Cra!g NRC/NRR/PGEB ATTACHMENT 1

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i WOG Presentation to the NRC on the Baffle Barrel '

Bolting G3) Program November 25,1997 i

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WOG Baille Barrel Bolting Working Group Chairman g

g Roger Newton, Wisconsin Electric Power i

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4s Dick Schwirian Barry Sloane.

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Meeting Objective

. E Provide an update / status of the WOG B3 program.

3 E Review WOG program plan schedule.

E Review WOG/NRC interactions for the program.

It E Present portions of the analysis methodology and risk-informed evaluation analysis j

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Topics for Discussion i

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E Introduction / Meeting Purpose j

E Status of Current WOG Programs E Review of Break Opening Time WCAP E Multiflex Version 3.0 1

E Baffle Barrel Region Dynamic Analysis and Modeling Approach

. E RiskInformed Evaluation E Review ofWOG/NRC Interactions / Schedule E Discussion / Summary I

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WOG June General Session E Formed WOG Working Group to manage B3 program 1

E Authorized tasks totaling $2.14 million for 1997 h

5 Directed additional tasks to be presented at the October Ger:eral i

Session L

e WOG October General Session j

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E Authorized tasks totaling $2.97 million for 1998 E Formed WOG Program Review Team to negotiate on behalf of the ji WOG the final terms, conditions, cost deliverables and shared data l

for the following tasks.:

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- Eaffle/ Barrel Bolt Information and Materials Data - EdF i

- Study of Behavior of Reactor Internals Materials -International l

IASCC Committee d

- Baffle /Former Bolt, Edge Boit Inspection and Dest uctive p

Testing i

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- Bolt Material Hot Cell Testing l

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B3 Program Strategy

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E Short term (by 3/1/98) 1.

Develop proactive program to address the issue 2.

Obtain WOG approval of the B' Program 3.

Present program to NRC to inform and identify resource needs and schedule 4.

Perform risk-informed assessment and present to NRC 5.

Provide basis and support for utilities to become lead plants (analysis, inspection and/or replacement) i m2m

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B3 Program Strategy (Con't)

'M Intermediate term (by 3/1/99) j 6.

Participate in EdF program to obtain Eurapean bolt inspection and operating

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i; information

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Determine acceptable number and configuration of bolts for 2,3, and 4-loop lead plant groups 8.

Develop, submit and have NRC approval on improved analytical and licensing methodology and lead plant analytical results l

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Determine lead plant bolt material fabrication and operating characteristics I

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Develop bolt performance assessment tool based on material properties and i

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11.

Develop decision analysis for lead plant inspection and testing work and for j

p response to lead plant results -

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Conduct lead plant bolt inspection, replacement and testing l

13.

Update risk-informed assessment based ois lead plant results l

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Develop long term utility action plan based on technical evaluations using updated bolt performance assessment tool i

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B3 Program Strategy (Con't) 1 j

i E Long term (> 3/1/99) i

15. Determine acceptable number and configuration of bolts for g

remaining plant groups

16. Participate in EdF and Japanese programs to develop a better bolt replacement material

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17. Test bolts from lead plants in hot cells to provide input to ll bolt performance assessment tool l

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18. Update utility action plan (14) to determine if and when -

bolts need to be inspected or replaced with new bolt i

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WOG Baffle Barrel Bolting Program Strategy

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LeadWOG Plants Remaining WDG Plants (5)

Obtain Detonine Plast

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Inspect Deterreine ActualPlant Develop Bet N 8nionned i

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Artion Plan Replace (11. u)

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WOG Working Group Program Plan Tasks O License mhanced analytical methods and criteria for acceptable bolt analysis.

S Perform risk informed evaluation of the baffle / barrel region to show that the core damage i

frequency resulting from assumed baffle / barrel region bolt failure probabilities is low when compared to other contributors to overall plant risk.

9 Perform acceptable boit analysis for WOG lead plant groupings.

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S Participate in donestic and foreign activities related to B2 j

G - Determine availability of B2 as-built information (Completed).

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ii 9 Prepare bid specification for high production inspection and replacement site activities.

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- 2 Loop Bid Specification (Completed)

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- 3 Loop Bid Specification (Under Review)

- Initial 4 Loop Group Meeting: November 17,1997 (analysis development only) l

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' S Program Review Team ll

- Baffic/ Barrel Bolt Informatica and Material Data (EdF)

- Study of Behavior of Reactor Internals Material (International IASCC Committee)

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- Bafne Bolt, Edge Bolt Inspection and On-Site Destructive Testing i

- Bolt Material Het-Cell Testing Iv24s7

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Identification of Lead Plants f-E Lead Plant Group 1 (347 SS) - Point Beach Unit 2 or R. E. Ginna Plant E Lead Plant Group 2 (316 CW SS) - Three Loop Plant (tentatively Farley 1) q E Lead Plant Group 3 - To be Based on Program Results j

Implementation Schedule

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- Group 1 Point Beach 2 (Fall 98) or R. E. Ginna (Spring 99)

- Group 2 Three Loop Plant (tentatively Farley 1 Fall 98)

- Group 3 To be based on program results i+

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Proposed Licensing Schedule

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E " Justification for Increasing Postulated Break Opening Times in Westinghouse PWRs" WCAP report submitted -

g January 1997 (NRC Approval-March 1998)

E Submittal of Bolt Analysis Methodology and Multiflex ji Version 3.0 WCAP Reports - January 1998 9

E Submittal of Lead Plant Group application of bolt analysis j

results

- 2 loop group March 1998

- 3 loop group April 1998 ij

- 4 loop group December 1998 (tentative) 1 i

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E NRC Approval of Methodology (including WCAP for l!

BOT) and Lead Plant Groups Results - September 1998 (Supports Outage Schedules) l:

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NRC REVIEW AND APPROVAL OF BOLT ANALYSIS APPROACH / METHODOLOGY.

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i E. Overview l

- BafHe Bolt Loads J

- LOCA Forces

- Grid Crush

- LOCA PCf E Methodology j

- Multiflex 3.0 i

- Increased Break Opening Time

- Leak Before Break 1

E Fuel Assembly Acceptance Criteria t

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- Bolt Configuration with no grid crush (no PCT Impact) l

- Bolt Configuration with peripheral assembly grid crush j

- Bolt Configuration with interior assembly grid crush i

E SeismicConsiderations l

- Use of plant current licensing basis cons:derations

- seismic + LOCA load combination methodology k

11/2497

1 ACCEPTABLE BOLTING ANALYSIS B

APPROACH OUTLINE l

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E Consider Effects of Reduced Bolting on Potentially

j Affected Normal / Upset and Faulted Conditions

- Thermal Growth (Iow cycle fatigue)-( Normal / Upset) y

- Flow-Induced Vibration (Normal) a ll

- Baffle-Jetting (Normal / Upset) g

- Bypass Flow (Normal / Upset) j

- Seismic (SSE) Event (Faulted)

- LOCA Event (Faulted) su2497 (1

Acceptable B3 LOCA Load Analysis Approach

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LOCA BREAK-OPENING TIME (BOT)

E Justification for Break-Opening Times (BOT)

Above 1 msec provided in WCAP 14748 E Bases for Justification

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- NUREG 0800, Section 3.6.2, Revision 1

. - Topical report evaluation of WCAP 8708

- Experimental data en crack propagation and BOT l

- Industry calculations on break opening area and BOT 1

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- Comparisons of MULTIFLEX with independent U.S. NRC l

calculations

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i LOCA BREAK-OPENING TIME (BOT)

V E NUREG 0800 (Standard Review Plan), Section 3.6.2, Revision 1, " Rupture Locations and Dynarnic Effects Associated with Postulated Rupture of Piping"

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"A RISE TIME NOT EXCEEDING ONE MILLISECOND SHOULD BE USED FOR THE INITIAL PULSE, UNLESS A COMBINED CRACK-i PROPAGATION TIME AND BREAK-OPENING TIME GREATER THAN ONE MILLISECOND CAN BE SUBSTANTIATED BY EXPERIMENTAL j

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I STRUCTURAL RESPONSE."

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E Proposed Westinghouse BOT Design Basis

- Longitudinal Breaks (Battelle Tests)

- Circumferential Breaks (Battelle 3

1 Tests /Schramm) t

- Higher BOT values than allowed in WCAP-14748: Use the guidelines in NUREG-0800, i

Sections 3.6.2 or the topical report WCAP-8708 e

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PROPOSED BOT DESIGN BASIS

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E CONCLUSIONS / RECOMMENDATIONS:

1) A design basis of 20 msecs breall-opening time for W i

primary coolant piping in LBLOCA applications is thoroughlyjustified by the results of analyses and tests performed by W, other NSSS vendors and independent l

technical organizations.

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2) For general breaks, a methodology for conservatively l

l, estimating break-opening times is presented in WCAP-l 14748.

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3) Larger break-opening times than those derived from (1) or (2) above l

can be justified by:

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n (A) demonstrating by analysis or test that subsequent LOCA load predictions will be conservative or j

a (B) demonstratin~g by analysis or test that the selected break-opening time is justified. These criteria are based en U.S. NRC reviews and guidance.

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Multiflex 3.0 Improvements over Multiflex 1.0 Multiflex 3.0 Multiflex 1.0 E Two dimensional E One dimensionalset of downcomer representation parallel downcomer legs i

E Non-linear boundary 5 Linear boundary conditions li conditions at core barrel /

at core barrel / vessel vesselinterface interface E. Vessel motion allowed

. E Vessel grounded o

l through relative modal j

analysis 1

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Multiflex 3.0 Improvement Validation E " Dynamic Analysis of A Hydro-Structure System" K.

l Takeuchi, ANS Toronto Meeting, ANS-tr. 23,287 (1976)

- Compared method of characteristics (Multiflex) calculations

, gainst theoretical results for a simple flexible wall calculation; a

l used in Multiflex 1.0 validation N " Hydraulic Force Calculation with Hydro-Structural Interactious", K. Takeuchi, Nuclear Technology 39,155 (1978) l

- Compared Multif1ex 1.0 model against experimental results of Bettis test and against theoretical results for a simple shaker model. Used in Multiflev 1.0 validation t

I-

Multiflex 3.0 Improvement Validation "One-Dimosional Network for Multi-Dimensional Fluid-Structure Interac*%ns", K. Takeuchi, Nuclear Science and EngineeriLr, M,322 (1979)

- Modeled the Fritz-Kiss (GE) shaker test using Multiflex 3.0 with d

a network downcomer representation.

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- Established criteria for Multiflex 3.0 for network i

representations of the two dimensional downcomer.

- Matched experimental data well, including the calculated j

in-water frequency.

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O Multiflex 3.0 Improvement Validation f

5 " Experiment and Analysis of Shell ModelIn-Water Frequencies of a 1/24th Scale Core Barrel Model" K.

Takeuchi and D. DeSanto, Nuclear Engineering and g

j Design 59,339 (1980)

- Again validated network downcomer Multiflex 3.0 model s

L against a simplified 1/24th scale vessel and core barrel model.

- Showed comparison between Multiflex 3.0 computed in-air and 1

in-water frequency of the core barrel for both beam and shell modes l

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Multiflex 3.0 Improvement Validation E " Analysis of HDR Blowdown Experiment for Fluid-Structure Interactions", K. Takeuchi, Nuclear Engineering and Design 70,357 (1982) i

- Analysis of full scale pressure vessel, core barrel and piping system conducted at the HDR facility.

- Used Multiflex 3.0 network downcomer representation and a

included relative modal analysis to represent the vessel motion.

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- Both beam and shell modes were modeled and compared.

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- The Multiflex 3.0 computed a'mplitude of the force was higher than the experimental result.

- Good agreement between the Multiflex 3.0 and HDR results were seen out to 70 milliseconds.

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E "Non-Linear Boundary Conditions for a Structure Under Fluid-Structure Interactions" K. Takeuchi, The

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8-th SMiRT Conference, B 1/6, (1985).

- Cross-code validation of Multiflex 3.0 non-linear boundary l

D condition model wall displacements against those predicted by t

WECAN.

- Modeled a 3 loop thermal sheild plant design for a one square foot break using the Multiflex 3.0 advanced beam model.

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- Multiflex 3.0 showed good agreement in calculated wall j

displacements under the same applied loads when compared to J

a more sophisticated WECAN structural model.

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Multiflex 3.0 Submittal i

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j E WCAP-9735 "Multiflex 3.0 A FORTRAN-IV Computer j

Program for Analyzing Thermal-Hydraulic-Structural System Dynamics (III) Advanced Beam Model"

- Originally submitted for NRC review in 1986

- Submittal withdrawn when Leak-Before-Break was approved

- WCAP-9735 revision 1 will include code verification test cases j

on the current Unix computing platforms to demonstrate good agreement with original WCAP cases on the CDC computing j

platform as Appendix B.

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- Revision 1 will also include updates of the test cases to the I

current model nodalization planned as Appendix C.

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- Intended WCAP-9735, Revision 1 submittal date is 1/31/98.

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B3 Dynamic Analysis Models l

M Finite Element models of the baffle-former-barrel region - 1/8th model E Forcing functions (pressure time histories) supplied by Multiflex 3.0 assuming rigid baffles E Fluid-structure interaction effects in ANSYS model to allow for the effects of the fluid on bolt loads and fuel

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Fluid-Structure Interaction Method i

4 E Fluid structure interaction analyses is done in two stages:

" Rigid wall" hydroacoustic analysis (Multiflex 3.0)

" Fluid-structure response" analysis (ANSYS dynamic model)

E Technique is dem( estrated in the following:

- Karabin & Schwirian,"Use of Spar Elements to Simulate Fluid Acoustic Effects and Fluid-Structure Interaction in the Finite Element Analysis of Piping System Dynamics", Nuclear Engineering and Design 66 (1981), pp. 47-59.

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- Belytschko and Lin," Modal Recovery Methods for Solution of Fluid Structure Problems with Rigid Wall Loads", Nuclear Engineering and Design 71 (1982),

pp.67-78.

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E " Fluid-structure" response model based on technique described in:

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- Riddell and Schwirian," Simulation of Fluid-Structure Interaction During a

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LOCA Using Existing Finite Element Formulations", PVP Volume 5, Advances i

m Fluid-Structure Interaction Dynamics, June,1983, pp. 57-67.

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References

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- WAPD-TM-1227," Hydraulic Pressure Pluses with Structural Flexibility: Test and Analysis", April,1976

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- WAPD-TM-1290, " Pressure Pulse Test Results and l

i Qualificati<.n of the FLASH-34 Structural Member Model with a Surge Tank Attached to the Test Vessel", August,1977 t

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Pressure Difference Across Former Plate 7 with Current Model

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L Risk-informed Assessment r

i Objective - Estimate risk associated with the possibility of degraded j

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bolts in the baffle-barrel region

+.

- Show conformance with draft NRC guidance l

> - provide a level of confidence that management attention i'

and actions already being taken through the WOG B3 program are adequate to address the issues

- otherwise f

a identify any important potential risk contributors help provide insights into ways in which the ongoing i

activities can address any potentially iroportant risk j;i contributors li l

l' un.

i b

4 Approach

~

E Use elements of risk analysis and decision analysis to estimate risk l

associated with B3 degradation and derive insights 1

- Use results of existing and ongoing WOG analyses as available for e

success criteria bases

- Review availabic results of bolt inspections in European plants of similar design (if appropriate)

- Use expert clicitations to develop and estimate frequencies of B3-reizted degradation scenarios which may be risk contributors l

- Employ project task team ofWOG utility personnel with diverse

?

backgrounds appropriate to this project for peer review of key

]

. assumptions and results ofwork in process l

t l

l E Greater reliance on expert input than in traditional PRA since the available experience and data are limited

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i Approach l

l M Complex issue with limited information regarding the status of the bolts in domestic

{

operating plants -

a several types (functions, materials, designs) of bolts ofinterest

}

a no real knowledge of condition of baffle former bolts in domestic plants l

(,

a foreign plant inspection data f

p are obtained from various sources t

• may not be directly applicable to WOG domestic plants l

E Compicxity ofissue requires:

l l

n starting with simplified approach, focus on manageable set of parameters I

a address the issue first for one set of plants and conditions l

n once initial step is completed, address additional parameters as necessary to L

resolve the issue sufficiently to support the project l

E Simplifying approach requires.

t 4

I a imposing some conservative simplification to bound results, rather than 1.

obtain best estimate; a generally results in overstatement of effects but provides bounding basis for I

identifying potentially important risk contributors for further attention i

i 4

l w2wi f

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n Approach 4

1 2

3 ForEach iftestration Of Defneindiaeng Develop g

h-General i

Events and Structuraland Documert Bases Process Degraded BBB TM Response I '8' " 5 Of Con 6guradons to Scenanos e RM as Risk h

be Considered 888 Failures Conenbutor o

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Consideration Omnkfy Risk l

I of Availableinfo l

Input to Scenanch'

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i Results l-I i

i e

l Evaluate Availette BBB inspection s

I De and Exiseng Evait Sources i

BBBAnalyses of in._ _--ny and 3

i Sensstivibes.

Document I

Assess Risk Results 1-__(norate as M:: - /,

impactand Develop insights ILQ497

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Assessment Status 1

l

• Important Bolt Failure Parameters Summarized j

i

• First Plant / Group selected: 2-Loop downflow 1

a Initiating Events of Potential Importance Defined, Frequencies

]

~

Estimated Success Criteria Defined (Bounding Basis) i a Bolt-Related Risk (CDF) Scenarios Defined a Set of Bolt Degradation Conditions with Potential for Adverse l

L, Consequences Defined j

t t

a Grid Impact Conditions Defined a Fuel Rod Integrity Chal:enge Scesarios Defined

- a ThermmVHydraulic Conditions Defined li a Additional peer reviews required

.il',

l

• Revised Project Schedule a Preliminary Group 1 results: Dec.10,1997 a Remaining assessments complete: February 1998

~

s, i-

~

i

[

Key Assumptions and inputs L

i -!

l

[

E Use best information currently available to derive useful insights from 1

this process even though there may be limited analyses and data l

!i

!i E Sensitivity and uncertainty assessment included as part of the process

I will test the variability of the results to changes in inputs and increase L

the utility of the derived insights.

L E Analyses supporting success criteria can use more realistic rather than j

i, more conservative assumptions, and may credit

- Use of Multiflex version 3.0 u

y

- Break opening times (BOT) > l millisecond

)

- Break sizes limited to Leak-Before:-Break (LBB) criteria break sizes j.

L F

i1/3497 i

l lI i

}-*

Key Assumptions and inputs E Success Criteria have been defined conservatively

- Believed to be a significant probability that sequence I

outcomes assigned as failure are not direct risk contributors

- Results represent " risk precursors" rather than risk I

a Loss of core coolable geomeiry is not anticipated for most f

(

" failure" sequences; impacts would likely be limited to small fraction of core

<1 l

a Bounding treatment of fuel assembly grid crush il/MT7 l

l

l Key Assumptions and Inpui:s 1

l l

E Theprocess used is intended to be generally consistent with NRC draft guidance on use of PSA l

in risk-informed decision-making relative to p

changes in plant current licensing bases (DG1061) i j'

E Acceptance criterion for incremental risk consistent l

with recent NRC guidance, within context of DG1061 " key principles" l

I u

1:

1-l.

- ~

. l

f i

Assessment Details i

j.

1 i-Specific Success Criteria l

I h

. - Failure conservatively defined:

>> inability to insert one or mere control rod (RCCA) for events j.,

in which control rod insertion is required; i

j]

a inability to adequately cool the core as measured by l

established analytical bases for safety limits per 10CFR50.46 for peak clad temperature (PCT > 2200 deg.F), clad exidation

(> 17%),

lI

>> inability to maintain fuel rod structural integrity (coo!able l

geometry) for any one or more fuel rod er assembly 3

i

- Sequence failure occurs if any of the success criteria are not met t

lt I

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I IM i.

lhb '

l l

Assessment Details Y

h E Sequence Model Development

- General conceptual model (Figure 1) of potential bolt risk developed i

in series of working sessions with multidisciplinary participants r

- Three main elements of general sequence development process that l

j.

require additional definition-i a Existing condition of the bolts when an event occurs; a Potentia! structural deformation given a particular initiatmg l

event and existing bolt conditions; and j

t j1 a Potential fuel and thermal / hydraulic response given an initiating P

event, existing bolt conditions, and structural deformation

.t

- Model refined as analysis proceeded, based on additional y

considerations and expert input at each step;

[,-

Working" model shown in Figures 2a,b y

l i

svan l

I.

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s Figure 1 Example litustration of Risk Evaluation Process, Combining Structural and Jhermal Assessment Elements BOLT COlOITIONS STRUCTURAL THUtMAL A s u stadF NT ASSESSRADIT AttFttaaFaiT

• 0taf and states indisses success (Le., no coes descw CLB lienits seet) ses cash

• ener cany.

t j.

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" Fail" and states indicate i

Potentist risk 6 i.

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Meeting Conclusions l

1 E NRC approval of BOT WCAP-14748 expected i

E Need feedback from NRC on analytical approach / methodology l

before reports are submitted in January 1998 l

- MultiGex 3.0 i

j

- Fuel Assembly Acceptance Criteria

)

- Application of LBB

- Connbination of Seismic /LOCA Loads E NRC review and approval needed to more clearly define current l

licensing basis i

E Submittal of Application of Methodology and Results

- Group 1 March 1998

.- Group 2 April 1998 E Group 3 Analyses to be Completed in 1998 i

E NRC Approval of Methodology and Results (both Groups 1 and 2)

Needed to Support Fall 1998 Plant Outages E Schedule / Subjects for Future Meetings 11/2497 i

PRELIMINARY FIGURE 1A Bok Degradation Assessment Event Tree 8 Bolts Degraded at Ortd Cron% Beyont Inl6 sting Bertel Bame Bolt Cach Location on Acceptable Limits Affected AssembBes Event Oscurs Degrada#on Worst Piete Avoidad ?

en Periphery Only 7 NONE OR NO StGNIFICANT PATTERN Success VES Success HALF OR FEWER YES Periphery NO MAINLY LEVEL 1 NO h

'YES Sueones MORE THAN HALF YES Periphery NO No trnanor YEr.

Sucomes

~

Hale OR FEWER YES Penphery NO MAINLY LEVELS 1&2

,l, intenor OR LEVEL 7 YES Euomens

~

MORE THAN HALF YES Periphery NO NO hearior YES Success HALF OR FEWER YES Fedphory NO MAINLY LEVELS 1,2,13 NO iremrior OR LEVELS 6&7 YES L-

~

MORE THAN $ ALF YES Periphery NO NO Ireenor i

YES Success HALF.O._R FEWER p

~

NO MAINLY LEVELS 2&3 NO keenor OR LEVEL 4 OR 5 OR 6 YES Succee6 MORE THAN HALF YES Periphe7 NO NO

_ _Irt.srior WOG B3 Risk-Informed Assessment Presenta5on to NRC 11-25-97 e~

-. ~

PRELIMINARY FIGURE 28 MECHANICAL Ah"J TH SEQUENCE ASSESSMENT MODEL Affected Aasenddes Uncompiomlood Estent of Not susceptMe to An Aut 1 RCCA Fuel Rod Structural 10CFR50A8 Grid Crush impe:1 Damage?

beertaWe?

kWegrtt/?

Crtteria Met?

NONE sussess YES sussess YES YES No Fes NO FA YES YES Swamena Yt' 94 0 NO*

FOR NO Fe4 PERIPHtRY ONLY YES Sueones YES YES NO Fe3 NO Fel

}

NO YES Suseses YES NO

-NO Fel e,

NO Fe8 YES Suecoms YES YES NO FL NO

_ Fe4 YES YES Sucomen YES NO NO Fe4 NO Fe4 NTEROR AND Pf.RIPHERY YES Success vrS YES NO Fe3 NO FeR NO YEs smee.

YES 94 0 NO fee NO Fe8

?

WOG B3 Ris'< triformed Assessment Presentation to NRC 11/25/97 4

~ ~ ~~_~ TT __ : _ _ _ -_:1_*

~

" ~ ' ' '

1 t

Initial Plant Assessment j

PRELIMINARY t

i

~

d

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Expert Elicitation Process t

- Assembled. relevant information for use by experts in forming opinions

- existing analyses available information from foreign inspections

- Assembled expert (s) to discuss one key issue per session

[

• met as a group when multidisciplinary issues ofinterest

• met with single expert when only 1 person with knowledge

- Experts helped structure the sequence model l

• iterative process

- general, multidisciplinary session to start

- focused areas as assessment proceeded IlfMT7

4

,e t

Initial Plant Assessment t

PRELIMINARY Expert Elicitation Process (continued)

- Employed traditional clicitation techniques used in decision analysis Well suited to abstract nature of much of the information needed For each topic, began with areas in which experts most familiar l

(from analysis or available data) then proceded to areas of greater uncertainty

• Mainly seeking ranges of conditions that could be expected Mainly eliciting value bounds rafher than specific values Extend results to probabilities fi < some cases using statistical methods

- Process, inputs and results being documented to be repeatable

- Peer review effort is included in the process l

11/24/97 i

(

t v \\

.j initial Plant Assessment PRELIMINARY E Factors Affecting Degradation Expert session to discuss / summarize bolt material-related factors that experts considered important in predicting bolt condition Experts indicated no clear relationship among the factors discussed and the international bolt failure experience data Expert recommendation to consider the following in assessing likelihood of BFB failures:

Increaring

1. High fluence & High temperature 3

2. Medium - High fluence & teraperature Probability

3. Low-Medium flu'ence & temgerature ofhFB

4. Low fluence & temperature Failure e

, t1/24&7

d cL-s

~

L 4

I First Plant Group Assessment PRELIMINARY

[

R Degraded Bolt Conditions L

- Expert session to define degrada' tion locations and patterns of potential concern given an initiating event Experts asked to focus on projecting where degradation most

?

likely to be found regardless ofimpacts

~

• this required some estimation of potential effects of degradation at different locations n Euperts asked to assume that least-supported baffle plates would be considcred in subsequent event sequence evaluation

• this translated to no consideration of edge bolt or adjacent plate support

>> Focused on bolt degradation scenarios involving single (least-supported) bafile-plate at each former level or combination of levels Experts asked to consider two conditions for each scenario: more than half bolts degraded; half or fewer bolts degraded

- Scenarios and expert rankings defined um

7 i

L Initial Plant Assessment-PRELIMINARY l

1 Initiating Events I.

Loss of Coolant accidents are the only potentially risk-significant I

events a Analyses for upset events (e.g., anticipated transients) show these do not inunce sufficient pressure differential across baffle plates to challenge bolt integrity j

l Consequences of bolt degradation during normal operation, l

heatup, ccoldown expected to manifest as " baffle jetting"

• expected result is controlled shutdown

• no significant fuel impact / no significant incremental CDF or LERF Analyses show seismic-induced loads on bolts are substantially lower than LOCA loads for scismic event (SSE) magnitudes representative of those anticipated for first group plants t

Remainder of first group plant assessment focuses on LOCA response l

i Large, Medium, Small LOCA defined consistent with typical PRA i

definitions in order to assign consistent initiating event frequencies

• Frequencies obtained using F/OG Risk-informed Inservice j

Inspection Methodology j

une97 i

I

F initial Plant Assessment PRELIMINARY Ta ble 2. S um ma ry of Assum ed B reak Size C ategories C a teg o ry D eterministic Analysis initiating Event R a tio n ale F req u e n cy S m all Represented by plant 2

• a n d s m a lle r The availa ble 2-loup LOCA behavior associated b re a ks plant analyses were with a 4" equivalent performed for a break diameter cold leg break in a 4' O D cold leg

(~ 12 In ')

pipe; resuits will bownd

[

those for 2" and i

sm alle r bre a ks i

l M e diu m Represented by plant Breaks between 2*

Break size taken as l

LOCA behavior associated and 6*

the upper limit of the with a 6" equivalent range used for j

diameter pipe break fre qu e n cy z

(-2 5-30 in )

L a rg e R epresented by pla nt 9 rea ks la rger than T h e a vaila ble 3-Ic o p I

LOCA behavior associated 6", up to th e la rg e st plant analyses were with a ~60 inz break break size p e rfo rm e d fo r a n 8 6 8e

• 1 associated with accumulator line break i-le s k-b e fo re-b re a k

(-10 in. lin e); an i

l crite ria equivalent line size for 2-loop plants was z

estimated to be ~60 in

(-8 in. lin e), which is larger than the break size no rm s.Ily a n alyze d fo r 2-lo o p pla n ts with lenk before break crite ria wwn

[

Initial Plant Asessment l

E Grid crush assessment

- function of LOCA size and bolt degradation condition j

- deflection of baffle plate (s) i E Structuralintegrity assessment l

r

.l

- ability to insert individual RCCAs for LOCA with degraded bolts

- ability of fuel rods to withstand impact without loss of fuel coolable geometry l

~

E Thermal hydraulic assessment

- analyses performed for large and small LOCA break sizes to l

support expert elicitation

~

11/2497

,e4

(;,

Initial Plant Assessment PRELIMINARY j

E Current Assumptions Under Review ll

- Gria Crush Impact Assessment a Assumed that any plate with degraded BFB would behave like a plate with no adjacent plate restraints and no edge bolt restraints

> Actually few baffle plates in first group plants unrestrained.by other plates

• In the Icad plant group, these unrestrained plates have edge bolts 1

• Edge bolt degradation could be noticed within a reasonabic time frame (bamejetting)

• Analyses show that including effects of adjacent plate restraint.into the plate impact force calculations results in reduced grid impact I

forca

- Bolt Degradation Assessment

>> Focused on BFB only, without factoring in probabilities of also having edge bolt degradation, even though the (subsequent) grid crush assessment took no credit for plate motion restraint from edge bolts Probabilities of BFB and Edge Bolt degradation expected to be lower than for BFB degradation only, due to possibility of discovery of edge bolt degradation

s 1;

- 4 s

a f

Initial Plant A, ssessment PRELIMINARY ll 4

l, Current Assumptions Under Review (continued) a

- Initiating Event Frequencies I

Based on WOG Risk-Informed ISI methodology, so nominally better estimate frequencies than IPE-type estimate a Currently based on 3 loop plant configuration 4

l O

It/WF7

4

,1 l1,

i i

Future Planned Actions l

H j

E Finalize assumptions

]

E Quantify results and perform sensitivity evaluation ll forinitialplant group i

E Finalize / review documentation of assessment E Define steps needed for other plant groups

~

E Perform and document remaining assessmenis e

)

l'

\\ NN

I o

4 0

Westinghouse Owners Group oc -

Mr. Nicholas Liparulo Westmghouse Electric Corporation MailStop ECE 415 P.O. Box 356 u

Pittsburgh, Pennsylvania 15320-0355 Mr. Hank Sepp Westinghouse Electric CGgwWumi Mail Stop ECE 4 07A P.O. Box 355 Pittsburgh, Pennsylvania 15320 0355 Mr. Andrew Drake Westinghouse Owners Group Westinghouse Electric Corporation Mail Stop ECE 5-16 P.O. Box 355 Pittsburgh, Pennsylvania 15320 0355 4

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