ML20217E886

Summary of 970925 Meeting W/Wog in Rockville,Md to Provide Staff W/Details Concerning WOG B3 Program,Including Program Plan Schedule,Analysis Methodology & risk-informed Evaluation Strategy.List of Attendees Encl

ML20217E886
Person / Time
Issue date:	09/30/1997
From:	Craig C NRC (Affiliation Not Assigned)
To:	Essig T NRC (Affiliation Not Assigned)
References
PROJECT-694 NUDOCS 9710070269
Download: ML20217E886 (65)

Text

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g' NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. SeteHOM September _30, 1997 l

MEMORANDUM TO: Thomas H. Essig, Acting Chief Generic issues and Environmental Projects Branch Division of Reactor Program Management Omco of Nuclear Reactor Regulation FROM:

Claudia M, Craig, Senior Project Manage A

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Generic lasues and Environmental Projects Branch Division of Reactor Program Management Omco of Nuclear Reactor Regulation

SUMMARY

OF MEETING WITH THE WESTINGHOUSE OWNERS GROUP (WOG) ON BAFFLE BARREL BOLTING (B') ISSUE The subject meetlag was held at the NRC offices in Rockville, Maryland on September 25, igg 7, between representatives of the WOG, Westinghouse, and the NRC staff. The meeting was a follow-on to the August 20, igg 7, public meeting between the staff and WOG on this same subject. The purpose of the meeting was to provide the staff with details regarding the WOGB8 program, including the program plan schedule, the analysis methodology, and risk-informed evaluation strategy. Attachmerit 1 is a list of attendees. Attachment 2 is a copy of the non-proprietary presentation materials.

Indications of cracking in bemo/former bolts were reported in the late ig80s in European plants.

Most of the indications came from four reactors. It appears the cause is age-related degradation, but the contributing factors have not yet been specifically identified. Although no indications have been observed in WOG plants by visual inspection, visual inspections may not detect cracking due to the locking' devices on the bolts. Based on the data to date, aging effects could potentially result in cracking of the bame barrel bolts, therefore, the WOG developed the B' program to maintain safety for WOG plants and proactively manage the program with a systematic approach.

The B8 program was developed to evaluate and assess the impact of potential bolt cracking on Westinghouse reactors, should it occur. The WOG provided an overview of the bama area C*

design and outlined some of the differences between the various U.S. Westinghouse plants.

The WOG B' program consists of a 5 phase approach, including lead plant group applications.

At the meeting, the WOG provided a summary of each of the program plan tasks, provided details of the analysis approach / methodology, and outlined the risk-informed safety assessment planned for bame barrel bolts.

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September 30, 1997 The WOG would like to schedule periodic meetings with the NRC over the next year. The WOG anticipates a number of submittals to the NRC will support the WOG approach:

WCAP 14748/9, ' Justification for increasing Postulated Break Opening Times in Westinghouse PWRs', was submitted in January 1997, and is currently under review; a Bolt Analysis Methodology and Multiflex_ Version 3.0 WCAP, to be submitted in January 1998; and lead plant group application of bolt analysis results to be submitted in March 1998. The WOG would like NRC approval by September 1998 to support the fall outage of the lead plant.

'Iha staff provided questions based on their understanding of the WOG approach, included in the staffs questions was an expression of interest in the root cause analysis. The WOG noted that it would be addressed after the contributing factors are identified and assessed, but was not a subject planned for this meeting. The staff and Westinghouse tentatively agreed to a meeting on November 20 to continue discussions on the overall B' program strategy, including all the activities that need to be accomplished before implementation, the preliminary results of the group 1 risk-informed assessment, and more detail: regarding the Multiflex Version 3.0 WCAP. The WOG would like NRC comments on the methodology / program earty in the procesa. The staff informed the WOG that they will continue the review of WCAP-14748/9 and notify the WOG of any review issues.

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

T. Essig -

2 September 30, 1997 The WOG would like to schedule periodic meetings with the NRC over the next year. The

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WOG anticipates a number of submittals to the NRC will support the WOG approach:

WCAP-14748/9, " Justification for increasing Postulated Break Opening Times in Westinghouse PWRs', was submitted in January 1997, and is currently under review; a Bolt Analysis Methodology and Multiflex Version 3.0 WCAP, ;o be submitted in January 1998; and lead plant group application of bolt analysis results to be submitted in March 1998. The WOG would like NRC approval by September 1998 to support the fall outage of the lead plant, l

The staff provided questions based on their understanding of the WO^2 approach. The staff and Westinghouse tentatively agreed to a meeting on November 2012 continue discussions on

. the overall B8 program strategy, including all the activities that need to be accomplished before implementation, the preli'ninary results of the group 1 risk-informed assessment, and more -

details regarding the Multiflex Version 3.0 WCAP. The WOG would like NRC comments on the methodology / program early in the process. The staff informed the WOG that they will continue the review of WCAP-14748/9 and aotify the WOG of any rev'mw issues.

Project No. 694 Attachments: As stated cc w/stts: See next page DISTRIBUTION-Gee attached page DOCUMENT NAME: 9_25_97. MIN *See previous concurrence OFFICE PGEB;,

m (A)SC:PGEB BC:EMEB NAME CCrakb MCase*

RWessman*

DATE 9 ()O /97 9/ 30 /97 9 / 30 /97 OFFICIAL RECORD COPY

4 DISTRIBUTION w/ attachments: Summary of Sept. 25,1997, meeting with WOG dated 9/30/97 Central FW j

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PGEB R/F Project File MCase CCraig E-Mail..

SCollins/F Miraglia RZimmerman JRoe BSheron GLainas l

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TEssig RWessman FGrubelich Slee, SPSB Slee, PDLR SMalik LLois CECarpenter RHermann KWichman BElliot JFlack LLund, RES AEl-Bassioni

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i WOG / NRC MEETING 4

SEPTEMBER 25,1997 MEETING PARTICIPANTS i -

NAME ORGANIZATION i

t Claudia Cmig NRC/NRR/PGEB Dick Wessman NRC/NRREMEB Francis Grubelich NRC/NRREMEB Barry Sloane Westinghouse

-Jim Barsic Westinghouse i

Dick Schwirlan Westinghouse i

Roger Newton WEPCoNVOG Tom Greene Southem Nuclear /WOG Kurt Cozens NEl Steve DiTommaso WOG Project Othee i

David Forsyth Westinghouse lf Jeff Bass.

Westinghouse Bob Borsum FTl

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Bill Gray FTl j

Mike Schoppman FP&L 2

Samuel Line NRC/NRR/SPSB l

Sarah Malik NRC/NRR/SPSB l

KarlJacobs NYPA/WOG Brad Maurer Westinghouse i

Ken Balkey Westinghouse Charlie Griffin CP& L (WOGEPRIJoBB) j Louise Lund NRC/RESEMMEB i

Lambros Lois NRC/NRR/SRXB Keith Moser '

Comed-CE Carpenter NRC/NRREMCB Sam Lee NRC/NRR/PDLR Robert Hermann NRC/NRREMCB Adel El-Bassioni NRC/NRR/SPSB Keith Wichman NRC/NRREMCB Barry Elliot NRC/NRREMCB John Flack NRC/NRR/SPSB ATTACHMENT 1

l WOG Presentation to the NRC on the Bame Barrel Bolting (B3) Program September 25,1997

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.- 1 WOG Chairman Tom Greene, Southern Nuclear WoD

• WOG Bame Barrel Bolting Working Group Chairman o

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c Roger Newton, Wisconsin Electric Power

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Westinghouse Jim Barsic Dick Schwirian Barry Sloane

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Meeting Objective E Provide an overview of the WOG B3 program.

E Review WOG program plan schedule.

E Review WOG/NRC interactions for the program.

E Present analysis methodology and risk informed evaluation strategy O

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i Topics for Discussion 1

E Introduction / Meeting Purpose E Overview E Executive Summary

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E NRC Review / Approval E CurrentWOGPrograms i

E Analytical Approach / Methodology E Acceptance Criteria E RiskInformed Evaluation E Establish WOG/NRC Interactions / Schedule E Discussion / Summary

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1. s-Baffle Plates, Former Plates and Bolting

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j Bolt Parameters l

E Bolts are made of 347 SS or 316 CW SS l

E Three types of bolt heads are utilized l

E Washer or bar locking devices at the head of the bolt j

are utilized to minimize loose parts potential l

E The number of bolts in a plant vary:

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- Bame/former bolts - 624 to 1988 i

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- Bame/Bame Edge bolts-176 to 1696

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- Barrel /former bolts -304 to 720 E Preloads vary from 45 to 205 ft-lbs.

E New Plants (21) have bolt cooling holes i

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Background===

E Indications of cracking in baffle /former bolts reported in some European plants from 1988 to present

- 90% of the indications from four reactors

- Percentage of bolts with indications versus total bolts inspected range from 0% to 10%

- No indications of cracking in bolts with former phte cooling holes E Indications of cracking reported in 316 CW stainless steel bolts

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E Some tests suggestIASCC as a cause

- Age related degradation

- Contributing factors not specifically identified

I WOG/ Westinghouse Activities 1

E Westinghouse informed of European experience early E In 1991, an initial assessment of safety significance performed concluded that the potential bolt degradation did not represent a substantial safety hazard for WOG plants.

E In 1992, the WOG initiated a multi-phase program to evaluate and assess the impact of potential bolt cracking on Westinghouse reactors, should it occur.

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WOG Multi-Phase Approach - (1992-1997'l m Phase 1 - Plant Categorization (#, size, material of bolts) m Phase 2 - Evaluation using approved licensing basis assumptions and the limiting European UT data that was available at that time a Phase 3 - Decision analysis to assess alternative strategies / approaches a Phase 4 -Industry Participation m Phase 5 - Evaluation with more severe bolt degradation distributions with alternative assumptions G

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i WOG Phases 2 and 5 Results Summary l

a PHASE 2 (Typical 3 Loop Plant)

- Conservative acceptance criteria is met if half of the bolts on each former level of each baffle plate are intact.

m PHASE 5 (Typical 2 Loop Plant)

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- Conservative acceptance criteria can be met for more severe bolt l

cracking distributions.

- Identified areas needing licensing basis methodology approval j

n Break Opening Time Greater Than 1.0 msec t

a MultifierVersion 3.0 l

I a Leak Before Break l

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Current Assessment-Industry Status E Four European plants represent most of the bolt indications experienced

(~ 90%) with a limited amount of bolt degradstion following multiple inspections.

E Cracking has been verified by metallographic testing and confirmed to be intergranular.

E Some tests have suggested IASCC as potential cause. Contributing factors have not been specifically identified (e.g water chemistry, load follow, fabrication differences).

E European plants have continued to experience additional cracks, although

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few in number, with cracking now found on the bottom former level.

E Foreign experience has not found indications of cracking in bolts with former plata cooling holes.

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Current Assessment - Impact on WOG Plants E No indications of baille/former bolt cracking have been observed in WOG plants as determined by visual inspections.

E Visual inspections may not detect cracking due to the locking devices on the bolts.To date no domestic plant has performed UT inspections.

E The operating time at some WOG plants are higher than at plants that have already experienced cracking in Europe. Design and operating characteristics may be different than those European plants.

- operatinghistory

- baffle / barrel region coolant flow direction

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- neutron panel vs. th'ermal shield

- bolts with former plate cooling holes

- bolt design / manufacture E UT inspection data is not available from WOG plants at this time.

E WOG plants are pursuing the availability of high production inspection /replaecment equipment.

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L WOG June GeneralSession E Formed WOG Working Group to manage B3 program E Authorized tasks totaling $2.14 million for 1997 E Directed additional tasks to be presented at the October General Session I

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B3 Program Objectives m Maintain safety for WOG plants a Proactively manage the program with a systematic approach

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WOG Working Group Program Plan Tasks l

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l 9 License enhanced analytical methods and criteria for l

acceptable bolt analysis.

l 9 Perform riskinformed evaluation of the bame/ barrel region to show j

that the core damage frequency resulting from assumed baffle / barrel i

region bolt failure probabilities is low when compared to other i

l contributors to overall plant risk.

i O Perform acceptable bolt analysis for three WOG Icad piant groupings.

O Participate in domestic and foreign activities related to B3 S Determine availability of B3 as-built information.

9 Prepare bid specification for high production inspection and j

replacement site activities.

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Identification of Lead Plants i

l E Lead Plant Group 1 (347 SS) - Point Beach Unit 1 or l

R. E. Ginna Plant i

E Lead Plant Group 2 (316 CW SS) - Three Loop Plant

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E Lead Plant Group 3 - To be Based on Program Results l

E Implementation Schedule l

Group 1 Point Beach 1 (Fall 98) or R. E. Ginna (Spring 99) l j

Group 2 Three Loop Plant Group 3 To be based on program results

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1 Proposed Licensing Schedule E Initial Methodology Meeting - September 1997 E NRC Periodic Meetings E " Justification for Increasing Postulated Break Opening Times in Westinghouse PWRs" WCAP report submitted -

January 1997 (NRC Approval-March 1998)

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

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E Submittal of Lead Plant Group application of bolt analysis results-March 1998 E NRC Approval of Methodology (including WCAP for BOT) and Lead Plant Group 1 Results - September 1998 (Supports Outage Schedule)

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SUMMARY

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i E Based on current data, baffle / barrel region bolt degradation does not l

represent a safety concern for WOG plants.

E Aging effects could potentially result in cracking of the baffle barrel bolts in the US Westinghouse plants before the end of their current license. The i

WOG report on the aging management of reactor internal components

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identifies these bolts as requiring additional aging management for license l

renewal.

i E To date, no domestic WOG plant has observed bailic/former bolt cracking using visual inspection techniques.

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E A proactive program plan has been initiated by the WOG which includes i

WOG/NRC interaction relative to the bolt analysis methodologies.

l E A lead plant has volunteered for inspection and/or replacement activities

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, with the lead plant field work scheduled for Fall 1998 or the Spring of1999.

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8 Overall Conclusions

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E Substantial conservatisms exist within the existing licensing basis and a large number of degraded bolts are required before challenging safety.

E Aging effects could potentially result in cracking of the baffle barrel bolts in WOG plants before the end of their current license.

E Additional aging management actions are warranted and are

- able to be implemented in a planned manner over the next several years.

E The WOG wishes to meet frequently wiGn the NRC to provide updates, obtain feedback and obtain approval oflicensing methodologies.

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NRC REVIEW AND APPROVAL OF BOLT ANALYSIS APPROACH / METHODOLOGY E Overview

- BafHe Bolt Loads

- LOCA Forces

- Grid Crush

- LOCA PCf M NewMethodology

- Multiflex 3.0 versus Multiflex 1.0

- Increased Break OpeningTime E Acceptance Criteria

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

- Bolt Configuration with peripheral assembly grid crush

- Bolt Configuration with interior assembly grid crush E SeismicConsiderations

- Use of plant current licensing basis considerations

- seismic + LOCA load combination methodology mn

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ACCEPTABLE BOLTING ANALYSIS APPROACH OUTLINE E Consider Effects ofReduced Bolting on Potentially Effected Normal / Upset and Faulted Cosditions:

- Thermal Growth -(Normal / Upset)

- Flow-Induced Vibration (Normal)

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- BafTle-Jetting (Normal / Upset)

- Bypass Flow (Normal / Upset)

- Seismic (SSE) Event (Faulted)

- LOCA Event (Faulted) e

ACCEPTABLE BOLTING ANALYSIS APPROACH OUTLINE E PREVIOUS EFFORTS HAVE ESTABLISHED THAT T11E LOCAEVENTIS LIKELY TO BE MOST LIMITING

- DEFINE ACCEPTABLE BOLTING DISTRIBUTIONS THROUGH PARAMETRIL STUDIES WITH THE LOCA EVENT

- ONCE THE DESIRED ACCEPTABLE BOLTING DISTRIBUTIONS ARE DEFINEL FOR THE LOCA EVENT, DETERMINE THEIR ACCEPTABILITY RELATIVE TO THE REMAINING NORMAIJUPSET AND F4ULTED CONDITIONS

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ACCEPTABLE BOLTING ANALYSIS APPROACH OUTLINE E Conditions To Be Metfor Bolting Acceptability Determination (LOCA/SSE)

- SRSS combination of LOCA and seismic loads (if applicable)

- Core coolability is maintained Intact" bolt stresses are below acceptable stress limits

, - Control rod insertibility is maintained

- Fuel assembly integrity is maintained (no fragmentation)

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Acceptable B3 LOCA Load Analysis Approach i

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o L'OCA BREAK-OPENING TIME (BOT'i E Justification for Break-Opening Times (BOT)

Above 1 msec provided in WCAP 14749 j

l E Bases for Justification i

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- NUREG 0800, Section 3.6.2, Revision 1 l

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- Topical report evaluation of WCAP 8708

- Experimental data on crack propagation and BOT

- Industry calculations on break opening area and BOT i

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

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LOCA BREAK-OPENING TIME @OTl-l E NUREG 0800 (Standard Review Plan), Section 3.6.2, Revision 1, " Rupture Locations and Dynamic Effects Associated with Postulated Rupture of Piping" "A RISE TIME NOT EXCEEDING ONE MILLISECOND SIIOULD BE USED FOR THE INITIAL PULSE, UNLESS A COMBINED CRACK PROPAGATION TIME AND BREAK-OPENING TIME GREATER TIIAN ONE MILLISECOND CAN BE SUBSTANTIATED BY EXPERIMENTAL DATA OR ANALYTICAL THEORY BASED ON DYNAMIC STRUCTURAL RESPONSE."

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LOCA BREAK-OPENING TIME (BOT)

E Proposed Westinghouse BOT Design Basis

- Longitudinal Breaks (BatteIIe Tests)

- Circumferential Breaks Glattelle Tests /Schramm)

- Higher BOT values than allowed in WCAt -

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14748: Use the guidelines.in NUREG-0800, Sections 3.6.2 or the topical report of WCAP 8708

a LOCA BREAK-OPENING TIME (BOT)

E CONCLUSIONS / RECOMMENDATIONS:

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1) A design basis of 20 msecs break-opening time for W primary coolant piping in LBLOCA applications is thoroughlyjusti;ied by the results of analyses and tests i

performed by W, other NSSS vendors and independent techr.ical organization.,

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2) For general breaks, a methodology for conservatively estimating break-opening times is presented in WCAP-14748.

3) Larger break-opening times than those derived from (1) or (2) above can bejustified by:

a (A) demonstrating by analysis or test that subsequent LOCA load predictions will be conservative or n (B) demonstrating by analysis or test that the selected break-opening time is justified. These criteria are based on U.S. NRC reviews and guidance.

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Multiflex 3.0 Improvements E Two dimensional downcomer representation E Non-linear boundary conditions at core barrel / vessel interface E Peak core barrel force reductions of up to 25% obtained relative to those obtained utilizing Multiflex 1.0 6

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Multiflex 3.0 Analysis Assumptions E Consistent with Multiflex 1.0 Assumptions f

- The breaks considered will be the most limiting branch lines not covered by leak before breax exclusions allowed in GDC-4 l

- The limiting hot leg break and the limiting co!d leg break will be f

analyzed

- Bounding minimum RCS temperatures (including uncertainty) will be modeled

- Bounding maximum RCS pressure (including uncertainty) will be modeled E Multiflex 3.0 is used to calculate the hydraulic loads resulting from a LOCA for the group 'of plants to be bounded by each analysis E Break Opening Time will be selected on the basis of WCAP-14748

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-l Multiflex 3.0 Grouping Assumptions E The most limiting reactor internal design will be identified and selected for the group of plants to be bounded by each analysis:

- Upflow vs. Downflow Barrel Baffle

- Thermal Shield vs. Neutron Pad Downcomer

- Other significant differences (pressure relief holes, etc.)

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Multiflex 3.0 Analysis Results i

E Multiflex 3.0 hydraulic data will be processed consistent

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- using LATFORC for horizontal loads on vessel, core barrel, and thermal shield l

- using FORCE 2 for vertical loads on the reactor vessel internals f

- using NSAPLOT to extract and process baffle /former pressures j

to generate AP E Thr~ee possible outcomes of fuel structural analyses:

- no fuel assembly grid crush

- peripheral fuel assembly grid crush

- interior fuel assembly grid crush l

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l Results Acceptability Criteria E In all cases, control rod insertability must be demonstrated for small break LOCA analyses which credit control rods E In all cases fuel assembly integrity will be maintained E 10 CFR 50.46 requirements are satisfied through Appendix K or Best-Estimate LOCA analysis considering fuel geometry W2257

l Results Acceptability with Grid Crush: Current Appendix K LOCA Analysis Methodology E Model Grid Crush using a Reduction in Hot Assembly Flow Area and as an Increase in Assembly Flow Resistance at the affected grid elevations E Demonstrate that 10 CFR 50.46 Acceptance Criteria continue to be met for assemblies affected by fuel grid crush

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Results Acceptability with Peripheral Assembly Grid Crush: Current Appendix K LOCA Analysis Methodology E For Peripheral Assembly Grid Crush, Reduced t

Assembly Power Limits can be established for peripheral assemblies. This reduction in maximum assembly average power is used to offset the penalties resulting from grid crush (flo.w area reduction and increased resistance) to show peripheral assemblies remain non-limiting 9/2257

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Results Acceptability with Peripheral Assembly Grid Crush: Current Appendix K LOCA Analysis Methodology Limitations E Appendix K LBLOCA methodology does not model multiple core regions, therefore peripheral assembly grid crush has required peripheral assembly power limits to demonstrate acceptability E Peripheral assembly grid crush would actually tend to force flow from low power assemblies into the other regions of the core resulting in beneficial flow increases which are not credited 9/2257 e

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i Results Acceptability With Peripheral Assembly Grid l

Crush in Best-Estimate LBLOCA Analysis M Peripheral Assemblies Only:

-Low power in affected assemblies

- Grid crush in these assemblies increases the resistance, increasing flow through high power assemblies

-This leads to no PCT penalty calculated 4

Results Acceptability with Interior Assembly Grid l

Crush: Current Appendix K LOCA Analysis Methodology E For grid crush on interior assemblies, the calculated LOCA consequences of an assembly with grid crush will be the limiting case E For many plants, complying with the 10 CFR 50.46 acceptance criteria while modeling interior assembly grid crush will result in more restrictive limits on peaking factors or power

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Results Acceptability with Interior Assembly Grid Crush:

Current Appendix K LOCA Analysis Methodology Limitations E Appendix K LOCA analyses ofinterior assemblies l

with grid crush are conservative because the assembly flow area reduction is applied to all elevations while only some elevations are subject to grid crush E Further, grid crush flow area reduction is conservatively modeled as applying to the entire assembly while only some rows of fuel rods are affected by grid crush

Results Acceptability With Interior Assembly Grid Crv.sh in Best-Estimate LBLOCA Analysis l

M Interior assemblies also affected:

- High power assembly may be affected

- Explicit calculation performed

- Flow resistance increased at affected elevations according to the extent of crushing in each of the 4 core regions modeled:

a 1) hot assembly n 2) Interior assemblies under guide tubes

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' n 3) Interior assemblies not under guide tubes a 4) Low power assemblies on periphery E If net effect is a PCT penalty, this is applied to the licensing basis PCT

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Two Phase Blowdown Loads E Historically, single phase acoustic and mass flow loads have been found to be limiting for LOCA structural analyses E However, given the possibility that two phase depressurization loads may be significant for specific enclesed vessel subregions (i.e. the barrel / baffle /former regian), analyses of these loads were performed to confirm structuralintegrity

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Two Phase Blowdown Loads (con't)

E While single phase loads are calculated with Multiflex 3.0, two phase blowdown loads have been calculated using the SATAN code E However, the piping break sizes considered after leak-before-break exemptions are credited fallinto the category ofintermediate break sizes E Although the SATAN break models are believed to be applicable to those intermediate break sizes, this has not been extensively validated E The NRC approved WCOBRA-TRAC code used in Best Estimate LOCA has a break model which is valid forintermedide break sizes 9/2197

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Two Phase Blowdown Loads Conclusions

' E Therefore, WCOBRA-TRAC BELOCA models of representative plant types will be used to generate the depressurization transient for calculating two phase loads E These analyses will confirm that two phase loads are highly non-limiting compared to the equivalent single phase loads 9/12/F7

l Leak Before Break E Evolution of Pipe Break Design Basis for Mechanical Equipment

- Initially, double ended pipe breaks postulated to occur anywhere

- Finite break locations established in WCAP-8082-P-A (January 1975)

- Limited area pipe breaks - have been used in the design and qualification of many RCS components and supports

- Leak Before Break (LBB)- portions of RCS design now based on LBB mun

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E Elimination of Pipe Breaks Based on Leak Before Break Approach

- Detectable Through-WaII Leakage Crack has Significant Margin to Catastrophic Failure

- Leak can be Detected and Plant Shut Down Before Crack Grows to Critical Size E 52 FR 41288, October 27,1987, Modification of GDC-4 Requirements (FinalRule)

E Westinghouse Plants have had LBB Analyses Performed for Reactor Coolant Loop Piping mn i

Leak Before Break E With LBB Technology, Dynamic Effects from Pipe Break Can Be Excluded from the Design Basis for Piping that meets Rigorous Acceptance Criteria.

E Dynamic Effects include:

- Pipe Whip, JetImpingement

- Missile Generation

- Dynamic Subcompartment Pressurization

- Decompression Waves Within the System

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Leak Before Break l

E Leak Before Break has been used in various applications:

- Whip Restraint Removal

- Steam Generator Snubber Reduction

- Fuel Mechanical Design

- Shield WaIIInspection Port Plug Redesign E LBB Final Rule Excludes:

- Containment design

- ECCS Thermal / hydraulic performance

- Environmental qualification of equipment m2m

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E Specifically for Reactor Components:

- Mechanical loads based on LBB are used in fuel structural i

integrity analysis

- Mechanical loads based on LBB used for qualification of reactor internals components, reactor vessel supports, and reactor vessel components

- LBB approach has been applied to 2,3, and 4 loop plants O

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Leak Before Break Conclusion N Use of Leak Before Break in the evaluation of the baffle barrel region bolts is consistent with the GDC-4 final rule, and with previous and current applications of Leak Before Break.

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l Risk-Informed Safety Assessment of Baffle-Barrel Bolts Objectives

- Assess the risk from potential baffle-barrel bolt degradation, and compare to other plant risk contributors

- Confirm that there is no significant risk related to the possibility of degraded bolts in the baffle region, so that actions already being taken through the WOG B3 Program adequately address these issues l

3

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mm

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l

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k' Risk-Informed Safety Assessment of Baffle-Barrel Bolts E General Approach

- Evaluate potential for B3 degradation to affect plant operation or response to events in ways that might increase plant risk

- Estimate risk associated with B3 degradation, using elements of risk and decision analysis, and relying on expert elicitation and available data

- Derive insights regarding risk due to potential B3 degradation

Risk-Informed Safety Assessment of Baffle-Barrel Bolts E This is Not a " traditional" application of probabilistic risk assessment, and detailed PRA models are not planned

- Limited information regarding B3 status :n domestic operating plants requires reliance on expert clicitation

- Basis for expert elicitation exists in extensive WOG analyses of B3 issues and foreign B3 inspection data M Similai to approach used in WOG ris!:-informed pilot program for RCS piping inservice inspection

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e

Flow Chart of RiskInformed Safety Assessment-1 2

3 For Each IllUStratI0f1 Of -

Define initia6ng Develop Getieral Events and Structuraland Docunent Bases Process Degraded BBB T/H Response r Sess, w Configurations to Scenarios with Reta' as Risk WeMew s

be Considered BBB Failures Contributor a

9 l

l 4

l P

l

& TaskTeam N PNs, j

ConM W Quantify Risk j

of Availableinfo, i

Contributing i

input to Scenarios, 3

l Review of l

Results l

0

^

l Evaluate Available l

BBB inspection s

y l

Data and Existing Evaluate Sources l

BBB Analyses of LJncertainty and j

Sensitivi6es, Document i

Assess Risk

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Results j

(Iterale as Necessary) gg Develop insights l

l

Comparison of Safety Assessment of Baffle-Barrel Bolts to Key Principles of Draft Reg. Guide DG-1061 NRC FSA Guidance Key Co.. % : - " ; Aspect of B' Rhls-Addad==d Neces Principles (per Drah Reg.

Inferneed Safety Assesnuent Guide DG-1961)

Freposed " change" to Ct,B No intent to change curred nest smeet current res a as===

reguistions te address B'insees Defense-in-Depth features usust Adequacy of D-in-D maintenance Adequacy of D-in-D and safety maargins be snaintained wlIl be evaluated in eteis meer mmena will be considered in aslesegrased naammer by espert penets during scenario de.J.

and esantification Sufficient safety neargins suust Sepperting analyses desmenstrale Success for sense scenarios niay require be niaistained neergins use of safety analysis 1-beysed CLB but witInn anticipated fwere scensing basis (perwher B'preernes etenients)

Increases in riskresults suest Fregrama plan requires sanaN B'-

If risk increase due se B" is not sniaE, be sanall,not esceed NRC related riskforsuccess reseles will be used se identify risk.

Safety Goal esseributors that nemy require additional mesentionin etteer parts of the B' prograsn Inspleinerttation & smenisering WOG B'prograne alreadyincludes Imeent is to atleast partially bound strategies should be proposed a strategy for obtaining operatisesi uncertainties due to limulted data and use se address asacertainties &

data siist een be fed backinto the of espert panels by applying CLB success providefeedback riskevahsation crlieria rather than FRA success criter's (e g., not exceed licensing analysis FCT I

lisnio m2m O

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RiskInformed Assessment Summary

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E Work to assess the risk from potential baffle-barrel bolt degradation relative to other plant risk contributors and confirm that there is no significant risk related to the possibility of degraded bolts.

E To provide risk informed insights relative to the B3 program E Schedule

- Preliminary Group 1 Results - Novemher 7,1997

- Final Results -December 31,1997 E No formal submittal planned at this time

l Meeting Conclusions i

E Need feedback from NRC on analytical approach /

methodology before reports are submitted in early 1998

- Multiflex 3.0 i

- Break Opening Time

- Fuel Assembly Acceptance Criteria l

- Application of LBB l

- Combination ofSeismic/LOCA Loads l

E Expect to obtain NRC input, review and approval to more l

clearly define current licensing bass i

E Submittal ofGroup1 Application ofMethodology and Results

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- March 1998

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E Group 2 & 3 Analyses to be Completed in 1998 L

E NRC Approval of Methodology and Results Needed to Support Fall 1998 Plant Outage j

E Identification of NRC Review Team Members

""" E Schedule / Subjects for Future Meetings

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O O.

Risk-Informed Safety Assessment of Baffle-Barrel Bolts 1

Development of Bolt-Related Risk Scenarios Each scenario will address:

. Initiating Event Pre-existing condition of bolts, for example:

- Quantity of degraded bolts

- Locations of degraded bolts

. Effect ofincrease in loading on bolts due to event

. Degree to which fuel grids may be affected by motion of internals given event and BBB failures

. Likelihood that core cooling, control rod insertability, or other Success Criteria are met Sept. 25,97 BACKUP MATERIAL: WOG-NRC Bame Barrel Bok Program Meeting

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PREIM MARY f

EXAMPLE OF BOLT STATUS LOGIC (Showing Case with No Barrel-or El "A Failures) 1-

.i k ' ***

i FesaurImets isMiWB 4

^

Benel-Emme-Walt Asummed Depad:sionat No.efBFB Isutissing EmmeEsit FenmerButt -

BBB ABimmed Fessur Depadedat

[

Eat Samens EdgeBak Shams Smaus r _ ' ^*

Imuis EachImanies 5

NomeDesuded OK

<Huir 1

Only1

-HmIf

.) '

- AN I

Imais 2&3 t

EdgeBehsOkay Omiy Sevessi,Ramduma f

l

-Hulf i

- As j

B s.s-B. m.

eme B.=

tme i.2.2 BehsOkay Psihees Oudy

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M f

AElmels, l-Randsmi f

4 BeneBen Ben Deynession Depadmien i.

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i BACKir MATERIAL:N SmMcBessel Bok Messius. Sept.25. N r

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'1 i

PRELDHNARY EXAMPLE OF MECHANICAL & T/H LOGIC (Case with No Barrel-or Edge-Bok Failures)

Bok Suess Grid Cash No EIIect en Is Event aIJrge No p wig AE Cnseria (Commesed fross Beyond Centret Red LOCA(Itad r

PmiousSessus W

lasernen Due to Immession Het Fuel Asseuddy forSuccessbe Tree) iimer Occurs?

Belt Failuses?

Requised)7 lateerwy7 niet?

END STATE YES No BEB Risk YES YES NO BBB Risk NO BBB Risk YES No BBB Risk YES YES YES NO BBB Risk NO -

NO BBB Risk NO BBB Risk NO No BBB Risk.

BACKUP MATERIAL: WOG-NRC BeNh4sevel Bok Prograsa Messing. Scyt. 25.1997

oc:

Mr. Nicholas Liparulo Westinghouse Electric Corporation Mall Stop ECE 415 P.O. Box 355 Putsburgh, PA 152300355 Mr. Hank A. Sepp Westinghouse Electric Corporation Mail Stop ECE 4 07A P.O. Box 355 Putsburgh, PA 152300355 Mr. Andrew Drne, Project Manager Westinghouse Owners Group Westinghouse Electric Corporation Mail Stop ECE 516 P.O. Box 355 Putsburgh, PA 152300355 Mr. Mark Beaumont l

Westinghouse Electric Corporation One Montrose Metro 11g21 Rockville Pike, Suite 450 Rockville, MD 20852 e

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