Risk-Informed Pressurized Water Reactor Vessel Inspection Interval Extension

1

• Risk-Informed Pressurized Water Reactor Vessel
  Inspection Interval Extension
• Cheryl L. Boggess, Nathan A. Palm
• Westinghouse Electric Company, LLC
• 12th International Conference on Nuclear
  Engineering
• Arlington, Virginia
• April 2004

2

• Risk-Informed Pressurized Water Reactor Vessel
  Inspection Interval Extension
• Authors Cheryl L. Boggess, Nathan A. Palm
• Bruce Bishop, Owen Hedden
• Westinghouse Electric Company, LLC

3
Agenda

• Introduction
• Code Case N-691
• NRC Pressurized Thermal Shock Risk Study
• WCAP-16168-NP Methodology
• Probabilistic Fracture Mechanics Computer tool
• Probabilistic Fracture Mechanics Inputs
• Results
• Applicability to Other Plants
• Conclusions

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Introduction

• Significant cost and man-rem exposure have been
  expended performing examination on ASME Category
  B-A, B-D, and B-J welds in the reactor vessel
  (RV)
• Requirements for inspection have been in effect
  since the 1989 Edition of the ASME Boiler and
  Pressure Vessel Code, Section XI
• Section XI requires 100 of RV welds to be
  inspected once each 10 years
• No service-induced flaws found to date

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Code Case N-691

• Provides the justification for extending the
  reactor vessel (RV) in-service inspection (ISI)
  interval from 10 to 20 years
• Applies to ASME Category B-A B-D welds in the
  RV and B-J welds to the RV nozzles
• Approved by ASME in November 2003
• Provides technical basis for WCAP-16168-NP

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NRC PTS Risk Study

• Significant work by the USNRC to investigate the
  effects of pressurized thermal shock (PTS) is
  ongoing
• Risk of RV failure due to embrittlement and other
  plant characteristics critical to PTS is
  calculated for three PWR designs
• Goal is to relax current PTS screening criteria
  using the latest information and risk-informed
  methods

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NRC PTS Risk Study
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WCAP-16168-NP Methodology

• Probabilistic Fracture Mechanics (PFM) methods
  are used to calculate a conditional vessel
  failure probability
• Methodology is consistent with NRC PTS Risk Study
• Pilot plants are the same as those for the NRC
  PTS Risk Study, Beaver Valley Unit 1 and
  Palisades (for Westinghouse and CE plant designs,
  respectively)
• Using ASME-XI flaw evaluation methods, a
  deterministic study identified beltline welds as
  the limiting locations

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Reactor Vessel Beltline Study
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PFM Computer Tool

• PROBSBFD Code developed to include the effects of
  fatigue crack growth (FCG ) and in-service
  inspection (ISI)
• Effects of ISI and FCG modeled in the same way as
  the NRC pc-PRAISE code and Westinghouse Win-SRRA
  tools for piping
• Effects are included in surface breaking flaw
  density file for input to the FAVOR Code
  developed for the NRC PTS Risk Study

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Key PFM Inputs

• PTS Transients
• Dominant PTS transients (each contributing ?1 to
  total PTS risk) considered in NRC PTS Risk Study
  for NSSS designs
• Risk contribution of transients considered
  contributes gt 99 of Total Plant Risk
• Transients which were considered resulted from
  loss of coolant accidents (LOCAs), stuck open
  relief valves, and large main steam line breaks
  (MSLBs)
• Operational Transients
• Design basis transients for pilot plants were
  reviewed and heatup/cooldown was determined to be
  the reference transient for the evaluation

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Key PFM Inputs

• Effectiveness of ISI
• Basis for probability of detection taken from
  studies performed at the Electric Power Research
  Institute NDE Center
• Pilot plant evaluations assumed future
  inspections conducted in accordance with ASME
  Section XI Appendix VIII
• Flaws detected were assumed to be turned to a
  flaw free state, thus maximizing the change in
  risk due to inspection

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Key PFM Inputs

• Probability of Detection

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PFM Computer Tool Process Flowchart
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Results - Flaw Growth
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Results - Flaw Growth
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Results Failure Frequency
Vessel Failure Frequency Results Vessel Failure Frequency Results Vessel Failure Frequency Results
Beaver Valley Unit 1 Palisades
Upper Bound 8.22E-9 4.95E-8
10-Year ISI Only 7.18E-9 3.93E-8
ISI Every 10 Years 6.71E-9 3.58E-8
Lower Bound 5.11E-9 2.81E-8
Difference in Risk 3.11E-9 2.14E-8
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Change in Risk - ?LERF

• Upper and Lower Bounds were determined by adding
  / subtracting 2 times the standard error to /
  from the 10 Year ISI Only and ISI Every 10
  Year cases respectively.
• Change in Vessel Failure Frequency is maximum
  difference between Upper and Lower Bounds.
• Reactor Vessel Failure is assumed to lead to a
  Large Early Release ? ?TWCF ?LERF
• ?LERF for pilot plants is an insignificant change
  per RG 1.174 if the mean value is ? 10-7 per year

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Results Summary

• Proposed ISI interval for Category B-A, B-D, and
  B-J welds is 100 inspection every 20 years after
  initial 10-year ISI
• Results support extension of reactor vessel
  inspection interval to 20 years after initial
  10-year ISI
• Methodology does not address dissimilar metal
  category B-F welds where PWSCC is a concern
• Interval provides for detection of any potential
  emerging degradation mechanisms
• Interval maintains defense in depth

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Applicability to Other Plants

• Pilot plant results apply to similar NSSS designs
  if the following parameters for a plant are
  bounded by the pilot plant
• Dominant PTS transients from NRC PTS Risk Study
  are applicable
• Degree of reactor vessel embrittlement, RTNDT
  (defined in the NRC PTS Risk Study)
• Frequency and severity of design basis transients
• Cladding layers (Single/Multiple)
• ISI meeting RG 1.150 or Section XI Appendix VIII

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Applicability to Other Plants

• A total of 24 plants with near term planned
  inspections have committed, through the
  Westinghouse Owners Group (WOG), to implement the
  methodology
• Current results of Oconee Unit 1 pilot plant
  demonstrates applicability of methodology to BW
  NSSS Design
• Westinghouse Owners Group working to determine
  schedule for fleet implementation

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WOG Implementation Schedule

• Summary of currently identified implementation
• Current 10 year interval inspection plans will
  provide yearly inspection results
• Implementation of 20 year interval inspection
  plan provides inspection results at least every 2
  years
• These results reflect implementation commitments
  from 50 of the WOG plants

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Conclusions

• The RV Beltline is the most limiting region
• Crack extension due to FCG during service is
  small for the vessels considered
• Change in risk is insignificant per RG 1.174
  guidelines
• Decrease in ISI frequency from 10 to 20 years
  maintains safety margins for defense-in-depth and
  provides a reduction in man-rem exposure
• Methodology expanded to BW NSS design
• Results of pilot plant analyses are being applied
  to other Westinghouse and CE plants
• NRC SE expected in 2004

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• Background

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RPV Inspection Scenarios