materials reliability program overview - san onofre safety ... · pdf...

© 2015 Electric Power Research Institute, Inc. All rights reserved.

Bernie RudellMRP Chairman, Exelon

Anne DemmaMRP Program Manager, EPRI

Technical Exchange Meeting on Materials, June 2-4, 2015

Materials Reliability Program Overview

2© 2015 Electric Power Research Institute, Inc. All rights reserved.

Contents

MRP History and Organization

Deliverables & Guidelines

Recent Issues & Research

Recent Research Areas

MRP/NRC Cooperative Research

2016 International Conference


Brief History

PWR specific materials issues in the late 1990s led to the formation of the EPRI Materials Reliability Program (MRP) within the Nuclear SectorThe objective of the MRP is to

resolve existing and emerging PWR materials issuesKey legacy deliverables:

Doc Number(EPRI PID) Document Title

MRP-126 Rev 0(1009561)

Generic Guidance for an Alloy 600 Management Plan

MRP-227-A(1022863)

Pressurized Water Reactors Internals Inspection and Evaluation Guidelines


MRP Membership

New members & participants since 2013: EDF Energy, Vattenfall and IHI

Rejoined in 2014: JAPC


Program Leadership

Materials Reliability Program (IC)

Rudell, ExelonHoehn-AmerenDemma, EPRI

Regulatory InterfaceRichter, NEIDyle, EPRI

Technical Support TACHoehn, Ameren

Childress, Duke EnergyMcDevitt, EPRI

Inspection TACFernandez, APS

Cordes, Southern NuclearSpanner, EPRI

Mitigation/Testing TAC

Gobell, EntergyEfsing, Ringhals

Smith, EPRI

Assessment TACSims, EntergyDeBoo, ExelonAmberge, EPRI

Technical Advisory Group

Hoehn-AmerenDemma, EPRIMaterials Review Visits

Crane, INPO


MRP Technical Advisory Committees

Assessment -- What needs to be inspected, when it needs to be inspected,

inspection options, how to disposition observed

degradation

Inspection -- How to inspect, what equipment

and techniques are available, what are the

associated uncertainties

Technical Support – Fatigue and reactor pressure vessel integrity, review and maintain guidelines, compile inspection results

Mitigation and Testing -- How can degradation be prevented or reduced, irradiated and non-irradiated material testing


ID Gap Description

P-AS-22Steam Generator Tubes & Internals Wear & High-Cycle Fatigue

P-AS-24Denting & SCC in Steam Generator Top of Tubesheet(TTS) Region

P-AS-26Steam Generator Tube Damage due to Loose Parts or Foreign Objects

P-AS-30ODSCC of Thermally Treated Alloy 600 Steam Generator Tubing

P-AS-31Safety Significance of Cracks in Steam Generator Divider Plate

P-AS-35 Steam Generator Sludge Deposits and Scale Buildup

P-I&E-15Steam Generator Tubing Eddy Current Technology Improvements

P-I&E-16NDE - Tools for Steam Generator Tubing Integrity Assessments

P-I&E-18Steam Generator Tube Eddy Current Data Analysis Software Improvements

P-I&E-20Steam Generator Foreign Object Detection and Evaluation Improvements

2013 PWR IMT High Priority GapsID Gap Description

P-AS-02Environmental Effects on Fatigue Life: Pressure Boundary Components

P-AS-09 SCC of Stainless Steels Exposed to Primary Water

P-AS-11PWSCC Crack Growth Rates for Alloys 600, 82, and 182

P-AS-12PWSCC Factors of Improvement for Alloys 690, 52, and 152

P-AS-13aThermal & Irradiation Embrittlement Synergistic Effects on CASS

P-AS-13bThermal & Irradiation Embrittlement Synergistic Effects on SS Welds

P-AS-14a IASCC Characterization: Generic Data NeedsP-AS-14b IASCC Characterization: Baffle Bolting

P-AS-17 Flow-Induced Vibration and Wear of Reactor Internals

P-AS-19 PWSCC Management for Ni-Alloy Reactor Internals

P-AS-27 Alternative ASME Section XI Appendix G Methodology

P-AS-28Neutron Embrittlement of Nozzle Forgings and Upper Shell Course

P-AS-38Fluence Impact on Stainless Steel Mechanical Properties (Fracture Toughness, Tensile Strength)

P-AS-46CASS Piping Component Thermal Aging Embrittlement & Long-Term Integrity Assess.

P-I&E-03 NDE Technology for J-Groove Weld LocationsP-I&E-12 NDE Technology for Examination of CASSP-I&E-21 Reactor Internals Generic Acceptance Criteria

P-RG-06NDE Qualification for Reactor Internals Inspection (VT Evaluation)

P-RG-09 Pipe Rupture Probability Re-Assessment (xLPR)


2014 Key Deliverables

1. Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375)

2. T-cold Reactor Vessel Upper Head Re-inspection Interval Technical Basis Re-evaluation (MRP-395)

3. Palo Verde #3 Bottom Mounted Nozzle Leak Boat Sample Destructive Examination Report (MRP-394)

4. Guideline for Nondestructive Evaluation of Reactor Vessel Upper Head Penetrations (MRP-384)

5. Irradiated Materials Welding Guideline (MRP-379)6. Damage-Based Modeling and Analysis of Reactor Vessels

with Numerous Base Metal Quasi-Laminar Indications (MRP-388)


MRP Inspection Results DatabaseAn Annual Publication


MRP Guidelines Active RequirementsDoc Number(EPRI PID) Document Title Date

Highest Requireme

nt LevelMRP-126 Rev 0(1009561)

Generic Guidance for an Alloy 600 Management Plan Nov 2004

Mandatory

MRP-146 Rev 1 (1022564)

Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines

Jun 2011

Needed

MRP-146S Rev 0(1018330)

Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines – Supplemental Guidance

Jan 2009

Needed

MRP-192 Rev 2(1024994)

Assessment of RHR Mixing Tee Thermal Fatigue in PWR Plants Aug 2012

Good Practice

MRP-227-A (R-A)(1022863)

MRP 227-A, Pressurized Water Reactors Internals Inspection and Evaluation Guidelines

Dec 2011

Mandatory

MRP 2014-006Rev 0

MRP-227-A Interim Guidance to inspection requirements of Westinghouse Control Rod Guide Tubes

Feb 2014

Needed

MRP-228 Rev 1 (1025147)

MRP-228 Inspection Standard for PWR Internals Dec 2012

Needed

MRP 2013-023Rev 0

MRP-228 Interim Guidance for Ultrasonic Examinations ofReactor Internal Baffle-Former Bolting

Oct 2013

Needed


Recent Industry Issues

Non-conservatism of BTP 5-3 for initial RTNDT and USE (MRP-401)*Thermal fatigue Operating Experience and in some cases

not in locations prescribed by the thermal fatigue guidelines*Topical report for peening mitigation of PWSCC (MRP-335)*

Reactor internals guidelines (MRP-227) revision*

*Details being presented today


Benefits of MRP-227, Rev. 1

Maintaining guidance current with lessons learned and research developmentsEnhancing nomenclature and detail of internal component

sketchesAdding specificity to coverage requirements/acceptability of

limitations Incorporating latest WCAP requirements for guide card wearAddressing lower support clevis/snubber OE Incorporate generic acceptance guidanceAddressing NRC SE A/LAIsAdding technical bases and associated exam scope

specifically for Core Barrel welds Incorporating new CASS criteria (if available – pending NRC

approval)


Mitigation of PWSCC by Peening

Benefits

– Mitigates PWSCC in alloy 600/82/182 components

– Reduces nuclear safety and leakage risks from PWSCC

– Proactive asset management solution

– Stress improvement for remaining operating life of plants

– Proven technologies with decades of implementation


Benefits of Thermal Fatigue Interim Guidance

Improving Thermal Fatigue Management Guidelines under Industry Initiative NEI 03-08– MRP-146, Management of Thermal Fatigue in Normally Stagnant Branch Lines,

Revision 1, June 2011– MRP-192, Assessment of RHR Mixing Tee Thermal Fatigue, Revision 2, August

2012Management through NEI 03-08 provides for streamlined response to

emergent events and actionable improvement opportunities– Operating Experience reporting and monitoring identified trend– PMMP created action to address apparent issue for PWR fleet (relevance was

considered for BWRs)– Thermal Fatigue Focus Group formed to assess experience against current

guidance and create interim guidance to address issuesOperating Experience has identified Thermal Fatigue Management

Guideline gaps that Interim Guidance will close– Adding emphasis on attention to and effects of transient stratification– Adding examination performance emphasis where OE identified weaknesses– Adding expanded pipe size range and examination zones for certain situations


Recent Research Areas

PWSCC of Alloy 690 and weld metals*xLPRWeld Residual Stress*Environmental Fatigue Irradiated material testing Inspection

*Details being presented today


MRP Research Area: Extremely Low Probability of Rupture (xLPR)

• Fully cooperative program with NRC Research to develop a probabilistic fracture mechanics approach to address GDC 4 regarding presence of PWSCC in piping previously approved for leak-before-break

• Further development could more broadly apply probabilistic tools to

• Evaluate reduction in risk with mitigation

• Other degradation mechanisms and materials

• Broader range of components

Modular ProbabilisticFramework

Stress intensityfactor

Crack Stability

Crack Initiation and growth

Inspection

Leak detection

Crack Coalescence

Frequency of failure

Geometry and Loads


MRP Research Area: Reactor Pressure Vessel

Extending research to address RPV integrity issues through a second license renewal– Generate high-fluence surveillance data to

support PWR operations to high fluence (PSSP)

Degradation modeling– Atom probe tomography on high-fluence

RPV surveillance specimens (w/ CRIEPI)– Support testing of PWR surveillance

materials in ATR-2 (Planned for 2015)– Develop embrittlement trend correlation for

Upper Shelf Energy (USE) decrease (2016)

Resolution of Current Issues– Appendix G “small flaw issue”: Work will

resume after FAVOR clad stress model improvements are completed by NRC

– BTP 5-3 (separate presentation)


MRP Research Area: Irradiated Materials Testing – Zorita Research Project (ZIRP)

Current Status Mechanical testing

– Tensile testing complete 10, 30, & 50 dpa @ RT & 320⁰

– CGR testing 10, 30, & 50 dpa complete Fractography in progress

– Fracture Toughness 10 dpa in air and PWR water

complete– Crack initiation testing to begin in

2015 Microstructural testing to begin in

2015

Jose Cabrera NPP “Zorita”Westinghouse design

1968 – 2006 (~26 EFPY)

Objective: Characterize the effects of neutron irradiation on mechanical and microscopic properties of stainless steel materials irradiated under service conditions to increase the understanding of fluence effects at end of service life (40, 60 years and beyond)

ParticipantsEPRI MRP, U.S. NRC, CSN, SSM, Tractebel, AXPOJapanese utilities/MHI (in-kind contribution)


MRP Research Area: Irradiated Materials Testing

Zorita Weld Testing Program• Specimen maching, material characterization, & tensile testing in 2015• CGR and fracture toughness testing of Zorita weld and HAZ materials

• CGR: 2 K levels and 3 temperatures as for ZIRP plate material • FT: air and PWR water; RT and 320⁰C

Effect of Lithium on IASCC Initiation• Flux thimble tubes from Ringhals 2: max. doses of 65 and ~100 dpa• Li levels of 2 and 8 ppm at 340 ºC with constant pH300C = 7.2• Preliminary results: high Li content increases susceptibility to IASCC

Gondole Void Swelling Program• Final irradiation cycle underway (total accumulated dose ~30 dpa)• TEM analyses to be conducted on irradiated specimen foils and from density

specimens• Preliminary results: only pre-irradiated specimens show swelling greater

than 0.5%; several specimens show negative swelling greater than -0.5%


PWR In-Vessel Inspection Course Project Snapshot

– Address requirements from the I&E Guidelines, ASME Section XI, and the three OEM PWR designs.

– Hands-on use of remote inspection devices – Course for NRC staff scheduled for August 17-19 and 19-21, 2015 at

AREVA NP, Lynchburg, VA Course Outline

– MRP-228 and MRP-227– B&W Internals Components– CE Internals Components– Westinghouse Internals Components– Lessons Learned– Inspection Results


EPRI MRP/U.S. NRC - Cooperative Research

CurrentA690 PWSCC Initiation & Crack Growth Testing Extremely Low Probability of Rupture (xLPR) Welding Residual Stress (WRS) FEA Model Validation Irradiation Material Testing: Zorita Internals Research Project

(ZIRP) Additional testing on Additional Zorita materials (weld and

HAZ)


Program Key Take-Aways

• MRP is focused on the resolution of materials issues for PWR primary components

• MRP has made significant contributions to the industry in nickel-base alloys, reactor internals, RPV integrity and fatigue areas

• Generating data, assessments, guidelines and closing gaps

• Continued proactive research is needed


2016 International LWR Materials ConferenceInternational Light Water Reactors

Material Reliability Conference & Exhibition 2016 August 1-4, 2016

Save The Date and Call for Abstracts Coming Soon

Hyatt Regency McCormick Place

Chicago, Illinois, USA


MRP Key Contact Information

Anne Demma, EPRI – Program Manager– (650) 855-2026, [email protected]

Bernard Rudell, Exelon – Chairman– (410) 495-4815, [email protected]

Elliott Flick, Exelon – Executive Sponsor– (302) 559-2287, [email protected]


Questions


Together…Shaping the Future of Electricity

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