© 2017 Electric Power Research Institute, Inc. All rights reserved.
Materials Reliability Program Overview
Mike Hoehn IIMRP Chairman, Ameren Missouri
Brian BurgosMRP Program Manager, EPRI
Technical Exchange Meeting on Materials May 23-25, 2017
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Contents
MRP History and Organization
Gaps, Deliverables & Guidelines
Recent Industry Issues
Recent Research Areas
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Brief History
PWR specific materials issues in the late 1990s led to the formation of the EPRI Materials Reliability Program (MRP) within the Nuclear SectorEPRI’s MRP supports efforts to
assess and implement countermeasures for degradation mechanisms impacting materials in PWR primary systemsProgram research provides utilities
and regulatory agencies with the information necessary to make technically sound and cost-effective decisions for managing degradation.
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MRP Membership
All U.S. PWR utilities In Europe
– EDF, including EDF Energy, in France & England
– Rolls-Royce in EnglandAll PWR utilities in SpainVattenfall/Ringhals in SwedenMiddle East
– ENEC In Asia
– KHNP in Korea– 3 Japanese PWR utilities– IHI in Japan– TaiPower in Taiwan
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MRP Program LeadershipMaterials Reliability Program
(IC) Hoehn, Ameren Missouri
Koehler, Xcel EnergyBurgos, EPRI
Regulatory InterfaceRichter, NEIDyle, EPRI
Technical Support TACChildress, Duke Energy
Vice Chair, OpenLong, EPRI
Inspection TACSmith, ExelonDoss, Duke
Spanner, EPRI
Assessment TACWells, Southern Nuclear
Petro, AEPCrooker, EPRI
Materials Review VisitsRobinson, INPO
Primary Systems Corrosion Research (PSCR)Cirilli, Exelon
Demma, EPRI
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MRP Technical Advisory Committees and PSCR
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
Primary Systems Corrosion Research -- How can degradation be prevented or reduced, irradiated and non-irradiated material testing
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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)
IMT Gaps being reassessed / updated in 2017 (MRP-205)
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2016 MRP Key Deliverables (1 of 2)
Title MRP Document #Materials Reliability Program: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement MRP-335, Rev. 3
Materials Reliability Program: Basis for ASME Section XI Code Case N-838—Flaw Tolerance Evaluation of Cast Austenitic Stainless Steel (CASS) Piping Components
MRP-362, Rev. 1
Materials Reliability Program: Effect of Lithium Concentration on IASCC Initiation in Irradiated Stainless Steel MRP-413
Materials Reliability Program: Specification Guideline for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement MRP-336, Rev. 1
Materials Reliability Program: Revised Technology for Reactor Vessel J-groove Weld Surface Examination MRP-410
Materials Reliability Program: Summary of JSME Thermal Fatigue Assessment Guideline and Comparison with MRP Management Guideline
MRP-408
Materials Reliability Program: Benchmark of Thermal Fatigue Management in France MRP-409
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2016 MRP Key Deliverables (2 of 2)
Title MRP Document #Materials Reliability Program: Environmentally Assisted Fatigue Testing of Stainless Steel Under Non-isothermal and Complex Loadings MRP-407
Materials Reliability Program: PWR Supplemental Surveillance Program (PSSP) Capsule Fabrication Report MRP-412
Materials Reliability Program: Basis for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement MRP-267, Rev. 2
Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines MRP-146, Rev. 2
Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion MRP-191, Rev. 1
Materials Reliability Program: PWSCC Growth Rate Testing of Alloys 690/52/152 Under Simulated Primary Water- An Update MRP-406
Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internal Components MRP-232, Rev. 2
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MRP Guidelines Active RequirementsDoc Number Rev Document Title Date Implementation
LevelMRP-126 0 Generic Guidance for an Alloy 600 Management Plan Nov
2004Mandatory
MRP-146 2 Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines
Sep 2016
Needed
MRP 2015-019 0 Implementation of NEI 03-08 Needed and Good Practice Interim Guidance Requirements for Management of Thermal Fatigue
May 2015
Good Practice
MRP-192 2 Assessment of RHR Mixing Tee Thermal Fatigue in PWR Plants Aug 2012
Good Practice
MRP-227-A A MRP 227-A, Pressurized Water Reactors Internals Inspection and Evaluation Guidelines
Dec 2011
N/A
MRP-227 1 Pressurized Water Reactor Internals Inspection and Evaluation Guidelines Oct 2015
Mandatory
MRP 2014-006 0 MRP-227-A Interim Guidance Modification to inspection requirements of Tables 4-3 and 5-3 for Westinghouse Control Rod Guide Tube Assemblies
Feb 2014
Needed
MRP-228 2 MRP-228 Inspection Standard for PWR Internals Dec 2015
Needed
MRP 2013-023 0 MRP-228 Interim Guidance Reactor Internal Baffle-Former Bolting Ultrasonic Examinations
Oct 2013
Needed
MRP-384 0 Guideline for Nondestructive Examination of Reactor Vessel Upper Head Penetrations
Sep 2014
Good Practice
MRP 2016-021 0 Transmittal of NEI 03-08 “Needed” Interim Guidance Regarding Baffle Former Bolt inspections for Tier 1 plants as Defined in Westinghouse NSAL 16-01
July 2016
Needed
MRP 2017-009 0 Transmittal of NEI 03-08 “Needed” Interim Guidance Regarding Baffle Former Bolt Inspections for PWR Plants as Defined in Westinghouse NSAL 16-01 Rev.1
Mar 2017
Needed
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Recent Industry Issues
Reactor Internals Baffle-Former Bolts (BFB)*Mitigation of PWSCC by PeeningThermal Fatigue Operating Experience (in some cases not
in locations prescribed by the thermal fatigue guidelines)*Carbon Macrosegregation*Reactor Internals Guide Card Wear*
*Details to be presented
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BFB Interim Guidance Inspection Requirement
Interim guidance issued in MRP-2017-009, dated 3/15/2017Baseline volumetric (UT) examination shall be performed as
follows: 1. NSAL-16-1 Rev.1 Tier 1 plants: per NSAL-16-1 Rev.1 and
MRP-2016-021, dated 7/25/20162. NSAL-16-1 Rev.1 Tier 2 plants: no later than 30 EFPY* 3. Remaining plants: no later than 35 EFPY
* Some Tier 2 plants have already performed the baseline UT exams between 2011 and 2016; therefore, any initial baseline UT exams performed prior to 1/1/2018 are considered acceptable even if performed later than 30 EFPY
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BFB Interim Guidance Inspection Requirement
Subsequent volumetric (UT) examinations shall be performed on an interval established by plant-specific evaluation, and shall not exceed 10-years – This evaluation is embedded within plant’s CAP program
A reduced re-inspection interval has been determined to be an appropriate response to atypical or accelerated BFB degradation and shall satisfy the following criteria:
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Testing Plan for BFBs
Short-Term Testing (2016)• Work to support Indian Point and Salem root cause
and operability analyses• Plant-funded work
Intermediate-Term Testing (2016-2017)• Testing with fleet-wide applicability resulting from
the OE
Long-Term Testing (2017+)• Characterize crack propagation mechanisms in
recently failed BFBs• Evaluate IASCC susceptibility of BFB materials with
respect to dose and time
What is the material condition of the bolt?What was the condition of the bolt at the onset of failure?Which mechanisms contributed to the failure of the bolts?What is the correlation between material condition and failure process?
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BFB Hot Cell Testing Summary
EPRI-funded hot cell testing results are consistent with prior findings– Hot cell work has not identified a different crack initiation or growth
mechanism– Defective or incorrect materials have not been observed for these
BFBs– Preliminary metallography does not indicate significant
microstructural features contributing to UT results
Utility-funded hot cell testing on DC Cook Unit 2 replacement bolts (Type 316, six-years old) indicate very high static (non-fatigue) stresses were imposed on the bolts– High loads may be attributable to failed original bolts which were not
replaced in 2010
Detailed review of BFB testing results to be presented July 2017
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BFB Operating Experience Database - Summary
Domestic plants with Westinghouse 4-loop, downflow configuration showed the highest number of bolt failures Early inspections at French plants (<10 EFPY) revealed cracking of
BFBs Shorter bolts (<1” shank) exhibited greater number of cracking incidents
than longer shank bolts and low installation torque shows higher number of cracked bolts– Bolting length is a function of plant design, and torque is related to bolt
material and bolt designWater chemistry effects are negligible compared to plant design effects;
although, elevated lithium content (> 4 ppm) may have a slight correlation
Database observations and trends are more correlated to plant design than any other external factors of the PWR environment
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Peening Mitigation of Primary Water Stress Corrosion Cracking (PWSCC) by Stress Improvement
Peening surface stress improvement (SSI) mitigates PWSCC by inducing compressive residual stress at the surface exposed to reactor coolant
– Initiation of PWSCC flaws requires tensile stress at the surface above a threshold
– Any existing flaws that are fully within the surface compressive normal plus operating stress zone cannot grow via PWSCC
Peening provides an option to mitigate reactor vessel closure head penetration nozzles instead of replacing the entire head Peening provides an option to mitigate
components that are not easily replaced or mitigated from outer surface using weld overlay or mechanical stress improvement (e.g., some reactor vessel inlet/outlet nozzles)
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US PWR Plant Applications of Peening- Planned or mitigated Alloy 600/182 components:
2016 - Byron-2, Braidwood-1, Wolf Creek, ANO-1
2017-2018 - Callaway, Byron-1, Braidwood-2, ANO-2
PWSCC Mitigation by PeeningMRP Program Complete
Technical Readiness & MRP R&D ASME Code NRC Safety
Evaluation Implementation
Peening LWRs in Japan• Both PWRs and BWRs
• PWRs mitigated during RFOs
• Laser and Water Jet Technologies
• Nozzles, J-Groove Welds and DMWs
MRP R&D Program Complete• PWSCC Initiation Testing
• Residual Stress Relaxation
• Vendor Technical Basis Information
Documentation and Guidance• Technical Basis - MRP-267, Rev 2 – published 2016
• Topical Report (MRP-335, Rev 3)- MRP-335 R3-A in published
in 2016- Submitted to NRC
• Utility Implementation Guidance
- MRP-336, Rev 1 – published 2016
Dissimilar Metal Butt- Welds (DMWs)
• Code Case N-770-5
Reactor Pressure Vessel Head Penetration Nozzles (RPVHPNs)• Code Case N-729-5
In-service peening of:• RV outlet and inlet nozzles
• Bottom-mounted nozzles
• Reactor vessel top head penetration nozzles– Three RPVHPN mockups for
post-peening UT inspectability demonstrations
MRP-335 Rev 3 Safety Evaluation (SE) for Optimizing Inspections after Mitigation
• Technical Documents submitted to NRC
• Fee Exemption and Acceptance Reviews
• Requests for Additional Information
• SE issued in 2016
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EPRI Materials Reliability Program Technical Documentation for PeeningEPRI MRP has prepared multiple documents in support of PWSCC mitigation by surface stress improvement (peening) MRP-267R2
Technical Basis for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement
– Provides background on peening methods and the technical basis for effectiveness of peening as a mitigation method
– Includes extensive data generated by peening vendors as well as confirmatory testing sponsored by EPRI
– Freely downloadable at www.epri.com, Product ID # 3002008083
MRP-335R3-A Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement
– Supports acceptance of peening as a mitigation method, including appropriate extension of the required inspection intervals following mitigation if performance criteria are met
– Freely downloadable at www.epri.com, Product ID # 3002009241
MRP-336R1Specification Guideline for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement
– Provides guidance to utilities regarding items that should be addressed by the utility / vendor
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MRP Thermal Fatigue Program
Original plant designs and inspection programs did not conceive of all potential thermal fatigue vulnerabilities– Thermal stratification– Thermal mixing
PWR OE during the mid 1980s alerted Industry to the need for management of thermal fatigue Industry responded - collaborative research led to a better
understanding of system behaviorMRP strategy focuses on component identification, inspection and
mitigation MRP initiated a Thermal Fatigue Focus Group to address OE MRP thermal fatigue management under NEI 03-08 is
implemented by:– MRP-146 Cyclic stratification in non-isolable RCS branch lines– MRP-192 Thermal mixing tees in RHR systems
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MRP Thermal Fatigue Program
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Recent OE Inspection Challenges (Thermal Fatigue)
RCS High Pressure Injection Nozzle cracking (3 inch OD)– Axial cracks on the nozzle side without craze cracking – Examination from the nozzle side may be difficult due to geometryNDE Challenge: single-side examination for axial crack detection
RCS Drain cracking (2 inch OD)– Weld cracks initiating in the heat-affected zone then propagating into
the weldNDE challenge: difficult to detect the crack tip inside the weldNDE challenge: weld complexity may trigger false calls
– Elbow skewed cracks NDE challenge: cracks with complex skewed paths difficult to
detect
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Inspection TAC Deliverables (Thermal Fatigue)
To respond to the OE, Inspection TAC plans include:– Revise the thermal fatigue examination procedureDevelop a cover sheet for use with PDI UT-2 Revise the generic thermal fatigue examination procedure
– Revise MRP 23, MRP 36 (CBT)– Fabricate 7 additional mockups– Add thermal fatigue mockups to Virtual UT System
Deliverables scheduled for December 2017
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Recent Research Areas
Reactor Pressure VesselEnvironmental Fatigue Irradiated Materials Testing
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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
– Develop revised prediction model for Upper Shelf Energy (USE) decrease (2017)
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Update on the MRP CRVSP & PSSP (1/2)Objective:
– Increase the quantity and quality of PWR surveillance data available to inform future embrittlement trend curves (ETCs) for operation to 60+ years
Coordinated Reactor Vessel Surveillance Program (CRVSP)– Developed 2010-2011; published as MRP-326– Identified the current license basis capsules whose withdrawal could
be deferred (no later than 2025) and achieve fluence >3E+19 n/cm2
If deferral would not increase to >3E+19, not included in program)– Identified 13 PWRs to change capsule withdrawal schedulesMRP-326 was issued as Needed guidance under NEI 03-08Needed action was to submit a request for schedule changeNRC staff were briefed multiple times & supportive
– After the program was implemented, the Needed aspect of the guidance was discontinuedCRVSP is not an ongoing program; rather, a one-time realignment
– 2 capsules were lost from program when SONGS 2 & 3 shutdown
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Update on the MRP CRVSP & PSSP (2/2)CRVSP increased the fluence levels of capsule data that was to
become available anyway, but did not increase the amount of dataPWR Supplemental Surveillance Program (PSSP) was designed
to obtain ~24 new high-fluence surveillance data points– Program designed/fabricated 2 supplemental surveillance capsules
containing previously-irradiated PWR materials– Specimens reconstituted per ASTM E1253– Irradiate each capsule ~10 years (some specimens up to 1.2E+20 n/cm2)– Surveillance materials were selected to fill projected data gaps and best
inform future ETCs (e.g., provide data for research, not for plant-specific licensing needs)
– Westinghouse fabricated 2 capsules: ALA-P; 14 materials (Host: Farley 1), inserted October 2016 CQL-P; 13 materials (Host: Shearon Harris) – to be inserted Spring
2018Specimen matrix, capsule loading diagrams
– Server, W., Burgos, B., Hall, B., Hardin, T., The EPRI PWR Supplemental Surveillance Program (PSSP) Final Design and Implementation, PVP2017-65307, ASME Pressure Vessels and Piping Conference, Hawaii, 2017.
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MRP Research Area: Environmentally Assisted Fatigue
Nuclear plants designed based on fatigue curves that were established using laboratory testing in air environment
– Fatigue cumulative usage factor (CUF) should be less than1.0
Licensees addressed environmentally assisted fatigue (EAF) for license renewal and now needed for SLR and for new plants
– Fatigue testing in water has shown reduced cyclic life– This effect has typically been addressed through
application of an environmental factor Fen based on USNRC-sponsored testing
– Significant challenge to demonstrate CUFen less than 1.0; often requiring substantial analysis, redesign & increased inspections
– CUFen limits have not been substantiated by plantexperience
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EAF Technical Approach - Two Parallel Paths
Analytical Identify and propose methods to reduce
conservatism in existing design rules Project led by fatigue practitioners Proposals presented to ASME code
for approval Two projects underway:
– Modification of Ke factor in ASME code
– Fatigue usage gradient factor Proposed changes would partially
offset environmental penalty
Experimental
Combine data and analysis to propose a modified approach to EAF that includes appropriate conservatism
Understand and characterize critical environmental variables
Reconcile lab data and operating experience
Three EPRI Projects to examine “separate effects” underway
Test results would be based on representative plant operation
Separate effects tests to be followed by large scale “component” tests beginning 2017
Testing to continue until about 2021
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MRP Research Area: Irradiated Materials Testing
Identify Assessment Gaps• Review and update Issue Management
Table gaps associated with long-term irradiation effects
Conduct Research• Validate aging management strategies for
current operations and provide the basis for degradation management for extended operations
Enhance Materials Models• Improve accuracy and technical robustness
of database for materials models
Characterize Margins• Better models lead to more accurate
predictions of long-term irradiation effects and better aging management strategies
Re-evaluate and Optimize Inspection Requirements• Confidence in aging management strategies
can lead to optimal inspection requirements
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Irradiated Materials TestingMRP Projects Using Zorita MaterialsProject Name Expected Results StatusZorita Internals Research Project
Increased understanding of irradiation effects on:•Tensile strength•Fracture toughness•Crack initiation and growth•Grain boundary chemistry and size•Void formation•Hydrogen and helium production•Project uses Zorita baffle plate material
• Tensile testing is complete• Crack growth rate (CGR) testing complete for
specimens at 10, 30, and 50 dpa; overall CGRs low
• Crack initiation testing complete; fractography in progress
• Fracture toughness testing of 10 dpa specimens in air and PWR water complete; testing of 30 and 50 dpa specimens underway
• Microstructural testing at MHI completed
Thermal and Irradiation Embrittlement and Environmental Effects Testing of Stainless Steel Welds
Determination of the combined effects of irradiation and exposure to elevated temperature on embrittlement of stainless steel welds and characterization of environmental effect on fracture toughness in irradiated stainless steel welds•Project uses Zorita core barrel weld material
• Machining of specimens complete• Tensile testing of weld and HAZ material
complete• Fracture toughness testing at room temperature,
elevated temperature, and intermediate shutdown temperature to began in 2016
CGR Testing of Irradiated SS Weld and HAZ Materials
Generation of IASCC CGR data in irradiated stainless steel weld and HAZ materials for comparison to existing data for base materials•Project uses Zorita core barrel weld material
• Machining of specimens complete• CGR testing to be conducted according to ZIRP
plate material protocol: 2 stress intensity levels and 3 temperatures
Determination of IASCC CGR, Initiation Rate, and Void Swelling in Zorita Material after Post-Reactor Irradiation
Evaluation of IASCC crack initiation and crack growth rates and degree of void swelling in highly-irradiated (near end-of-life conditions) stainless steel base metal and welds•Project uses Zorita baffle plate & core barrel weld material
• Testing and additional irradiation of weld and HAZ material began in 2016 as part of Halden Research Program
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Irradiated Materials TestingMRP Projects Beginning in 2017
Project Name Expected Results StatusEffect of Lithium on the SCC Initiation in Irradiated Stainless Steel
•Determination of the effect of lithium (Li) on the rate of IASCC initiation.•Work is a continuation of earlier study (MRP-413, Product ID 3002008082)
• Test matrix (dose, specimen type, loads, Li content, etc.) currently being defined
• Work to begin in 2017
Thermal Aging Analysisof Stainless Steel Weld Material at High and Low Neutron Irradiation Dose
•Study the combined effect of thermal aging and irradiation on cast and welded stainless solidification structures (austenite/ferrite)•Analyses include atom probe-field ion microscopy (AP-FIM), SEM, and EBSD on both deformed and undeformed material to characterize spinodal decomposition, G-phase precipitation, phase boundary compositions, etc.
• This work supports a PhD student thesisproject
• Project is already underway
IASCC Behavior of Baffle Former Bolt Materials
•Characterize the IASCC behavior of BFB materials•Study cracking mechanisms and crack morphologies of BFBs extracted during recent outages
• Work begins in 2017
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Irradiated Materials TestingMRP Collaborative Projects
Project Name Expected Results StatusGondole Void Swelling Irradiation and Testing
•Increase accumulated dose to ~30 dpa for virgin samples•Provoke swelling on other materials to determine kinetics of swelling•Investigate possible existence of threshold temperature for swelling
• Final irradiation cycle has been completed
• Additional TEM work on 3 to 5 specimens (either irradiated TEM foils or samples made from density specimens)
• Project ends in 2018Crack Growth Tests in Halden under PWREnvironment
Generation of IASCC CGR data in irradiated stainless steel materials in a variety of PWR conditions (effects of hydrogen, lithium, zinc additions, etc.)
• Continued participation in Halden Research program
• Current program planned for 2015-2017 and includes Zorita baffle plate, weld, and HAZ materials
Dynamic Strain Effects on IASCC Initiation Rates
•Compare existing results from static-loaded teststo tests conducted using dynamic loads representative of PWR transients to better understand EDF baffle bolt experience and IASCC test observations
• EPRI report on characterization of transients affecting U.S. PWR fleet issued in 2014 (MRP-393, Product ID 3002003085)
• Testing by CEA and SCK-CEN suggest time to initiation decreases in the presence of transients
• For 2017, EPRI to calculate changes in loads on reactor internals bolting materials resulting from transients
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Issue Program Key Takeaways
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
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MRP Key Contact Information
Brian Burgos, EPRI – Program Manager– (724) 610-8559, [email protected]
Mike Hoehn II, Ameren Missouri – Chairman– (314) 225-1543, [email protected]
Brad Adams, Southern Nuclear – Executive Sponsor– (205) 992-5181, [email protected]
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Questions?
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Together…Shaping the Future of Electricity