materials reliability program overview · materials reliability program: pwscc growth rate testing...

© 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

2© 2017 Electric Power Research Institute, Inc. All rights reserved.

Contents

MRP History and Organization

Gaps, Deliverables & Guidelines

Recent Industry Issues

Recent Research Areas


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.


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


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


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


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)


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


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


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


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


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


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:


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?


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


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


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)


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


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


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


MRP Thermal Fatigue Program


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


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


Recent Research Areas

Reactor Pressure VesselEnvironmental Fatigue Irradiated Materials Testing


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)


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


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.


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


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


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


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


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


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


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


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]


Questions?


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