Primary Systems Corrosion Research Program
Transcript of Primary Systems Corrosion Research Program
© 2017 Electric Power Research Institute, Inc. All rights reserved.
Anne DemmaProgram Manager, EPRI
Technical Exchange Meeting on MaterialsRockville, MD May 23, 2017
Primary Systems Corrosion Research
Program Overview
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Materials Advisory StructureMaterials Action Plan Committee
Executive Chair: Fadi Diya, Ameren MissouriTechnical Chair: Scot Greenlee, Exelon
EPRI PM: Kurt Edsinger
Exec SponsorBrad Adams, SNC
Exec SponsorJoe Donahue
MRPPWR Materials
Reliability ProgramIC Chair
Mike HoehnAmeren
Brian BurgosEPRI PM
SGMPSteam Generator Mgmt Program
IC ChairJohn ArharPG&E
Helen CothronEPRI PM
PWR Materials Management Program (PMMP)Exec Chair: David Czufin, TVA
EPRI PM: Robin Dyle
Exec SponsorTom McCaffrey, Entergy
BWRVIPBWR Vessel &
Internals Program
Exec ChairTim Hanley
Exelon
IC ChairDrew Odell
Exelon
Andy McGehee EPRI PM
WRTCWelding & Repair
Technology Center
IC ChairDan Patten
FENOC
Greg Frederick EPRI PM
PSCRPrimary Systems
Corrosion Research
ChairJim Cirilli, Exelon
Vice ChairPål Efsing, Ringhals AB
Anne DemmaEPRI PM
PSCR Technical Staff: Peter Chou, Raj Pathania, Cem Topbasi, Jean Smith, and David Steininger
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Mission and Scope for PSCR• Overall responsibility for materials testing,
characterization, and associated modeling within the Materials Programs– PSCR Strategic Plan integrated with the other
Materials Programs
• Provide the key engagement point for Base members that do not participate in the other Materials Programs
• Act as a focal point for materials-related Technology Innovation scope
• Support other areas of EPRI Nuclear Power Sector with materials expertise
Interactions between PSCR and the other Materials Programs are similar to the way these programs interact with EPRI’s Chemistry Program
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List of PSCR Members
All EPRI Nuclear Full Members are PSCR Members- 22 US- 18 International
PSCR
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PSCR Research Focus Areas
RFA Description
A Irradiated Materials Testing of Stainless Steels and High-Strength Alloys
B Fatigue Testing
C Low Alloy Steels Testing
D Non-Irradiated Nickel-Base Alloys Testing
E Non-Irradiated Testing other than Nickel-Base Alloys (Stainless Steels, CASS, etc.)
F Next Generation of Materials, Irradiation, and Testing Techniques
G Materials Strategic Tools and Training
Atom probe tomography (APT) examination of high fluence surveillance specimens
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Key Accomplishments
1. Published Materials Degradation Matrix (MDM)– Supports NEI 03-08 Materials Initiative for the Management of Materials Issues– Provides a comprehensive listing of potential degradation mechanisms for
existing LWR primary system components– Proactively identifies potential challenges to avoid surprises– Covers multiple plant designs
2. Published Materials Handbook for Nuclear Plant Pressure Boundary Applications – Provides accurate mechanical and physical properties data on structural materials used in nuclear applications– Provides a concise source of information on materials performance in light water reactor service
3. Developed irradiation assisted stress corrosion crack (IASCC) crack growth rate (CGR) models stainless steels used in BWR and PWR internal components, and submitted the models for ASME code case development– Developed crack growth models and disposition curves for application to BWR and PWR internals– Transferred to MRP and BWRVIP for guidelines and ASME code development
1.E-09
1.E-08
1.E-07
1.E-06
1.E-05
1.E-04
1.E-03
0 10 20 30 40Stress Intensity Factor K, MPa√m
da/d
t (32
5°C
, 700
MPa
), m
m/s
Normalized Low-ECP Data Ranked < 3New Upper-bound to 75% of Low-ECP DataMRP-227-A PWR Curve, Upper-bound to 43% of Data
PWR CGR Curve at 325°C and700 MPa (~4.3 dpa, Solid Curve)
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PSCR Publishes and Updates the MDMExample of MDM Results
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PSCR Publishes and Updates the Materials Handbook
SECTION I — BASE MATERIALS FOR PIPING AND PRESSURE VESSEL PRESSURE BOUNDARIES
1. Carbon and low alloy steels for pressure vessels2. Carbon and low alloy steel piping 3. Stainless steel for piping, components, and pressure vessels4. Nickel-base alloys for pressure vessels, components, and piping
SECTION II — HIGH STRENGTH MATERIALS FOR BOLTING, VALVE STEMS, SPRINGS
SECTION III — TUBING ALLOYS
SECTION IV — PUMP AND VALVE TRIM MATERIALS
SECTION V — NON-METALLIC MATERIALS
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Top Technical Issues for PSCR in the Next 3 Years• Irradiated Materials Testing and Modeling of Stainless
Steels– Participate in global research efforts to better understand the
role of key parameters associated with IASCC of reactor materials and to develop improved materials for reactor vessel internals components for replacement in existing plants or for new plants
• Environmentally Assisted Fatigue (EAF) Testing and Modeling– Perform testing and develop models to more accurately predict
EAF in light water reactor environments for existing and new plants
• Potassium Hydroxide Materials Testing– Perform materials testing to qualify potassium hydroxide to
potentially replace lithium hydroxide in certain PWR designs
IASCC Initiation Test Set-up
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2017 Research Focus Areas and Example of Project Value (1/2)Project Schedule Issue Programs RFAs
Applicability Value Recipient of the ValueRFA Description 2017 Project(s) 2017 2018 2019 IMT
gaps MRP BWRVIP SGMP WRTC
RFA A: Irradiated
Materials, Testing of Stainless Steels, and High-Strength
Alloys
Support Core Shroud cracking investigation
(end in 2018)TI-funded in 2017
(TI)BWRRFA-
2
Applicable to irradiated 304SS components in
BWRs at ~3 dpa.
The material is the base metal from a core-shroud boat sample. APT observations will be analyzed for
additional insight into the cracking of this core shroud. SCC initiation testing will be funded by TI in 2017. Subsequent CGR testing will be co-funded by
DOE LWRS. The latter laboratory tests will quantitatively assess the IASCC susceptibility
of the material.
Utility members of PSCR and BWRVIP.
RFA F: Next generation
of materials, irradiation and
testing techniques
Development of Radiation Resistant Material (ARRM)
Phase-I (end in 2017)Co-funded by DOE and
BNPC.
P-RR-08
MRFA-1
MRFA-2
Applicable to reactor internals in PWRs,
BWRs, VVERs, CANDUs, and new plants. The project will identify more radiation resistant alloys for both low strength structural
applications (e.g. baffles, formers,
core barrel, core shroud and top
guide) as well as high strength
applications (bolts and springs) in
the core.
The final report on Phase 1 to be issued in mid-2018 will identify materials that have good
resistance to irradiation induced degradation after proton irradiation and are suitable for neutron
irradiation and post irradiation testing in Phase 2. For example, testing of several high-strength alloys has shown that age hardened Alloy 725
is more resistant to IASCC in BWR and PWR environments than age hardened Alloy 625 that
is currently used in some PWRs. This is co-funded by DOE and BMPC.
Utility members of PSCR, MRP, BWRVIP, and
ANT programs.
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2017 Research Focus Areas and Example of Project Value (2/2)Project Schedule Issue Programs RFAs
Applicability Value Recipient of the ValueRFA Description 2017 Project(s) 2017 2018 2019 IMT gaps MRP BWR
VIP SGMP WRTC
RFA F: Next Generation
of Materials, Irradiation, and
Testing Techniques
(TI)Rapid Simulation of High Fluence -- Ion radiation of
LWR irradiated FTT(end in 2019)
TI-funded in 2017
P-AS-15 MRPRFA-1
Applicable to PWR and VVER
reactor internals irradiated
materials testing.
Develop and validate an approach based on heavy ion irradiation with He/H implantation to
simulate the irradiation damage at very high fluence to simulate void swelling. Could offer a cost and time efficient method to estimate
void swelling in stainless steel core internals in extended operation.
Utility members of PSCR and MRP.
RFA B:Fatigue Testing
Short crack behavior in EAF of stainless steel
(starting 2017) MRP
RFA-10BWR
RFA-10
Applicable to PWRs, BWRs,
VVERs, CANDU, and new plants
piping components.
This work will investigate the cause of CGRs being different for short fatigue cracks of less than 0.1 mm and long engineering cracks of
greater than 1 mm. Generating accurate fatigue CGR data for prototypical environmental
conditions over both short and long crack length ranges will help in understanding the results
obtained from the prototypical full scale component test. Additionally, the data are
required to support the development of a total fatigue life evaluation of plant components
starting from an initially small crack length through the engineering long crack growth phase and thus
leading to component leakage.
Utility members of PSCR, MRP, BWRVIP, and ANT programs.
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On-Going Collaborations1. EPRI-DOE LWRS Co-fund Study on IASCC
Mechanisms– To develop a more complete understanding of IASCC– To determine the roles of key solute additions and
microstructures on both crack initiation and crack growth
2. Advanced Radiation Resistant Materials (ARRM) program
– To develop the next generation of materials for in-core structural components and fasteners
– Phase-I started in 2012 with co-funding from DOE-LWRS and KAPL
3. EPRI-MAI Collaborative Projects on SCC and IASCC
4. EPRI-CANDU Owners’ Group on PWSCC of Nickel-base Alloys to Carbon Steel Welds in CANDU Primary Heat Transfer System
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Collaborations with EPRI Programs
• Other Materials Programs– MRP, SGMP, BWRVIP, and WRTC
• Advanced Nuclear Technology (ANT)• Chemistry• Long-Term Operations (LTO)• Flexible Operations• Non-Destructive Evaluation (NDE)
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Upcoming EPRI Meetings in 2017
Event Date Location
PSCR Meeting in coordination with the MRP and BWRVIP Meetings June 12-13 Pittsburgh, PA
PSCR Meeting during the EPRI Nuclear Power Council Meeting Week August 28 Hollywood, FL
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Program Key Take-awaysThe PSCR Program will:• Continue to develop collaboration with
external R&D programs• Continue to focus on understanding of the
degradation mechanisms• Coordinate and integrate materials testing
among the materials programs• Increase R&D efforts to address materials
issues in VVER and PHWR
Collaboration, Coordination, and Integration