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

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

Mission/PurposePSCR

3© 2017 Electric Power Research Institute, Inc. All rights reserved.

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


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


List of PSCR Members

All EPRI Nuclear Full Members are PSCR Members- 22 US- 18 International

PSCR


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


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)


PSCR Publishes and Updates the MDMExample of MDM Results


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


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


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.


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.


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


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)


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


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


Together…Shaping the Future of Electricity

primary systems corrosion research program

Documents