primary systems corrosion research program update · pwr and vver designs and heavy water reactor...

© 2018 Electric Power Research Institute, Inc. All rights reserved.

Anne Demma Program Manager, EPRI

Jim CirilliPSCR Chairman, Exelon

Technical Exchange Meeting on Materials in Rockville, MD Tuesday, May 22, 2018

Primary Systems Corrosion Research Program Update

May 21, 2018

2© 2018 Electric Power Research Institute, Inc. All rights reserved.

Topics

PSCR Mission, Scope and Membership Recent PSCR Key Deliverables Top Technical Issues for PSCR in 2018-19 2018-19 PSCR Research Focus Areas External and Internal Collaborations Selected Project Updates Upcoming PSCR Meetings in 2018 Conclusion


PSCR Mission, Scope and MembershipMission and Scope• Overall responsibility for Materials testing,

characterization and associated modeling within the 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

PSCR Members 22 US and 19 International

– One new international member: SNPTC in ChinaPSCR Members- 22 US- 18 International

Interactions between PSCR and the other Materials Programs are similar to the way these programs interact with EPRI’s Chemistry Program


Recent PSCR Key Deliverables In 2017, summarized the results for the Advanced Radiation Resistant Materials

Program to develop the next generation of materials for in-core structural components and fasteners. The project performed microstructural, mechanical and stress corrosion cracking studies on proton/ion irradiated alloys to identify promising alloys for further evaluation in neutron irradiation conditions. Revised the Materials Handbook for Nuclear Plant Pressure Boundary

Applications in 2018 which provides materials properties and performance data for pressure boundary materials. Utility engineers can use these mechanical and physical properties data and service performance information to assist them in making informed failure analysis, repair and replacement decisions. Revised the Materials Degradation Matrix in 2018 to update scientific knowledge in

degradation mechanisms for the materials used in Light Water Reactors including BWR, PWR and VVER designs and Heavy Water Reactor primary systems components– Report sent to EPRI Publications in May; expect publication in June– Then Materials Programs will revise the BWR and PWR Issue Management Tables

2018 PSCR planned Deliverables are listed in the additional slides


Top Technical Issues for PSCR in 2018-19

• 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

• Potassium Hydroxide Materials Testing– Perform materials testing to qualify potassium hydroxide to

potentially replace lithium hydroxide in certain PWR designs

• Environmentally Assisted Fatigue Testing and Modeling– Perform testing and develop models to more accurately predict

EAF in light water reactor environments for existing and new plants

IASCC Initiation Test Set-up


2018-19 PSCR Research Focus Areas

RFA Description

Irradiated Materials Testing of Stainless Steels and High-Strength Alloys

Fatigue Testing

Low Alloy Steels Testing

Non-Irradiated Nickel-Base Alloys Testing

Non-Irradiated Testing other than Nickel-Base Alloys (Stainless Steels, CASS,…)

Next Generation of Materials, Irradiation, and Testing Techniques

Materials Strategic Tools and Training

Atom probe tomography (APT) examination of high fluence surveillance specimens


External Collaborations

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 growth

Advanced Radiation Resistant Materials (ARRM) program– To develop the next generation of materials for in-core

structural components and fasteners– Phase 1 with co-funding from DOE-LWRS and KAPL

EPRI-MAI Collaborative Projects on SCC and IASCC

EPRI-CANDU Owners’ Group on PWSCC of Nickel-base Alloys to Carbon Steel Welds in CANDU Primary Heat Transfer System


Collaborations with other 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)


KOH Qualification Project Driven by availability & long term cost issues for Lithium-7

– Increased use of non-isotope-specific Li, e.g. in electric batteries

– Potential geopolitical restrictions on supply of LiOH to USA

– Potential future increased usage of Li in GenIV (molten salt) reactors

– Need for demonstration of an alternative pH-balancing alkali for PWRs

Chemistry and Radiation Safety (2016–2019)– Work is ongoing and will be completed in 2019

Fuel (2016–2020)– Vendor reviews in progress; Loop testing to start in mid-2018 and complete in 2020

Materials Effects (2017–2020)– Crack initiation testing of non-irradiated nickel-base alloy in progress (nominal chemistry)

– Other testing scheduled to begin in 2018

Irradiated Stainless Steel Testing

Non-irradiated Alloy 600 and Stainless Steel Testing

Plan has Qualification effort completed in 2020, with start of Plant Trial in 2021

GAO-13-716, “Managing Critical Isotopes: Stewardship of Lithium-7 Is Needed to Ensure a Stable Supply”, Sep. 2013.

Press Release, House Committee on Science, Space, & Technology, “GAO Raises Questions about Adequate Supply of Lithium-7 for Nuclear Power Reactors”, Oct 9, 2013.


Overview Plan for Qualifying KOH

PSCR and MRP

FRP

Chem

RS

• Initiation & Crack Growth Rate Testing– Non-irradiated testing

• Stainless Steel, Alloy 600– Irradiated testing

• Stainless Steel

• Fuel Vendor Assessment• Experimental Loop Testing • Fuel Exams

• Activation species and dose pathways• Effect on plant radiation fields• Effluent and radioactive waste handling

• System review and impacts• High temperature chemistry (MULTEQ)• Purity specifications• Multiple alkali (Li & K) modeling and

control

• 2017-2020

Schedule

• 2017• 2018-2020• Plant Trial*

• 2017-2019

• 2016-2019

* Plant Trial to follow Qualification (3 cycles of operation with KOH)

Materials Testing

Fuels Testing and Exams

Radiation Fields and Radwaste

Chemistry/pH Control

Qualification Actions Required Prior to Start of Demonstration

On schedule to complete in 2020


IASCC for Ti Containing Stainless Steels in VVERResearch Gaps VVERs use Ti-stabilized austenitic stainless steel (similar to

Type 321) which is affected by the same irradiation-induced degradation mechanisms as stainless steels used in Western PWRs EPRI PWR IASCC database does not include VVER materials Differences between PWR and VVER primary water chemistry

(LiOH vs. KOH) As an example, expected doses of VVER 440 reactor internals

after 40 years of operation:

Core Barrel and Bottom

Core Shroud/Baffle/Former

Protective Tube Unit

Component VVER 440Core Barrel 8 dpa

Basket 15 dpaBaffle 41 dpa


IASCC Data and Models for Ti Containing Stainless Steels in VVER

IASCC Initiation Testing– IASCC Initiation Gaps: initiation data in VVER water at end-

of-life fluence Conduct interrupted slow strain rate (SSRT)

on basket and core barrel materials in VVER water to determine threshold stress

Compare results with PWR water/materials Limited materials from plants

Crack Growth Rate Testing– Compare CGRs of basket and core barrel

materials in VVER (KOH) and PWR (LiOH) environments

Literature review and data comparison on IASCC of Ti-stabilized stainless steel (08Ch18N10T) UJV has completed review EPRI Technical Report to be published in 2018

Conduct IASCC tests of VVER materials in KOH and LiOH environments Testing to begin in 2018 and finish in 2020


Advanced Radiation Resistant Materials (ARRM) Program Long Range Program to test and identify:

– A degradation-resistant alloy or alloys within the current commercial alloy specifications– An advanced alloy with superior degradation resistance

Program Phase 1: 2013-2018– Co-funded by BMPC and by DoE LWRS program– Work conducted at University of Michigan Using proton/ion implantation exposures to rapidly… Assess the relative effects of alloy type/composition on irradiated material properties

Program Phase 2: 2019–2025– Neutron irradiations and post irradiation testing of down-selected alloys from Phase 1 IASCC initiation and crack growth testing Void Swelling evaluations Microstructural characterizations

Program Status– Phase 1 work completed – report in review– Down selected the materials to be included in Phase 2– Phase 2 program proposal submitted into NSUF funding round Passed pre-proposal evaluation Full proposal submitted in May


Rapid Simulation of Void Swelling in Stainless Steels Using Ion Irradiation - Status

– PWR Internals will experience high irradiation doses exceeding 120 dpa at certain locations during long-term operation. – There is a need for data and validated models to predict the degree of irradiation damage, particularly void swelling, in

austenitic stainless steel internals at these elevated doses.– Rapid Simulation Project incorporates neutron-irradiated samples harvested from a PWR as starting material for subsequent

ion irradiation which is a novel approach and validates the ion irradiation technique by benchmarking it against the neutron irradiation.

– Specific samples from FTTs were fabricated and are ready for shipment from Studsvik. Required shipment permits were obtained from Swedish, French and US authorities, and contracts are in place with UM and ORNL for the experimental work.

– Experimental work will start in 2018 and be completed in 2021


EAF Component Testing Project– The difference between predicted component fatigue life and the international LWR fleet

operating experience suggest that the analytical approach to fatigue evaluation based on classically-generated laboratory data does not represent the conditions of components operating in the field.

– Using large, component-type specimens, this program will: Increase the understanding of the effect of light water environments on the fatigue life

and cumulative usage factor (CUF) of plant materials Reconcile the differences between the current CUF methodology results and the

nuclear fleet operating experience with respect to EAF failures Provide the technical basis for an improved CUF analytical methodology for EAF

– Status: Identified Supplemental funders for this project Selecting the laboratory


Upcoming PSCR Meetings in 2018

Event Date Location

PSCR Meeting during the EPRI Nuclear Power Council Meeting Week August 27, 2018 Washington, D.C.

PSCR Technical Meeting October 10-12, 2018 Charlotte, NC


Conclusion

The 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 IntegrationCollaboration, Coordination, and Integration


Together…Shaping the Future of Electricity


2018 Planned DeliverablesTitle Planned Completion Date Deliverable Status

Irradiation-Assisted Stress Corrosion Cracking of Ti-Stabilized Austenitic Stainless Steel: IASCC in VVER Reactor Environments 11/01/2018 On Schedule

Progam on Technology Innovation: Toward Predictive Modeling and Mitigation of Irradiation Assisted Stress Corrosion Cracking of Austenitic Steels in LWR Core Components 12/01/2018 On Schedule

Irradiation-Assisted Stress Corrosion Cracking (IASCC) Susceptibility and Evolution of Microstructure in Austenitic Alloys after Proton Irradiation: 2018 Interim Report 12/01/2018 On Schedule

Materials Handbook for Nuclear Plant Pressure Boundary Applications (2018) 02/25/2018 Completed

Evaluation of Microstructure and Performance for Alloy 600TT Steam Generator Tubing 12/01/2018 On Schedule

Relevance of Long Range Ordering in Nickel-Chromium-Iron Alloys to Pressurized Water Reactors 11/01/2018 On Schedule

Atom Probe Tomography Round Robin on Irradiated Stainless Steels 12/01/2018 On Schedule

Characterization and of the Dilution Zone of Alloy 690/52/Carbon Steel Welds of CANDU Feeder Pipe Flow Elements 12/01/2018 On Schedule

Crack Initiation of Alloy 600 in PWR Water 12/01/2018 On Schedule

primary systems corrosion research program update · pwr and vver designs and heavy water reactor...

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