ML20245E186
| ML20245E186 | |
| Person / Time | |
|---|---|
| Issue date: | 09/30/2020 |
| From: | Harris B, Raj Iyengar, Jeffrey Poehler, Wendy Reed Office of Nuclear Regulatory Research |
| To: | |
| B. Harris | |
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| RIL 2020-09 | |
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Text
RIL 2020-09 INTERNATIONAL WORKSHOP ON ADVANCED NONLIGHT-WATER REACTOR MATERIALS AND COMPONENT INTEGRITY Date Published: September 2020 Prepared by:
B. Harris W. Reed J. Poehler R. Iyengar Research Information Letter Office of Nuclear Regulatory Research
Disclaimer This report was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any employee, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third partys use, or the results of such use, of any information, apparatus, product, or process disclosed in this publication, or represents that its use by such third party complies with applicable law.
This report does not contain or imply legally binding requirements. Nor does this report establish or modify any regulatory guidance or positions of the U.S. Nuclear Regulatory Commission and is not binding on the Commission.
ABSTRACT On December 9-11, 2019, the U.S. Nuclear Regulatory Commission (NRC) met with representatives from the international nuclear community to discuss the state of knowledge, operating experience, and research activities related to high-temperature metallic and nonmetallic materials, coolant chemistry, reactor component integrity, and applicable codes and standards. In addition, members of the public participated by teleconference.
Engineers and scientists from the United States, United Kingdom, Japan, Canada, China, Malaysia, Italy, Turkey, and the Netherlands attended the workshop. Organizations represented at the workshop included regulators, national research organizations, national laboratories, vendors, and universities.
All presentations and information related to this public meeting (e.g., meeting notice, agenda, and presentations) are available in the NRCs Agencywide Documents Access and Management System (ADAMS), Accession No. ML20030B755.
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TABLE OF CONTENTS ABSTRACT ------------------------------------------------------------------------------------------- iii ABBREVIATIONS AND ACRONYMS --------------------------------------------------------- IX DAY 1 PRESENTATIONS ----------------------------------------------------------------------- 1-1 Session 1: Vendor Overview ------------------------------------------------------------------------------- 1-1 Salt Composition, Corrosion Testing, and Tritium Control for the Kairos Power Fluoride-Cooled High-Temperature Reactor ------------------------------------------------------ 1-1 Overview of the Liquid Fluoride Thorium Reactor: Materials and Components ---------- 1-1 Continuing High-Temperature Gas Reactor Development in the United StatesFamily of Framatome High-Temperature, Gas-Cooled Reactors -------------------------------------- 1-2 Nonlinear Mechanical Modeling in Materials Subject to Property Changes due to Neutron Irradiation 1-3 Materials Selection and Development for Advanced Reactors at TerraPower ------------ 1-3 Session 2: Technical and Research ActivitiesOverview A ------------------------------------ 1-4 U.S. Nuclear Regulatory Commission Readiness for Licensing Advanced Reactors Materials and Component Integrity Research ----------------------------------------------------- 1-4 Overview of Office for Nuclear Regulation Activities for Regulating Advanced Modular ReactorsA Structural Integrity Perspective ------------------------------------------------------ 1-5 Reviewing Innovative Reactor DesignsA Canadian Nuclear Safety Commission Perspective 1-5 Overview of Materials Research at the U.S. Department of Energys Office of Nuclear Energy 1-6 International Atomic Energy Agency Activities on Fast Reactor Technology 1-7 Session 3: Technical and Research ActivitiesOverview B ------------------------------------ 1-8 Overview of Reactor Materials Investigation with Modeling and Experimental Studies at Canadian Nuclear Laboratories --------------------------------------------------------------------- 1-8 Research and Development Activities for Advanced Reactors at the Japan Atomic Energy Agency ------------------------------------------------------------------------------------------ 1-8 Material and Properties Development Needs for Design and Development of Commercially Viable Gen IV Reactors ------------------------------------------------------------ 1-9 Advanced Reactor Technology Irradiation Programs at the Nuclear Research and Technology Group --------------------------------------------------------------------------------- 1-9 Structural Materials Research within the Gen IV International Forum ----------------------1-10 DAY 2 PRESENTATIONS ----------------------------------------------------------------------- 2-1 Session 1: Graphite Materials A -------------------------------------------------------------------------- 2-1 Applying Brittle Materials Design Concepts in the Manufacture of Graphite and Ceramics for 21st Century Nuclear Reactors 2-1 United Kingdom Experience of Graphite-Moderated Reactors 2-1 American Society of Mechanical Engineers Considerations for High-Temperature Reactor Graphite and Composite Components 2-2 Graphite Research and Qualification Activities at Nuclear Research and Technology Group 2-3 v
Session 2: Graphite Materials B -------------------------------------------------------------------------- 2-3 Understanding Graphite Behavior in Nuclear Reactor Environments for Lifetime Predictions 2-3 The Chemistry of Graphite in Fluoride-Salt-Cooled, High-Temperature Reactors and Molton Salt Reactors 2-4 Graphite Electrode Behaviors and Its Application for Salt Purification 2-4 Session 3: Materials Qualification Challenges A ---------------------------------------------------- 2-5 Metallurgical Challenges Associated with Using Grade 91 Steels at Elevated Temperature 2-5 Advanced Structural Materials for Nonlight-Water Reactors 2-6 Challenges in Qualifying Advanced Manufacturing Technologies for High-Temperature Nuclear Service 2-6 Qualification of Materials for Elevated Temperature Nuclear Components 2-7 Session 4: Materials Qualification Challenges B ---------------------------------------------------- 2-7 Challenges with American Society of Mechanical Engineers Code Qualifying Graphite Irradiation Effects in Test Reactors 2-8 Potential Material Issues for the Canadian Nuclear Safety Commission To License Advanced Reactors 2-8 Realizing the Nuclear Materials Discovery and Qualification Initiative 2-9 DAY 3 PRESENTATIONS ----------------------------------------------------------------------- 3-1 Session 1: Inspection, Monitoring, and Surveillance ---------------------------------------------- 3-1 Development of the Technical Basis for In Situ, Passive Surrogate Materials Surveillance for Advanced Nonlight-Water Reactors 3-1 Progress Toward Bridging Harsh Environment Online Monitoring Gaps for Advanced Reactors 3-2 A Vision for Advanced Nondestructive Examination Methods for Nonlight-Water Reactor Components 3-2 Sodium-Cooled Fast ReactorsJapans Experience and Future 3-3 Nondestructive Evaluation in Advanced Reactors ----------------------------------------------- 3-4 Session 2: Molten Salt Chemistry ------------------------------------------------------------------------ 3-4 Research Reactor Experiments To Study Materials and Fuel Salt Performance --------- 3-4 Molten Fluoride Salt Chemistry 3-5 Fundamental Properties and Removal of Impurities in Molten Salt Systems -------------- 3-6 Fluoride Salt Chemistry and Properties 3-6 Online Monitoring of Molten Salt Reactors --------------------------------------------------------- 3-7 Session 3: Metallic Materials: Environmental Effects --------------------------------------------- 3-8 Overview of Environmental Issues and Material Property Gaps for Commercial Viability of Advanced Reactors 3-8 Environmental Effects in Liquid Metal Systems 3-9 Environmental Effects in High-Temperature, Gas-Cooled Reactor Environments 3-10 An Overview of Molten Salt Corrosion Research at ORNL -----------------------------------3-11 WORKSHOP
SUMMARY
4-1 Overview of Vendor Designs (Monday, December 9, Session 1) ------------------------------- 4-1 vi
Technical and Research Activities (Monday, December 9, Sessions 2 and 3) ------------- 4-1 Graphite (Tuesday, December 10, Sessions 1 and 2) ----------------------------------------------- 4-2 Materials Qualification (Tuesday, December 10, Sessions 3 and 4) --------------------------- 4-3 Inspection and Monitoring (Wednesday, December 11, Session 1) --------------------------- 4-3 Molten Salt Chemistry (Wednesday, December 11, Session 2) --------------------------------- 4-4 Environmental Effects on Metallic Materials (Wednesday, December 11, Session 3) ---- 4-5 APPENDIX A WORKSHOP ATTENDEES ------------------------------------------------------- A-1 APPENDIX B PRESENTATION SLIDES -------------------------------------------------------------- B-1 vii
ABBREVIATIONS AND ACRONYMS ACRS Advisory Committee for Reactor Safeguards ADAMS Agencywide Documents Access and Management System AGR advanced gas-cooled reactor AID Assessment Integration Division ANL Argonne National Laboratory ANLWR advanced nonlight-water reactor ASME American Society of Mechanical Engineers B&PV boiler and pressure vessel BEIS Department for Business, Energy and Industrial Strategy BYU Brigham Young University C Celsius CH4 methane CHX compact heat exchanger CNL Canadian Nuclear Laboratories CNSC Canadian Nuclear Safety Commission CNWRA Center for Nuclear Waste Regulatory Analyses CO carbon monoxide CO2 carbon dioxide CRP Coordinated Research Projects DOE U.S. Department of Energy DOE U.S. Department of Energy Basic Energy Sciences EPRI Electric Power Research Institute FHR fluoride-salt-cooled, high-temperature reactor FLiBe lithium fluoride and beryllium fluoride mixture FLiNaK lithium fluoride, sodium fluoride, potassium fluoride mixture Gen IV Generation IV GIF Generation IV International Forum H2 hydrogen H2O water HFR High Flux Reactor HTGR high-temperature gas reactor HTR high-temperature reactor IAEA International Atomic Energy Agency INL Idaho National Laboratory JAEA Japan Atomic Energy Agency JSME Japan Society of Mechanical Engineers LFTR Liquid Fluoride Thorium Reactor LMR liquid metal cooled reactor LWR light-water reactor MIT Massachusetts Institute of Technology MSR molten salt reactor MWe Megawatt electric MWt megawatt thermal NDE nondestructive evaluation ix
NMDQi Nuclear Materials Discovery and Qualification Initiative NNC National Nuclear Corporation NRC U.S. Nuclear Regulatory Commission NRG Nuclear Research and Technology Group NRU National Research Universal ONR Office for Nuclear Regulation (UK)
ORNL Oak Ridge National Laboratory OSU Ohio State University PNNL Pacific Northwest National Laboratory R&D research and development RES Office of Nuclear Regulatory Research RPI Rensselaer Polytechnic Institute SFR sodium fast reactor SMR small modular reactor TWR Traveling Wave Reactor U uranium UC University of California UK United Kingdom USNIC United States Nuclear Industry Council VHTR very-high-temperature reactor Virginia Tech Virginia Polytechnic Institute and State University x
DAY 1 PRESENTATIONS Session 1: Vendor Overview Several vendors summarized their proposed reactor designs and provided details on materials for structural and internal applications and testing to address data and knowledge gaps for these materials.
Salt Composition, Corrosion Testing, and Tritium Control for the Kairos Power Fluoride-Cooled High-Temperature Reactor George Young, Kairos Power Dr. George A. Young is the Manager of Structural Materials at Kairos Power. Formerly, he was a consultant scientist and manager at the Knolls Atomic Power Laboratory. Dr. Young has a B.S. in Materials Engineering from Rensselaer Polytechnic Institute and an M.S. and Ph.D. in Materials Science from the University of Virginia. He has over 25 years of experience in the nuclear power industry and is an expert in materials selection and performance for both conventional and advanced nuclear power systems. Dr. Young is also a faculty research associate at Oregon State University, an adjunct professor at Clarkson University, and author of over 50 peer-reviewed articles and book chapters.
Presentation Summary Dr. Young gave an overview of the Kairos Power high-temperature, fluoride-cooled reactor being designed and built with 316H and its weld filler metal 16-8-2 as its primary structural materials (ADAMS Accession No. ML20030B777). As shown in the Molten Salt Reactor Experiment, a lithium fluoride and beryllium fluoride mixture (FLiBe) salt is compatible with both iron and nickel-based alloys when impurity elements in FLiBe are properly controlled. This presentation provided an overview of Kairos Powers efforts on salt purification, corrosion testing, and tritium management. Dr. Young compared historic FLiBe compositional guidelines to Kairos Power targets and discussed the benefits of redox control. He then summarized the results to date on the corrosion of 316 stainless steel in FLiBe and developmental work on tritium barrier coatings.
Overview of the Liquid Fluoride Thorium Reactor: Materials and Components Kurt Harris, Flibe Energy Dr. Kurt Harris received his Ph.D. in Mechanical Engineering from Utah State University, where his research focused on advanced measurements systems and materials. With experience at Idaho National Laboratory (INL), Sandia National Laboratories, and Westinghouse, his interests in advanced nuclear energy technologies as a solution to energy poverty led Dr. Harris to join Flibe Energys growing team.
Presentation Summary Dr. Harris provided an overview on Flibe Energys development of the Liquid Fluoride Thorium Reactor (LFTR), a two-fluid thermal spectrum molten salt reactor (MSR) that fissions uranium 1-1
(U)-233 to generate thermal energy and breed replacement fuel (ADAMS Accession No. ML20030B768). Materials of interest include thorium as fuel, U-233 as the fissile inventory catalyst, and graphite as a moderator and to provide flow channels. Other materials of interest include Hastelloy N metal for salt-contacting vessels and piping, FLiBe-based salts with highly depleted lithium, and supercritical carbon dioxide (CO2) as the working fluid in the power conversion system.
Continuing High-Temperature Gas Reactor Development in the United States Family of Framatome High-Temperature, Gas-Cooled Reactors Farshid Shahrokhi, Framatome Inc.
Dr. Farshid Shahrokhis entire professional career has been in the nuclear power industry, beginning at the Oak Ridge National Laboratory (ORNL) as a graduate student and an adjunct research assistant. He then joined Framatome 39 years ago, where he has held a variety of technical leadership positions. He oversees Framatomes development of the high-temperature gas reactor (HTGR) concept. Previously, as the safety and licensing manager, he established the licensing strategy for Framatome HTGRs. He led the HTGR component design and various high-temperature material selection tasks. Under his technical leadership, Framatome completed the design of a multiloop helium test facility capable of simulating high-temperature conditions in the reactor environment to test HTGR materials and components. Dr. Shahrokhi is currently the Director of HTGR Technology at Framatome.
Presentation Summary Dr. Shahrokhi proposed developing a prismatic HTGR and provided an overview of the basis for selecting this technology (ADAMS Accession No. ML20030B769). The Framatome reference SC-HTGR is a modular, graphite-moderated, helium-cooled, high-temperature reactor (HTR) with a nominal thermal power of 625 megawatts thermal (MWt) and a nominal electric power capability of 272 megawatts electric (MWe). It produces high-temperature steam suitable for numerous applications, including industrial process heat and high-efficiency electricity generation. The safety profile of the SC-HTGR allows it to be collocated with industrial facilities that use high-temperature steam. This can open a major new avenue for nuclear power use.
The modular design allows plant size to be matched to a range of applications.
The SC-HTGR concept builds on Framatomes past experience of HTGR projects, as well as on the development and design advances that have taken place in recent years for modular HTGRs. The overall configuration takes full advantage of the work performed on early modular HTGR concepts, such as the General Atomics MHTGR and the HTR-MODUL.
The design is scalable from the reference 625 MWt down to 315 MWt, 180 MWt, 54 MWt, and 7MWt, using the same reference fuel blocks arranged in annular or standard core configurations. The fuel form and the entire primary system components and materials are common in all scaled variances of the reference plant. The secondary side of the plant uses the Rankin cycle configuration for electricity production with or without the high-temperature process steam distribution option. Only the smallest variant (7 MWt) uses the Brayton cycle for electricity production.
The Framatome family of HTGRs uses currently available materials to manufacture all major components. Framatome presented a component-by-component list of key materials.
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Framatome is currently in discussions with potential agencies, investors, and early adopters for funding opportunities to build the first commercial-scale plant to demonstrate design completion, licensability, cost, and schedule.
Nonlinear Mechanical Modeling in Materials Subject to Property Changes due to Neutron Irradiation Paul Kirchman, X-Energy Mr. Paul Kirchman is a graduate of Virginia Polytechnic Institute and State University (Virginia Tech) with a B.S. in Engineering. He spent much of his career as an aerospace structural analyst at the National Aeronautics and Space Administrations Goddard Space Flight Center, working on projects including space shuttle payloads, geosynchronous weather satellites, and high-altitude arctic balloon instruments. Mr. Kirchman specializes in developing unique approaches to technically challenging mechanical analysis problems, including work for NASTRAN as a developer. He is presently working at X-Energy on nonlinear material modeling for high-radiation environments.
Presentation Summary Mr. Kirchman discussed graphite structural changes and described how finite element models were developed and studied to examine predicted structural changes in graphite exposed to irradiation (ADAMS Accession No. ML20030B797). The X-Energy reactor design is a pebble bed reactor. Mr. Kirchman presented the technical aspects of nuclear reflector blocks undergoing significant changes in physical properties due to radiation damage over time. The radiation-induced changes in component size are significant and need to be accounted for in tolerance stacking for mechanical analysis and bypass flows for computational fluid dynamic analysis. Materials properties test data, along with temperature and fluence fields, are put into the finite element model using scripts developed in the open source Python programming language. Mr. Kirchman presented example results of the finite element model on graphite reference blocks.
Materials Selection and Development for Advanced Reactors at TerraPower Greg Vetterick, TerraPower Dr. Greg Vetterick is a principal materials engineer at TerraPower, responsible for material selection, testing, and qualification for both the sodium-cooled, metallic-fueled Traveling Wave Reactor (TWR) and the liquid-salt-cooled/fueled Molten Chloride Fast Reactor. Before joining TerraPower, he received his Ph.D. in Materials Science from Drexel University and his B.S. and M.S. in Materials Engineering from Iowa State University.
Presentation Summary Dr. Vetterick presented information on both the TWR and Molten Chloride Fast Reactor designs (ADAMS Accession No. ML20030B789). TerraPower is developing advanced reactor designs, including the sodium-cooled TWR, metallic-fueled TWR, and the liquid-salt-cooled/fueled Molten Chloride Fast Reactor. A common feature of these reactor projects is the need to base the design on materials for which a considerable amount of data is available. By using established materials, designers hope to mitigate the technical, programmatic, and regulatory risks of getting a new reactor to market. Consequently, new reactor designs do not 1-3
take advantage of the potential benefits of using more advanced materials. This presentation covered the decisionmaking process for materials selection, the steps that TerraPower is taking to fill in the gaps in the existing database to make the case for licensing its reactor designs, and the expectations that are placed on new materials for use in a first-of-a-kind reactor plant.
Session 2: Technical and Research ActivitiesOverview A These presentations summarized research programs related to advanced reactor materials development underway at major national and international organizations.
U.S. Nuclear Regulatory Commission Readiness for Licensing Advanced ReactorsMaterials and Component Integrity Research Raj Iyengar, NRC Office of Nuclear Regulatory Research Dr. Raj Iyengar is the NRCs Branch Chief for the Component Integrity Branch, Division of Engineering, Office of Nuclear Regulatory Research (RES). He has worked at the NRC since 2009 as an RES Branch Chief, senior materials engineer, and technical assistant; an executive technical assistant in the Office of the Executive Director for Operations; and a project manager in the Office of Nuclear Material Safety and Safeguards. Before joining the NRC, Dr. Iyengar held corporate management positions in the automotive industry, where he led product development and applications efforts, and research positions at BattelleColumbus and the University of Pennsylvania. Dr. Iyengars publications cover topics in reactor component integrity, materials degradation, development of advanced high-strength steels, structural optimization, and computational fracture mechanics.
Dr. Iyengar holds a Ph.D. in Solid Mechanics and an Sc. M. in Applied Mathematics from Brown University, an M.S. in Mechanics and Materials Science from Rutgers University, and an M.S. in Metallurgy from the Indian Institute of Science.
Presentation Summary Dr. Iyengar provided an overview of the Executive Director for Operations initiative to make the NRC a more modern, risk-informed regulator and the need for the agency to rapidly increase its knowledge base in advanced nonlight-water reactors (ANLWRs), such as the need for tools related to neutronics, fuels, and severe accidents (ADAMS Accession No. ML20030B780). He discussed leveraging the expertise of the U.S. Department of Energy (DOE) to minimize costs of regulatory research and addressed the impacts of advanced manufacturing techniques on ANLWRs. Dr. Iyengar presented an overview of work being done in RES, including the potential endorsement of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, and tools needed for the probabilistic fracture mechanics of graphite. Dr. Iyengar also discussed the evaluation of creep and creep-induced fatigue as potential life-limiting factors for components in HTRs. Dr. Iyengar explained the work performed on high-temperature corrosion and molten salt compatibility, along with technical challenges and opportunities.
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Overview of Office for Nuclear Regulation Activities for Regulating Advanced Modular ReactorsA Structural Integrity Perspective Russ Green, United Kingdom Office for Nuclear Regulation Mr. Russ Green started his career after graduating from the University of Liverpool with a bachelors degree in Metallurgy and Materials Science in 2002. He joined the United Kingdoms (UKs) National Nuclear Corporation (NNC) as a graduate technical consultant and studied part time to achieve a postgraduate masters degree in Corrosion Control Engineering from the University of Manchester. He had a career of more than 11 years at NNC (taken over by global engineering consultancy AMEC in 2005). Mr. Green worked across the United Kingdoms nuclear industry, predominantly on advanced gas-cooled reactor (AGR) and Magnox sites but also including Sellafield experimental accelerator sites. He became a chartered engineer and a professional member of the Institute of Materials, Minerals and Mining in 2010. In 2013, Mr. Green left the private consultancy market and joined the UK Office for Nuclear Regulation (ONR), training as a nuclear safety inspector and specializing in structural integrity. For 5 years, he worked across the United Kingdom operating reactor fleet, supporting planned and emergent site inspections, interventions, and investigations. In 2018, he moved over to ONRs New Reactors Division as the structural integrity lead for advanced nuclear technologies, supporting the UK-government-funded project on small modular reactors (SMRs) and nonlight-water SMR technologies.
Presentation Summary Mr. Green provided an overview of ONR activities for regulating advanced modular reactors (ADAMS Accession No. ML20030B783). In December 2017, the UK governments Department for Business, Energy and Industrial Strategy (BEIS) launched its Nuclear Sector Deal as part of the national Clean Growth Strategy. One of the objectives is to set out a new framework to support the development and deployment of SMRs and the innovative technologies that support them. As part of this objective, BEIS is conducting a feasibility and development project, inviting eight advanced modular reactor vendors to submit proposals to receive a share of £40 million (about $50 million) of funding to support the development of their designs.
To support this initiative, ONR was tasked with developing a focused program to review these designs from a nuclear safety perspective, with respect to the feasibility of potential future licensing in the United Kingdom. Fundamental to achieving this aim was the requirement for ONR to build technical capability and capacity related to Generation IV (Gen IV) technologies, focused on liquid metal fast reactor, HTGR, and MSR technologies.
Reviewing Innovative Reactor DesignsA Canadian Nuclear Safety Commission Perspective Stephen Cook, Canadian Nuclear Safety Commission Mr. Stephen Cook is currently a lead integrator for the Assessment Integration Division (AID) within the Directorate of Assessment and Analysis at the Canadian Nuclear Safety Commission (CNSC). The AID staff coordinates and integrates specialist input for planning, implementation, and technical assessment of regulatory safety and control areas as applied to refurbishments, vendor design reviews, proposed new nuclear power plants, and new major nuclear facilities in Canada.
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Mr. Cook is a Professional Engineer with a bachelors degree in Mechanical Engineering from Memorial University of Newfoundland in 1989 and a masters degree in Nuclear Engineering from McMaster University in 2008. He joined CNSC in 2000. His CNSC experience includes 8 years in his current position with AID as well as work with the Personnel Certification Division as an examination officer, the Research Facilities Division as a project officer, and the Systems Engineering Division as a specialist responsible for the regulatory oversight of nuclear power plant maintenance programs.
Previously, Mr. Cook spent 7 years at the Chalk River Nuclear Laboratories National Research Universal (NRU) research reactor and completed the senior reactor shift engineers training program. Before working with NRU, he spent 3 years at New Brunswick Powers Point Lepreau Nuclear Generating Station as a system engineer for the nuclear auxiliary systems and fire protection systems.
Presentation Summary Mr. Cook gave the CNSC perspective on reviewing innovative reactor designs (ADAMS Accession No ML20030B761). This presentation provided a high-level overview of CNSCs mandate, regulatory framework, and readiness preparations for the regulation of advanced reactor designs (SMRs). It included a status of CNSCs prelicensing vendor design reviews, as well as some of the challenges that CNSC recognizes in terms of demonstrating safety claims supporting novel designs. He followed this with an overview of CNSCs internal research and support program for the development of regulatory positions.
Overview of Materials Research at the U.S. Department of Energys Office of Nuclear Energy Sue Lesica, U.S. Department of Energy Ms. Susan Lesica is a materials engineer in the DOE Office of Materials and Chemical Technologies. She plans and implements materials research relating to the Office of Nuclear Energys missions, including developing advanced nuclear fuel cycle technologies and the next generation of nuclear reactors.
Ms. Lesica is a graduate of the University of Texas at Austins Materials Science and Engineering masters program and Rensselaer Polytechnic Institutes Materials Engineering bachelors program. She began her career in 1995 as a participant in the DOE Technical Leadership Development Program. Through the program, she worked at DOE headquarters, Rocky Flats Environmental Technology Site, Nevada Operations Office, and Argonne National Laboratory (ANL), supporting the Offices of Nuclear Energy, Environmental Management, Defense Programs, and Science, respectively.
Ms. Lesica returned to the Office of Nuclear Energy to fill a variety of positions, including manager of a project to prepare an environmental impact statement, project engineer for the electrochemical treatment of sodium-bonded spent nuclear fuel, and project manager for advanced waste forms development and nuclear materials research.
Presentation Summary Ms. Lesica provided an overview of DOEs advanced materials research and development program supporting advanced reactors (ADAMS Accession No. ML20030B762). DOE focuses 1-6
on providing the technical basis needed to support the regulatory requirements for structural materials required for fast reactors, gas-cooled reactors, MSRs, and microreactors that could be deployed in the near-to-midterm horizon. Ms. Lesica discussed specific research performed at various national laboratories, along with funded Nuclear Energy University Programs and Small Business Innovation Research projects. Research on gas-cooled reactors is focused on high-temperature design methodologies, graphite, and data management to maintain a Gen IV materials handbook. Focus areas for fast reactor research include qualifying new materials such as Alloy 709 and qualifying existing material (Grade 91) to support licensing and long-term planning. MSR research focuses on salt and materials interactions, materials surveillance development, and materials under ASME B&PV Code,Section III, Division 5, Class B. Lastly, microreactor research focuses on the core block code case for heat pipe reactors and elevated temperature cyclic properties of advanced manufacturing materials.
International Atomic Energy Agency Activities on Fast Reactor Technology Vladimir Kriventsev, International Atomic Energy Agency Dr. Vladimir Kriventsev is Team Leader of the Fast Reactor Technology Development Team of the International Atomic Energy Agency (IAEA). After graduating from the Moscow Engineering Physics Institute in 1984, Dr. Kriventsev worked in the field of nuclear engineering and fast reactor technology in various international technical organizations. He participated in design studies, thermal-hydraulic analysis, and safety assessments of the BN-600, BN-800, Monju, JSFR, ESFR, Astrid, MYRRHA, and other fast reactors.
Dr. Kriventsev, who has worked at IAEA since 2016, serves as scientific secretary for the IAEA activities related to fast reactors, such as Coordinated Research Projects (CRPs), topical technical meetings, and education and training workshops.
Dr. Kriventsev obtained his Ph.D. from the Obninsk Institute for Nuclear Engineering in 1994 and a Doctor of Engineering from Tokyo Institute of Technology in 1999. His fields of interest include turbulence modeling, computational fluid dynamics, fast reactor thermal-hydraulics, safety analysis, and simulation of severe accidents in fast reactors, including core disruptive accidents.
Presentation Summary Dr. Kriventsev presented IAEA activities on fast reactor technologies (ADAMS Accession No. ML20030B770). He gave an overview of the IAEA Technical Working Group on Fast Reactors role in providing guidance and a link between IAEA and the international community.
He discussed various CRPs on fast reactor technology, such as the neutronics benchmark of the China Experimental Fast Reactor. The China Experimental Fast Reactor is a sodium-cooled fast reactor (SFR) with a nominal power of 65 MWt or 20 MWe that reached criticality in 2010. This reactor generated electricity at 100-percent power in December 2015 and operated for more than 40 effective full-power days. Dr. Kriventsev also described a new CRP on a benchmark analysis of fast flux test facility loss of flow without scram. He concluded his discussion by providing information on various upcoming international technical workshops and advised the participants on ways to stay engaged within the international community on standards for salt purity and composition.
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Session 3: Technical and Research ActivitiesOverview B These presentations are continuations from Session 2, summarizing research programs related to the development of advanced reactor materials underway at major national and international organizations.
Overview of Reactor Materials Investigation with Modeling and Experimental Studies at Canadian Nuclear Laboratories Mike Welland, Computational Techniques, Canadian Nuclear Laboratories Dr. Mike Welland graduated with a Ph.D. in Nuclear Engineering from the Royal Military College of Canada and completed research fellowships at ANL and the Institute for Transuranium Elements in Germany. He is the head of Mesoscale and Transport Methods in the Computational Techniques Branch at Canadian Nuclear Laboratories (CNL) and primary investigator for multiple projects, including Fundamental Mechanisms of Fuel and Cladding Performance and the Clean Energy Demonstration, Innovation, and Research park.
His work includes applying mesoscale models to processes in the nuclear and energy industries to enhance the understanding, safety, and efficiency of related technologies. His research interests include thermodynamics; transport processes; and mesoscopic phenomena in nuclear fuels, structural materials, and nanoparticles for energy storage.
Presentation Summary Dr. Welland provided an overview of reactor materials investigations with modeling and experimental studies at CNL (ADAMS Accession No. ML20031D418). Reactor and advanced materials research activities at CNL contribute to a diverse range of areas, including reactor sustainability, SMR and Gen IV reactor technologies, and hydrogen storage. CNL is currently leveraging opportunities through a harvest of reactor and structural components from the NRU reactor to characterize properties of materials that have been irradiated in an operating reactor environment to an unprecedented, but well-recorded, degree. Projects underway codevelop multiscale modeling tools with advanced material characterization techniques to enhance the value of these and other investigations. He reviewed CNLs work in materials modeling and experiments, focusing on recent successes such as work on material properties and behavior of molten salts, micromechanical testing on reactor components, hydriding of zirconium alloys, and analysis of harvested irradiated reactor components.
Research and Development Activities for Advanced Reactors at the Japan Atomic Energy Agency Tai Asayama, Japan Atomic Energy Agency Dr. Tai Asayama is Deputy Director General of the Fast Reactor Cycle System Research and Development Center of the Japan Atomic Energy Agency (JAEA). He has more than 30 years of experience evaluating elevated temperature materials and developing design methodologies for fast reactors. He is engaged in codes and standards development in the Japan Society of Mechanical Engineers (JSME) and, until 2017, chaired two working groups of the ASME B&PV Committee and led the development of Code Case N-875, which provides alternative inservice inspection requirements for liquid metal reactors (LMRs) under the System-Based Code concept.
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Presentation Summary Dr. Asayama provided an overview of ongoing research and development activities on fast reactors and HTGRs at JAEA, focusing on materials and structural integrity (ADAMS Accession No. ML20031D423). Activities for fast reactors include the development of elevated temperature design methodologies and fitness-for-service evaluation schemes that could be applied to a wide spectrum of next generation fast reactors. With regard to HTGRs, he presented research and development (R&D) accomplishments associated with the high-temperature test reactor and future R&D plans on high-burnup fuel.
Material and Properties Development Needs for Design and Development of Commercially Viable Gen IV Reactors Mike Burke, Electric Power Research Institute Dr. Mike Burke is a technical executive at the Electric Power Research Institute (EPRI) with responsibility for near-core and structural materials for operating and advanced nuclear plants.
Before joining EPRI in 2017, Dr. Burke was the consulting engineer for materials at Westinghouse Electric Company. In 2011 and 2012, he was a professor of nuclear materials at the University of Manchester (UK) and technical director of the UK Nuclear Advanced Manufacturing Research Center. Before these appointments, he had a career of over 30 years at the Westinghouse research center, where he managed programs in the development and testing of materials for fossil and nuclear power plants. His doctoral degree is in Metallurgy from the University of Sheffield (UK), and his M.S. is in Metallurgical Engineering from the University of Pittsburgh.
Presentation Summary Dr. Burke presented an overview of material properties development needs for the design and development of commercially viable Gen IV reactors (ADAMS Accession No. ML20031D420).
Dr. Burke discussed the extensive amount of data describing the behaviors of established light-water reactor (LWR) materials in water and irradiation environments, along with postmortem investigations on reactor surveillance capsules and materials harvested from operating reactors. He suggested that, while there are some simple materials properties to support concepts for advanced reactors, many gaps need to be overcome in supporting advanced reactor technologies. Dr. Burke discussed three categories needed in materials development. The first is the need for materials properties to support initial design and attain ASME code acceptance for construction, to include high-temperature and time-dependent properties. Next is the effect of realistic levels of neutron irradiation and the response of microstructural property stability. The last category is the materials response in realistic environments and effects on mechanical behavior.
Advanced Reactor Technology Irradiation Programs at the Nuclear Research and Technology Group Uazir Bezerra de Oliveira, Nuclear Research and Technology Group Dr. Uazir Bezerra de Oliveira is a materials engineer with 20 years of experience in diverse industries such as aerospace, metallurgy, and oil and gas. He has an M.Sc. from Chalmers University (Sweden) in high-temperature corrosion and a Ph.D. from Groningen University (Netherlands) in direct energy deposition. During his career, he has gained experience in 1-9
various roles such as project manager for product development (Sandvik Steel, Tata), technical authority for subsea materials (Totals Kaombo project at Heerema Marine Contractors) and product manager for modern steel converters (Danieli Corus). In September 2019, he joined the Nuclear Research and Technology Group (NRG) as program manager for MSR and in the project management arena for Blackstone. (Blackstone carries out graphite irradiation experiments to extend the life of EDF [Électricité de France] Energys AGRs.)
Presentation Summary Dr. Oliveira provided an overview of advanced reactor technology irradiation programs at NRG (ADAMS Accession No. ML20031D425). He provided insights on the operations and technical capabilities at the High Flux Reactor (HFR) in Petten in the Netherlands. The HFR and adjacent hot cell facilities allow for research and qualification activities in support of ANLWR development. For more than 60 years, a wide range of funded and commercial programs has studied the effects of neutron irradiation on materials and fuels. Dr. Oliveira discussed recent and current R&D programs and bilateral qualification projects on non-LWRs. Examples of those programs include the characterization of modern graphite grades for HTR applications, an MSR irradiation program, graphite qualification for the lifetime extension of advanced gas-cooled reactors for EDF Energy, and HTR fuel qualification for Tsinghua Universitys Institute of Nuclear and New Energy Technology. The presentation highlighted the research capabilities that are necessary and available at NRG to facilitate materials and fuels research and described different irradiation programs, with a focus on ANLWRs.
Structural Materials Research within the Gen IV International Forum Bill Corwin, Advanced Reactor Materials LLC Mr. Bill Corwin graduated with a B.S. and M.S. in Materials Science and Metallurgy from the Massachusetts Institute of Technology (MIT) and is currently president of Advanced Reactor Materials LLC. Mr. Corwin is Chair and DOEs Representative on the Gen IV International Forums (GIFs) Very High Temperature Reactor [VHTR] Project Management Board.
Previously, Mr. Corwin was DOEs Materials Technology Lead for the Advanced Reactor Technology Program. He managed LWR and advanced reactor materials R&D at ORNL for 37 years and served as National Technical Director for Materials for the Gen IV Reactor Program.
Presentation Summary Mr. Corwin discussed structural materials research for advanced HTRs, which is conducted within the framework of the GIF (ADAMS Accession No. ML20031D421). GIF was established in 2003 with the explicit purpose of helping to develop and share interdisciplinary types of R&D needed to design and qualify HTRs. Currently, 11 countries and the European Union participate in GIF.
GIF includes six advanced reactor systems cooled by a variety of high-temperature coolants (helium, liquid sodium, molten salt, liquid lead (or lead-bismuth), and supercritical water). The multiple Gen IV reactor systems are envisioned to be safer, more efficient, more economical, and more proliferation resistant than the current fleet of LWRs, but since they all operate in the creep range of structural alloys, inelastic design methods and materials must be qualified for their construction. Additionally, designs must also address environmental challenges associated with the compatibility of structural materials with the various coolants and the 1-10
potentially higher doses of irradiation. Finally, many of the HTRs will likely include nonmetallic structural materials (e.g., graphite and ceramic composites) for core and internal components, so those materials and their design rules must also be developed and qualified.
GIF has formed specific committees that address the various types of HTRs being developed. Within those committees, international project arrangements have been formally developed to provide the legally binding structures for the sharing of R&D results for critical technologies, including structural materials and design methods. Mr. Corwin presented an overview of the GIF structure and relevant groups developing and sharing structural materials R&D. He included the relevant topics for metals, graphite, and ceramic composites being examined and shared among its members.
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DAY 2 PRESENTATIONS Session 1: Graphite Materials A The presentations in this session discussed the use of graphite in advanced HTGRs, design considerations, ASME code considerations, and international research and development activities with regard to graphite.
Applying Brittle Materials Design Concepts in the Manufacture of Graphite and Ceramics for 21st Century Nuclear Reactors Makuteswara Srini Srinivasan, NUMARK Associates, Inc.
Dr. Makuteswara Srini Srinivasan is the principal analyst of Materials Matter, a consulting services company he founded in 2015, after retiring as a senior materials scientist in the NRC's RES, where he served for almost 18 years. He is also a senior executive consultant at NUMARK Associates, Inc.
In 1964, he earned a B.S. in Chemistry (Honors) from the University of Madras, India, and in 1967, a Bachelor of Engineering Degree with Distinction in Metallurgy from the Indian Institute of Science, Bangalore, India. He earned his M.S. Met., E (1967) and Ph.D. (1972) in Metallurgical Engineering from the University of Washington, Seattle, where he was appointed assistant professor in 1972. From 1974 through 1978, he was a staff scientist of the Carbon Products Division of Union Carbide Corporation at the Parma Technical Center in Ohio, specializing in nonlinear elastic fracture mechanics and nondestructive evaluation (NDE) of graphite, leading to the production of several grades of graphite with resistance to thermal shock. From 1979 through 1989, he worked at the Carborundum Company in Niagara Falls, NY, where he directed the research for four operating divisions (Refractories, Electrical Products, Structural Ceramics, and Insulation). From 1989 through 1997, he was the founder and principal of Materials Solutions International, Inc., in Grand Island, NY. He joined the NRC in 1997.
Presentation Summary Dr. Srinivasan focused on the importance for both designers and manufacturers of having a continual dialogue through the materials development process and described a five-pronged approach for doing so (ADAMS Accession No. ML20030B782). Dr. Srinivasans approach suggested holding interactions between designers and manufacturers at every life stage of the reactor, from the design of the facility, through operation and inspection, to final demolition and decommissioning. He emphasized the importance of cost effectiveness for these processes; commercial success can be realized with the overall objective of achieving the lowest possible cost of component manufacture with the highest quality and an installation of the core assembly that can be repeated time and again.
United Kingdom Experience of Graphite-Moderated Reactors Russ Green, United Kingdom Office for Nuclear Regulation Mr. Russ Green started his career after graduating from the University of Liverpool with a bachelors degree in Metallurgy and Materials Science in 2002. He joined the UKs NNC as a 2-1
graduate technical consultant and studied part time to achieve a postgraduate masters degree in Corrosion Control Engineering from the University of Manchester. He had a career of more than 11 years at NNC (taken over by global engineering consultancy AMEC in 2005).
Mr. Green worked across the UK nuclear industry, predominantly on AGR and Magnox sites but also including Sellafield experimental accelerator sites. He became a chartered engineer and a professional member of the Institute of Materials, Minerals and Mining 2010. In 2013, Mr. Green left the private consultancy market and joined the UK ONR, training as a nuclear safety inspector and specializing in structural integrity. For 5 years, he worked across the UK operating reactor fleet, supporting planned and emergent site inspections, interventions, and investigations. In 2018, he moved over to ONRs New Reactors Division as the structural integrity lead for advanced nuclear technologies, supporting the UK-government-funded project on SMRs and nonlight-water SMR technologies.
Presentation Summary Mr. Green described the UK graphite-moderated reactors, summarized nuclear graphite grades used in the UK reactors, and gave an overview of graphite behavior under high-temperature operation and irradiation. Mr. Green also discussed how the United Kingdom is learning from operational experience, acknowledging that irradiated graphite exhibits complex changes in properties and that the region of turnaround is an important consideration in component design.
He pointed out that the dose gradients in components result in dimensional changes strains, which can be significant and result in cracking. Mr. Green also talked about radiolytic oxidation, stating that it is not a concern for known Gen IV designs that use helium coolant.
American Society of Mechanical Engineers Considerations for High-Temperature Reactor Graphite and Composite Components William Windes, Idaho National Laboratory Dr. William E. Windes is a Distinguished Scientist at INL, and his research interests are in the areas of processing, characterization, and analysis of novel material systems for both nuclear and nonnuclear applications. These include materials for use in high-temperature, irradiation, and other extreme environments. He is the DOE technical lead for the Advanced Reactor Technology Graphite R&D program, responsible for thermomechanical testing of nonirradiated and irradiated graphite and composites and for test standards and code case development for determining material properties of nuclear graphite and composites. He holds a Ph.D. in Material Science and Engineering from the University of Idaho, an M.S. in Nuclear Engineering from the University of Illinois, and a B.S. in Nuclear Engineering from the University of California at Santa Barbara.
Presentation Summary Dr. Windes provided an overview of the unique properties of graphite, including its response to irradiation (ADAMS Accession No. ML20030B772). He highlighted the properties that will have most impact on ASME code development; in particular, the aspects of the ASME B&PV Code that will need to be considered for graphite and composite materials. He also discussed the new probabilistic design methodology, failure criteriacomponent strengthstress buildup, and, finally, how changes from environmental effects (irradiation and oxidation) during operation affect design considerations. Dr. Windes briefly mentioned the challenges with in situ inspections for continuous operating designs, such as those for MSRs.
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Graphite Research and Qualification Activities at Nuclear Research and Technology Group Tjark van Staveren, Nuclear Research and Technology Group Mr. Tjark van Staveren has been a materials scientist and project leader at NRG for 9 years, having joined NRG after obtaining an M.Sc. in Materials Science at the Technical University of Delft (the Netherlands). In his role as a materials scientist/project leader, he has been active in different irradiation programs (e.g., for AGR lifetime extension (graphite) and material qualification for the International Thermonuclear Experimental Reactor (steel)). His experience covers the different activities that are part of irradiation qualification programs (i.e., the design and fabrication of irradiation facilities, radioactive materials characterization, and nuclear transports).
Presentation Summary Mr. van Staverens presentation identified the different design, construction, and operational aspects of graphite irradiation facilities that are fundamental to successfully acquiring representative irradiation data for ANLWRs (ADAMS Accession No. ML20030B781). Graphite is a life-limiting component, susceptible to neutron damage. Mr. van Staveren described the behavior of graphite under irradiation conditions, including the NRG work relating to the extended operation of AGRs in the United Kingdom. The results from this project were then compared to AGR actual inspection and monitoring data. The actual and experimental data were found to be of similar magnitude. Mr. van Staveren also highlighted the NRG capability to combine irradiation and oxidation testing and discussed the Materials Test Reactor capabilities.
Session 2: Graphite Materials B This session discussed graphite behavior under irradiated conditions, as well as the effects of molten salt on graphite chemistry and electrochemical behavior.
Understanding Graphite Behavior in Nuclear Reactor Environments for Lifetime Predictions Anne Campbell, Oak Ridge National Laboratory Dr. Anne Campbell is a research associate in the Fundamentals of Radiation Effects group at ORNL. Dr. Campbell received a B.S. in Nuclear Engineering from Purdue University in 2006, an M.S. in Nuclear Engineering and Radiological Sciences from the University of Michigan in 2009, and a Ph.D. in Nuclear Engineering and Radiological Sciences from the University of Michigan in 2014.
Dr. Campbells research focuses on understanding the correlations between microstructural and bulk property changes (including developing fundamental understandings of how the microstructural changes in graphite result in the observed bulk property changes) and irradiation creep in nuclear fuels and structural materials in nuclear environments. She is the primary investigator at ORNL for the graphite code qualification irradiation programs.
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Presentation Summary Dr. Campbell discussed the overall trends of the irradiation-induced changes to the physical and thermal properties of nuclear graphite during her presentation (ADAMS Accession No. ML20030B784). She explained the need to understand the behavior of graphite in advanced reactor environments, including irradiation and volumetric effects. A novel aspect of graphite is that irradiation-induced creep acts as a stress relaxation mechanism. One of the main questions is about graphite behavior during the lifetime of graphite in the core.
Dr. Campbell also discussed the lifetime limitations for various reactor types and the concerns with using graphite in MSRs, including salt wetting in pores and fission gas retention.
The Chemistry of Graphite in Fluoride-Salt-Cooled, High-Temperature Reactors and Molton Salt Reactors Raluca Scarlat, University of California at Berkeley Dr. Raluca Scarlat is an assistant professor at the University of California (UC) at Berkeley in the Department of Nuclear Engineering. She has a Ph.D. in Nuclear Engineering from UC Berkeley, a certificate in Management of Technology from the Hass School of Business, and a B.S. in Chemical and Biomolecular Engineering from Cornell University.
Dr. Scarlats research focuses on the chemistry, electrochemistry, and physical chemistry of high-temperature inorganic fluids and their application to energy systems. She has experience in the design and safety analysis of fluoride-salt-cooled, high-temperature reactors (FHRs),
MSRs, and HTGRs. Dr. Scarlats research includes safety analysis, licensing and design of nuclear reactors, and engineering ethics.
Presentation Summary Dr. Scarlat began by presenting an initial overview of the considerations in the Molten Salt Reactor Experiment for graphite integrity and described experiments that her group had performed on the effect of the contact angle of the molten salt with the graphite (ADAMS Accession No. ML20030B756). With regard to salt intrusion, Dr. Scarlat stated that FLiBe is not expected to intrude into pores of graphite at ambient pressure. As to salt exposure experiments, her group did not observe salt intrusion but it did observe fluorination, with some material deposition on the surface. It concluded that the material is likely beryllium that was reduced by carbon. Dr. Scarlat ended the presentation with a list of questions to be investigated on the use of graphite in FHRs and MSRs.
Graphite Electrode Behaviors and Its Application for Salt Purification Jinsuo Zhang, Virginia Polytechnic Institute and State University Dr. Jinsuo Zhang is a professor of nuclear engineering at Virginia Tech. He was previously an associate professor in nuclear engineering at the Ohio State University and a staff scientist at Los Alamos National Laboratory. Dr. Zhang earned a Ph.D. and a B.S. in Engineering Mechanics from Zhejiang University (China) in 2001 and 1997, respectively.
Dr. Zhangs research interests relate to advanced nuclear coolant (liquid metal and molten salt) chemistry and control, material corrosion and control, and advanced nuclear structural and fuel materials.
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Presentation Summary Dr. Zhangs presentation focused on the graphite-electrochemical behaviors in a molten salt system (ADAMS Accession No. ML20030B795). He began by describing experiments looking at the anodic effect on a graphite electrode in molten lithium fluoride, sodium fluoride, and potassium fluoride (LiF-NaF-KF) or FLiNaK salt mixture. Dr. Zhang also highlighted the difference in cyclic voltammograms that were produced using a tungsten versus a graphite electrode. The study indicated that a graphite electrode can be applied as an active electrode to separate lanthanum (as a carbide) from molten LiF-NaF-KF and as an inert electrode to deposit europium metal from potassium chloride-lithium chloride. The presentation also discussed the use of graphite as a sacrificial electrode to separate oxygen by forming carbon monoxide (CO)/CO2 and as an inert electrode to separate iodine from molten LiF-NaF-KF.
Professor Zhang also discussed the galvanic corrosion of a metal alloy by graphite. Nickel was used to do the test because it is an inert metal. However, when contacted with graphite, the nickel corroded significantly.
Session 3: Materials Qualification Challenges A This session addressed some of the challenges identified with materials qualification at the high temperatures associated with advanced reactor types and the selection of materials for use in these reactors.
Metallurgical Challenges Associated with Using Grade 91 Steels at Elevated Temperature Jonathan Parker, Electric Power Research Institute Dr. Jonathan Parker is a senior technical executive at EPRI. He received his B.S. (with honors) in Physical Metallurgy from the University of Wales in 1972 and his Ph.D., also from the University of Wales, in 1977. He is a member of several professional associations, including ASME. He has earned a number of awards for his work, including the 2016 Advanced Materials Lifetime Achievement Award, for research achievements in materials engineering with particular emphasis on improvements with the electricity supply chain.
Presentation Summary Design codes for alloys used in high-energy applications typically assume that the components fabricated will exhibit homogeneous composition, microstructure, and properties. Dr. Parker explained that this is not always the case (ADAMS Accession No. ML20030B765). A good understanding and control of the manufacturing process and lingering effects are needed, since variability and homogeneity are linked to manufacturing. EPRI has studied Grade 91 (and 92) steel for over 20 years. It found that damage tolerance can depend critically on design, such as geometry or the crack-arresting step weld. Additionally, Dr. Parker discussed how the materials are imaged can be misleading; it may appear as tempered martensite in optical metallography when it is not.
Dr. Parker also talked about advanced manufacturing; specifically, powder metallurgyhot isostatic pressing and how it provides possible opportunities to Grade 91 components. It may also be possible to reduce welds and inhomogeneity. He mentioned the need for engagement among all stakeholders and to have good design rules in place.
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Advanced Structural Materials for Nonlight-Water Reactors Steve Zinkle, University of Tennessee, Knoxville Dr. Steve Zinkle is a Governors Chair Professor in the Nuclear Engineering and Materials Science & Engineering Departments at the University of Tennessee, Knoxville, with a joint appointment at ORNL. He earned a Ph.D. in Nuclear Engineering from the University of Wisconsin, Madison, in 1985, and from the same university has an M.S. in Materials Science (1985), M.S. in Nuclear Engineering (1982), and B.S. in Nuclear Engineering (1980).
Dr. Zinkles research interests include deformation and fracture mechanisms in structural materials; advanced manufacturing; and radiation effects in ceramics, fuel systems, and metallic alloys for fission and fusion energy systems. He has written over 290 peer-reviewed publications and is a member of the National Academy of Engineering.
Presentation Summary Dr. Zinkle presented some key considerations for identifying and selecting reference structural alloys for several non-LWR concepts (ADAMS Accession No. ML20030B792). New tools are now available to examine the thermodynamic stability of materials and new data on radiation effects, and it is common knowledge that properties can be improved with fine-scale (about 10 nanometer) precipitates. Thermomechanical treatment variation also improves creep strength. If more creep strength is needed, oxide-dispersion strengthened steels can be used.
For mitigation of both void swelling and irradiation hardening of materials used in high-temperature, high-radiation environments, a high density of uniformly distributed nanoscale precipitates (high sink strength) is desirable. Dr. Zinkle noted that most of these materials are still in the research stage, not the production stage. Their use requires ASME code qualification.
Challenges in Qualifying Advanced Manufacturing Technologies for High-Temperature Nuclear Service Mark Messner, Argonne National Laboratory Dr. Mark Messner is a principal mechanical engineer at ANL. He earned a Ph.D., M.S., and B.S. in Civil and Environmental Engineering from the University of Illinois at Urbana-Champaign in 2014, 2011, and 2010, respectively.
Dr. Messners research focuses on modeling, simulating, and designing high-temperature materials and structures. He specializes in mesostructural modeling, structural and material design and optimization, machine learning for design problems, and the development of simulation and engineering design methods. He participates as a member and chair of several of the ASME B&PV Code,Section III, working groups responsible for high-temperature design methods and has published more than 30 peer-reviewed articles on high-temperature materials and design.
Presentation Summary Dr. Messner began his presentation by highlighting some of the benefits of using advanced manufacturing technologies to make ANLWR components: complex geometries, cladded parts, functionally graded materials, and built-to-suit spares can all be manufactured (ADAMS 2-6
Accession No. ML20030B757). Dr. Messner then discussed challenges associated with qualifying advanced manufacturing components, including variability, the fact that short-term tests may tell very little about long-term behavior, statistical variability, weld resilience, and the sparsity of data for the behavior of advanced manufacturing materials at high temperatures.
Dr. Messner explained the traditional approach for qualifying high-temperature materials, which involves extrapolating results from shorter term mechanical property tests to longer times.
Qualification challenges specific to advanced manufacturing components include greater variability of properties, many processing parameters, and a wide variety of technologies. For advanced manufacturing materials, the current emphasis with regard to qualification is more on repeatability than specific properties. In his conclusion, Dr. Messner stated that qualifying advanced manufacturing materials for high-temperature nuclear service will have some unique challenges, but options are available. It would be best to start now with likely technologies and materials, given the reliance of high-temperature design on long-term material properties.
Qualification of Materials for Elevated Temperature Nuclear Components Richard Wright, Idaho National Laboratory Dr. Richard Wright is Laboratory Fellow Emeritus at INL. For the past 10 years, he has participated in work sponsored by the DOE Office of Nuclear Energy to qualify additional materials for elevated temperature nuclear components. He received his Ph.D., MS., and B.S.
in Metallurgical Engineering from Michigan Technological University and has a Masters of Business Administration from Idaho State University.
Presentation Summary Currently, the ASME B&PV Code qualifies a very limited number of metallic materials for construction of elevated-temperature nuclear components. Dr. Wrights presentation outlined the qualification process, describing recent experience with the qualification of nickel-based Alloy 617, which supports HTGR/VHTR applications (ADAMS Accession No. ML20030B774).
He described additional requirements that might be needed outside the scope of ASME B&PV Code,Section III, Division 5, including evaluating inservice deterioration that may occur as a result of corrosion, for example.
Dr. Wright outlined the different regulatory paths to the acceptance of new materials, including developing a code or standard or following processes delineated in the NRCs regulations.
Dr. Wright concluded by discussing several methods to accelerate the qualification of materials, including extrapolation by a factor of 3 to 5, limiting initial design life while testing continues, and simulations.
Session 4: Materials Qualification Challenges B This session continued the theme of qualification challenges, further addressing the code qualification of materials, from both the U.S. and Canadian perspectives, and efforts being made to streamline development and qualification of new alloys for advanced reactor applications.
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Challenges with American Society of Mechanical Engineers Code Qualifying Graphite Irradiation Effects in Test Reactors Anne Campbell, Oak Ridge National Laboratory Dr. Anne Campbell is a research associate in the Fundamentals of Radiation Effects group at ORNL. Dr. Campbell received a B.S. in Nuclear Engineering from Purdue University in 2006, an M.S. in Nuclear Engineering and Radiological Sciences from the University of Michigan in 2009, and a Ph.D. in Nuclear Engineering and Radiological Sciences from the University of Michigan in 2014.
Dr. Campbells research focuses on understanding the correlations between microstructural and bulk property changes (including developing fundamental understandings of how the microstructural changes in graphite result in the observed bulk property changes) and irradiation creep in nuclear fuels and structural materials in nuclear environments. She is the primary investigator at ORNL for the graphite code qualification irradiation programs.
Presentation Summary The ASME B&PV Code includes minimum requirements for the temperature and neutron fluence ranges, properties, and sampling requirements for measurements before and after irradiation, and references American Society for Testing and Materials International standards to be used for testing. Dr. Campbells presentation discussed the requirements called out in ASME B&PV Code,Section III, Division 5, Subsections HAB and HHA, on graphite used as reactor core internals, some examples of the requirements/limitations from the relevant standards, and how these requirements result in challenges related to obtaining the irradiation-induced property changes through campaigns in materials test reactors (ADAMS Accession No. ML20030B760). The challenges identified included the need for multiple temperature/fluence combination specimens and the ability to make replicate measurements.
This leads to multiple capsules or zones per temperature/fluence combination. Because not all specimens can be irradiated at one time, this will prolong the time needed to collect irradiation data.
Potential Material Issues for the Canadian Nuclear Safety Commission To License Advanced Reactors Xuejun Wei, Canadian Nuclear Safety Commission Dr. Xuejun Wei is a technical specialist in the Engineering Design Assessment Division of CNSC. He obtained his Ph.D. in Material Science and Engineering and currently works for CNSC on assessing engineering design, material selection, and aging management of nuclear power plants. He is also a member of several workgroups for developing ASME B&PV Code,Section III, Division 5, for HTRs. From 2009 to 2017, he chaired the Multinational Design Evaluation Programme Codes and Standard Working Group on harmonization of international code requirements for mechanical design.
Before joining CNSC, Dr. Wei worked on CANDU fuel channel fabrication, chemistry control, and material degradation at the Chalk River Laboratory in Canada. He also conducted research in the United States on material degradation and mitigation at the Pennsylvania State University.
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Presentation Summary Dr. Weis presentation identified some of the potential materials issues or knowledge gaps in materials and component integrity that CNSC identified in the context of licensing advanced reactors (ADAMS Accession No. ML20031D419). Dr. Wei began his presentation by giving an overview of the CNSC vendor design review process. The purpose of this review is to provide early identification of potential regulatory and technical issues in licensing a vendors design and to identify additional regulatory research that may be needed to inform regulatory requirements in specific cases.
Dr. Wei stated that the majority of the vendors propose to use ASME B&PV Code,Section III, Division 5, for advanced reactor designs but that there are gaps in the code with regard to deterioration that may occur in service as a result of radiation effects, corrosion, erosion, thermal embrittlement, or instability of the material. His presentation focused on discussing these gaps and material development and qualification.
Realizing the Nuclear Materials Discovery and Qualification Initiative Matthew Kerr, Idaho National Laboratory Dr. Matthew Kerr is currently Department Manager for Nuclear Materials in the Nuclear Science and Technology Directorate at INL. The department has a broad focus with scientists and engineers researching structure-property relationships for nuclear materials, concepts for accelerated materials development, and tritium transport behavior for fusion applications.
Before joining INL in 2018, Dr. Kerr held positions of increasing technical responsibility at the NRC and the Naval Nuclear Laboratory at the Knolls Site, where he supported a wide range of plant and core materials systems. At Knolls, Dr. Kerr received the 2018 Engineer and Scientist Award (highest technical recognition) for leadership and technical contributions that supported fleet support and new design needs.
At INL, Dr. Kerr is recognized for his unique background, which has led to involvement in the Laboratory Directed Research and Development program as the mission area point of contact for the Advancing Nuclear Energy initiative. Dr. Kerr holds a Ph.D. in Materials Engineering from Queens University (2009), where he was a member of the Nuclear Materials Group, and an M.S.E. (2004) and B.S.E. (2002) in Materials Science and Engineering from Arizona State University. He has participated in leadership development programs at the NRC and the Naval Nuclear Laboratory.
Presentation Summary Dr. Kerrs presentation focused on the work being performed at INL under the Nuclear Materials Discovery and Qualification Initiative (NMDQi) (ADAMS Accession No. ML20031D422). He described how NMDQi aims to accelerate development and qualification of new nuclear materials and fuels for future advanced reactor technologies. He provided examples of instances where complex new alloys had been developed using computational materials design and suggested how the current nuclear materials (e.g., alloys) development cycle can be optimized. Dr. Kerr described how the NMDQi program takes a Grand Challenge approach to accelerate the development and qualification of new nuclear materials and fuels for future advanced reactor technologies. Dr. Kerr also emphasized the importance of collaboration and 2-9
highlighted both internal collaborations and those between other government entities and universities.
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DAY 3 PRESENTATIONS Session 1: Inspection, Monitoring, and Surveillance Development of the Technical Basis for In Situ, Passive Surrogate Materials Surveillance for Advanced Nonlight-Water Reactors Sam Sham, Argonne National Laboratory Dr. Sam Sham is the Program Lead of Advanced Reactor Materials in the Applied Materials Division of the Energy & Global Security Directorate at ANL. His technical specialty is the deformation and failure of advanced materials and structural mechanics technologies for HTRs.
He is the technology area lead of the advanced materials R&D activities for the Advanced Reactor Technologies Program, Office of Nuclear Reactor Deployment, in the DOE Office of Nuclear Energy. The R&D portfolio cross-cuts the three reactor campaigns on fast reactors, gas-cooled reactors, and MSRs. He also participates in R&D activities for microreactors.
Dr. Sham is a member of the ASME B&PV Committee on Construction of Nuclear Facility Components (III) and the B&PV III Executive Committee. He chairs the Subgroup on High Temperature Reactors, which, together with seven subtier working groups and task groups, is responsible for ASME B&PV Code,Section III, Division 5. He was elected an ASME Fellow in 2000.
Before he joined ANL in 2015, Dr. Sham was a Distinguished R&D Staff Member at ORNL, held senior positions with AREVA NP Inc. and Knolls Atomic Power Laboratory, and was a tenured faculty member at Rensselaer Polytechnic Institute. He holds a B.Sc. (First Class Honour) degree in Mechanical Engineering from the University of Glasgow (Scotland) and an M.S. and Ph.D. in Mechanics of Solids and Structures, as well as an M.S. in Applied Mathematics, from Brown University.
Presentation Summary Dr. Sham opened his presentation by suggesting that the technology maturity of MSR systems is substantially less than other ANLWRs such as gas-cooled reactors and SFRs (ADAMS Accession No. ML20030B758). Information on material degradations due to irradiation, corrosion, elevated temperature exposure, and creep-fatigue loading during MSR operations is limited. The establishment of an in situ surrogate materials surveillance program that would allow the collection of information on these material degradation mechanisms would be an important pathway in supporting the timely licensing of MSRs.
Dr. Sham then described the progress and necessary future work to develop passive, in situ materials surveillance programs for ANLWRs. A proof-of-concept test has been successfully completed on a method to load specimens passively in situ using differential thermal expansion.
Near-term future tasks include developing a prediction tool to select test article geometry and driver material, fabricating new test articles based on the prediction tool results, and conducting thermal cycling tests. Further, more work is needed on methods for fabricating smaller test articles that can conform to the space limitation of an operating reactor. Future issues include identifying methods to extract the data from the in situ specimens, acceptance criteria, and developing methods to measure stress and strain.
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Progress Toward Bridging Harsh Environment Online Monitoring Gaps for Advanced Reactors S.W. (Bill) Glass, Pacific Northwest National Laboratory Mr. S.W. (Bill) Glass has focused his 40-year career on inspection and robotic technologies mostly related to nuclear power plant operation, plant life extension, and high-temperature sensing. Following 35 years with Framatome in Virginia and France, he joined Pacific Northwest National Laboratory (PNNL) in 2015 as a technical advisor, continuing his interest in inspection and robotics. His current focus includes cold spray inspections, high-temperature NDE for advanced reactors, and cable NDE. He has authored more than 100 technical and scientific papers, holds seven patents, and is a Licensed Professional Engineer.
Presentation Summary Mr. Glass provided an overview of some of the ANLWR technologies and the advantages and concerns with these types of reactors (ADAMS Accession No. ML20030B788). A principal advantage is the lower predicted cost per kilowatt of electricity generated. For MSRs and LMRs, advantages include high temperature (versus other fluid-fueled systems), ability to configure as breeders/waste burners, low-pressure operation, stability of liquid under radiation, and high solubility of uranium and thorium (in fluoride salts). Concerns include material corrosion susceptibility, which can be partially addressed by monitoring and inspection, and accessibility for inspection, repair, and maintenance under high temperature and radiation conditions.
Mr. Glass discussed PNNLs development of a high-temperature cold spray magnetostrictive electromagnetic acoustic transducer, a technique that provides a way of applying a permanent ultrasonic sensor to components that can be used while the plant is on line for flaw detection and as a noninvasive flow meter. Mr. Glass described progress by PNNL on online heat exchanger monitoring, virtual penetrations using ultrasonics, fiber and optical guides for motion temperature, and material identifications, as well as an advanced reactor sensor Web site and database. Mr. Glass also discussed PNNLs support for the development of the reliability and integrity management concept in ASME B&PV Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Division 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants.
A Vision for Advanced Nondestructive Examination Methods for Nonlight-Water Reactor Components Yarom Polsky, Oak Ridge National Laboratory Dr. Yarom Polsky is a distinguished member of the R&D staff and currently leads the Sensors and Embedded Systems Research Group at ORNL. This group develops novel or specialized sensor technologies, complex instrumentation systems that require special attention to sensor integration details, and embedded microelectronics, and it conducts feedback control-related modeling and algorithm R&D to advance the state of the art of measurement technologies, methods, and their use. Some of his R&D activities over the course of his career include the modeling and characterization of thermomechanical warpage of electronic packages during manufacturing processes, the design and construction of oilfield services equipment, the use of neutron imaging and neutron diffraction to experimentally study fluid flow and material strain under triaxial loading conditions, and the development of improved image reconstruction 3-2
methods for nondestructive evaluation using phased array ultrasonics. He has written over 50 technical publications and has 11 issued patents. He holds a B.A. in Physics from Rice University, an M.S. in Mechanical Engineering from Rice University, and a Ph.D. in Mechanical Engineering from the Georgia Institute of Technology.
Presentation Summary Dr. Polskys presentation provided an overview of the NDE challenges associated with ANLWRs (ADAMS Accession No. ML20030B786). NDE of ANLWRs will be challenging due to harsh environmental conditions and extended operating cycles (e.g., no traditional refueling outages),
which limit opportunities for inspection. Both NDE and structural health monitoring, which continuously monitors a subset of components, may need to be considered for ANLWRs that may have longer refueling cycles. Under structural health monitoring, component selection would be based on contribution to risk and limited accessibility. Structural health monitoring will require high-temperature and radiation-resistant sensors. Dr. Polskys presentation also discussed NDE of nonmetallic components in ANLWRs, which can build on techniques used in LWRs. ANLWRs will also need methods to characterize the material state nondestructively to allow predictions of remaining life.
Sodium-Cooled Fast ReactorsJapans Experience and Future Tai Asayama, Japan Atomic Energy Agency Dr. Tai Asayama is Deputy Director General of the Fast Reactor Cycle System Research and Development Center of JAEA. He has more than 30 years of experience evaluating elevated temperature materials and developing design methodologies for fast reactors. He is engaged in codes and standards development in JSME and, until 2017, chaired two working groups of the ASME B&PV Code Committee and led the development of Code Case N-875, which provides alternative inservice inspection requirements for LMRs, based on the System Based Code concept.
Presentation Summary Dr. Asayamas presentation focused on operating experience with material surveillance programs, online monitoring, and the systems-based monitoring concept from Japans SFRs, Monju, and Joyo (ADAMS Accession No. ML20030B775). With respect to operating experience, no significant corrosion or stress-corrosion cracking has been observed in Japans SFRs. Fatigue and creep are the major degradation mechanisms of concern and are accounted for through the design to JSME standards. The in situ surveillance program monitors irradiation damage of stainless-steel reactor vessel and core support materials. Continuous monitoring for sodium leaks supports leak-before-break searches and is considered the main method for inservice inspection, with visual examination and surveillance programs complementary to the continuous monitoring. Inservice inspection of reactor vessel welds and primary piping is also conducted by robotic devices, which can perform ultrasonic testing, electromagnetic acoustic transducer techniques, and visual inspections. Dr. Asayama also discussed the JSME code requirements for SFRs, which mainly require continuous monitoring (satisfied by sodium leak detection) rather than volumetric or surface examinations required for LWRs. Finally, Dr. Asayama described the system-based monitoring concept, in which target reliability is determined and appropriate margins established accordingly, whereas the conventional approach accumulates margins but may result in excessive margins.
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Nondestructive Evaluation in Advanced Reactors Greg Selby, Electric Power Research Institute Mr. Greg Selby is a senior technical executive at EPRI. He holds strategic leadership responsibilities within the NDE Program and its staff of 50, the worlds leading independent resource for NDE in the nuclear power industry. The program delivers NDE technology, technology transfer, training, and qualifications for all U.S. nuclear utilities and for about two dozen international utilities.
Mr. Selbys research has focused on developing ultrasonic techniques for examining reactor piping, pressure vessels, pressure vessel internals, bolting, and turbines. Since 1993, he has led EPRIs successful, strategic drive for acceptance and adoption of phased array ultrasonic technology in the nuclear power industry. Mr. Selby also led EPRIs support of inspection issues for the Boiling Water Reactor Vessel and Internals Project from the mid-1990s through 2006. He served as director of the NDE Program from 2006 through 2013.
Mr. Selby joined EPRI in 1983 as a senior engineer in the NDE Program. Before joining EPRI, Mr. Selby worked at PNNL in Richland, WA, as a senior engineer in the NDE group. His research in NDE reliability led to his employment at EPRI and to his initial responsibilities developing training and qualification programs for NDE of reactor coolant piping systems.
Mr. Selby received a B.S. in Nuclear Engineering from Oregon State University.
Presentation Summary Mr. Selbys presentation focused on two areas: nondestructive methods for determining material microstructure and NDE for compact heat exchangers (CHX) (ADAMS Accession No. ML20030B766). Nondestructive methods assess material susceptibility to degradation due to mechanisms such as void swelling. EPRI is investigating both ultrasonic testing and resistivity methods for assessing void swelling. Mr. Selby also discussed NDE for assessing fracture toughness, for which EPRI is investigating ultrasonic testing and instrumentation indentation methods. NDE could also assess microstructure in P91 and P92 materials, in which damage susceptibility is linked to variability in microstructure.
Conventional NDE methods for shell and tube heat exchangers will not work for CHX because CHX are essentially solid metal blocks with many small hot and cold flow channels. Mr. Selby discussed EPRI participation in a DOE project directed at an ASME B&PV Code,Section III, code case related to CHX construction for advanced reactors. Mr. Selby also presented some ultrasonic testing results by EPRI on a failed CHX. Finally, Mr. Selby discussed the use of embedded sensors, such as strain gauges, to enable online monitoring of CHX.
Session 2: Molten Salt Chemistry Research Reactor Experiments To Study Materials and Fuel Salt Performance Uazir Bezerra de Oliveira, Nuclear Research and Technology Group Dr. Oliveira is a materials engineer with 20 years of experience in diverse industries such as aerospace, metallurgy, and oil and gas). He has an M.Sc. from Chalmers University (Sweden) in high-temperature corrosion and a Ph.D. from Groningen University (Netherlands) in direct 3-4
energy deposition. During his career, he has gained experience in various roles such as project manager for product development (Sandvik Steel, Tata), technical authority for subsea materials (Totals Kaombo project at Heerema Marine Contractors), and product manager for modern steel converters (Danieli Corus). In September 2019, he joined NRG as program manager for MSR and in the project management arena for Blackstone. (Blackstone carries out graphite irradiation experiments to extend the life of EDF Energys AGRs.)
Presentation Summary Dr. Oliveiras presentation described the Dutch molten salt program (ADAMS Accession No. ML20031D426). NRG conducts nuclear technological research in various fields, including the molten salt program. This research comprises (among others) trial irradiations on nuclear materials and nuclear fuels in the HFR in Petten in the Netherlands, together with accompanying postirradiation experiments, inspection methods, and high-quality simulations of behavior displayed by nuclear components and reactors.
The molten salt program at NRG, in collaboration with the European Joint Research Centre and Technical University of Delft, has expanded considerably since 2015 and comprises research into (among others) relevant construction materials, fission product stability and behavior in molten salt, and processing of fuel waste. The research aims at investigating the combined effects of high temperature, strong neutron fields, temperature gradients, pressure, and salt flow on materials and fuels. Dr. Oliveira stated that the unique capabilities at the integrated site allow NRG to perform research in an independent and autonomous manner, which helps bring MSRs closer to reality.
Dr. Oliveira then summarized some of the specific research programs underway at NRG, including the Salient program, which is investigating interactions of molten salt fuel with graphite and molten salt with metallic materials. The SAGA program simulated the radiolytic production of fluorine and uranium hexafluoride in MSRs through gamma irradiation of a fuel salt at low temperature. The ENICKMA program is investigating embrittlement of nickel-based alloys in helium. With respect to waste, the overall approach is conversion of the salts to recognized, acceptable chemical forms by direct oxidation or aqueous processing. The MSR loop (LUMOS) would be an in-pool loop positioned next to the HFR core wall and would contain an actinide-bearing FLiBe salt.
Molten Fluoride Salt Chemistry Raluca Scarlat, University of California at Berkeley Dr. Raluca Scarlat is an assistant professor at UC Berkeley, in the Department of Nuclear Engineering. She has a Ph.D. in Nuclear Engineering from UC Berkeley, a certificate in Management of Technology from the Hass School of Business, and a B.S. in Chemical and Biomolecular Engineering from Cornell University.
Dr. Scarlats research focuses on the chemistry, electrochemistry, and physical chemistry of high-temperature inorganic fluids and their application to energy systems. She has experience in the design and safety analysis of FHRs, MSRs, and HTGRs. Dr. Scarlats research includes safety analysis, licensing and design of nuclear reactors, and engineering ethics.
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Presentation Summary Dr. Scarlats presentation covered complex ion formation in molten salts, the construction and use of reference electrodes, and the results of electrochemical studies of hydrogen and oxides in FLiBe (ADAMS Accession No. ML20031D427). She also highlighted the need for standards and standard methods of elemental analysis.
Dr. Scarlat concluded with several additional questions to be investigated. First, fluoroacidity, speciation diagrams, and activity coefficients should be developed as tools for predicting corrosion drivers, solubility and volatility. Second, the role of oxide anions and other anions on complexation and activity coefficients needs to be determined. Ion-specific reference electrodes can enable the measurement of acidity, the determination of activity coefficients, and the generation of speciation and Pourbaix diagrams. Demonstrating reference electrode stability and reliability is important, and the electrodes performance is melt specific and composition specific (i.e., an electrode that works in one salt composition might not work in another salt composition). Reference electrodes enable the development and calibration of electrochemical sensors (e.g., for oxide content quantification).
Fundamental Properties and Removal of Impurities in Molten Salt Systems Jinsuo Zhang, Virginia Polytechnic institute and State University Dr. Jinsuo Zhang is a professor of nuclear engineering at Virginia Tech. He was previously an associate professor in nuclear engineering at the Ohio State University and a staff scientist at Los Alamos National Laboratory. Dr. Zhang earned a Ph.D. and a B.S. in Engineering Mechanics from Zhejiang University (China) in 2001 and 1997, respectively.
Dr. Zhangs research interests relate to advanced nuclear coolant (liquid metal and molten salt) chemistry and control, material corrosion and control, and advanced nuclear structural and fuel materials.
Presentation Summary Dr. Zhangs presentation covered new methods for redox potential measurements, exchange current density measurements, and lanthanides separation, as well as online corrosion measurement, and the applications of these methods for both molten fluoride and chloride salts (ADAMS Accession No. ML20031D428). Impurities that can exist in molten salts include tritium, noble gases, halogens (iodine), alkaline metals, rare earths, noble metals, tellurium and antimony, actinides, corrosion products, and oxygen and moisture. Dr. Zhang described the redox potential measurement process and showed an example of results for corrosion products.
Next, he showed how the exchange current density model is being fitted to the experimental data. Dr. Zhang then described the conditions for electrochemical separation in fluoride salts.
Fluoride Salt Chemistry and Properties Matthew Memmott, Brigham Young University Dr. Matthew Memmott is an assistant professor in the Chemical Engineering Department at Brigham Young University (BYU). He received a B.S. in Chemical Engineering from BYU in 2005, and an M.S. and Ph.D. in Nuclear Science and Engineering from MIT in 2007 and 2009, respectively. His research focuses on advanced nuclear reactor design, nuclear safety, and 3-6
system modeling. Following his graduation from MIT, Dr. Memmott worked as a senior engineer in the advanced reactor group at Westinghouse Electric Company. Dr. Memmott's research at BYU focuses on MSR technologies, including thermodynamics, chemistry, and thermophysical properties for clean, actinide bearing, and fission-product-bearing molten salts.
He also focuses on MSR design concepts that have unique and enabling features based on the functional aspects of molten salt behavior.
Presentation Summary Dr. Memmott opened his presentation by outlining some of the licensing needs related to molten salt chemistry (ADAMS Accession No. ML20031D417). These include developing an understanding of source terms, transport system models, thermophysical properties, and thermodynamic behavior. Transport system models involve a complex web of interactions including reaction with contaminants, bulk oxidation, homogeneous reduction, reduction of the surface, corrosion, and surface oxidation. With respect to determining thermophysical properties, molten salt is challenging because it is anaerobic, anhydrous, prone to wall creep, has the toxicity of beryllium, involves high temperatures, and is susceptible to impurity retention.
Characterizing properties of actinide-bearing salts is challenging because purification standards have not been set, only limited clean salt data exist, and salts are expected to rapidly pick up impurities. Further, there are literally millions of possible combinations of experiments for all the different combinations of salts and materials.
Thermodynamic modeling would involve the investigation of equilibrium potentials for actinides (thorium, uranium) or fission products at low concentrations, analysis of thermodynamic data, and assessment of trends and behaviors, as well as making phase predictions using a modified quasichemical model.
The overall conclusions of Dr. Memmotts presentation are that salt experiments are needed to inform MSR design and licensing processes, knowledge of actinide and fission-product-bearing salts is essential for system modeling but massive in scope, thermodynamic assessments inform experiments and can reveal correlations, understanding trends in ion-ion interactions can minimize experimental load, and improved characterization capability is needed, with neutron scattering (PDF (atomic pair distribution function)) analysis a promising technique.
Online Monitoring of Molten Salt Reactors Nathaniel Hoyt, Argonne National Laboratory Dr. Nathaniel Hoyt is a principal chemical engineer in the Chemical and Fuel Cycle Technology Division at ANL. His work includes the development of electrochemical technologies to enable the operation of MSRs and molten salt fuel reprocessing systems. Much of this research involves the production of sensors to provide crucial process monitoring and safeguards capabilities for molten salt media.
Presentation Summary Dr. Hoyt opened his presentation with the thought that chemical considerations within fuel salts and secondary salts demand high-fidelity process monitoring capabilities (ADAMS Accession No. ML20031D415). Parameters that will require monitoring include concentrations of reactants 3-7
in the core (neutronics), the fission product removal system (in situ processing), noble metal deposition, structural corrosion, and safeguards.
Dr. Hoyt showed a sample of results from voltammetry, which is one technique that can be used to monitor actinide concentrations in coolant. It allows rapid online measurements and the sensors are robust. Dr. Hoyt described the development of a multielectrode array sensor that can measure salt redox potential, species concentration, and salt levels. Some challenges for the development of electroanalytical sensors in real molten salt systems include electrical noise, highly concentrated salts, nonideal electrochemical behavior, materials stability, seals, automation of acquisition, and automation of analysis.
Dr. Hoyt discussed long-duration testing of sensors in the thermal convection loop at ORNL.
This testing is being conducted at 750 degrees Celsius (C), and the sensor provides long-term quantitative measurements of salt potential, metal ion corrosion products, and impurities. The sensor has provided measurements for more than 650 hours0.00752 days <br />0.181 hours <br />0.00107 weeks <br />2.47325e-4 months <br /> of testing.
In conclusion, Dr. Hoyt stated that robust electroanalytical sensors have been demonstrated for durations longer than a year and a half. They provide high stability, longevity, and accuracy combined with very low uncertainty. Dr. Hoyt indicated that future work will be directed toward application of the sensor to additional nuclear-relevant fluoride and chloride salts, shakedown testing in noisy electromagnetic interference environments, standardization of sensor construction, and forced convection studies.
Session 3: Metallic Materials: Environmental Effects Overview of Environmental Issues and Material Property Gaps for Commercial Viability of Advanced Reactors Mike Burke, Electric Power Research Institute Dr. Mike Burke is a technical executive at EPRI with responsibility for near-core and structural materials for operating and advanced nuclear plants. Before joining EPRI in 2017, Dr. Burke was the consulting engineer for materials at Westinghouse Electric Company. In 2011 and 2012, he was a professor of nuclear materials at the University of Manchester (UK) and technical director of the UK Nuclear Advanced Manufacturing Research Center. Before these appointments, he had a career of over 30 years at the Westinghouse research center, where he managed programs in the development and testing of materials for fossil and nuclear power plants. His doctoral degree is in Metallurgy from the University of Sheffield (UK), and his M.S. is in Metallurgical Engineering from the University of Pittsburgh.
Presentation Summary Dr. Burkes presentation focused on the data gaps that must be filled to qualify materials for service in ANLWRs (ADAMS Accession No. ML20030B764). Dr. Burke presented a figure showing a progression from mechanical properties, to environmental effects, to irradiation properties, all of which must be determined to qualify materials. There is a lot of white space (e.g., data gaps) in this figure for ANLWRs compared to LWRs. Dr. Burke emphasized that without these data, life estimates cannot be performed, and without life estimates, economic viability cannot be determined. The presentation included a table of materials knowledge gaps by reactor type, component, and material. For example, for Type 316 stainless steel core support materials in MSRs, gaps include proof of resistance to long-term corrosion in properly 3-8
controlled salt environments; time-dependent properties for certification under ASME B&PV Code,Section III, Division 5; and demonstration of resistance to environmentally assisted cracking under loading in salt.
In his summary, Dr. Burke listed the following major categories necessary for development:
materials properties to support initial design and to attain ASME code acceptance for construction, including high-temperature and time-dependent properties; response of candidate materials to neutron irradiation; and materials response to environment, with effects measured in realistic environments. The initial focus will need to be on the development of code-required mechanical properties to support designing and building prototypes. Future needs will include better knowledge of the composition of coolants after operation to enable realistic environmental testing of materials, more knowledge of critical variables in the most significant degradation mechanisms, rig and loop testing in simulated environments of previously neutron-exposed materials, and prototyping and postmortem analyses to identify key assessment variables.
Finally, Dr. Burke highlighted the need for investment in materials testing facilities, including specialized test rigs, dedicated loops, and reactors for irradiation exposure.
Environmental Effects in Liquid Metal Systems Todd Allen, University of Michigan Dr. Todd Allen is a professor at the University of Michigan and a senior fellow at Third Way, a think tank based in Washington, DC, supporting its Clean Energy Portfolio. He was the Deputy Director for Science and Technology at INL from January 2013 through January 2016. He was previously a professor in the Engineering Physics Department at the University of Wisconsin, a position held from September 2003 through December 2012 and again from January 2016 to December 2018. From March 2008 to December 2012, he was concurrently the scientific director of the Advanced Test Reactor National Scientific User Facility at INL. Before joining the University of Wisconsin, he was a nuclear engineer at ANL-West in Idaho Falls, ID. His Ph.D. is in Nuclear Engineering from the University of Michigan (1997), and his B.S. is in Nuclear Engineering from Northwestern University (1984). Before his graduate work, he was an officer in the U.S. Navy nuclear power program.
Presentation Summary Dr. Allens presentation concentrated on the environmental effects in SFRs and lead fast reactors (ADAMS Accession No. ML20030B791). Dr. Allen discussed the potential susceptibility to corrosion and methods of corrosion control in both types of reactor, which are very different. In SFRs, which use liquid sodium as the coolant, corrosion is controlled by limiting dissolved oxygen, thus limiting the dissolution of chromium. Operating experience from the Monju SFR in Japan showed that no corrosion of stainless steel was observed after the sodium was drained from the primary tank after 35 years. Chromium and molybdenum also have very low solubility in sodium. Iron and nickel do not form stable oxides in sodium so they could potentially undergo corrosion (however, the Monju experience did not appear to show this was a problem).
There is less experience with lead fast reactors. Russia has used lead alloys for military applications, but there is little experience in the United States, outside of test loops. In lead fast reactors, the primary corrosion control strategy is to limit dissolved oxygen in a band such that a stable thin oxide is formed and maintained. Lead and lead-bismuth alloys are being considered as the coolant. Silicon-containing structural alloys form a more stable oxide in lead, and 3-9
aluminum coatings are being studied in Europe for higher temperature applications. Between 300 degrees C and 470 degrees C, with sufficient oxygen, austenitic and ferritic steels form a protective oxide but not with lower oxygen or higher temperatures. With respect to mechanical properties in lead, austenitic steels appear to be little affected but mechanical properties of ferritic-martensitic steels can be affected and must be chosen carefully.
Environmental Effects in High-Temperature, Gas-Cooled Reactor Environments Kumar Sridharan, University of Wisconsin, Madison Dr. Kumar Sridharan is a professor in the Departments of Engineering Physics and Materials Science & Engineering at the University of Wisconsin, Madison. He has participated in research on materials surface modification technologies and corrosion effects in reactor-relevant environments for over 20 years. He is the author of over 300 publications in these areas, including seven invited book chapters, journal articles, conference proceedings, technical abstracts, and technical reports. He presently serves on the editorial committee of the journal Advanced Materials and Processes. Dr. Sridharan was elected as a Fellow of the American Society for Materials and as a Fellow of the Institute of Materials, UK, for distinguished contributions in the field of materials science and engineering, applications, and education.
Presentation Summary Dr. Sridharans presentation focused on corrosion control in HTGRs (ADAMS Accession No. ML20030B793). There are five decades worth of experience with this topic, from the HTGRs in the United Kingdom and the United States (Ft. St. Vrain). The ASME B&PV Code approves only five materials for HTRs, with a code case for Alloy 617. HTGRs use helium as the coolant, which will not corrode metals, but impurities in the helium such as oxygen and water (H2O), hydrogen (H2), methane (CH4), CO2, and CO can cause corrosion. Dr. Sridharan discussed the different regimes of corrosion in helium as a function of oxygen and carbon activation. Shifts in impurities at the parts-per-million level can change the corrosion mechanism. Chromium is the most important participant in corrosion reactions.
Dr. Sridharan then discussed the sources of impurities in HTGRs. Graphite plays a big role in reacting with other impurities to produce CO, H2, and CH4. CH4 can also come from leakage of oils. To minimize corrosion, it is necessary to form a thin, dense, adherent, thermodynamically and mechanically stable oxide layer, which prevents further oxidation and acts as a barrier for carburization and decarburization. The consequences of corrosion in HTGRs can include effects on mechanical properties, increased wear, and emissivity changes. However, no instances of mechanical failures have been observed due to corrosion in HTGRs.
Dr. Sridharan concluded with the following key points:
- Controlling total impurity content in helium to below 10 parts per million should be targeted.
- Molecular sieves are effective, but they cannot capture CO and H2; gas can be flowed over copper oxide to convert CO to CO2 and H2 to H2O.
- Back streaming of oils must be minimized.
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- Graphite in an HTGR core plays a dominant role in corrosion.
- The effect of flow-velocity and system pressure should be considered in corrosion prediction.
- Data mining and analysis of the large amount of literature on HTGR-helium corrosion of 800H and 617 will be very valuable.
- Since the alloy composition cannot be altered, surface modification approaches to promote passive oxide layer formation (making corrosion immune to the regime) are attractive options.
An Overview of Molten Salt Corrosion Research at ORNL Stephen Raiman, Oak Ridge National Laboratory Dr. Stephen Raiman is an R&D associate in the Corrosion Science and Technology Group at ORNL. He is interested in understanding the degradation of materials in extreme environments.
He graduated from the University of Michigan in 2016 with a Ph.D. in Nuclear Engineering and Radiological Sciences. He also holds a B.S. in Physics from the University at Buffalo.
Presentation Summary Dr. Raimans presentation focused on the important variables in the corrosion of metals in molten salt, from an experimental perspective (ADAMS Accession No. ML20030B787).
Dr. Raiman first made the point that metals degrade in molten salt by chromium depletion. At first glance, the data for mass change for a wide variety of materials and salt compositions show no observable trends. However, it seems that container material does not matter but salt purity does. Dr. Raiman then talked about experimental techniques. It is important to have flowing salt with a thermal gradient to accurately reproduce MSR conditions. Dr. Raiman showed a schematic of the thermal convection loop at ORNL and discussed recent experiments conducted in the loop featuring FliNaK salt and Type 316 stainless steel specimens at 540-650 degrees C. Mass loss tended to increase as the temperature increased. Dr. Raiman next discussed how to prepare good salts for experiments. Techniques include chemical dehydration and chemical chlorination to purify the salts. Salt purification, moisture and redox control additives, and capsule material all affect the mass change that was measured in experiments. Dr. Raiman finished with the thought that while consistent experiments and control of variables are important, salt chemistry matters most. When salt chemistry is good, other data emerge. Dr. Raiman noted that DOE can provide characterized salts for research under certain circumstances.
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WORKSHOP
SUMMARY
Overview of Vendor Designs (Monday, December 9, Session 1)
Several vendors summarized their proposed reactor designs and provided details on materials for structural and internal applications and testing to address data and knowledge gaps for these materials. MSRs, LMRs, and HTGRs are under development. A common feature of these advanced reactors is their operation at much higher temperatures than current commercial LWRs.
Challenges:
- No material is ready for off-the-shelf use; additional understanding is needed to satisfy design requirements.
- Regulatory and technical approaches need to focus on ensuring that safety goals are met. Each combination of reactor design and material poses unique issues.
Key Takeaways:
- A variety of nuclear technologies are being developed to fulfill the unique energy strategies being pursued by each vendor.
Technical and Research Activities (Monday, December 9, Sessions 2 and 3)
These sessions summarized research programs underway at major national and international organizations. The NRC and DOE summarized research supporting ANLWR material use and development. ONR and CNSC discussed regulatory and technical frameworks being used (or developed) for licensing ANLWRs. IAEA described activities to support ANLWR development and knowledge management.
In summary, significant technical and regulatory efforts are underway internationally, and the need for technical and regulatory flexibility is universally recognized. CNL discussed developing modeling and simulation tools using the evaluation of and benchmarking with NRU component and material properties and performance. JAEA summarized research activities and discussed a risk-informed structural integrity design framework. NRG summarized HFR capabilities and current research activities. EPRI and DOE described strategies for addressing research gaps and collaborating internationally.
Challenges:
- At present, ANLWRs provide much less commercial operating experience than conventional LWRs.
- Specific expertise in each ANLWR reactor type and the specific materials selected for use is needed to ensure an appropriate safety focus.
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- It is unclear how much collaboration and information sharing will be possible, given the proprietary nature of designs.
- Some research gaps (e.g., irradiation performance) are complex, costly, and lengthy to address.
Key Takeaways:
- Significant technical and regulatory efforts are underway internationally.
- The need for technical and regulatory flexibility is recognized.
- Leveraging international ideas, knowledge, and capabilities is a foundational strategy that can benefit all countries and organizations.
- Prioritizing research is important for efficiently using both resources and time.
Graphite (Tuesday, December 10, Sessions 1 and 2)
These sessions discussed the use of graphite in advanced reactor concepts, including the long and successful use of graphite as a core-support material in gas-cooled reactors. ONR summarized extensive graphite experience gained through Magnox reactors and AGRs in the United Kingdom. The lack of operating experience of graphite use in MSRs was also addressed. UC Berkeley provided an overview of the chemical behavior of graphite under molten fluoride salt exposure, and Virginia Tech explored graphite electrochemical behavior in a molten salt environment. NRG identified important characteristics for irradiated graphite studies and unique features of existing facilities. Both INL and ORNL discussed the qualification of graphite under the ASME code. INL described the philosophy and guiding tenets behind the ASME graphite and composite code, while ORNL explained how irradiation-induced property changes can affect lifetime predictions. NUMARK argued that the successful application of graphite materials should be used as a basis for addressing future needs and challenges.
Challenges:
- There is limited available operating experience with graphite in MSRs.
- Due to highly vendor-specific and proprietary manufacturing methods, it may not be possible to standardize graphite through the conventional approaches used for metallic materials.
- Performance-based, risk-informed standards for graphite may be necessary for ANLWR applications.
- Understanding irradiation effects on graphite, particularly dimensional changes, and chemical behavior of graphite in a molten salt environment are key gaps.
- Fundamental corrosion and irradiation mechanisms are not entirely understood.
Key Takeaways:
- Graphite has a long and successful performance history in gas-cooled reactors.
- Graphite has a finite lifetime.
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- Understanding both thermal activity and galvanic corrosion effects is important to develop effective mitigation strategies for commercial use
- Understanding irradiation effects is necessary to ensure reactor core safety over the intended plant lifetime.
Materials Qualification (Tuesday, December 10, Sessions 3 and 4)
These sessions discussed the challenges with qualification of materials for use in advanced reactors. Advanced materials qualification methods, such as physics-based modeling and data analytics, may provide an approach to accelerate the typical 10-year (or more) time cycle for qualification of structural materials for nuclear reactors.
Challenges:
- The design framework needs improvement to account for and address uncertainties in material performance due to operating in higher temperatures.
- Gaps in knowledge about high-temperature, long-time performance are challenging to fill with experimental data alone.
- Nonlinear and time-dependent material performance complicates model validation.
- Strategies to work effectively with both nuclear power plant vendors and material suppliers are needed.
Key Takeaways:
- Qualification success stories exist.
- Prior nuclear and nonnuclear experiences provide some foundational knowledge.
- Qualification will be a more difficult and lengthy process for the first new ANLWR materials until the process become more familiar and refined.
- Coupled and strategic use of modeling and experiments will likely be necessary to span the parameter space.
- Advanced materials qualification methods, such as physics-based modeling and data analytics, may provide an approach to accelerate the typical 10-year (or more) cycle for qualification of structural materials for nuclear reactors.
Inspection and Monitoring (Wednesday, December 11, Session 1)
NDE of ANLWR components will not be possible in the same way it is performed in LWRs due to the design of these components and the fact that ANLWRs will not have refueling outages like LWRs. Components such as CHXs will require new NDE techniques. Techniques such as online monitoring may be necessary to provide reasonable assurance of component integrity.
Japans Joyo and Monju SFRs provide some useful operating experience relative to non-LWR inspection. For ANLWRs, it may be necessary to implement a systems-based approach, such 4-3
as has been used in Japan and is incorporated into ASME B&PV Code,Section XI, Division 2, rather than a traditional inservice inspection program. In situ surveillance programs will also be important to monitor changes in materials properties due to irradiation, creep, and fatigue. The detailed methods for these surveillance programs do not currently exist, which represents a gap that should be addressed to support ANLWR operation.
Challenges:
- The conventional notion of using a refueling outage to conduct inspections is likely not possible.
- Long-term operation under harsh conditions will be required, making NDE reliability paramount.
Key Takeaways:
- Current uncertainties in material and component performance can be substantially reduced through effective inspection, monitoring, and surveillance programs.
- Vendors need to integrate these concepts directly into the reactor designs (e.g., online monitoring).
Molten Salt Chemistry (Wednesday, December 11, Session 2)
This is an evolving and complex field of study since various vendors are proposing several different salts, along with several different structural materials. NRG has a robust molten salt chemistry and irradiation testing program. ANL discussed the development of online monitoring of molten salt chemistry. Continued research is important to further understand molten salt chemistry, including both salt properties and modeling methods, and to refine monitoring techniques.
Challenges:
- Careful experimental design and control are needed to ensure that the targeted behavior is measured.
- Obtaining reliable data on thermophysical properties is necessary to inform models.
- Continued development of experimental techniques is needed to both understand and control molten salt chemistry.
Key Takeaways:
- Electrochemical methods show promise for chemistry control and monitoring.
- Salt composition is constantly changing during operation.
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Environmental Effects on Metallic Materials (Wednesday, December 11, Session 3)
The presentations in this session summarized the knowledge of environmental effects, and strategies for managing environmental effects, in LMRs, HTGRs, and MSRs. EPRI summarized the importance of coordinated test programs and use of harvested materials to address knowledge gaps (analogous to LWR materials). Knowledge gaps are greater for higher temperatures and for MSR environments. Uncertainty in performance requirements and environments for reactor applications makes it challenging to prioritize research activities.
Challenges:
- The biggest knowledge gaps relate to higher temperatures and MSR environments.
- Uncertainty in performance requirements and environments for reactor applications makes it challenging to prioritize research needs.
Key Takeaways:
- The research framework used to address environmental effects in LWRs remains applicable for addressing structural material performance in ANLWRs.
- While unique environmental effects exist for each reactor type, the understanding of LWR performance provides a good basis for understanding effects, especially in LMRs and HTGRs.
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APPENDIX A : WORKSHOP ATTENDEES First Name Last Name E-Mail Organization University of Todd Allen traumich@umich.edu Michigan Tai Asayama asayama.tai@jaea.go.jp JAEA Ronald Ballinger hvymet@mit.edu ACRS Benjamin Beasley benjamin.beasley@nrc.gov NRC Michael Benson michael.benson@nrc.gov NRC Bezerra de Uazir Oliveira bezerra@nrg.eu NRG Bud Brust bbrust@emc-sq.com EMC2 Mike Burke mburke@epri.com EPRI Anne Campbell campbellaa@ornl.gov ORNL Carolyn Fairbanks carolyn.fairbanks@nrc.gov NRC Alexander Chereskin alexander.chereskin@nrc.gov NRC Ganesh Cheruvenki ganesh.cheruvenki@nrc.gov NRC Anne Co co.5@osu.edu OSU Steve Cook stephen.cook@canada.ca CNSC Bill Corwin advrxmatls@gmail.com David Diec david.diec@nrc.gov NRC Cyril Draffin cyril.draffin@usnic.org USNIC Molly-Kate Gavello molly-kate.gavello@nrc.gov NRC Bill Glass bill.glass@pnnl.gov PNNL Hipolito Gonzales hipolito.gonzalez@nrc.gov NRC Russ Green russell.green@onr.gov.uk ONR Brian Harris brian.harris2@nrc.gov NRC Nathan Hall nathan.hall@swri.org CNWRA James Hammelman james.hammelman@nrc.gov NRC Kurt Harris kurt.harris@flibe-energy.com Flibe Energy Michelle Hayes michelle.hayes@nrc.gov NRC Nathaniel Hoyt nhoyt@anl.gov ANL Gabriel Ilevbare gabriel.ilevbare@inl.gov INL Jay Wallace jay.wallace@nrc.gov NRC Andrew Johnson andrew.johnson@nrc.gov NRC Colin Judge colin.judge@inl.gov INL Matthew Kerr matthew.kerr@inl.gov INL Paul Kirchman pkirchman@x-energy.com X-Energy Paul Klein paul.klein@nrc.gov NRC P Krishnaswamy kswamy@emc-sq.com EMC2 A-1
First Name Last Name E-Mail Organization Vladimir Kriventsev v.kriventsev@iaea.org IAEA Stephen Lam stlam@mit.edu MIT Christina Leggett christina.leggett@nrc.gov NRC Amanda Leong aleongsw@vt.edu Virginia Tech Walter Leschek walter.leschek@nrc.gov NRC Sue Lesica sue.lesica@nuclear.energy.gov DOE Zhian Li zhian.li@nrc.gov NRC Jie Lian lianj@rpi.edu RPI Louise Lund louise.lund@nrc.gov NRC Timothy Lupold timothy.lupold@nrc.gov NRC Sandia National David Luxat dlluxat@sandia.gov Laboratories Makuteswara Srini Srinivasan msrinivasan@numarkassoc.com NUMARK Shah Malik shah.malik@nrc.gov NRC Matthew Memmott memmott@byu.edu BYU Mark Messner messner@anl.gov ANL Seung Min seung.min@nrc.gov NRC Carol Moyer carol.moyer@nrc.gov NRC Lauren Ning laurenkillian.ning@nrc.gov NRC Carol Nove carol.nove@nrc.gov NRC Mark Nutt mark.nutt@pnnl.gov PNNL Jonathan Parker jparker@epri.com EPRI Patricia Paviet patricia.paviet@pnnl.gov PNNL Tom Pham tom.pham@nrc.gov NRC Jeffrey Poehler jeffrey.poehler@nrc.gov NRC Yarom Polsky polskyy@ornl.gov ORNL Pat Purtscher patrick.purtscher@nrc.gov NRC Stephen Raiman raimanss@ornl.gov ORNL Raj Iyengar raj.iyengar@nrc.gov NRC Appajosula Rao appajosula.rao@nrc.gov NRC Wendy Reed wendy.reed@nrc.gov NRC Weiju Ren renw@ornl.gov ORNL Ali Rezai ali.rezai@nrc.gov NRC Marcos Rolon Acevedo marcos.rolonacevedo@nrc.gov NRC David Rudland david.rudland@nrc.gov NRC Raluca Scarlat scarlat@berkeley.edu UC Berkeley Frank Schaaf treecode@cs.com SRC Greg Selby gselby@epri.com EPRI A-2
First Name Last Name E-Mail Organization Farshid Shahrokhi f.shahrokhi@framatome.com Framatome Sam Sham ssham@anl.gov ANL Taran Sondhu taran.sondhu@flibe-energy.com Flibe Energy University of Kumar Sridharan kumar.sridharan@wisc.edu Wisconsin Casper Sun casper.sun@nrc.gov NRC Cheng Sun cheng.sun@inl.gov INL Cem Topbasi ctopbasi@epri.com EPRI Ricardo Torres ricardo.torres@nrc.gov NRC Robert Tregoning robert.tregoning@nrc.gov NRC John Tsao john.tsao@nrc.gov NRC Tjark van Staveren vanstaveren@nrg.eu NRG Lorenzo Vergari lorenzo_vergari@berkeley.edu UC Berkeley John Vetrano john.vetrano@doe.science.gov DOE-BES Greg Vetterick gvetterick@terrapower.com TerraPower Lori Walters lori.walters@cnl.ca CNL University of Wisconsin, Jun Wang jwang564@wisc.edu Madison Xuejun Wei xuejun.wei@canada.ca CNSC Michael Welland michael.welland@cnl.ca CNL Haley Williams haley_williams@berkeley.edu UC Berkeley Will Windes william.windes@inl.gov INL Richard Wright richard.wright@inl.gov INL Andrew Yeshnik andrew.yeshnik@nrc.gov NRC George Young young@kairospower.com Kairos Power Jinsuo Zhang zjinsuo5@vt.edu Virginia Tech University of Tennessee, Steve Zinkle szinkle@utk.edu Knoxville A-3
APPENDIX B : PRESENTATION SLIDES SALT COMPOSITION, CORROSION, AND TRITIUM CONTROL IN THE KP -FHR G EO R G E A . YO U N G , G U S M E R W I N , M I C H A E L H A N S O N ,
M I C H A E L Z U PA N , A N D S T E V E N H UA N G KAIROS POWER A L A M E DA C A U.S. NRC December 2019 1
Kairos Powers mission is to enable the worlds transition to clean energy, with the ultimate goal of dramatically improving peoples quality of life while protecting the environment.
U.S. NRC December 2019 2
Outline
- Salt chemistry: Flibe
- Impurity effects on materials performance
- Corrosion Testing & Analysis Strategy
- Redox Control
- Materials for Tritium Management
- Summary U.S. NRC December 2019 3 Kairos Power High Temperature Fluoride Cooled Reactor (KP-FHR)
Approximate Temperatures
- Primary Flibe Salt Temperatures 550-650°C
- Intermediate Nitrate Salt Temperatures 360-600°C
- Steam Generator Temperatures 300-585°C U.S. NRC December 2019 4
Flibe Composition Guidelines U.S. NRC December 2019 5
Allowable Impurities in Flibe Impurity MSRE General Specification Allowable Concentration1 (wt.%) Expected Kairos Power Oxygen (Water) 0.1 Cu 0.005 Fe 0.01 Ni 0.0025 S 0.025 Cr 0.0025 Al 0.015 Si 0.01 B 0.0005 Na 0.05 Ca 0.01 Mg 0.01 K 0.01 Li (natural) 0.005 Zr (natural) 0.0025 Cd 0.001 1Shaffer, J.H. (1971). Preparation and Handling of Salt Mixtures for the Molten Salt Reactor Experiment, Oak Ridge National Laboratory.
U.S. NRC December 2019 6
Allowable Impurities in Flibe
- It may preferable to set an integrated limit on impurities
- Kairos Power is assessing controlling several corrosive impurities (table to the right) in total, i.e. impurities < ### wt.
ppm
- Many of these are dissolved fluorides that can result in chromium oxidation via:
2 2
+ +
U.S. NRC December 2019 7 Corrosion & Redox Control U.S. NRC December 2019 8
Initial Corrosion Testing (University of Wisconsin, FCL-0)
- Material: 0.5 thick, hot rolled 316H Plate (ASTM)
- Condition: Annealed at 1052°C (1925°F) min. and water quenched
- Machined coupons (0.58x0.48x0.0625)
- Coupon orientation L-T
- Exposed 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> / T = 700°C (hot), 650°C (cold),
Flow = 2m/s U.S. NRC December 2019 9 Initial Corrosion Data: 316, 700°C, Flibe, Flowing Loop (FCL-0)
- Sample shows Cr depletion rates Temperature Dependence of Corrosion, 316 in Flibe consistent with the literature 977oC 838oC 727oC 636oC 560oC 496oC 441oC Relatively impure salt 3 1000 LOG Corrosion Rate (LOG m/year)
Keiser 1979 Not steady state corrosion rate Kondo 2009 316 Coupled to Zheng 2015 Corrosion Rate (m/year) graphite (Zheng) Koger 1972 Flibe+ThF4+UF4 Rates based on weight change 2 100 Kairos Power (wt. loss) 2019 consistent with literature Q ~56 kJ/mol 1 10 See Data of Zheng et al., JNM 2015, testing in 700°C Flibe - 0 1 The Kairos data Be redox control Are consistent with this (Keiser) study
-1 0.1 0.00 0.00 0.00 0.00 0.00 0.00 0.00 08 09 10 11 12 13 14
-1 Reciprocal Temperature (K )
U.S. NRC December 2019 10
Conceptual Overview - Flibe Chemistry Control
- The Flibe in KP-1 will be buffered by dissolved Be metal. The this method of corrosion protection was demonstrated in the MSRE program
- Beryllium will react with oxidants and impurities to See Keiser, DeVan, and Lawrence, JNM Vol.
protect structural alloys 85&86, 1979. With Be addition, 316 corrosion
+ = + rates < 2 microns per year Ex: +2 = +
- The concentration of Be in the Flibe will be controlled Weight loss of 316 stainless steel at the KP-FHR chemistry control system, monitored by 650oC in Flibe with and without beryllium metal additions an electrochemical probe
- Maintaining a controlled concentration of dissolved Be will fix the redox potential of the Flibe and control the chemistry of radionuclides (tritium, activation products, etc.)
J.R. Kaiser, J.H. DeVan, E.J. Lawrence, Compatibility of molten salts with type 316 stainless steel, J. Nucl.
Mat., 85/86, 295-298.
U.S. NRC December 2019 11 Corrosion Coupon Characterization - WDS Mapping
- Cr loss as expected
- But, uniform attack at 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, no IG penetration
- Mo rich phase present in corroded layer Mo rich Laves
- Cr loss likely a conservative corrosion metric Cr Depletion Cr Fe Ni Mo U.S. NRC December 2019 12
Corrosion Analysis Zheng, G., et al. (2015). "Corrosion of 316 Stainless Steel in High Temperature Molten Li2BeF4 (FLiBe) Salt." Journal of Nuclear Materials 461: 143-150.
- XRD shows notably higher ferrite on Flibe exposed As-machined the Flibe exposed corrosion surface
- Zheng et al. (Wisconsin) also reported high ferrite in 316 exposed to Flibe (700°C/3000 hours), attributed to the corrosion environment
- How do you lose a ferrite stabilizer (Cr) and form more ferrite?
computational thermodynamics U.S. NRC December 2019 13 Phase Predictions from Corroded Layer
- Thermodynamic modelling of corrosion layer composition via JMatPro and FactSage Fe-10.92Ni-8.26Cr-2.03Mo-0.47Mn-0.40Cu-0.13Si-0.02V
- Key points At 700°C, austenite is stable, ferrite likely forms on cooling 2 phase region below about Corrosion tests run here Laves is Mo rich ~Fe2Mo U.S. NRC December 2019 14
Electron Backscatter Diffraction of Corroded Layer
- Annealed, equiaxed 316H sample
- EBSD Confirms ~50% ferrite (red) in the corroded layer
- Corroded layer has significant plastic strain (machining damage?)
U.S. NRC December 2019 15 Materials for Tritium Management U.S. NRC December 2019 16
Tritium Generation in KP-FHR
- Tritium will be generated due to neutron reaction with Flibe
- KP-FHRs Tritium Management Strategy is to assure that all environmental releases are monitored and comply with licensed pathways and limits
- The chemistry control system in the KP-FHR will keep the tritium in a reduced state as T0 or T2 The formation of tritium fluoride will be mitigated through reactions with beryllium metal to avoid corrosion
- Kairos Power is developing several methods by which tritium transport will be controlled, e.g. low permeability cladding
- Kairos Power is working with MIT to model tritium production, diffusion pathways and release via the TRIDENT code
+ +
+ + +
+ + +2
+ + ( = 0.8 )
U.S. NRC December 2019 17 Materials Development: Cladding
- Currently evaluating several coatings (carbide, oxide, metallic) and methods (thermal spray, cold spray, explosion bonding, etc.)
- Note Kairos / ANL (Messner & Sham) GAIN to develop ASME rules for corrosion resistant cladding Explosion Bonded W U.S. NRC December 2019 18
Summary
- Kairos Power has understanding and experience with impurity control for Flibe Ability to meet or exceed historic composition guidelines Currently purifying salt for upcoming corrosion and stress corrosion testing programs
- Scoping corrosion testing indicates 316 SS rates comparable with literature data Note rates in nominal Flibe, conservative definition of corrosion rate via Cr loss Redox control of the salt provides additional benefit Aware of potential metallurgical complications to testing
- Materials for Tritium Management Developing several methods of mitigation (trapping, permeation, recovery, etc.)
Working with ANL to develop ASME rules to use corrosion resistance cladding Working with MIT to model tritium production and release Goal is to achieve comparable release (or less) relative to conventional PWRs U.S. NRC December 2019 19
Overview of LFTR: Materials and Components Kurt Harris, PhD Flibe Energy, Inc.
kurt.harris@"ibe-energy.com Nuclear Regulatory Commission Rockville, Maryland December 9, 2019 Three Nuclear Options
The Thorium Reactor Company Located in Huntsville, Alabama
Gen-4 Molten Salt Reactor Concept The MSRE successfully operated for over 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, from 1965-1969.
MSR Material Compatibility At ORNL, scientists and engineers made excellent materials decisions, culminating in the MSREs demonstration of materials compatibility in a neutron environment between FLiBe salts, graphite, and Hastelloy N.
The feasibility of a nuclear reactor that used LiF-BeF2 "uoride salt as a coolant with UF4 as a nuclear fuel was demonstrated in the United States in 1965.
The nuclear fuel heated a coolant salt in a simple heat exchanger.
The reactor had the potential for high performance because it could generate high temperatures at low pressure in a compact unit.
The reactor and its primary systems were compact.
Operators enjoyed the fact that the reactor was inherently safe and easy to control.
Successful operation for over 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> was demonstrated.
Liquid fuels enhance safety options The reactor is equipped with a "freeze plug"an open line where a frozen plug of salt is blocking the "ow.
The plug is kept frozen by an external cooling fan.
In the event of total loss of power, the freeze plug melts and the core salt drains into a passively cooled con"guration where nuclear "ssion and meltdown are not possible.
Drain tank
Design constraints
fuel salt drain tank must be designed to be prepared to receive the entire inventory of hot fuel salt immediately after shutdown
insigni"cant neutron "ux, strongly subcritical due to no graphite
Non-emergency use parameters
to help keep drain tank at operating temperature, plan to utilize volume for decay of noble gases
cooling requirement for noble gas decay is similar to "ssion-product decay heating immediately after shutdown
small amount of fuel salt held up in drain tank for decay, prior to transfer to chemical processing system
Drain tank represents the "emergency core cooling system" MSRE salt drain tank design shown, b td t t LFTR d i
Two-Fluid MSBR Reactor Module and Core Cutaway Coolant Choices for a Nuclear Reactor atmospheric high-pressure pressure operation operation Metal Water moderate temperature (250-450 C)
Salt Gas high temperature (650-900 C)
Fluoride salts are safe and versatile Chemically stable in air and water Unpressurized liquid with 1000 C range of temperature
Fluorine has extremely high electronegativity, allowing "uoride salts to retain most "ssion products.
Free energies of formation of oxides at 1000 K 50 Free Energy of Formation (kcal/mol O2 )
100 150 200 250 300 Th Gd Pr Nd La Ce Sr U Zr Ba Pu Cs
Free energies of formation of "uorides at 1000 K 100 Free Energy of Formation (kcal/mol F2 )
150 200 250 300 Ba Sr La Gd Pr Ce Nd Cs Th Pu U Zr Free energy differences (oxides/"uorides) at 1000 K 0
Free Energy Differences 50 100 150 200 Cs Ba Sr La Pr Ce Gd Nd Pu Th U Zr In "uoride form, all actinides and alkaline "ssion products -
most notably cesium and strontium - remain in "uoride salt form in the presence of air, and dont form volatile species. In molten salts, the "rst barrier to "ssion product release is the chemical form of the fuel salt, rather than the mechanical integrity of the fuel pin.
LiF-BeF2 "uoride salt is an excellent carrier for uranium (UF4 ) nuclear fuel.
Element Absorption Cross-Sections Only a few elements have neutron absorption cross sections low enough to be of interest, along with suitable chemistries.
heavies Ce Pb Bi transitions Sn Zr group 16 Se S O group 15 As N P 15 group 14 Si C group 13 Al 11 group 17 Cl 37 F group 2 Ca Mg Be group 1 Li K Na aRb HD 7 hydrogen H 2 102 101 100 101 102 103 104 105 Neutron Absorption Cross-Section (barn)
Melting Temperatures of Salt Constituents Individual "uoride salt components tend to have melting points too high (>600 C) to be useful in practical reactors.
PuCl3 PuF3 plutonium UCl4 UCl3 UF4 uranium ThCl4 ThF4 thorium PbCl2 PbF2 lead ZrF4 zirconium RbCl RbF rubidium CaCl2 CaF2 calcium CaCO3 KNO3 KCl KF potassium K2 CO3 MgCl2 MgF2 magnesium NaNO3 NaCl NaF sodium BeCl2 BeF2 Na2 CO3 beryllium LiNO3 LiCl Li2 CO3 LiF lithium 200 300 400 500 600 700 800 900 1,000 1,100 1,200 1,300 1,400 1,500 Temperature ( C)
Lithium Fluoride Melting Temperatures 1200 CaF2 1110 1100 NaF UF4 1036 ThF4 996 1000 BeF2 Melting Temperature ( C) 900 848 800 chromium migrates in Hastelloy-N 700 600 551 500 400 300 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 Composition
Neutron reactions in LiF-BeF2 salt n,
7 3 Li 83 Li 84 Be 42 He 0.045 b 0.842 sec 67 asec n ,2n 9
4 Be 84 Be 42 He 0.160 b 67 asec n,
19 9F 209 F 20 10 Ne 0.009 b 1.1 sec n,
9 4 Be 104 Be 105 B 0.009 b 1.6 Myr n,
6 3 Li 31 H 32 He 941 b 12.3 yr n ,2n 6 n, 3 7
3 Li 3 Li 1 H 32 He 68 b 941 b 12.3 yr n , n, 9
4 Be 63 Li 31 H 32 He 0.051 b 941 b 12.3 yr Neutron reactions in LiF-BeF2 salt n,
6 3 Li 31 H 32 He 941 b 12.3 yr n ,2n 6 n, 3 7
3 Li 3 Li 1 H 32 He 68 b 941 b 12.3 yr n , n, 9
4 Be 63 Li 31 H 32 He 0.051 b 941 b 12.3 yr Residual lithium-6 forms tritium in a neutron "ux, and even if it could be eliminated, fast neutron reactions in lithium-7 and beryllium recreate lithium-6. Tritium shall be captured to the maximum degree practical at each stage in the reactor system, with the summation of these capture techniques minimizing tritium release to a degree that is acceptable from a licensing basis.
Graphite undergoes dimensional change Under sustained irradiation, the lattice structure of graphite changes. First it contracts, then it expands steadily. This effect is enhanced by an increase in temperature. Ultimately, it leads to a need for graphite replacement.
To meet the challenges of liquid-"uoride salt mixtures, metallurgists at Oak Ridge developed a high-nickel alloy they called "INOR-8", but which is now more commonly known as "Hastelloy-N".
Structures Fabricated from INOR-8 From this material they fabricated all the structures that would be in contact with the "uoride salt mixture.
Sustainable Use of Thorium is the Ultimate Goal recycle "ssion transmute Uranium Thorium Uranium 233 232 233 neutrons decay Uranium-233 alone produces suf"cient neutrons per thermal neutron to match or exceed its consumption. In this way, U-233 is the "catalyst" to sustainable energy production from thorium and can almost eliminate transuranic production.
Fluoride Thermal Reactor Material Flows Fluoride Thermal 1000 MW Reactor "ssion thorium products uranium-233 Potential Steady-State Waste Pro"le
Conclusions
Flibe Energy is developing the Liquid Fluoride Thorium Reactor (LFTR)
Two-"uid MSR with chemical processing, off-gas treatment and sequestration, and sCO2 PCS
FLiBe salts (HD Li), graphite channels also moderate, and thin-walled Hastelloy N vessel and piping for low pressures and corrosion
Minimal stored energy or driving forces for radionuclide release, inherent safety features
Fuel cycle potential: no mining, no enrichment, minimal "ssile inventories, minimal "ssile transportation, no excess "ssile production, minimal "ssion product inventory during operation, steady-state and manageable waste stream
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Non-Linear Mechanical Modeling in Materials Subject to Property Changes due to Neutron Irradiation NRC 2019 Materials Workshop Paul Kirchman Principal Structural Analyst 09DEC2019
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined.
Background
Nuclear reactor reflector blocks undergo significant changes in their physical properties due to radiation damage over time.
The conditions that induce material property changes in NBG (Nuclear Block Graphite) are dependent on the cumulative radiation exposure and the temperature at which the exposure occurs The temperature and fluence vary throughout any given component leading to complex stress/strain fields Radiation induced changes in component size are significant and need to be accounted for in tolerance stacking for mechanical analysis and bypass flows for CFD analysis
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 2
Assumptions Uniform radiation induced dimensional change creates strain without stress in the manner of thermal expansion. Stress in this scenario results from non-linear gradients of temperature, dimension change, and external forces.
The stress relief effect of mechanical creep is critical to the calculation of stresses to be evaluated for failure probability The blocks in a typical design are otherwise lightly stressed due to their own weight, pressure due to fuel pebbles and seismic loading.
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 3 Assumptions (continued)
The temperature and rate of radiation dosing remains constant at each location allowing for a linear solution at each cumulative dose level as the properties undergo non-linear variation.
The available property data represents cumulative change so that time integration is entrained in the input curves.
The reactor is anticipated to operate in the nominal predicted condition for the majority of its active life with only transitory variations making it possible to sum full power equivalent time for analytical evaluation.
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 4
Approach From material test data, create a continuous numerical model of the changing properties using standard math libraries.
Interface numerical model to any FEM solver through object oriented programming interface, in this case FEMAP.
- Calculate unique material properties for each analysis element based on input temperature and fluence fields
- Apply new properties to elements and execute FEM solver Automate this process for use on multiple load cases, candidate materials and model geometries.
Include iterative capability to step through sequential cases making a time history - note that maximum stresses occur during maximum gradient conditions which are not necessarily at end of life
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 5 Numerical Properties Model Application Flow
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 6
Material Property Raw Input Material Property test data input as xy pairs with no restrictions on order or number of pairs Params Values 300C Dim wg 400C Dim wg 500C Dim wg 750C Dim wg 300C Dim ag 400C Dim ag 500C Dim ag 750C Dim ag
- properties 6 0 0.01 0 0.01 0.01 0 0.01 0 0.01 0 0.01 0 0.01 0 0.01 0
- temps 4 8.9 -1.1 6.5 -1 3.7 -0.6 1.8 -0.2 10.4 -2.1 6.1 -1.2 2.5 -0.5 1.8 -0.1
- extrap 4 17 -1.9 14.1 -2.1 13.5 -2.1 2.2 -0.3 19 -3.4 13.1 -2.9 3.5 -0.7 2.3 -0.2 iterations 1 25.4 -1.8 19.7 -1.9 20 -1.4 2.5 -0.3 26.6 -3.6 18.2 -3.5 13.4 -3.2 2.7 -0.3 dpa Begin 60 33.9 -0.1 20.9 -1.6 25.3 0.6 3.4 -0.6 34.3 -2.4 19.4 -3.5 19.8 -3.4 3.8 -0.4 dpa End 61 16.8 -1.6 30.2 1.7 36.4 8.2 3.8 -0.6 17.1 -3.1 29 -2.1 25.5 -2.4 4.1 -0.5 Prop 1 Dim wg 27.6 -1.4 7.4 -1.2 3.3 -0.5 4.5 -0.8 27.7 -3.6 7.5 -1.5 3.4 -0.7 4.7 -0.5 Prop 1 p Dim ag 38.5 0.4 15.5 -2.1 7.9 -1.6 4.7 -0.9 38.5 -2.5 14.7 -3.2 11.2 -2.8 4.7 -0.5 Prop 2 E wg 47.5 3.1 22.8 -1.5 15.2 -2 8 -1.3 47.8 -0.4 21.9 -3.4 18.5 -3.2 5.1 -0.7 Prop 2 p E ag 17.1 -1.8 24.1 -1 10.4 -1.8 11.2 -1.6 17.2 -3.2 23.2 -3.2 24.7 -2 5.6 -0.5 Prop 3 CTE wg 27.8 -1.4 33.7 2.7 16.8 -2 12.2 -1.5 27.7 -3.7 32.6 -1 34.4 2.5 10.1 -0.8 Prop 3 p CTE ag 13.9 -1.6 3.3 -0.5 3.7 -0.6 12.2 -1.6 14.8 -2.8 3.9 -0.7 8.5 -2.2 10.3 -0.6 Order P1 3 23.9 -2 6.3 -1.2 13 -2.2 16.4 -0.6 25.1 -3.9 7.7 -1.8 16 -3.4 13.3 -0.7 Order P1p 3 34.7 -0.4 12 -2.1 20.5 -1.4 16.8 -0.8 35.4 -3 13.7 -3.2 3.7 -0.7 13.8 -0.3 Order P2 3 45.8 2.2 13.2 -2.2 26.8 1 17.1 -1.1 45.9 -1.5 14.8 -3.3 3.4 -0.7 14.6 0.2 Order P2p 3 16.9 -2 21.4 -1.7 3.3 -0.6 22.6 1.8 16.4 -3.2 22.5 -3.2 12.7 -3 17.8 0.9 Order P3 5 27 -1.7 10.4 -1.7 3.2 -0.6 22.4 4.3 26 -3.6 10.5 -2.3 20.2 -3.1 18.2 1.9 Order P3p 5 36.5 -0.5 19.7 -2 5.4 -1 36 -3.1 19.5 -3.6 26.4 -1.8 19.7 3.7 Ezero 9.00E+10 46.2 2.2 20.6 -1.7 11.6 -2 46.7 -1.4 20.5 -3.5 3 -0.6 CTEzero 4.43E-06 11.7 -1.5 28.1 0.3 6.7 -1.4 12.9 -2.6 28.6 -2.4 3 -0.6 Temp 1 300 21.7 -2.2 7.4 -1.2 13.6 -2.4 23.2 -3.8 7.9 -1.7 10 -2.3 Temp 2 400 32.3 -0.8 16.4 -2.2 26.4 0.2 33.4 -3.3 17.2 -3.5 15.9 -3.3 Temp 3 500 42 1.7 17.4 -2.2 6.7 -1.4 43.9 -1 18.1 -3.5 6 -1.4 Temp 4 751 16.1 -2 24.5 -0.6 13.6 -2.2 15 -2.8 24.4 -2.8 12.6 -2.9 ExTemp1 300 25.6 -1.9 6.6 -1.1 9.7 -1.9 23.8 -3.6 6.9 -1.5 7.4 -1.9 ExTemp2 400 34.9 -0.7 14.2 -2.2 15.6 -2.3 33.6 -3.4 14 -3.1 14.5 -3.2 ExTemp3 500 44.6 1.9 19.8 -2.1 20.2 -1.7 43.9 -1.8 19.1 -3.6 7.4 -1.9 ExTemp4 750 5.3 -0.9 9.5 -1.8 5.8 -1.2 14.5 -3.3 10.6 -1.8 14.8 -2.4 11.8 -2.5 9.7 -2.3 17.3 -2.2 19.4 -2 18.7 -3.3 15.6 -3.2 5.7 -0.9 9.7 -1.9 6.2 -1.3 20.5 -0.7 11.6 -1.9 16.1 -2.3 12.7 -2.8 9.5 -2.2 18.4 -2.3 7.7 -1.7 19.7 -3.6 14.8 -3.2
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 7 Material Property Curve Fit Data points fitted to polynomial curves and plotted as a diagnostic
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 8
Scripts Developed in Python Object oriented, powerful, compact, and simple coding.
- Analysis speed bound by program interface Open source with straightforward syntax and extensive specialty libraries.
Naturally interfaces with an object oriented FEM pre and post processing toolkit like FEMAP.
Has libraries available with the numerical analysis capability of MATLAB Provides a single language for:
- the development of the numerical model
- Application to FEM
- Post processing
- Results evaluation and report generation.
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 9 FEMAP Interface Object oriented programming interface with access to every aspect of FEM model creation, load application, solver execution, and results evaluation.
Accepts geometry from any CAD program through STEP files.
Solver independent since it maintains FEM models in a generic object space.
Creates input files compatible with dozens of commercial FEM solvers.
Extensive post-processing capabilities once FEM results are imported into the FEMAP database.
Results are accessible by the Python script making automated failure probability reporting possible.
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 10
Simplified Test Model This allows both a fast turnaround for example runs as well as checks against textbook cases.
A ten by ten by 20 cm model supported kinematically along the 20 cm axis passed checks for no stresses generated by a uniform free expansion as well as expansion with a linear gradient.
Free expansion displacements of a uniformly applied dimensional change are also zero Free expansion of a linear expansion gradient result in zero stresses in this simplified case Stresses from a high thermal gradient (shock) match results from Roarks Formulas for Stress and Strain
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 11 Load Application Temperature and radiation dose loads are presently accepted in table form in cylindrical coordinates centered along the axis of the reactor vessel.
These loads have rotational symmetry and reduce to a function of the radial distance from the center and the vertical station.
Input values for the region being evaluated are fitted to a polynomial and applied by location to the model
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 12
Calculation and Application Material properties are adjusted based on conditions unique to the location of each element.
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 13 Calculation of Intermediate Property Values The dots overlaid on the property model curves represent individually computed solutions based on input conditions. Values to the far right represent the core facing side grading to the outward facing side on the left.
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 14
Verification of creep model Coupon from creep testing described in Reference 2, Section 6
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 15 Verification of creep model Coupon from creep testing described in Reference 2, Section 6 500C Tensile Creep test vs model 4
3 2
percent dim change 1
0
-1
-2
-3
-4 0 5 10 15 20 25 30 35 40 dpa data stressed model unstressed data unstressed appendix model stressed .1725
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 16
Test Case FEM Results Progressive material change as radiation dose increases at approximately one year per frame.
Displacements exaggerated to highlight dimensional change over time and show turnaround.
© 2017 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 17 Block Interaction Block interaction is complex and requires non-linear contact effects
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 18
Reference Blocks Generic Reference Blocks for Independent Review
- A set of generic graphite blocks using ATR2-E properties was generated and solved for displacements and stresses using notional HTGR fluence and temperature loading fields.
- To be exchanged with an external entity for validation and documented in Reference 3
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 19 Reference Blocks Generic Reference Blocks for Independent Review
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 20
Reference Blocks Example Temperature and Fluence 1.2 Test case Simple Blocks : Temperature and Dose profiles 600 Dose 1 Y Temp 1 500 0.8 400 Temperature (°C)
Dose (dpa)
Temperature = -65747x4 + 373207x3 - 792903x2 + 746730x - 262414 0.6 R² = 0.9949 300 0.4 200 0.2 Dose = 130615e-9.955x 100 R² = 1 0 0 1.15 1.2 1.25 1.3 1.35 1.4 1.45 1.5 1.55 1.6 Radial Position
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 21 Reference Blocks Example Results
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 22
Reference Blocks Example Results
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 23 References Reference 1 - Properties of ATR-2E Graphite and Changes due to Fast Neutron Irradiation Gerd Haag, Berichte des Forschungszentrums Jülich ;
4183, ISSN 0944-2952, Institut für Sicherheitsforschung und Reaktortechnik Jül-4183 Reference 2 Reactor System Coupon Model Reference Analysis Report X-Energy XE01-TS3-Sw-D 000105 Jun 2018 Reference 3 Graphite Reference Block Analysis Benchmark Report X-Energy XE01-N-R-Z-D 000110 Aug 2018
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined. 24
End of Presentation NRC 2019 Materials Workshop
© 2019 X Energy, LLC, all rights reserved Nuclear Energy. Reimagined.
Materials Selection and Development for Advanced Reactors at TerraPower Greg Vetterick COPYRIGHT© 2019 TERRAPOWER, LLC. ALL RIGHTS RESERVED.
TerraPowers Reactors Shared Benefits Deep-burn open fuel cycle with no reprocessing Strong inherent safety features and passive safety systems Flexible siting Production of high temperature heat Reduced waste TWR MCFR Based on high-readiness Based on novel sodium cooled fast molten chloride salt reactor technology technology 2 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Key Materials Challenges TWR© MCFR High endurance materials to enable Structural (e.g. ASME B&PV Code) materials breed-and-burn see fuel salt Swelling and creep Fuel/coolant-component interactions Clever design choices and good project planning are needed to drive down cost and reduce time to market 3 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Material Selection Driven by Timeline The market is looking for reactors around 2030. Heres a hypothetical timeline for an SFR:
Formal Design First First PSAR Concrete FSAR Criticality Start 2019 2020 2021 2022 2023 2024 2025 2026 2027 2028 2029 2030 Conceptual Design Preliminary Design Final Design Planning and BD PSAR Prep Licensing Review Construction, Licensing Review Commissioning Operation Core Assembly Fabrication Early design assumptions (core size, outlet Core Component Fabrication Core temperature, refueling schedule, economics)
Materials Raw Material Fabrication are driven by the core material choices Materials testing for Selected Core Component Manuf. Qual. qualification & licensing is expensive and time-Core Component Manuf. Development and Long core component lead time means vendors consuming Testing must be already working with the selected Matl. Fab material early in design process Neutron Irradiation 15 dpa 30 dpa 45 dpa 60 dpa 75 dpa 90 dpa 105 dpa 120 dpa 135 dpa 150 dpa Creep/Aging 9k hrs 17k hrs 26k hrs 35k hrs 43k hrs 52k hrs 61k hrs 70k hrs 78k hrs 87k hrs 4 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Timeline for Commercialization Materials selection has a strong impact on early, major design decisions Testing has to start early Momentum is important to maintain Common, well-understood materials take priority Demo Reactor Commercial Prototype 30 - 150 MWth Reactor Integrated Effects Test Class 104 License & 600 - 2500 MWth e.g., Separate Effects Tests (IET) & Materials/ Component Scale up/ Class 103 License MCFR Component testing Materials Validation
& Materials Scoping Today 2019 -2022 2020-2027 2025 - 2030s Phase 1 - TRL 1 to 5 Phase 2 - TRL 5 to 7 Phase 3 - TRL 8 to 9 5 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Fuel Salt Development Materials selection for the MCFR is fundamentally dependent on the fuel salt composition The MCFR is fast spectrum chloride rather than a thermal spectrum um fluoride We are focused upon high-U salts with low melting temperatures es Multiple fuel salt synthesis pathways demonstrated.
Along with fundamental fuel research, understanding how each salt composition interacts with available structural materials is key 6 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Thermophysical Property Measurement Melting temperature, heat capacity, viscosity, and density are critical measurements for reactor physics TerraPower has procured and custom-built equipment to measure these properties Heat capacity Also collaborating with ORNL for thermal conductivity TerraPower Radiochemistry Lab Density Viscosity 7 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Materials Selection and Testing for MCFR Multiple rounds of static corrosion completed for up to 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> at 650 and 750°C.
Nearly three dozen natural circulation flow loops operated at up to 750°C and 10,000 hrs Austenitic and FM steels, Inconels, Hastelloys, and refractory alloys Static Corrosion Testing Corrosion Test Results Natural Circulation Loops 8 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Scaling Up for Separate Effects Testing Salt loops are being scaled up for separate effects testing (e.g. thermohydraulics)
Testing being extended to forced flow Incidental materials testing through operational experience Isothermal Coolant Salt Loop (run complete) Polythermal Coolant Salt Loop 9 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Construction of IET Underway Goals:
Provide technical support for design and licensing of MCFR test reactor Increase technical readiness of all MCFR technologies Provide thermal-hydraulic performance data Operate in steady and transient modes 10 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
TWR Materials The TWR reactor leverages DOE program (e.g. FFTF, EBR) design experience
- Extensive R&D and downselection of reactor materials
- Documented operating experience ASME Code Case materials for
- Pressure boundary components
- Supports for pressure boundary components
- Components welded <2t from the pressure boundary
- Attachments that form a structural function (e.g., CSS)
Non-ASME Code materials for
- Internal baffles
- Non-structural or non-welded attachments
- Valve and pump internals
- Reactor core (fuel assemblies and blankets) 11 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Selection of HT9 Pros Extensive operational experience in both the EBR-II and FFTF sodium fast reactors
- FFTF ACO3 assembly reached ~155 dpa over 6 years Large body of irradiation data for FM steels Demonstrated excellent swelling performance No reduction in creep rupture strength from irradiation Simple, predictable failure mechanism (ballooning of cladding tube)
HT9 can fulfill needs for transition from SFR to TWR Cons Higher than desirable DBTT Lower thermal expansion coef.
Not ASME approved (OK for clad/duct)
Difficult material sourcing 12 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
HT9 Development Program Optimize chemistry and processing parameters
- Improve the swelling resistance
- Improve the mechanical properties (e.g. creep, fracture toughness)
- Material must stay within original specification HT9, Neutrons Testing of HT9 to benchmark performance against legacy data FFTF HT9, Neutrons FFTF HT9, Ions FFTF HT9, Ions
- Accelerated (ion) irradiation to high doses for scoping TP HT9 #3, Ions Volumetric Swelling (%)
- Neutron irradiation for licensing case TP HT9 #4, Ions FFTF Steel TP Steel Typical SFR Design Limit
-ferrite TP HT9, Ions TP HT9, Ions 0 200 400 600 Dose (dpa) 13 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
The Supply Chain is Key Must Have:
A manufacturing process capable of consistently achieving the desired microstructure A thorough understanding of how process variability affects performance Domestic and international suppliers with the capability to manufacture bar, tube, wire, and ducts ROD WIRE ROUND BAR BILLET FOR DUCT PLATE CLADDING TUBE Melt (Ingot) Forge (Billet) Hot Roll Cold Work and Shape Manufacturing Welding and Fabrication 14 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Need for Qualification Testing of HT9 The HT9 licensing basis will rely heavily on historical (FFTF) HT9 data; however, additional testing is required for several key reasons:
- To fill in missing gaps of knowledge (e.g. friction, fracture toughness)
- To provide additional confirmatory data points to enable qualification of the existing data for licensing purposes
- To prove to the regulator that the manufacturers can produce HT9 with consistent properties that meet or exceed historical performance
- To take advantage of additional design margin provided by the optimized TerraPower HT9 material Most long-lead testing has started and is producing favorable results 15 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Qualification Summary Materials Data Requirement Basis U.S. NRC 10 CFR Appendix A to Part 50 General Design Criteria for Nuclear Power Plants Data needs based upon the obligation to meet Regulatory Acceptance Criteria (RAC) for a given reactor system TerraPower Principle Design Criteria (PDC)
RACs are flowed down from 10 CFR Appendix A Regulatory Compliance Plans (RCPs)
Two internal documents outline required testing Regulatory Acceptance Criteria (RAC)
Qualification of existing data Four methods are available for qualifying existing data Materials Data Requirements for Licensing the TWR-300 SSC Class 1 safety component material data should use at Material Specific Qualification Plan least two qualification methods RAC Confirmatory Testing SSC Test Plan Datasets Datasets Datasets Datasets Properties Test Specification Data Corroboration Test Procedure Material Specific Qualification Report Peer Review Test Report QA Program Equivalency Acceptance 16 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Material Property Needs for Licensing Properties Unirradiated Irradiated
- Physical Properties (CTE, Poissons, specific heat, etc) -- Complete --------------------- Planned
- Thermal Conductivity ----------------------------------- Complete --------------------- Planned Complete = testing complete, may
- Strength (Yield, Ultimate, Fracture, Elongation) -------- Complete --------------------- In Progress require Dedication
- Fracture Toughness ------------------------------------ Complete --------------------- In Progress
- Thermal Creep ------------------------------------------ In Progress -------------------- In Progress In Progress = equipment fabrication
- Fatigue, Creep Fatigue--------------------------------- Planned ---------------------- Planned and/or testing underway
- Thermal Aging ----------------------------------------- In Progress ------------------- NP Planned = testing scheduled based
- Youngs Modulus --------------------------------------- Planned ----------------------- NP on technical or licensing need. In
- Corrosion ---------------------------------------------- Planned ---------------------- NP some cases, equipment has been
- SCC ----------------------------------------------------- Planned ---------------------- NP designed and/or preliminary testing has been performed.
- Erosion ------------------------------------------------ Planned ---------------------- NP
- Wear Rates --------------------------------------------- In Progress ------------------- NP NP = technical or licensing need not
- Coefficient of Friction ---------------------------------- In Progress ------------------- NP expected
- Irradiation Creep --------------------------------------- N/A --------------------------- In Progress N/A = does not apply (e.g. swelling is
- Stress-Free Swelling ----------------------------------- N/A --------------------------- In Progress only a phenomena in irradiated
- Stress-Enhanced Swelling ----------------------------- N/A --------------------------- In Progress materials)
- FCCI ---------------------------------------------------- N/A --------------------------- In Progress
- ACCI --------------------------------------------------- N/A --------------------------- In Progress 17 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Qualification Data Organization Materials Database SQL based data management system with a Django front-end Stores raw data from samples irradiated in BOR60 Mat_Props V&Ved temperature-dependent material properties Directly compatible with ARMI© (Python)
Alchemy Fuel Model Fuel Properties Materials Handbook Document that mirrors Mat_Props Constitutive Models Testing ARMI Reactor Model Time/dose dependent properties Datasets Literature Review Datasets Constitutive Models Datasets Datasets Mat_Props (Database) Materials Handbook Materials Database Correlations YAML Files 18 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Summary Two Reactor Designs, Similar Challenges No. 1 challenge is meeting timeline for reactor construction Early design decisions require material selection -> need off-the-shelf solutions MCFR Reactor Needs Code or near-Code materials that can withstand chloride salts TerraPower makes and tests salts in-house Multi-scale corrosion testing is being used to screen candidate materials TWR Reactor Needs well-understood materials that can survive long core lifetimes Focus has been on establishing supply chain for HT9 TerraPower is ~4.5 years into a ~10 year qualification plan for its HT9 material Most long lead testing has started and is producing favorable results 19 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Q&A 20 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
Qualification Summary Materials Data Requirements for the TWR Design 77 RACs identified related to materials development for the TWR, for example:
RAC SSC System Materials Data Types Qualification Method Class Number 4.2-1.1 Stress, strain, or 1 31.1 HT9
- Creep (thermal & irradiation) Data corroboration loading limits for all fuel 31.2 316H
- Yield strength (thermal & irradiation)
- Confirmatory testing system components IN718
- Youngs modulus
- Peer review shall be established. [9] Welds
- Poissons ratio
- Fracture toughness (thermal & irradiation)
- Thermal aging
- Thermal expansion coefficient
- Thermal conductivity
- Density
- Irradiation swelling
- Stress enhanced swelling Qualification Plan for HT9 Materials Data for Licensing the TWR-300 Identifies the material properties required (tied to associated RACs)
Recommends test methods and standards Recommends qualification method 21 Copyright© 2019 TerraPower, LLC. All Rights Reserved.
NRC Readiness for Advanced Reactors Licensing - Materials and Component Integrity Research Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop Raj Iyengar Office of Nuclear Regulatory Research U.S. NRC December 9, 2019 Disclaimer The views expressed in this presentation are those of the authors and do not reflect the views of the U.S. Nuclear Regulatory Commission. This material is declared a work of the U.S. Government and is not subject to copyright protection in the United States.
Approved for public release; distribution is unlimited.
2
Advanced Reactor Landscape Micro Liquid Metal Cooled Coole Fast High ig gh-gh h-Temperature Temperature Te p Gas Ga as-as Molten n Salt Reactors Rea Reactors Reactors Cooled oled Reactors Reacto (
(MSR) )
(
(LMFR) ) (
(HTGR) )
Oklo GE-H (VTR) X-energy Kairos Others TerraPower General Atomics ARC Framatome Stationarry Stationary Westinghouse StarCore Sodium-Cooled TRISO Fuel Terrestrial Others Westinghouse TerraPower Columbia Basin Elysium Mobile Hydromine Thorcon Lead-Cooled Muons Flibe Alpha Tech Liquid Salt Fueled 3
Framework for Rightsizing Regulatory Processes NRCs Policy Statement on the Regulation of Advanced Reactors
- Identifies desired characteristics of advanced reactors (enhanced margins of safety; use of simplified, inherent, passive, or other innovative means to achieve safety)
- Address implications for safety and regulatory processes Developer Goals - Meet the Advanced Reactor Policy Statement through innovation Regulatory processes should assure safety and provide predictability, and not be a barrier to innovation 4
Framework for Rightsizing Regulatory Processes Principles of Good Regulation Independence Openness Efficiency Clarity Reliability 5
Framework for Rightsizing Regulatory Processes Modern Risk-informed Regulator Accepting Risk in Decision Making Utilizing Technology Innovating how we work Attracting and Retaining Talent 6
Readiness for Advanced Reactors and Tomorrows Technologies
- Staff knowledge and capacity
- Analytical capability
- Modern licensing approaches for advanced reactor design
- Consensus codes and standards
- Technology inclusive mindset Strategy Document on Non-light Water Reactor Readiness
- International and domestic collaborations 6
7 Thimble Plugging Device Advanced Reactors:
Materials/Component Integrity Goal:
Assess performance needs and issues for materials/component integrity Support development of a regulatory framework Approach:
International Operating Experience Technical issues identification and resolution Flexible approaches to material qualification Coordination with DOE, EPRI, and International Counterparts 8
Advanced Reactors:
Materials/Component Integrity Recent NRC Reports:
- International operational experience with SFRs and HTGRs, focused on materials and component integrity (ADAMS ML18353B121)
- Technical Gap Assessment for Materials and Component Integrity Issues for Molten Salt Reactors (ADAMS ML19077A137)
Ongoing Activities:
Potential endorsement of ASME Section III Div. 5 Graphite: source dependency of properties; molten salt intrusion/infiltration Evaluation of creep and prevention of creep failures of structural alloys High temperature corrosion (HTGR, SFR) and molten salt corrosion of structural alloys (MSR)
Molten salt purity, redox control, and standards for corrosion experiments 9
Graphite Graphite is a material that presents a number of unique design considerations for ANLWRs (Contract support Numark Associates Inc)
Assess graphite design criteria in ASME Section III, Division 5 Code rules Assess graphite properties and degradation including source dependency
- Review experimental data and operational experience relevant to the performance of graphite
- Perform a gap analysis on standards, regulatory guidance, and test procedures for evaluating graphite properties and degradation 10
Creep and Creep-Fatigue Creep-induced cracking and creep-fatigue are potential life-limiting factors for components in high-temperature reactors (contract support: ANL)
Survey to identify gaps in current creep and creep-fatigue design procedures in ASME Code and other codes Develop post-processing tools to aid in executing the ASME Section III, Division 5 Code rules
- Take FEA input and compare with code rules 11 High Temperature Corrosion /
Molten Salt Compatibility Molten Salt Compatibility of Structural Materials and Graphite High Temperature Corrosion/Erosion/Oxidation of Structural Materials Assessment of Applicability of Existing Regulatory Guides and Standards Endorsed by Regulatory Guidance to Liquid-Fueled MSRs Planned Reports:
Recommendations for molten salt corrosion testing/test plan Recommendations for high temperature corrosion/oxidation Summary of technical and regulatory gaps on molten salt chemistry 12
Technical Challenges/Opportunities High Temperature Materials Aging and Degradation Radiation; Corrosion; Mechanical Reactor Surveillance Programs; Accelerated Testing Welding and Joining Additive/Advanced Manufacturing Probabilistic Fracture Mechanics of Graphitic Components Materials Performance in Realistic Environments Lead-time for Materials Characterization & Qualification Development of New Materials - High-Entropy Alloys 13 Forging The Future Connect, Create, Contribute*
- (ACL-OAM 2019 Theme) 14
References
- NRC Vision and Strategy (ML16356A670)
- Implementation Action Plans (IAPs) (ML17165A069 and ML17164A173)
- Regulatory Review Roadmap including prototype guidance (ML17312B567)
- RG 1.232, "Guidance for Developing Principal Design Criteria for Non-Light Water Reactors" (ML17325A611)
- SECY-18-0011, Advanced Reactor Program Status (ML17334B217)
- SECY-19-0009, Advanced Reactor Program Status (ML18346A075)
- SECY-18-0076, Option and Recommendations for Physical Security for Advanced Reactors (ML18052B032)
- SECY-18-0096, Functional Containment Performance (ML18114A546)
- SECY-18-0113, Proposed Rule: Emergency Preparedness for Small Modular Reactors and other New Technologies, (ML18134A086)
- NEI-18-04, Risk-Informed Performance-Based Guidance for Non-Light Water Reactor Licensing Basis Development, (ML18271A172)
- DG 1353, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Approach to Inform the Content of Applications, (ML18264A093) 15
Overview of ONR Activities for Regulating Advanced Modular Reactors A Structural Integrity Perspective Russell Green, Structural Integrity, ONR, UK 9 December 2019 Objectives
- Scope of UK regulatory interest in Gen IV reactors and materials development
- Areas for further consideration
- Potential knowledge gaps for Gen IV reactor designs against ONR expectations 2
Scope of UK Regulatory Interest in Gen IV Designs
- In 2017, the UK Government launched a clean growth strategy setting out the intention to invest in new nuclear technology
- One of the objectives included investment of up to £12 million
(~$15 million) to ensure that the regulators (ONR and Environment Agency) develop the capability and capacity to regulate Advanced Nuclear Technologies (ANTs), including SMRs (light water and non-light water types [AMRs])
- Regarding AMRs the focus is on four of the six GIF technologies:
9 Sodium Fast Reactors (SFR) 9 Lead Fast Reactors (LFR) 9 Molten Salt Reactors (MSR) 9 High Temperature Gas Reactors (HTGR) x Gas Fast Reactors (GFR) x Super Critical Water Reactors (SCWR) 3 Objectives of our work on ANTs Develop ONR capability and capacity Review Engage with ONRs ANT industry guidance and processes ONRs objectives for ANTs Increase Advise BEIS engagement AMR with feasibility and international development regulators programme 4
Regulatory Activities to Build Capability and Capacity
- Knowledge capture and transfer through:
- Attendance at international forums, workshops, conferences and working groups
- Targeted technology-specific training courses
- Review suitability of existing regulatory assessment guidance and arrangements for regulating ANTs In addition, ONR undertook the following:
- Modernisation of UK ONR Generic Design Assessment (GDA) process
- Provided advice to the UK Government Department for Business, Energy and Industrial Strategy (BEIS) in the Advanced Modular Reactor Feasibility and Development Project 5
AMR Feasibility and Development Project
- In 2017, UK government department for Business, Energy and Industrial Strategy (BEIS) launched a £44 million (~$55 million)
Feasibility and Development Project
- With over 20 initial applicants for phase 1, BEIS selected 8 designs do for consideration in phase 1 - 7 fission designs (plus 1 fusion)
- 1 SFRs, 2 LFRs, 3 HTGRs, 1 MSR:
- Advanced Reactor Concepts LLC
- Westinghouse Electric Company UK Limited
- LeadCold
- U-Battery Developments Ltd
- Ultra Safe Nuclear Corporation
- DBD Ltd
- Moltex
- Tokamak Energy Ltd*
6
ONR Activities on Materials Challenges for Advanced Modular Reactors
- Production of safety consideration reports (complete)
- Production of targeted materials knowledge capture reports (in process)
- Identification of structural integrity challenges for the four targeted Gen IV technologies focused around:
- materials development/selection
- understanding of ageing and degradation mechanisms
- development of codes and standards, and
- international approach to regulation
- Ongoing scope of ONR involvement is dependant on UK government strategy for engagement with AMR industry 7
Challenges for Advanced Modular Reactors: Structural Integrity
- Using ONR SAPs and TAGs, identification of structural integrity challenges for ANTs based on information gathered from knowledge capture activities, key findings are presented under three topics:
- Design of SSCs important for safety
- Materials selection and development
- Accessibility for Inspection and Maintenance
- Note that regulatory expectations are highly dependent on reactor design and safety claim placed on components and structures
- Current stage of design maturity means that safety claims may not be fully developed yet for Gen IV structures and component
- Proposed use of advanced manufacturing and inspection techniques
- how to provide confidence for defect prediction and detection?
8
Sodium Fast Reactors
- Determination of any cliff edge effects that may undermine extrapolation of existing OPEX Materials
- Understandingg materials performance p to resist localised thermal fatigue/striping in above core components p - development p of high cycle fatigue modelling/assessment methodologies.
- Understand the level of reliance on chemical control of primary coolant (oxygen, Na impurities) for materials performance
- Application pp of leak-before break methodology for low pressure, ductile components Design
- Importance p of robust material selection,, design and fabrication to avoid/minimise likelihood of IHX/steam generator failures inter-cooling r cooling loop barriers
- Ability to detect and isolate leaks particularly for inter-Inspection
- Capability and reliability of under-sodium viewing technology 9
Lead Fast Reactors
- Performance and application of advanced coating technologies and importance for demonstrating component integrity through life
- Use of coatings to reduce reliance on active primary circuit Materials chemistry control
- Availability of research and test facilities (especially for irradiation)
- Understanding of ageing and degradation mechanisms as a result of aggregated operating conditions (chemical, flow, thermal, mechanical, irradiation)
Inspection
- Challenges faced by inspection of core and primary circuit internal structures and components
& Design 10
High Temperature Gas Reactors
- Understanding g effects of high g temperature, p , chemical environment and radiation on graphite and chemical effects on primary/secondary interfaces (S/GS, ((i)H/Xs) i)H/Xs)
Materials
- Availabilityy and applicability pp p y of relevant codes and standards for graphite design, manufacture and in-service in-service inspection
- Tribological g effects of mechanical interaction of graphite/metallic structures and components in He primary circuits
- Effect of graphite/carbon dust generation Inspection
- Inspection of concealed in-core components such as:
- graphite moderator, reflector, thermal shields and
- metallic core support structure, gas baffles etc)
Design
- Reliance on plant p control and instrumentation to ensure materials operate within the design safe operating envelope (temperatures, high cycle fatigue) 11 Molten Salt Reactors
- Active monitoring and managing primary circuit chemistry to minimise degradation of primary circuit components
- Control of corrosion by avoidance of contaminants - safety classification of chemical/volume control systems?
Materials
- Shielding of structural materials to limit effects of thermal/irradiation degradation mechanisms
- Materials performance in varying/mixed chloride/fluoride salt environments
- Potential for crack initiation and growth requiring validated fatigue assessment procedures Design
- Design of fuel dump tanks to be tolerant of thermal shock 12
Summary of Identified Knowledge Gaps
- Use of advanced manufacturing techniques
- Performance of engineered barriers and design features to control ageing Design and degradation
- High temperature assessment methodologies/codes
- Use of novel heat exchanger/steam generator technology (e.g. printed circuit, microchannel, plate-type)
- Performance of materials in excess of current OPEX - regulatory expectations for extrapolation
- Use of codes and standards to demonstrate appropriateness of material Materials selection
- Development of a standardised approach to materials testing
- Materials selection in areas where OPEX is limited
- Multidiscipline approach to materials selection 13 Summary of Identified Knowledge Gaps
- Accessibility of key structural components and capability to inspect
- Reliance on performance/condition monitoring to supplement limitations of inspectability Inspection
- Application of risk informed inspection methodology
- Expectations and demonstration of claims on in-service inspection capability to demonstrate through life reliability
- Strength of condition monitoring to support claims of through life integrity
- Basis for maintenance strategy and accessibility in design 14
Currently Ongoing
- Review of ONR guidance to evaluate applicability to new Gen IV reactor technologies.
- Review of advanced manufacturing techniques and associated application within the UK nuclear industry.
- Gather OPEX/LFE of non-light water reactor designs.
- Build knowledge of developments in codes and standards for use of high temperature materials.
- Continued engagement with safety regulatory forums e.g. in NEA WGSAR, IAEA and bilateral meetings with other nuclear regulators to develop expectations and share experience/learning.
- Continued attendance at workshops/conferences/seminars to support knowledge transfer and build technical capability/capacity of ANT designs.
15 Next Steps
- Dependent on BEIS decisions for future phases of ANT project
- Possible engagement with Feasibility and Development project participants
- Consider outcome of ONR guidance review and develop a strategy to implement findings.
- Build understanding of how advanced manufacturing is expected to be used within the UK nuclear industry and engage with supply chain to communicate regulatory expectations.
- Continue to support BEIS as necessary.
16
Thank you for Listening Any Questions?
Russell Green Inspector - Structural Integrity Office for Nuclear Regulation russell.green@onr.gov.uk
Reviewing Innovative Reactor Designs
- A CNSC Perspective Stephen Cook Technical Specialist Assessment Integration Division Canadian Nuclear Safety Commission International Workshop on Advanced Reactor Materials and Component Integrity December 9-11, 2019 Rockville, MD nuclearsafety.gc.ca E-doc: 6024881 Canadian Nuclear Safety Commission OUR MANDATE REGULATE IMPLEMENT DISSEMINATE the use of nuclear energy and Canada's international objective scientific, technical and materials to protect health, safety, commitments on the peaceful use regulatory information to the public and security and the environment of nuclear energy OVER 70 YEARS OF REGULATORY EXPERIENCE E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 2
Legal Basis for Risk-Informed Regulation The purpose of this Act is to provide for Section 3(a) the limitation, to a reasonable level and in a manner that is consistent with Canadas international obligations, of the risks to national security, the health and safety of persons and the environment that are associated with the development, production and use of nuclear energy and the production, possession and use of nuclear substances, prescribed equipment and prescribed information; E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 3 Basis for a Licensing Decision Section 24 (4) No licence may be issued, renewed, amended or replaced unless, in the opinion of the Commission, the applicant:
(a) is qualified to carry on the activity that the licence will authorize the licensee to carry on; and (b) will, in carrying on that activity, make adequate provision for the protection of the environment, the health and safety of persons and the maintenance of national security and measures required to implement international obligations to which Canada has agreed E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 4
Safety Case and the Licence Application Safety and Control Areas (SCAs) 3/4 Technical topics the CNSC uses to Management System assess, review, verify and report on Human Performance Management Operating Performance regulatory requirements and Safety Analysis performance across all regulated Physical Design Fitness for Service facilities and activities Radiation Protection Conventional Health and Safety 3/4 Regulatory framework documents Environmental Protection exist for each Safety and Control Emergency Management and Fire Protection Waste Management Area (SCA)
Security Safeguards and Non-Proliferation 3/4 Application comprises the safety Packaging and Transport case, and is part of the licensing Other regulatory areas Public information and disclosure basis for the regulated activity Engagement with indigenous peoples E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 5 Technical Topic - Physical Design REGDOC-2.5.2 Design of Reactor Facilities 3/4 Sets out design requirements and guidance 3/4 Aligned with international practice (IAEA SSR-2/1) 3/4 Outlines safety objectives and concepts to be applied 3/4 Includes requirements for safety management of the design Important sections for new materials and components:
3/4 Proven engineering practices 3/4 Design rules and limits 3/4 Alternative approaches E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 6
Industry Codes and Standards REGDOC-2.5.2, Section 5.4 Proven Engineering Practices (Codes and Standards):
3/4 Design authority shall identify the modern codes and standards that will be used for the plant design 3/4 Codes and standards need to be evaluated for applicability, adequacy, and sufficiency with respect to the design 3/4 Where needed, codes and standards shall be supplemented to ensure that the final quality of the design is commensurate with safety function 3/4 Structures, Systems and Components (SSCs) important to safety shall be of proven design, and shall be designed according to the standards and codes identified for the Nuclear Power Plant (NPP)
E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 7 Novel Designs or Features REGDOC-2.5.2, Section 5.4 Proven Engineering Practices (New design or feature):
3/4 Adequate safety shall be demonstrated by a combination of supporting research and development programs and by examination of relevant experience from similar applications 3/4 An adequate qualification program shall be established to verify that the new design meets all applicable safety requirements 3/4 New designs shall be tested before being brought into service and shall be monitored while in service so as to verify that the expected behaviour is achieved 3/4 Where the design has to accommodate an SSC failure, preference shall be given to equipment that exhibits known and predictable modes of failure, and that facilitates repair or replacement E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 8
Engineering Design Rules and Limits REGDOC-2.5.2, Section 7.5 Design Rules and Limits:
3/4 The design authority shall specify the engineering design rules for all SSCs. These rules shall comply with appropriate accepted engineering practices 3/4 The design shall also identify SSCs to which design limits are applicable 3/4 The engineering design rules for all SSCs should be determined based on their importance to safety (safety classification) and include:
Material specifications Conservative safety margins Reliability and availability Equipment qualification Operational considerations E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 9 Appropriate Margins 3/4 Understanding margins to failure is tantamount to maintaining safety 3/4 Design Margin, analysis margin, safety margin, operating margin, etc.
3/4 Materials properties knowledge is fundamental 3/4 Margins should cover uncertainties, aging and wear, and unknown unknowns, etc.
3/4 Failure limits need to be established for the physical barriers based on experimental data obtained under sufficiently representative conditions set close to the values indicating non-failed state (rather than close to data for failed states) especially where the experimental data are limited, account for measurement uncertainties and data scatter 3/4 Margins need to consider the entire lifecycle of the material or component It is recognized that, at the same time, efficiency of operation requires optimization of safety margins E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 10
Vendor Design Review AN OPTIONAL PRE-LICENSING PROCESS THAT ASSESSES:
- 1. How the vendor is addressing Canadian requirements in their design and safety analysis activities
- 2. Key issues emerging in a design that could impact a licensing process for a potential future project referencing the vendors design This process does not approve a generic design and is not site specific.
The conclusions of any design review do not bind or otherwise influence decisions made by the Commission.
E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 11 CNSC Pre-Licensing Vendor Design Reviews Company Reactor type / output per unit VDR Status Terrestrial Energy Molten salt integral / 200 MWe PHASE 1 compl eted 2018, PHASE 2 in progress PHASE 1 compl eted 2019, Advanced Reactor Concepts Sodium pool fast spectrum /100 MWe PHASE 2 s ervi ce a greement under development Moltex Energy Molten salt fast spectrum / a300 MWe PHASE 1 in progress, pha se 2 pending PHASE1 in progress SMR, LLC. (A Holtec International Company) Pressurized water / 160 MWe PHASE 2 s ervi ce a greement under development Pha s e 1 completed 2018 UltraSafe Nuclear/Global First Power High-temperature gas prismatic block / 5 MWe PHASE 2 project s tart pending mid-2020 NuScale Power Integral Pressurized Water / 50 MWe PHASE 2 project s tart pending U-Battery High temperature gas prismatic block / 4 MWe PHASE 1 project s tart pending GE-Hitachi Boiling Water / 300 MWe PHASE 2 s ervi ce a greement under development X Energy High temperature gas pebble bed / 75 MWe PHASE 2 s ervi ce a greement under development Westinghouse Electric Co. eVinci Micro Reactor / < 25 MWe PHASE 2 s ervi ce agreement under development LeadCold Molten lead pool fast spectrum / 3 - 10 MWe PHASE 1 on hol d a t vendor request StarCore Nuclear High-temperature gas prismatic block / 10 MWe PHASE 1 a nd 2 s ervice agreement on hold E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 12
Determination of Proven 3/4 What level of evidence is necessary for a regulator to make the determination that an aspect of the design is proven enough?
3/4 Are uncertainties understood and adequate safety margins applied?
3/4 Are additional safety and control measures needed?
Prototypical Collect specific scientific/ Low state of proven-ness - risks and experiments engineering information on (proof uncertainties are higher - additional safety of concept) and control measures needed Demonstration Demonstration of integrated Varying amounts of OPEX - proving in reactor / First-of-a components / systems and progress- varying risks and uncertainties to be Kind collection of OPEX to refine addressed - some additional safety and design for nth of a kind control measures needed where uncertainties are high Nth-of-a-Kind Commercial operation - High state of proven-ness - uncertainties information used to improve generally well understood and ongoing R&D operational performance supports management of uncertainties E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 13 Supporting Safety Claims 3/4 Effective Management Systems 3/4 Quality-assured scientific and engineering processes 3/4 Adequate and relevant R&D experimental or field-derived data 3/4 Relevant operating experience 3/4 Application of codes and standards or alternatives 3/4 Verified and validated computer models 3/4 Identification and characterization of uncertainties 3/4 Lifecycle considerations taken into account (margin deterioration)
E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 14
Regulatory Research Program Generates knowledge and information to:
3/4 Support timely, science-based regulatory judgments and decisions 3/4 Allow the regulator to perform a challenge function when discussing safety claims (what questions to ask) 3/4 Aid in the development of safety standards 3/4 Contribute to the independence of the regulator 3/4 May include supplemental training, technical papers and Workshops 3/4 Supported by the Federal Nuclear Science and Technology program E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 15 Conclusions 3/4 CNSC continues to take steps to improve its readiness to regulate SMRs 3/4 A diversity of technologies with innovations are under review 3/4 Supporting research is underway to fill regulatory knowledge gaps 3/4 International collaboration is an important contributor to regulatory efficiency 3/4 SMR design and safety analysis must be supported by R&D activities 3/4 Effective management systems and quality assurance must be used 3/4 The development and application of codes and standards will be crucial in supporting innovation E-Doc: 6024881 Canadian Nuclear Safety Commission - nuclearsafety.gc.ca 16
Questions Thank You!
As Scientists and Engineers, we must hold the balance between irresponsible optimism and crippling caution. Gwen Martin 2011 Connect With Us Join the conversation nuclearsafety.gc.ca
Materials R&D Sue Lesica Office of Nuclear Energy Advanced Materials R&D Program
- The ART Advanced Materials R&D Program supports the Advanced Reactor Pipeline
- Focused is on providing the technical basis needed to support the regulatory requirements for structural materials required for fast reactors, gas-cooled reactors, molten salt reactors and microreactors that could be deployed in the near-to-mid-term.
High Temperature Existing Design Materials Methodologies Support Licensing Innovative New Materials Materials Solutions 2 energy.gov/ne
NEUP, IRP, NSUF, GAIN, FOA, SBIR Programs Contribute to Additional Advanced Materials R&D GAIN, iFOA, SBIR NEUP, Advanced IRP, NE Reactor NSUF Missions Pipeline DOE Labs Advanced Materials 3 energy.gov/ne Gas-cooled Reactor ART Advanced Materials WPs (1/3)
- Support ASME Division 5 Code committees participations (7 Lab staff)
- Complete Alloy 617 Code Case balloting
- Support update of Division 5 code rules for Class A and Class SM components
- Extend qualified lifetime and temperatures of existing Class A materials
- Develop new EPP+SMT creep-fatigue design methods
- Notch and multiaxial stress state effects on creep rupture (Alloy 617)
- Develop and initiate testing on improved weld consumables for Alloy 800H
- AGC and HDG irradiation campaigns
- Graphite material properties
- Baseline properties, irradiation properties, oxidation (acute and chronic), irradiation damage study, mechanisms and model development
- Support ASME Division 5 Code committees participations and ASTM graphite standard (D02.F) activities (5 Lab staff)
- Data Management (ORNL)
- Maintain GEN IV Materials Handbook operations to support GIF VHTR Materials PMB 4 energy.gov/ne
Fast Reactor
- ART Advanced Qualifying Materials New Material WPs 709
- Conduct mechanical testing (tensile, fatigue, creep, creep-fatigue, SMT) to support a new Alloy 709 Code Case
- Fabrication processes for different Alloy 709 product forms
- Develop matching filler metal compositions and welding parameters to optimize Alloy 709 weldment properties
- Qualifying Existing Material to Support Licensing and Long-term Plant Operations - Grade 91(ANL)
- Study carburization and de-carburization behavior of Grade 91 in sodium 5 energy.gov/ne Molten Salt Reactor
- Salt and Materials Interactions (ORNL)
- Evaluate corrosion performance of candidate structural materials by conducting exposures static exposures and flowing tests in natural circulation loops
- Evaluate compatibility of candidate graphite grades with relevant molten salts
- Evaluate effects of environmental variables such as salt chemistry on degradation of structural materials and graphite
- Conceive a roadmap and conduct targeted experiments in support the development of predictive physics-based corrosion modeling and component lifetime prediction
- Materials Surveillance Development (ANL)
- Develop in-situ, passive surveillance specimen designs and fabricate down-selected scoping test articles
- Develop methods to correlate damage to time-dependent indentation test data from surveillance specimens
- ASME Section III Division 5 Class B Materials (ANL)
- Introduce existing high nickel alloys to ASME Section III Division 5 Class B designs
- Develop design methods to introduce the design lifetime concept to Class B rules 6 energy.gov/ne
Microreactor
- ART CoreAdvanced Block CodeMaterials Case for WPs Heat (2/3)
Pipe Reactor - Grade 91 (ANL, ORNL)
- Extend EPP strain limits and creep-fatigue code cases to Grade 91
- EPP primary load code case for Grade 91
- EPP+SMT creep-fatigue design method for Grade 91
- Notch sensitivity correlation for Grade 91, using crystal plasticity finite element method
- Unified viscoplastic material model for Grade 91
- Include provisions in the Code Case to account for the effects of irradiation
- Elevated Temperature Cyclic Properties of Advanced Manufactured Materials (INL)
- Evaluate elevated-temperature fatigue and creep-fatigue properties of 316L stainless steel manufactured using powder metallurgy (PM) hot isostatic pressing (HIP)
- Pursue procurement of 316H and Grade 91 (wrought products already qualified for ASME Section III Division 5) manufactured using PM HIP 7 energy.gov/ne Other Structural Materials Related Efforts
- High Fidelity Ion Beam (LANL)
- High Fidelity Simulation of High Dose Neutron Irradiation led by the University of Michigan
- Objective: Utilizing unique facilities and capabilities, demonstrate the capability to predict the evolution of microstructure and properties of structural and cladding materials in a nuclear reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations
- NEUP-IRP ASME Nuclear Code Case for Compact Heat Exchangers (Led by University of Wisconsin, Madison)
- Develop the required information to certify the use of compact heat exchangers in the ASME Code for nuclear applications
- Develop an understanding of operational issues in operating these heat exchangers in nuclear systems
- NEUP projects
- SBIR projects - Bimetallic structures and SiC/SiC composite fabrication 8 energy.gov/ne
DOE Office of Basic Energy Sciences
- Basic Energy Sciences (BES) supports fundamental research to understand, predict, and ultimately control matter and energy at the electronic, atomic, and molecular levels in order to provide the foundations for new energy technologies and to support DOE missions in energy, environment, and national security.
- The two BES programs that fund the bulk of basic research underpinning nuclear technologies are Heavy Element Chemistry (Philip Wilk) and Mechanical Behavior and Radiation Effects (John Vetrano)
- A major tool for strategic planning at BES is the Basic Research Needs workshop series (https://science.osti.gov/bes/Community-Resources/Reports )
- In August, 2017 BES held a workshop entitled Basic Research Needs for Future Nuclear Energy with input from DOE-NE on the technological needs.
- These workshops, consisting of close to 100 members of the basic and applied research community, develop a series of priority research directions that help to guide the basic science research directions BRN for Future Nuclear Energy - Priority Research Directions Enable design of revolutionary molten salt coolants & liquid fuels How can we characterize and predict the structure, dynamics, and energetics of molten salts including evolving chemical composition across length and time scales?
Master the hierarchy of materials design and synthesis for complex, reactor environments How do we design, synthesize, and process superior materials able to function and perform over decades in the extreme environments of advanced nuclear reactors?
Tailor interfaces to control the impact of nuclear environments How can the multitude of inextricably linked chemical and physical processes that occur at interfaces be controlled?
Reveal multiscale evolution of spatial and temporal processes for coupled extreme environments How can computational and experimental techniques be integrated to bridge spatial, temporal, and energy scales that underpin materials behavior and chemical transformations in coupled extreme environments?
Identify and control unexpected behaviors from rare events and cascading processes How do we identify, anticipate and control rare events that initiate cascading processes and cause aberrant properties and materials responses?
https://science.osti.gov/-/media/bes/pdf/reports/2017/BRN-FNE_rpt.pdf
Workshop hop on n Advan Advanced vannced Non Noon-on-Light n Light Li ht W Water Rea Reactors s-Materials s and Component nt Integrity Atoms for Peace and Development nt 9 - 11 December 2019 U.S. NRC, Rockville, USA IAEA Activities on Fast Reactors Technology Vladimir Kriventsev, Chirayu Batra Fast Reactor Technology Development Team Nuclear Power technology Development Section Division of Nuclear Power email:l: FR@IAEA.ORG Department of Nuclear Energy International Atomic Energy Agency Advanced Non-Light Water Reactors -
https://www.iaea.org/topics/fast-reactors Materials and Component Integrity Workshop 1
Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA IAEA Fast Reactor Technology Development Team Division of Nuclear Power Nuclear Power T Technology ology Development Section n NPTDS Atoms for Peace Fast Reactor Reacttor Technology Tech and Development nt nt Development Team Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 2
Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
Nuclear Power Technology Development NPTDS:
Tasks & Activities es Fast Reactors Technology hnology Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop hop 3
Vladimir Kriventsev, 9 Dec 2019, U.S.NRC,, USA IAEA Technical Working Group on Fast Reactors (TWG G-G-FR)
The Driving Force Members of the IAEA Technical Working g Group on Fast Reactors Full Members Belarus Brazil China Belgium Czech Republic France Germany India Italy Japan Kazakhstan Korea, republic of Netherlands Russian Federation Slovakia Sweden Switzerland Ukraine UK USA
- Provide advice and guidance
- Forum for information exchange and knowledge sharing European Commission OECD/NEA OECD/
- Link between IAEA activities and national communities Observers
- Provide advice in planning and implementing of CRPs Argentina Belgium
- Develop and review selected documents Czech Republic Mexico
- Contribute to status report, technical meetings, topical conferences
- Identify important topics for SAGNE Romania Spain
- Encourage participation of young professionals in IAEA activities Generation-IV International Forum (GIF)
Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 52st TWG-FR Meeting, Pitesti, Romania, 10-14 June 2019 4
Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
IAEA Technical Working Group on SMR
- To advice and support IAEA programmatic planning and implementation in areas related to technology development, design, deployment and economics of SMRs
- 1st meeting in 2018 with 14 Member States
- Now 20 Member States and two International Organizations: European Commission and OECD-NEA as invited observers:
Scientific Secretary:
Mr Frederik Reitsma F.Reitsma@iaea.org
- Three technical subgroups established in 2018 / 2019:
- SG-1: Development of Generic Users Requirements and Criteria (GURC)
- SG-2: Research, Technology Development and Innovation; Codes and Standards
- SG-3: Industrialization, design engineering, testing, manufacturing, supply chain, and construction technology
- 1st TWG Meeting held on 23 - 26 April 2018 in Vienna
- 2nd Meeting : 8 - 11 July 2019 in Vienna TWG-SMR TWG SMR Chair:
- 3rd scheduled for 29 June - 2 July 2020 in Vienna Mr Marco Ricotti Advanced Non-Light Water Reactors - President of CIRTEN Materials and Component Integrity Workshop 5
Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA Fast Reactors:: Coordinated Research Projects CRPs on Fast Reactors Technology On-Onn-going CRPs New Proposals PSFR R Source Term - Total Instantaneous Blockage Radioactive Release Under of SFR Fuel Assembly Severe Accident Conditions Simulation of CLEAR R-R-S Neutro Neutronics utro tro onnics Benchmark Benc of CEFR Loss-of-Flow Experiment Startr -Up Tests (299 participants) rt-rt Benchmar Benchmark marrkk A Analysis of Benchmark Analysis o off FF FFTF FTF F Loss STELLA A- LOHS/LOF Tests A-2 off Flow Flow Without Witho Scram Test (255 participants)
NAPRO - Na Properties and Safe Operations of Exp. Facilities Ended in Sept 2018 2 TECDOCs in Publishing Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 6
Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
New IAEA CRPs on Fast Reactors (Started in 2018)
Neutronics s Be Benchmark ench hmark o of Benchm Benchmark mark Analysis Analysi of CEFR Start rt-rtt-Up Tests FFTF ULOF Test Country Organization(s)
Belgium SCK*CEN SCK C China CIAE, INEST, SNERDI, XJTU France CEA Germany KIT, HZDR, GRS Hungary BME, EK Country Organization(s)
India IGCAR China CIAE, NCEPU, INEST, XJTU Italy NINE/UNIPI France CEA Japan JAEA Germany KIT, HZDR Rep. of Korea KAERI, UNIST India IGCAR, ISSSA Mexico ININ Italy NINE, Sapienza Romania RATEN Japan JAEA Russia IPPE, IBRAE, SSL, Kurchatov Inst. Rep. of Korea KAERI Slovakia VUJE Netherlands NRG Switzerland PSI Russia IPPE, IBRAE Ukraine KIPT Spain CIEMAT UK Cambridge Sweden KTH USA ANL, NRC, INL Switzerland PSI 17 Countries 29 Organizations USA ANL, PNNL, TerraPower, NRC, TAMU Advanced Non-Light Water Reactors - 13 Countries 25 Organizations Materials and Component Integrity Workshop 7
Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA New CRP: CR Neutronics Benchmark of CEFR Start-Up Tests
- China Experimental Fast Reactor
- Sodium-cooled fast reactor with nominal power of 65MW(th), 20MW(e)
- Reached the first criticality in 2010
- Generated electricity at 40% full power and was connected firstly to the grid in July 2011
- Generated electricity at 100% power in December 2015 and operated for more than 40 effective full power days Advanced Non-Light Water Reactors -
1st, Kick-Off RCM: June 2018, Vienna Materials and Component Integrity Workshop 8
Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
CEFR Start-Up: Tests and Simulations
- 2. Measurement 3. Reactivity
- 1. Fuel loading 4. Irradiation test of control rod effects and criticality of foils worth measurement Experiments Temperature Reaction rate Net criticality Net criticality core reactivity coefficient distribution Pressure reactivity effect 250(cold state) Cross-section ratio Operation Flowrate reactivity loading core effect 360 (hot standby)
(250) Neutron spectrum Sodium void reactivity effect Operation loading Nuclear heating start core (300) Core SA exchange Absolute nuclear point reactivity effect power Planned duration: 2018 - 2022 Set up Perform blind Comparison with Refined simplified neutronics experimental Simulations; model m o of CEFR calculations data, models Publications core independently update of the results Kick-off RCM: 11-14 June 2018 (27 Participants from 17 MSs)
Advanced Non-Light on-Light Water Rea Reactors ctors t - 2nd RCM: 28 October - 1 November, Beijing Materials and d Component Integrity Workshop 9
Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA New CRP:
CR Benchmark Analysis of FFTF Loss of Flow Without Scram Test
- FFTF Reactor:
- 400 MWth sodium cooled fast test reactor
- Mixed UO2-PuO2 (MOX) fuel
- Loop type plant, axial and radial reflectors
- Prototypic size
- ~1m3 core volume
- ~91 cm high, ~120 cm diameter
- Series of Passive Safety Tests
- Demonstrated passive safety of SFRs
- Demonstrated efficacy of negative reactivity insertion safety devises (GEMs)
PNNL/ANL at Consultants Meeting November 2017, Vienna Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 10 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
CRP Benchmark: FFTF ULOF Test Coupled Neutronic, Thermalhydraulic and System Codes Plant Data Modelling Simulations Comparison ULOF Transient from 50% Power; Reactor Survives (thanks to GEM)
Blind Sharing of results Uncertainties Publication of Benchmark Simulations and comparison Qualification IAEA Specifications performed with Experimental and Results independently Data Validation document Planned duration: 2018 - 2022 Knowledge Storing Preservation CRP Advanced Non-Light Water Reactors -
Kick-off RCM: 22-25 October 2018 data Materials and Component Integrity Workshop 11 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA TM on Benefits and Challenges of Fast SMRs Country Participants 24-27 September 2019, Milan, Hosted byy CIRTEN:
/Papers Consortium of Italian Nuclear Universities Six Technical Sessions:
Belgium 4/1
- Heavy Liquid Metal Cooled Fast SMRs France 1/1 Germany 2/0
- Safety Investigations India 1/1
- Technology and Research in Support of Fast SMRs Italy 13/5 Three Group Discussions:
Japan 3/2
- In-factory construction Thanks to advanced coolants, Fast Korea, Rep. of 2/3
- Technological Challenges design Netherlands 1/1 TECDOC Proceedings to be published in 2020 BUT:
Russia 3/2
- Fast construction (in- factory) is Slovakia 1/0 required to win economic competition; Switzerland 1/1 Sweden 1/1
- Extended R&D are needed to fit technological gaps USA 1/1
- LFRs require more R&D to prove p
EC/JRC 3/1 material compatibility and an Total: 16 40/23 develop new materials Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop
- Licensing challenges Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA 12 12
Liquid Metal cooled Fast SMRs Latest reactor designs presented at TM on Fast SMRs in Milan ALFRED SFRs 150 MW(e) PGSFR CLFR-300 14 MW(e) CLEAR- Rep. of Korea 125-250 MW(e) -100 (LBE)
China M10d EU Russia China LFRs 3-10 MW(e) SEALER LFR-AS-200 MW(e)
Sweden 55 MW(e) SEALER-UK Luxembourg Transportable LFR-TL-5 MW(e)
Sweden 50 MW(e) SMFR 300 MW(e) SFR Luxembourg Advanced Non-Light Water Reactors - Japan Japan Materials and Component Integrity Workshop Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA 13 13 TM on Structural Materials for Country Participants
/Papers Heavy Liquid Metal Cooled Fast Reactors Belgium 2/1 Three TechTechnical China 7/4 15-17 October 2019, Vienna Czech Rep. 3/1 Sessions:
Germany 1/1
- HLM Compatibility with Italy 5/3 Structural Materials Korea, Rep. of 4/1
- Corrosion Mitigation Luxembourg 1/0 Measures Netherlands 1/0
- Qualification Programmes Romania 2/2 of Structural Materials Russia 3/3 Three Group Slovakia 1/1 Discussions:
Sweden 2/1
- Outstanding Research UK 1/0 Challenges Ukraine 1/1
- New Materials and EC/JRC 1/1 Coating Techniques TECDOC Proceedings to be published in 2020 Total: 14 34/20
- Technology Readiness Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 14 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
TM on LFR Materials: Sessions and Topical Discussions
- General Chair: Daniela Gugiu g
- IAEA Secretariat: Vladimir Kriventsev, Chirayu Batra Papers/
Session Title Chair(s) Presentations HLM Compatibility with Structural Materials: Phenomena, I Kamil Tucek (EC/JRC) 6/6 Modelling and Operational Experience Peter Szakalos (KTH),
Corrosion Mitigation Measures: Coating, New Structural II Alfons Weisenburger 6/8 Materials, Environmental Conditioning (KIT)
Qualification Programmes of Structural Materials for HLM Fast III Bin Long (CIAE) 3/5 Reactors Group Discussion Title Moderator I Outstanding Research Challenges Kamil Tucek II New Materials and Coating Techniques Alfons Weisenburger III Industrialization Erich Stergar Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 15 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA Fast Reactors Safety: y: Joint GIFIF-F-IAEA Workshops on Safety of LMFRs 1st : June 2010 2nd : Dec 2011 3rd : Feb. 2013 4th : June 2014 5th : June 2015 6th GIF-IAEA Workshop on Safety of SFR November 2016 7th Joint GIF-IAEA Workshop on LMFR Safety March 2018
- Final Review of GIF Report on Safety Design Guidelines on Safety Approach & Design Conditions for GEN-IV SFRs 8th GIF-IAEA Workshop on LMFR Safety 20-22 March 2019
- Discussion of GIF Report on Safety Design Guidelines on Structures, Systems and Components for Gen-IV SFRs 9th GIF-IAEA Workshop on LMFR Safety Advanced Non-Light Water Reactors - 18-20 March 2020 Materials and Component Integrity Workshop 16 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
Online Catalogue on LMFNS Experimental Facilities Experimental Facilities in support of Development and Deployment of Liquid Metal cooled Fast Neutron Systems Includes an overview as well as detail detailed information on 150 experimental facilities under design, construction or operation 19 institutions from 14 IAEA Member States contributed Freely Available at iaea.org:
Search for IAEA LMFNS Updated August 2019!
Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 17 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA Joint ICT ICTP TP-T P-IA IAEA AEA Workshops on A
Innovative Nuclear Energy Systems
- In 2016 and in August 2018 Trieste, Italy
- Contributed by NPTDS, INPRO, GIF, and other external experts
- Next Workshop: 13-17 July 2020 Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 18 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
FR09 >> FR13 >> FR17 >> FR21 Conferences IAEA International Conferences on Fast Reactors and Related Fuel Cycles FR21 June 2021 Vienna?
Yekaterinburg g 2017
~600 Participants from 27 IAEA Member States 6 International Organizations 460 Technical Papers 10 Invited Plenary Speeches Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 19 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA M
Main IAEA IAAE EA Activities Activ on Fast Reactor Technology in n 2018 8 - 2019 CRPs/Benchmarks/Studies
- Technical Working Group on Fast Reactors
- 51st TWG-FR Meeting in Hefei, China,
- NAPRO CRP (2013 - 2018) 21-25 May 2018
- 3 Ongoing g g CRPs: - 52nd TWG-FR Meeting in Romania,
- PSFR Source Term ((2016 - 2020) 10-14 June 2019
- Neww CEFR Start rt-rtt-Upp Tests (2018 - 2022)
- Joint IAEA-GIF Workshops on LMFR Safety
- Neww FFTF ULOF Test (2018 - 2022) - 7th GIF-IAEA Workshop on LMFR Safety:
- 2 New CRPs proposed (to start in 2021): 27-29 March 2018
- Modelling of Total Instantaneous Blockage of SFR F/A - 8th 8th h GIF GF GI F-IAEA
-IA AEA A AW Workshop Works on LMFR Safety:
- Benchmarking LOF transient test in CLEAR-S HML Pool 20-20 0-22 2 March h 2019 Facilityy
- LMFNS Experimental Facilities Database
- Study on n Passive Shutdown Systems for Fast Reactors
- Training Courses and Workshops (completed p in 2017,, NES to be p published in 2019)
- Joint ICTP-IAEA Workshops on the Physics and
- TM on n Benefits and Challenges g of Fast SMRs (2019)
( Technology of Innovative Nuclear Energy Systems
- TM on n Structural Materials for HLM Reactors (2019) ( ) (2016, 2018, 2020 in Trieste, Italy)
- TM on n Economic (or Industrial) Optimization of Liquid Metal
- 3rd Workshop: 13 -17 July 2020 cooled Fast Reactor Designs g (2020)
( ) - Regional Workshop on Advances in Modelling &
Simulation of Thermal Hydraulics in LMFRs
- TM on n Proliferation Resistant Re Features of Fast Reactors and
- 6-10 April 2020, India Related Fuel Cycles (2020)
Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 20 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
FRs: Planned Events and Activities in 2020-2021 Year Title Location 2020-2021 53rd and 54th Meetings of the Technical Working Group on Fast Reactors (TWG-FR) 4th RCM of CRP on Radioactive Release from PSFR under Severe Accident 2020 Vienna Conditions 2020-2021 3rd and 4th RCMs of CRP on Neutronics Benchmark of CEFR Start-Up Tests 2020-2021 2nd and 3rd RCMs of CRP on Benchmark Analysis of FFTF ULOF Test 2021 New CRP on Benchmark Analysis of Loss-Of-Flow Test at CLEAR-S Pool Facility China 2021 New CRP on Benchmark Analysis of STELLA-2 LOHS (or LOF) Test Rep. of Korea 2021 New CRP on Computational Modelling of Total Instantaneous Blockage of SFR S/A India 2020 Joint ICTP-IAEA Workshops on Physics and Technology of Innovative NESs Trieste, Italy 2020 Regional WS on Advances in Modelling & Simulation of Thermal-Hydraulics in LMFRs GNCEP, India 2020-2021 Training Course with PC-based SFR Simulators for Educational Purposes 2020-2021 TM on Economic and Industrial Optimization of LMFR Designs 2020-2021 TM on Proliferation Resistant Features of Fast Reactors and Related Fuel Cycles 2020-2021 TM on Status of the IAEA Fast Reactor Knowledge Preservation Initiative 2020-2021 Joint IAEA-GIF Workshops on LMFR Safety Vienna 2021 IAEA International Conference on Fast Reactors and Related Fuel Cycles (FR21)
Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop 21 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA Thank you!
Advanced Non-Light Water Reactors - email:l: FR@IAEA.ORG Materials and Component Integrity Workshop 22 Vladimir Kriventsev, 9 Dec 2019, U.S.NRC, USA
Overview of reactor materials investigation with modelling and experimental studies at CNL Mike Welland, PhD, PEng Research Scientist, Computational Techniques Head of mesoscale and transport methods 2019 International Workshop on Advanced Reactor Materials &
Component Integrity Dec 9-11, 2019 UNRESTRICTED / ILLIMITÉ Canadian Nuclear
- 9,100 acres with 200 acres of lab complex Laboratories (CNL) is among
- 17 nuclear facilities, 70 major buildings the largest hot cell and
- 3,000 employees nuclear R&D facilities worldwide.
- 1,600 engineering, scientific & technical staff UNRESTRICTED / ILLIMITÉ
Overview of research areas
- New investments and projects
- NRU component harvest
- Inventory of material from well characterized irradiation conditions
- Tritium handling laboratory
- SMR siting
- Molten salt behaviour
- Material properties, FG retention, corrosion
- Degradation of reactor components during irradiation
- He bubbles in Ni, H in Zr UNRESTRICTED / ILLIMITÉ NRU Artefacts for Research & Investigation National Research Universal reactor operated from 1957-2018 Wide range of temperature (35 oC to 310 oC), flux and neutron spectra.
The NRU inventory includes:
- Structural materials steels, Inconels, zirconium, aluminum, concrete;
- a thermal graphite column;
- flux detectors;
- equipment (including pumps);
- elastomers/seals;
- electrical cables.
UNRESTRICTED / ILLIMITÉ NRU Artefacts for Research & Investigation Stainless steel header cups Study weldability of 304 stainless with high helium content Graphite sections Microstructural degradation of graphite irradiated at low temperatures 44 year operation, <230 °C 7.1x1022 n/cm2 (E<0.625 eV)
UNRESTRICTED / ILLIMITÉ Tritium Processes and R&D
- Tritium Processing Expertise:
- Detritiation of heavy water for CANDU
- Detritiation of light water for emission control and environment
- Air clean-up (recombination and scrubbing)
- Process modeling
- Tritium R&D:
- Thermal cycling adsorption isotope separation absorbents
- Immobilization of tritiated water for waste disposal
- Permeation of tritium through materials at high temperatures
- Tritium tolerant electrolytic cells
- Beta-voltaic power sources
- Organically-bound tritium in the environment UNRESTRICTED / ILLIMITÉ SMR siting progress STAGE 1: STAGE 2: STAGE 3: STAGE 4:
Prequalification Due Diligence Negotiation Project Execution VENDOR A VENDOR B
- Have not publically announced their participation.
UNRESTRICTED / ILLIMITÉ Molten salt properties Thermal conductivity, Corrosion, FP release UNRESTRICTED / ILLIMITÉ Thermal conductivity of Molten Salt
- Determining thermal diffusivity of molten salts
- Co-development of model led to improved crucible design
- Concurrent prediction of thermal conductivity from MD Crucible Lid Molten salt Crucible Crucible Salt Crucible UNRESTRICTED / ILLIMITÉ Thermal conductivity of Molten Salt
- Determining thermal diffusivity of molten salts
- Co-development of model led to improved crucible design
- Concurrent prediction of thermal conductivity from MD Crucible C
Crrucib ucib ucible le Lid Lid d
Molten Molt lten s lt salt alt lt Crucible Crucible Cr C ucible Salt UNRESTRICTED / ILLIMITÉ First principle calculations of behaviour Molten salt: Ni-alloy + Th-based salt.
Salt properties Interface properties Alloy properties MD model DFT model Chemical reactions UNRESTRICTED / ILLIMITÉ Scoping tests of FP release
- Six tests at 1000 °C, exposed to argon, air and steam
- Release of Cs, I, Ru measured in oxidizing conditions
- Uncertain results from argon tests due to chemistry control 100 1400 Argon Air Argon 90 1300 80 1200 Ru Release Percentage 70 1100 Temperature (K) 60 1000 50 900 40 800 30 700 20 600 10 500 0 400
-10 300 0 2000 4000 6000 8000 10000 Time (s)
MS5 F-Na-K-U Ru-103 497 keV T (K)
UNRESTRICTED / ILLIMITÉ Material degradation during irradiation Embrittlement, grain boundary perforation, Hydride formation UNRESTRICTED / ILLIMITÉ Embrittlement of Ni by He Single crystall Grain boundaryy Strain He effects (a)
Grain boundary (b)
He embrittlement ttleme of nickel Impurities (c)
Tensile deformation Induced dislocations UNRESTRICTED / ILLIMITÉ Perforation of Ni grain boundary by He bubbles Included phase model for bubble behaviour Intergranular bubble PDF differs from LSW substantially.
UNRESTRICTED / ILLIMITÉ Perforation of grain boundary by He bubbles Included phase model for bubble behaviour Simulation Results UNRESTRICTED / ILLIMITÉ Zirconium hydride TSSP simulation Predicting the TSSP with a modern computational approach
- Elucidate the source of -ZrHx precipitation / dissolution hysteresis from -Zr
- Revisit elastic penalty thesis with a modern, sophisticated phase-field approach
- Multiphysics phase-field model incorporating bulk thermodynamics (CALPHAD)
- 3D anisotropic elasticity, concentration-dependant properties
- Anisotropic interfacial energy (coherent lattices)
- Stress driven diffusion of H in both phases Predicted phase diagram Underlying thermodynamics UNRESTRICTED / ILLIMITÉ Zirconium hydride TSSP simulation Energy barrier to precipitation in a pristine material (defect free)
Seans plots UNRESTRICTED / ILLIMITÉ Thank You!
Michael.Welland@cnl.ca Mike@mikewelland.com WWW.CNL.CA UNRESTRICTED / ILLIMITÉ
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Material and Properties Development Needs for Design and Development of Commercially Viable GEN IV Reactors Mike Burke David Gandy Technical Executives EPRI Nuclear Sector International Workshop on Advanced Reactor Materials and Component Integrity Rockville Md.
Tuesday December 10th 2019 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Development and Validation for LWR Applications In-reactor, In-environment Mechanical Response, Long Time Fleet Performance IASCC, Rad. Fatigue Life Prediction from Assessment of Aged Materials via Stress Corrosion Cracking representative lab testing of Pre-irradiated Material IASCC Moderate-time Multi-unit Environmentally Assisted Fatigue (EAF) Stress Irradiation Effects on Creep -
Performance data Amplitude vs Life, Crack Growth Rates Creep Rates, Creep Life Laboratory test data for Mechanical Properties, Stress Corrosion Quantitative Irradiation interacting variables generating Toughness in Cracking Stress vs Life, Effects on Hardness, durability data for lifing Environment Crack Growth Rates Toughness etc.
Prototype Performance Data Time Dependent Creep Cyclic Properties Fatigue EAC Effects ?
Properties Life Cyclic / Curves Intermediate-time durability test data for simple effects Tensile Properties, Hardness, Qualitative Irradiation Effects Microstructure Corrosion Resistance Construction Code Data
- Hardening Embrittlement with margin of safety on Mechanical Properties Environmental Effects Irradiation Effects short time durability data 2 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Development and Validation for LWR Applications In-reactor, In-environment Mechanical Response, Long Time Fleet Performance IASCC, Rad. Fatigue Life Prediction from Assessment of Aged Materials via Stress Corrosion Cracking representative lab testing of Pre-irradiated Material IASCC Moderate-time Multi-unit Environmentally Assisted Fatigue (EAF) Stress Amplitude vs Life, Irradiation Effects on Performance data Crack Growth Rates Creep - Creep Rates, Creep Life Laboratory test data for Mechanical Stress Corrosion Quantitative interacting variables generating Properties, Toughness in Cracking Stress vs Irradiation Effects durability data for lifing Life, Crack Growth on Hardness, Environment Rates Toughness etc. Prototype Performance Data Time Dependent Creep Cyclic Properties Fatigue EAC Effects ?
Properties Life Cyclic / Curves Intermediate-time durability Tensile Properties, Hardness, Corrosion Resistance Qualitative Irradiation Effects test data for simple effects
- Hardening Embrittlement Microstructure Construction Code Data with margin of safety on short time durability data Mechanical Properties Environmental Effects Irradiation Effects 3 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
What is the Data-driven Paradigm that Support Design of New LWRs and Life Extension of Operating Plants?
Extensive data that describes the behavior of established LWR materials in water + irradiation environments but
- LWRs are still designed on simple properties
- Service Aging Effects Identified 3/4 Measurements of the effects of irradiation - hardening and embrittlement 3/4 Measurements of the effects of environment - SCC rather than simple corrosion
- Discovery of key life limiting factors during service 3/4 Subsequent simulation testing of materials in laboratories and hot cells
- Post mortem investigations on reactor surveillance capsules and materials Harvested from operating reactors This extensive database has been built up over 60 years of plant operations and increasingly sophisticated materials testing.
4 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Development and Validation for Advanced Reactor Applications In-reactor, In-environment Reactor Structures Mechanical Response, Life Prediction and IASCC, Rad. Fatigue Safety Validation Data Stress-Corrosion Cracking Behavior of Previously Irradiation Effects on Lifing Irradiated Material Parameters Environmentally Affected Cyclic Deformation Response
- Stress Amplitude vs Life, Crack Growth Rates etc. Structure/Component Irradiation Effects on Creep :
Creep Rates, Creep Life Data for Realistic Lifing Mechanical Properties, Stress Corrosion in Environment Cracking Stress vs Life, Crack Growth Rates Irradiation Effects on Material Viability/
Mechanical Properties Cyclic Properties Fatigue Selection for Life Cyclic / Curves Components Irradiation Effects on Microstructure Time Dependent Corrosion Behavior Swelling, He Production, Matrix Mechanical Properties Weight Loss, IGA Attack Hardening, Phase Changes Licensing and Environmental Effects Irradiation Effects Qualification Mechanical Properties Construction Code Data New Materials Development: Composition Control, Processing, Microstructural Optimization Scale Up, Standardization and Commercial Availability 5 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Data Generation and Material Testing for Structures in Advanced Reactors In-reactor, In-environment Long Time Fleet Performance Mechanical Response, Life Prediction from IASCC, Rad. Fatigue Assessment of Aged Materials via representative lab Stress-Corrosion Cracking testing Irradiation Effects on Lifing Behavior of Previously Parameters Irradiated Material Moderate-time Multi-unit Environmentally Affected Cyclic Performance data
? Deformation Response : Stress Amplitude vs Life, Crack Growth Rates etc. ? Irradiation Effects on Creep :
Creep Rates, Creep Life Laboratory test data for interacting variables Time Dependent Cyclic Stress Corrosion Cracking Stress Irradiation Effects on generating durability data Properties Mechanical Properties for lifing vs Life, Crack Growth Rates
? Mechanical Properties,
?
Prototype Performance Data
? Time Dependent Mechanical Properties in Environment Corrosion Behavior Irradiation Effects on Intermediate-time durability test data for Weight Loss, IGA Attack Microstructure Phase Changes Construction simple effectsCode Data Mechanical Properties Environmental Effects Irradiation Effects with margin of safety on New Materials Development: Composition Control, Processing, short time durability Microstructural Optimization Properties data 6 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
How Well Can We Follow The LWR Paradigm To Support The Development Of Advanced Reactors? (2)
Lots of white space compared to the data that supports LWR reactors
- Immediate need for new materials/properties to be developed, validated and standardized High temperatures, irradiation and corrosion data
- Some simple properties are available to support concepts - but there are many gaps in supporting data - long time data, time dependent properties, severe environment service
- Progression from properties under combined conditions- Stress +
Environment, Cyclic at Temperature, Strain + Irradiation Design and build of prototype systems can validate materials selections and identify key data needed to support longer term lifing predictions 7 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Gaps identified for Advanced Reactors (MSR, LFR, VHTR/GFR, SFR) in EPRI Material Gap Studied - 1 of 2 GEN IV REACTOR COMPONENT MATERIAL NEEDED R&D Molten Salt Reactors Core Support/ 316 Proof of resistance to long time corrosion in properly controlled salt environment. Time dependent Structural Materials properties for ASME code Sec III Div 5 certification. Demonstration of performance (resistance to EAC) in salt under loading.
Stainless Steels Ferritic-Martensitic & LAS Development and demonstration of cladding (Mo rich) for protection Nickel-based Alloys Hastelloy N and Demonstration of radiation tolerance of Hast N variants (Proper understanding of chemistry Other (Graphite, Ceramics) variants microstructure properties Development of properties for ASME code Sec III Div 5 certification Corrosion Cladding Coolant Salt Development of salt chemistry (and impurity) control. Demonstration of Te control Moderator Graphite Development of long time properties in salt etc. for the specific type of graphite to be employed High temperature Core Support/ 316 and Code approval of time dependent properties - creep, creep-fatigue Gas Reactor Structural Austenitic Alloys Materials 316FR Code qualification properties for ASME code Sec III Div 5 for 316FR including time dependent properties Vessel LAS Time dependent and fatigue properties for ASME code Sec III Div 5 Moderator Graphite Development of long time properties etc. for the specific type of graphite to be employed Na SFR Vessel and Core 316 Stainless Extend code properties to include time dependent behavior (Creep. Creep fatigue)
Support Structure D9 Development of for ASME code Sec III Div 5 properties (including time dependent properties) for D9 Development of swelling behavior at long times under realistic conditions - demonstrate adequacy Core Support Ferritic Prove adequacy of swelling resistance at high fluence Structure and Martensitic Development of fabrication technology and proof pf performance of welds Cladding 8 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Gaps identified for Advanced Reactors (MSR, LFR, VHTR/GFR, SFR) in EPRI Material Gap Studied - 2 of 2 GEN IV COMPONENT MATERIAL NEEDED R&D REACTOR GFR Core Ferritic Martensitics Demonstration of adequate resistance to swelling at high dpa. Time dependent properties for support ASME code Sec III Div 5. (include development of fabrication technologies - and demonstrate properties of joints)
Cladding Ceramics For advanced GFR - SiC-SiC, Zr3Si need materials endurance data for these materials and reflector Lead Fast Structural 316 (code qualified already) but need creep and creep fatigue data to be added into code.
Reactor Materials/ Need corrosion data/demonstration of resistance to lead corrosion Stainless Steels Vessel Type 15-15Ti stainless Verification of swelling resistance Ferritic-Martensitic &
Development of code properties for 15-15Ti material design LAS Nickel-based Alloys Near core Ferritic Martensitics Demonstration of adequate resistance to swelling at high dpa.
Other (Graphite, structures Time dependent properties for ASME code Sec III Div 5. (include development of fabrication Ceramics)
Corrosion and technologies - and demonstrate properties of joints)
Cladding cladding Demonstration of resistance to lead corrosion/development of corrosion data Development of fabrication and effective joining methods High Structural Alumina Forming Austenitic Demonstration of resistance to lead corrosion Temperature Materials/ Stainless Steels Demonstration of adequate resistance to irradiation/swelling at expected high dpa Lead Reactors Vessel Development of processing and joining of alumina forming austenitic stainless steels Cladding SiC-SiC Development of SiC-SiC structures Demonstration of resistance to lead corrosion Development of properties and support to code qualification 9 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Properties Needed for Design and Development of Advanced Reactors Note: Some mechanical 316H SS Extend BPV-III Div 5. Code properties to include time Mechanical Properties dependent behavior (Creep. Creep fatigue) at increased Demonstration of adequate resistance to swelling at high properties serve more than temperatures (600C and greater)
Development and demonstration of cladding (Mo rich) for Ferritic- fluence range. one reactor development protection Martensitic Time dependent properties for ASME Code Sec III Div 5.
316FR SS Code qualification properties for ASME code Sec III Div 5 9 Cr Development of fabrication and effective joining for 316FR including time dependent properties methods Hast N & Demonstration of radiation tolerance of Hast N Type 15- Verification of swelling resistance Demonstration of adequate resistance to swelling at high Variants variants (Proper understanding of chemistry Ferritic- fluence range. microstructure properties 15Ti SS Development of code properties for 15-15Ti material Martensitic Time dependent properties for ASME Code Sec III Div 5. Development of properties for ASME Code Sec III design Demonstration of adequate resistance to --12Cr Development of fabrication and effective joining Div 5 certification Alumina methods with mechanical integrity 800H Summary Document of Properties (High Forming irradiation/swelling at expected high dpa Ferritic Validation of commercial reliability - Properties and 617 temperature, time dependent)
SS Development of processing and joining of alumina forming austenitic stainless steels Martensitic sensitivity to heat treatment/local microstructures D9 Development of for ASME Code Sec III Div 5 properties Generic Response to fabrication processes - welding practices Stainless (including time dependent properties) for D9 LAS Time dependent and fatigue properties for ASME code Development of swelling behavior at long times under Sec III Div 5 to higher temperatures Steel realistic conditions - demonstrate adequacy Demonstration of long time resistance to molten salt Corrosion Properties Demonstration of sustained mechanical behavior in molten salt (Salt equivalent of SCC) Extended corrosion data/demonstration of resistance 316 Demonstration of sustained mechanical behavior of Ferritic- Demonstration of resistance to lead Hast N & to molten salt Demonstration of sustained mechanical behavior in CW/Irrad. material mechanical behavior in molten salt Martensitic corrosion/development of corrosion data Variants molten salt (Salt equivalent of SCC)
Data/demonstration of resistance to lead corrosion 9Cr Demonstration of sustained mechanical properties in Demonstration of sustained mechanical behavior of Sustained mechanical behavior in molten lead lead CW/Irradiated material mechanical behavior in molten Proof of resistance to long time corrosion in properly Ferritic- Demonstration of resistance to lead salt 316H controlled salt environment. Martensitic corrosion/development of corrosion data Proof of resistance to long time corrosion in properly Demonstration of performance (resistance to EAC) in salt 12Cr Demonstration of sustained mechanical properties in controlled salt environment.
Alloys under loading. lead Demonstration of performance (resistance to EAC) in 800H salt under loading.
Alumina Demonstration of resistance to Lead corrosion Demonstration of resistance to molten salt corrosion and 617 Comparison of behavior vs 316 and Hastelloy N Forming Determination of the effects of potential EAC SS 10 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Development of Materials Technology for Advanced Reactors Design and Development Three categories of development needed:
- Materials properties to support initial design and to attain ASME code acceptance for constructions - including high temperature and time dependent properties
- Response of candidate materials to neutron irradiation - effects of realistic levels of irradiation on microstructural and property stability
- Materials response in environment - effects measured in realistic environments: stand alone & effects on mechanical behavior (equivalent to IASCC)
Initial Focus on development of code required material properties
- Support for design and build of prototype(s) Å Simple properties
- Initial design for short life (predict prototype performance) Å Analysis incorporates some time dependent property data
- Development of extended time dependent properties & properties of appropriately aged and exposed materials (T,t + n,t) 11 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Caveat :Materials Robustness for Advanced Reactors Development There is ALWAYS a need to understand the potential for variability in materials properties:
development materials sampling vs production scale up Which of these points would be obtained in a survey program to support or deny the use of a material for an advanced reactor application?
New Materials Development: Composition Control, Processing, Microstructural Optimization Scale Up, Standardization and Commercial Availability 12 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
TogetherShaping the Future of Electricity 13 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
ADVANCED REACTOR TECHNOLOGY IRRADIATION PROGRAMS AT NRG- PETTEN REACTOR NRC Advanced Reactor Materials Workshop Uazir Bezerra de Oliveira 9 December 2019 2
OUTLINE
- Introduction to NRG
- Infrastructure
- Ensuring Nuclear Performance Life extention of AGR reactors HTR PM Qualification The Dutch Molten Salt Program
- Take away message
Introduction to NRG 3 MISSION & STRATEGY NRG responds to the social need for High-quality nuclear research and innovation.
Mission Safe and reliable Services to nuclear isotope organizations production. working with nuclear technology.
Advancing Nuclear Ensuring Nuclear Strategy Medicine Performance Activities
- Material & Fuel Qualification
- Research & Innovation Introduction to NRG 4 WHERE ARE WE?
} Petten NRG~ 650FTE All site ~ 1600 FTE
}
Arnhem 1600 employees on the Energy & Health Campus
Introduction to NRG 5 CAMPUS: PRESENT AND FUTURE Future Processing facility Transport and Logistics Pallas Waste High Reactor Moly Processing Hot Cell Treatme Flux Facility Lab nt 1600 employees on the Energy & Health Campus Reactor FieldLab Energy & Health Campus Development Existing Under der developm development Infrastructure 6 CAPABILITIES
Infrastructure 7 THE HIGH FLUX REACTOR 45 MW thermal power Stable and constant flux profile in each irradiation position 31 operation days per irradiation cycle, 9 cycles a year Infrastructure 8 THE HIGH FLUX REACTOR (HFR)
Infrastructure 9 FROM HFR TO PALLAS
- HFR is projected to operate until 2026, but has no fixed end-of-life date
- PALLAS is taking over the roles of HFR from ~2026 in a seamless cross-over
< 2026 > 2026 Ensuring Nuclear Performance 10 ENSURING NUCLEAR PERFORMANCE We will consolidate and grow our existing business, by supporting new generation reactors, conduct materials and fuel research, implement measures to safely extend the lifetime of existing nuclear operations. We perform certifications and define safety protocols, measurement and compliance, performance optimization, implement smart software solutions, support decommissioning, decontamination, waste management, and medical safety.
NUCLEAR TECHNOLOGY ACTIVITIES
- Nuclear Compliance
- Radiation Protection
- Software solutions (ROSA)
- Decommissioning D
- MMaterial & Fue Fuel el Qualifica Qualification n H HTR
- Research & Innovation MSR M
Ensuring Nuclear Performance 11 LIFETIME EXTENSION AGR REACTORS Supporting long life of Advanced Gas Cooled Reactors EDF Energy operates Advanced Gas Cooled Reactors, supplying
~ 20% of electricity in the United Kingdom Graphite cores age with time due to neutron damage and radiolytic oxidation Accelerated ageing tests to determine graphite properties ahead of actual AGR core structures Project Future AGR Blackstone data operation Weight loss Neutron dose Ensuring Nuclear Performance 12 HTR-PM FUEL QUALIFICATION Client: HTR-PM reactor INET, China.
Scope:
- From Design to Manufacturing.
- Fuel irradiation in HFR.
- Monitoring of fission gas release (gamma spectroscopy).
- Monitoring temperatures (in-situ thermocouples).
- Transport to JRC- Itu for Küfa tests, by which the qualification is completed.
Ensuring Nuclear Performance 13 HTR-PM FUEL QUALIFICATION
- 5 pebbles are placed in graphite samples holders
- Double containment
- A total of 48 thermocouples for accurate temperature registration
- Online gas monitoring
- Including neutron fluence registration
- Self Powered Neutron Detectors
- Activation monitor sets Ensuring Nuclear Performance 14 THE DUTCH MOLTEN SALT PROGRAM
- NRG = Enabler of MSR Technology due to nuclear know-how, infrastructure, international network.
- Collaborations with competence centers:
JRCs, TUDelft, FUBerlin and CV Rez.
- Objectives:
- 1. Obtain operational experience
- 2. Safety
- Investigate FP management methods
- 3. Material investigation:
- Material properties of irradiated containment materials
- In-pile corrosion / deposition of metal alloys and SiC
- 4. Waste:
- Provide a waste route for spent molten salt fuel
- 5. Integral Demonstration:
- Feasibility of experimental Molten Salt loop for the HFR Petten n
Ensuring Nuclear Performance 15 THE DUTCH MOLTEN SALT PROGRAM
- Focus on irradiation technology
- Focus on generic topics
- Ambitious program open for collaborations 16 TAKE AWAY MESSAGE EU DuC = N Goods labeled with an EU DuC (European Dual-use Codification) not equal to N are subject to European and national export authorization when exported from the EU and may be subject to national export authorization when exported to another EU country as well. Even without an EU DuC, or with EU DuC N, authorization may be required due to the final destination and purpose for which the goods are to be used.
No rights may be derived from the specified EU DuC or absence of an EU DuC.
Structural Materials Research within the Gen IV International Forum William Corwin Advanced Reactor Materials LLC Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop U.S. NRC Headquarters, Rockville, MD December 9-11, 2019 Generation IV Concept GIF Framework Agreement implemented 2005 Goals - Sustainability, Safety & Reliability, and Proliferation Resistance & Physical Protection Nearly 100 reactor designs were evaluated and down selected to 6 most promising concepts x Sodium Fast Reactor x Gas-Cooled Fast Reactor x Lead Fast Reactor x Very High Temperature Reactor x Supercritical Water Cooled Reactor x Molten Salt Reactor System Steering Committee for each reactor concept x Plan and integrate R&D Projects x Project Arrangements are technology-specific
Fourteen Current Members of Generation IV International Forum Australia
- Inactive members Generation IV Evolution 4
Sodium Fast Reactor Integral part of the closed fuel cycle x Can either burn actinides or breed fissile material R&D focus x Analyses and experiments to demonstrate safety approaches x High burn-up minor actinide bearing fuels development x Develop advanced components and energy conversion systems BN-800 operating in Russia & CEFR in China No specific materials R&D projects active, but extensive work on high-dose-tolerant fuel claddings 500-550°C Lead Fast Reactor Lead is not chemically reactive with air or water and has lower coolant void reactivity Variants include both lead and lead-bismuth cooled systems LFR MOU working towards 480-800°C SteeringCom & technical projects Principal materials issue is materials compatibility with lead or lead-bismuth coolant x Precise oxygen-chemistry control or protective cladding layers needed Europes ELFR lead-cooled system, Russias BREST-OD-300 and the U.S.
SSTAR system are actively being developed
Molten Salt Reactor Two design options: fuel dissolved in MS coolant, solid fuel with MS coolant Variants include both thermal & fast designs, with fluoride or chloride salts Key technical focus x Neutronics x Materials and components x Safety and safety systems x Liquid salt chemistry and properties x Salt processing MOU working towards SteeringCom & technical projects Materials compatibility with MS
- Salt chemistry controls
- Stable alloys w/o oxide layers 700-800°C
- High-dose-tolerant, fine-grained graphites that avoid salt intrusion Gas Fast Reactor High temperature, inert coolant and fast neutrons for a closed fuel cycle Fast spectrum enables extension of uranium resources and waste minimization High temperature enables non-electric applications Very advanced system Passive safety challenges Requires advanced materials and fuels x Key technical focus: SiC-clad carbide fuel x Lack of graphite will impact helium chemistry
§°C High temperature materials issues include in VHTR Materials Project
Supercritical Water Cooled Reactor Merges GEN-III+ reactor technology with advanced supercritical water technology used in coal plants x Operates above the thermodynamic critical point (374°C, 22.1 MPa) of water x Fast and thermal spectrum options 510-625°C Includes pressure vessel and pressure tube variants Key technology issues:
x Materials, water chemistry, and radiolysis x Thermal hydraulics and safety to address gaps in SCWR heat transfer and critical flow databases x Fuel qualification Materials and Chemistry R&D project focused on corrosion and SCC of alloys in SCWR conditions Very High Temperature Reactor High temperature He-cooled reactor enables both electrical generation and process heat applications Goal - VHTR outlet temperature of 950-1000°C, with near term (HTGR) focus on 750-850°C 750-950°C Reference configurations: prismatic &
pebble bed cores Japan HTTR & China HTR-10 in operation; China HTR-PM demo plant nearing completion Includes strong materials R&D focus x Graphite x Metals & Design Methods x Ceramics & Composites x Contributions shared via Gen IV Materials Handbook
Advanced Reactors Need Additional Limited Options for Approved Alloys for Elevated Temperature Construction Only six alloys are qualified in ASME Sec III Division 5 for service in inelastic temperature range x Two ferritic steels: 2 1/4Cr-1Mo and 9Cr-1Mo steels x Two stainless steels: 304 and 316 x Two high-temperature alloys: Alloy 800H and Alloy 617 Hastelloy N (and similar foreign alloys, GH3535) are not yet approved for liquid salt service Advanced alloys (e.g., high-entropy, ODS, TMT, Ni, etc.) need development and qualification Corrosion resistant alloys for lead and SCW compatibility (Modified 310 SS, high Si F/M steels, FeCrAl alloys, new Ti & Zr alloys) need development and qualification Beryllium for compact reactor reflectors Graphites, Ceramics, and Composites Need Qualification and Code Coverage for Advanced Reactors Graphite used for core supports Graphite Core in HTGRs, VHTRs & MSRs Supports Ceramics & composites used as specialized reactor internals Special issues x Insufficient material standards x Lack of ductility x Need for statistically set load limits x Coupled irradiation and environmental effects Composite Reactor Now included in ASME Code InternalsSection III Division 5 Additional qualification required Ø1.5 m Hot Gas Ducts
Very High Temperature Alloys (800H, 617, Hast X(R) & Hast N) for IHXs and SGs Are of Greatest Concern for HTRs Temperatures 700 up to 950°C Corrosion and creep-fatigue damage 800H & 617 are ASME Code qualified to 762°C HTR PM Heat Exchanger
& 950°C Alloys X & X(R) suitable Printed Circuit Board but not yet Code Heat Exchanger qualified 2 1/4Cr-1Mo Code-qualified for lower SG temperatures Improved high temperature design methodology essential THTR-300 Steam Generator (Courtesy Heatric)
High Temperature Design Methodologies Need Updating Weldments x Weldment evaluation methods, metallurgical & mechanical discontinuities, transition joints, tube sheets, validated design methodology Aging & environmental issues x Materials aging, irradiation & corrosion damage, short-time over-temperature/load effects Creep and fatigue x Creep-fatigue (C-F), negligible creep, racheting, thermal striping, buckling, elastic follow-up, constitutive models, simplified &
overly conservative analysis methods Multi-axial loading x Multi-axial stresses, load combinations, plastic strain concentrations
High Temperature Code Materials and HTDM Need Updating (cont)
Materials allowables x Elevated temperature data base & acceptance criteria, min vs ave props, effects of melt & fab processes, 60-year allowables Failure criteria x Flaw assessment and LBB procedures Analysis methods and criteria x Strain & deformation limits, fracture toughness, seismic response, core support, simplified fatigue methods, inelastic piping design, thermal stratification design procedures Rules for design/use of clad structures for high temperature service, including efficacy and reliability of coatings or cladding for corrosion prevention Construction rules for CHEs for high temperature service SiC C/SiC C /SiCCC Composites Are Potenti Potentially Applicable to Many Advanced Reactor Concepts Reactor Operating Project / Design Possible Application Concept Condition Examples Deployment
- He, Pb-Li
- ARIES
- Blanket structures Fusion
- 400-900°C
- EU-PPCS
- Long-term
- Various functions
- >50 dpa
- DREAM
- He HTGR
- Reaction control systems
- SC-HTGR
- 600-1100°C
- Near-term VHTR
- Core support
- GT-HTR300C
- Up to ~40 dpa
- Channel box
- Water
- Mid-term?
- Grid spacer
- 300-500°C
- Fuel cladding * ~10 dpa
- Liquid salt FHR
- Core structures
- AHTR
- ~700°C
- Long-term AHTR
- >10 dpa
- Liquid sodium
- Core structures SFR
- 500-700°C
- Long-term
- Fuel cladding/support
- >100 dpa
- He
- Core structures
- CEA GFR
- 700-1200°C
- Long-term
- Fuel cladding/support
- GA EM2
- >100 dpa
Structural Composites Are Being Developed & Qualified for Advanced Reactors Boron Carbide Composite Compact SiC-SiC and C-C are best candidates for non-metallic control rods Control Rod C-C composites also evaluated for structural reactor internals applications at lower doses Irradiation, corrosion, architecture, manufacture & testing standard development all needed Lateral Core Restraint Rings Core Tie Rod Alloys & Graphite in High Temperature Coolant Environments Must Be Qualified Oxidation, carburization, and decarburization of metallic components SCC and LME effect Microstructural stability &
strength impacts during long-term aging Mass and strength loss in graphite and composites Impact of coolants on metallic tribology High temperature strength, Alloy environmental compatibility 617 Aged
& corrosion in Hi Temp 100 coolants Chronic Hours graphite at oxidation 1000°C in air
Environmental Compatibility Is Especially Challenging for MSRs, LFRs, and SCWRs Protective oxide films not formed in molten salts Lead coolants require very tight oxygen control & may cause liquid metal embrittlement Enhanced stress corrosion cracking in supercritical water plus impact of irradiation-induced free radicals Development of coatings, GIF SCWR ROUND ROBIN EXERCISE #2 Canadian Nuclear Laboratories Reference # 900-511300-STD-001 claddings & associated design methods needed Qualification of new materials may be required 1.4970 steel exposed to LBE at
- Modified 310 SS, high Si F/M 600°C A. Weisenburger et. al.
IAEA Tech Mtg on structural materials steels, FeCrAl alloys, advanced for heavy liquid metal cooled fast reactors, Vienna, 15-17 October hi temp nickel alloys, new Ti & Zr alloys Compact Heat Exchanger Usage in HTRs Requires Qualified Design and Construction Rules Complex channels &
Sabharwall et al.2013 corners result in stress concentrations x Transfer external boundary conditions transfer to internals x Significant pressure & thermal stress redistribution Southall et al., ICAPP 08 MicroChannel x Chemically etched integral flow channels x Good for high pressure PlateFin x Corrugated plate fin sandwiched between two flat plates or shims x More efficient use of materials and Southall et al., larger flow passages ICAPP 09
Each Advanced Reactor Requires Materials for Its Own Temperature, Dose & Coolant Compatibility Needs Materials R&D in the Gen IV International Forum Is Generated and Shared Some Project Arrangements for Materials R&D are in place x Very High Temperature Reactor x Gas-Cooled Fast Reactor x Supercritical Water Cooled Reactor Other reactor systems have identified R&D needs x Lead Fast Reactor x Molten Salt Reactor x Sodium Fast Reactor R&D needs for metals, graphite, ceramics, composites, and high temperature design methods span multiple systems Environmental challenges related to coolant compatibility, irradiation doses, and service temperatures are reactor specific
Applying Brittle Materials Design Concepts in the e Manufacture of Graphite and Cerami Ceramics for 21st Century Nuclear Reactors MATERIALS MATTER 2M Dr. Makuteswara Srinivasan, Analyst, Materials Matter, Germantown, Maryland, U.S.A.
msrini.2M@gmail.com Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop U.S. Nuclear Regulatory Commission, Rockville, MD December 9-11, 2019 Designer and Manufacturer Interactions D&M For Safety Performance D&M For Ease of Inspection Pre and Post Installation Online Monitoring In-Service Shutdown D&M For Ease of Cost-Effective Decommissioning For Life Waste Minimization Base represents 10 CFR Part 50, Appendix B, QA Requirements ASME NQA-1; ASME Div. 5 D&M For Ease of Dis-Code, HAB Requirements Assembly, Ease of ISO 9000 (QMS)/9001 Replacement (Documentation)/TQM/
Lean Six Sigma D&M For Quality Manufacturability © M. Srinivasan, Materials Matter, 2019.
MATERIALS MATTER 2 2M
ASME Considerations for HTR Graphite and Composite Components DOE - Advanced Reactor Technologies William Windes - Idaho National Laboratory Advanced Non-Light Water Reactors - Materials & Component Integrity Workshop December 9-11, 2019 NRC Headquarters, Rockville, MD (US)
Discussion points Underlying graphite behavior behind ASME code development Inherent graphite behavior Degradation effects on properties and behavior Irradiation and oxidation effects ASME code - as written to address graphite behavior Probabilistic versus Deterministic failure assessments Failure criteria Full assessment and simple assessment Unirradiated versus irradiated data requirements Oxidation requirements Gaps in ASME graphite code Preliminary gap analysis results on ASME Code
Graphite Manufacture & Unique Properties
- All graphite grades are proprietary. Only limited/general fabrication data is known.
It isnt like metals - there are no ASTM specified alloy compositions and fabrication processes These are closely guarded, proprietary formulae for each grade And no, graphite suppliers are not willing to give up their private recipes to the nuclear community
- Remember: no nuclear graphite has been ordered in decades
- But the good news is that all grades react similarly under nuclear core conditions Specific changes are dependent upon individual grade
- To understand graphite behavior need to know unique manufacturing processes Graphite is a porous material (15-20%) - By design!
Porosity provides thermal and irradiation stability But porosity alone doesnt predict behavior
- Large flaws (cracks/pores) are built in from fabrication
- But irradiation stability also comes from small (nm) pore structure Filler particles, binder phase, filler-binder ratio, fabrication method
- All affect performance 3 Graphite Manufacture & Unique Properties
- Properties and performance influenced by both PBMR (600MW) ~ 0.85 dpa/yr SMR (200 MW) ~ 0.3 - 0.5 dpa/yr raw materials and processing Micro (5 - 50 MW) ~ 0.1 - 0.25 dpa/yr Isotropic (or near isotropic) material response Full ~ 22 yrs High thermal stability > 3000°C SMR ~ 38 yrs High heat capacity (thermal sink) Micro ~ 76 yrs High thermal conductivity (better than metal) Full ~ 18 yrs SMR ~ 30 yrs Neutron moderator - thermal reactor Full ~ 12 yrs Micro ~ 60 yrs Easy machinability / cheap material SMR ~ 24 yrs Micro ~ 48 yrs Ceramic like material response
- Low fracture toughness (~ 1-2 MPa m)
- Quasi-brittle cracking
- Low tensile strength (ceramic composites used)
- Decent irradiation response Dimensional change (life-limiting mechanism)
- Multiple decades of safe operation
- And even longer at lower temperatures Graphite generally gets stronger with irradiation Isotropy stays relatively constant
- A small reduction possibly M.C.R. Heijna, S. de Groot, J.A. Vreeling, "Comparison of irradiation behaviour of HTR graphite Thermal stability and capacity are unaffected grades", Journal of Nuclear Materials 492 (2017) 148e156 4
Irradiation Effects on Graphite Properties
- Irradiation induced changes must be considered in design
- Significant changes occur during normal operation in:
Component dimensions
- Components actually shrink
- Until Turnaround when they begin to expand until failure Density
- Components become more dense
- After Turnaround dose they decrease in density Strength and modulus
- Graphite gets stronger with irradiation
- Until Turnaround dose is achieved. It then decreases Thermal conductivity
- Decreases almost immediately to ~30% of unirradiated values Coefficient of thermal expansion
- Initially increases but then reduces before Turnaround until saturation Irr. creep effects
- Significant changes do not typically occur in the following properties:
Oxidation rate, neutron moderation, specific heat capacity, emissivity, heat capacity
- No Wigner energy release if components irradiated above 300°C.
5 What are the main parameters needed to develop ASME code?
- 1. There is no single nuclear grade of graphite We cant design around a specific nuclear grade as metals can (i.e., 316, 316L, 617, etc.)
- 2. Graphite has significant flaws (pores/cracks) - by design We do not want to eliminate these flaws
- 3. Graphite is not ductile Brittle or quasi-brittle fracture behavior
- 4. Irradiation significantly alters the graphite behavior Behavior is completely different before and after Turnaround dose is achieved.
6
ASME Code Considerations 50X
- Because all graphite has significant flaws Some amount of failure (i.e., a crack) is certain
- Therefore, core components need to be designed to 100X accept some amount of failure. 200X Probability of failure approach is taken Based upon overlap of applied stresses and inherent strength of the nuclear grade used 500X
- Probablistic verses deterministic design approach 7 Deterministic is generally too limiting for a brittle material A distribution of possible strengths in a material is needed for quasi-brittle materials (i.e., flaw size for graphite).
Probability of failure in component based upon inherent strength of graphite grade and applied stresses during operation.
- New graphite grades are consistent and ready for codification Unfortunately, historical nuclear grades are no longer available From Dr. Mark Mitchell - PBMR Inc. We also lack significant quantitative data on new graphite behavior at higher temperature and high dose applications 7 Need to correlate defined material changes to assist in failure analysis.
How the graphite (and composite) ASME Code works Three methods are provided for assessing structural integrity Structural Maximum
- 1. Deterministic Reliability Probability Simplified conservative method based on ultimate strength derived from Weibull Class of Failure statistics.
Irradiation changes well contained within the operational envelope SRC-1 1.00E-04
- 2. Full Analysis Method SRC-2 1.00E-02 Detailed structural analysis taking into account stresses, temperatures, irradiation SRC-3 1.00E-01 history, and chronic oxidation effects.
Weibull statistics used to predict failure probability Maximum allowable probability of failure defined for three Structural Reliability Classes (SRCs), which relate to safety function
- 3. Qualification by Testing Full-scale testing to demonstrate that failure probabilities meet criteria of full-analysis method.
The graphite code is a process. Not just picking a preapproved material The applicant must demonstrate the graphite grade selected will consistently meet the component requirements.
Getting the material property proof is responsibility of the applicant 8
Flow diagram for the Full Assessment ASME Section III.5 Subsection HH Subpart B - 2017 Full Assessment: 3 parameter Weibull (Probabilistic Analysis)
Calculate the POF of the Define the Material Estimate 3 parameter graphite core component Reliability Curve by fitting a Weibull parameters using the Material Reliability POFcomponent 3 parameter Weibull model using MLEs .
to the measurement data. Curve and stress distribution (So, m095% and Sc095%) in the component.
(ref. HHA-II-3200 pg. 417)
(ref. HHA-3217 pg. 393)
Determine the allowable POF Evaluate the acceptability of the design from the Structural Reliability Class (SRC), and Service Level POFallowable --4 POFcomponent < POFallowable Design Loading.
(ref. HHA-3230 thru HHA-3237 pg. 397) 9 Flow diagram for the Simple Assessment ASME Section III.5 Subsection HH Subpart B - 2017 Simple Assessment: 2 parameter Weibull (Deterministic Analysis)
Using m* and Sc*
Estimate the scale and Using m95% and Sc95% Calculate the determine the Weibull shape of a 2 parameter determine the design ratio of parameters Weibull using a linear allowable stress as a flexural to Sg(P) corresponding to a fit to measured function of POF = 10-4, tensile Rtf 95% confidence property data 10-3, 10-2 and 5x10-2 strength interval m* and Sc* Sg(P) Rtf m95% and Sc95%
(ref. HHA-II-3100 pg. 414) (ref. HHA-II-3300 pg. 418) (Ref.HHA-II-2000 pg. 412)
(ref. eq.6 and eq.7 pg. 417)
Perform a stress Cm = Combined Membrane Stress Evaluate the acceptability of the design analysis of the Cb = Combined Bending Stress - Cm < Sg(P) graphite F = Peak Stress - Cm + Cb + F < Rtf
- Sg(P) component Rtf = ratio of flexural to tensile strength (ref. HHA-3220 pg. 394)
(ref. HHA-3215 pg. 392 and HHA-3216 pg. 393) 10
And then the hard part
- Fundamental material properties change with irradiation/oxidation must be addressed Applicant must assess changes to design of component due to Irradiation effects
- New cracks formed after Turnaround
- Internal stresses from dimensional change. Need creep response, too
- Changes to density, strength, CTE, thermal conductivity Turnaround G. Haag, Properties of ATR-2E Graphite and Property Changes due to Fast Neutron Irradiation, Juel-4183, 2005 Applicant must assess changes to design due to Oxidation degradation
- Changes in density, strength, CTE, neutron moderation, and thermal conductivity.
11 Finally a word on code improvements No one has used the new graphite or composite code Very little feedback from vendors or applicants
- In-situ inspections for continuous operating PB & MSR designs is difficult Split disk development lop pme ment nt Some discrepancy between ASME code and available ASTM testing Currently there are no standard test methods for Mechanical testing of small (sub-sized) specimens as needed for irradiation testing No mechanical testing of oxidized specimens Performing NDE techniques on large graphite components Effects of oxidation on full-scale components Current test standards compare graphite grades Nothing to address the effects on large components Fatigue - does it apply?
No studies on fatigue of graphite components U.K. shows low cycle - large stress events (fatigue) promote crack formation in bricks Underlying mechanisms are not well understood Affects the code and how it is applied Will lead to standardized nuclear graphite grades 12
13 13 GRAPHITE ACTIVITIES AT NRG Research and qualification overview T.O. van Staveren Washington, 12 December 2019 2
OVERVIEW
- Graphite research and qualification activities at NRG
- Graphite irradiation program for Advanced Gas-Cooled Reactors
3 GRAPHITE IS LIFE-LIMITING
- Graphite in nuclear reactors
- Moderator
- Core structural integrity
- Life-limiting component
- Impact on economics
- Neutron-damage to graphite
- Changes in material properties increase stresses
- Increased stresses lead to graphite failure
- Moment of failure depends on graphite material properties
- High quality graphite irradiation data needed for graphite selection, design and safety cases 4
GRAPHITE IRRADIATION BEHAVIOUR
- Volume change
- Induces stresses in graphite components
- Length change in AG and WG direction
- Induces anisotropy
- Induces stresses in graphite components
- Coefficient of Thermal Expansion
- Induces stresses in graphite components
- Youngs Modulus
- Response of graphite to load and deformation
- Irradiation creep
- Deformation and stress relief of graphite under irradiation and load
- Strength
- Determination of failure under load/stress
- Stress build-up leads to graphite component failure
- How and when > requires accurate data
5 GRAPHITE PROGRAMS AT NRG
- Research for GenIV reactors (2001-2015)
- Study of modern graphite grades for High Temperature Reactors
- Work conducted within European Framework programs
- EDF Energy AGR life time extension (from 2006)
- BLACKSTONE: irradiation of AGR graphite under oxidising conditions
- ACCENT: irradiation creep irradiation
- Irradiations for current graphite suppliers and reactor developers
- Graphite irradiation as part of MSR program 6
HTR GRAPHITE IRRADIATION AT NRG
- 5 INNOGRAPH irradiations
- Modern-day HTR graphite candidates
- European framework programs (HTR-M, RAPHAEL, ARCHER)
- Low, medium, and high dose ranges to cover graphite behaviour beyond cross-over
- Re-load of active A-phase samples in B-phase for time-efficient acquisition high dose data Schematic of INNOGRAPH irradiation campaign 2001-2006 2006-2011 2011-2015
7 DIMENSIONAL CHANGE AND IRRADIATION TEMPERATURE Heijna et al., Comparison of irradiation behaviour of HTR graphite grades, Journal of Nuclear Materials 492:148
- May 2017 8
USE OF MTR DATA Irradiation
- Application of graphite within reactor core can be limited by the range and accuracy with which the material is characterised in an MTR program, e.g.:
- Temperature
- Dose
- Stress
- Specimen environment (inert / oxidising etc.)
- Control of these parameters is performed by design and operation of the irradiation facility, in order to:
- Meet target irradiation conditions
- Minimise deviations, drift and uncertainties
- There is no qualitative or quantitative specification in international standards on the accuracy and method by which the irradiation conditions (e.g. temperature, stress) should be obtained Material characterisation
- Measurements are typically performed on small sized specimens, which may require additional validation efforts
9 LIFETIME EXTENSION AGR REACTORS Supporting long life of Advanced Gas Cooled Reactors EDF Energy operates Advanced Gas Cooled Reactors, supplying
~ 20% of electricity in the United Kingdom Graphite cores age with time due to neutron damage and radiolytic oxidation Accelerated ageing tests to determine graphite properties ahead of actual AGR core structures Project Future AGR Blackstone data operation Weight loss Neutron dose 10 BLACKSTONE INTRODUCTION Aim >
- Collect material property data in advance of the operating AGR stations, specifically at high dose and weight loss
- Demonstrate lack of cliff-edge at end of generation conditions Approach >
- Irradiate specimens High Flux Reactor (HFR) in Petten using accelerated conditions relative to the AGRs
- Use source material from operating AGRs Key succes factors >
- Representative irradiation conditions
- Generating similar graphite property evolution in AGR and HFR
- Accurate measurements on small specimens
- Ensuring quality for application of data in reactor safety cases
11 BLACKSTONE PROJECT FLOWCHART Samples drilled from AGR reactor core and shipped to NRG Petten Machining of specimens from extracted discs in hot cell laboratories (HCL)
Characterisation of radioactive graphite specimens in hot cell laboratories 12 BLACKSTONE PROJECT FLOWCHART Blackstone capsules are assembled in hot cell laboratories Blackstone capsules are loaded into the HFR and connected to a dedicated gas handling system that controls graphite oxidation After irradiation the capsules are dismantled, and post-irradiation characterisation on the specimens is carried out
13 REPRESENTATIVE IRRADIATION CONDITIONS (1/2)
- Temperature
- Validated thermomechanical models using MCNP input, using true dimensions of specimens and capsule materials
- Online temperature measurement and control
- Temperatures were well controlled, generally within
+/-20°C of target with minimal drift
- Dose
- MNCP models using predicted and acual core loadings
- Validation of model predictions by measurement of neutron activation monitor sets
- Environment: weight loss / oxidation control
- Weight loss controlled by monitoring chromatography on the capsule outlet and adjusting the inhibitors (mainly CO, CH4, C2H6)
- Final weight loss calculated for each specimen using the starting weight loss and measured mass 14 REPRESENTATIVE IRRADIATION CONDITIONS (2/2)
Blackstone Capsule 1
- Microstructural characterisation to AGR demonstrate similarity (or otherwise) of aging mechanisms 3/4 The AGR and Blackstone microscopy images have been compared and are broadly similar
- Data analysis to demonstrate consistency and tie in with AGR data 3/4 The measurements from the different Blackstone capsules and oxidation rates have been compared and are broadly consistent.
3/4 The Blackstone measurements have been compared to the AGR inspection and monitoring data and are of similar magnitude
15 VALIDATION OF MEASUREMENTS
- Comparison with international (ASTM) standards
- Investigation of the influence of non-conformities with standards
- Perform measurements on standard materials with known properties
- Perform measurements on similar materials (thermally oxidised specimens with comparable weight losses)
- Show repeatability of test method
- Prove independence of operator / experimental conditions and specimen size
- Independent review of work procedures (e.g. NPL)
- Tie-in with inspection data from AGRs
- Historical MTR data 16 CONCLUSIONS
- Graphite research and qualification activities at NRG, supporting
- High Temperature Reactors
- Advanced Gas-cooled Reactors
- Molten Salt Reactors
- Recent graphite irradiations at NRG are well controlled and have successfully supplied data for:
- Analysis of modern graphite grades for HTR application
- Life time extension of AGR reactors
- Apply lessons learned on control of irradiation conditions and validation of measurements in support of development of new reactor types and graphite qualification
Understanding Graphite Behavior in Nuclear Reactor Environments for Lifetime Predictions Anne A. Campbell, Ph.D.
Oak Ridge National Laboratory Oak Ridge, TN campbellaa@ornl.gov US-NRC Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop, December 9-11 ORNL is managed by UT-Battelle, LLC for the US Department of Energy Irradiation Effects in Graphite
- Irradiation damage in graphite causes changes to all the See Fig.1 from following paper:
physical, thermal, and Heijna et al. Journal of Nuclear Materials, v492 (2017) p148 mechanical properties of graphite
- Property change trends are similar between grades, but the net change is different for each
- Example is the plot of volume change to the right showing 4 different grades 2 Heijna et al. JNM v492 p148
Volumetric Change
- Graphite dimensional changes are a result of crystallite dimensional change and graphite texture.
- Swelling in c-direction is initially accommodated by aligned microcracks that form on cooling during manufacture.
- Therefore, the a-axis shrinkage initially dominates and the bulk graphite exhibits net volume shrinkage.
- With further irradiation, incompatibilities in crystallite strains causes the generation of new porosity and the volume shrinkage rate falls eventually reaching zero.
3 Campbell, et al., Carbon, 109 (2016) 860-873 Neutron Fluence and Temperature Effects on Properties
- Volume change follows semi-parabolic shape for increasing fluence
- Higher temperature reduces net contraction and time for swelling onset
- Strength, Youngs modulus, and shear modulus have trends that appear to mimic the density change
- Following density curve is seen for isotropic fine-grained graphite and is different behavior than seen in coarser grades
- Poissons ratio does not change 4
Campbell, et al., Carbon, 109 (2016) 860-873
Neutron Fluence and Temperature p Effects on Properties p
- Thermal conductivity
- Coefficient of thermal expansion
- No density dependence like mechanical properties 5
Campbell, et al., Carbon, 109 (2016) 860-873 Irradiation-Induced Creep
- For most nuclear applications irradiation-induced creep is not desirable
- For graphite the internal stress buildup is a function of neutron fluence and localized temperature
- Would quickly surpass the UTS of graphite (15-25 MPa)
- Process is actually internal stress relaxation, unlike irradiation creep of torqued baffle bolts 6
Irradiation Creep/Stress Relaxation
- Volume change is controlled by irradiation temperature and flux See Figure 6 from following paper:
- In large components the temperature and flux Tsang, B.J. Marsden, Journal gradients can be quite severe of Nuclear Materials, 350
- Internal stress formation and buildup (2006), 208-220
- Can quickly surpass the graphite strength
- Irradiation-induced creep acts as a stress relaxation mechanism See Figure 8 from following See Figure 9 from following paper: paper:
Tsang, B.J. Marsden, Journal Tsang, B.J. Marsden, Journal of Nuclear Materials, 350 of Nuclear Materials, 350 (2006), 208-220 (2006), 208-220 7
D.K.L. Tsang, B.J. Marsden, JNM, 350 (2006), 208-220 Creep Effects on Dimensional Change See Figure 42, from the following report: See Figure 43, from the following report:
G. Haag, Properties of ATR-2E Graphite and G. Haag, Properties of ATR-2E Graphite and Property Changes due to Fast Neutron Property Changes due to Fast Neutron Irradiation, Jül-4183 (2005) available at: Irradiation, Jül-4183 (2005) available at:
http://juser.fz- http://juser.fz-juelich.de/record/49235/files/Juel_4183_Haag.p juelich.de/record/49235/files/Juel_4183_Haag.p df df ATR-2E Graphite (WG), Tirr = 550ºC, 5 MPa compressive stress ATR-2E Graphite (WG), Tirr = 500ºC, 5 MPa tensile stress 8
Haag, G., Jül-4183, (2005).
Irradiation Creep Behavior in Graphite
- Complex to measure
- Need both a stressed specimen and a reference specimen to account for non-stressed dimensional change
- Behavior is also complex
- Atypical temperature dependence 9
Campbell, Carbon, 139 (2018) 279 Lifetime Limiting Questions
- Does graphite need to be removed from the core?
- Refueling
- Replacement
- Is strength life limiting property? Is dimensional change life limiting property?
- ASME B&PV Section III Division 5 Code states Material that exceeds this fluence limit is considered to provide no contribution to the structural performance (stiffness and strength) of the Graphite Core Component.
This fluence limit shall be set to the fluence at which the material experiences a +10% linear dimensional change in the with-grain direction.
- Maybe final tensile strength should be limit.
- Is life limiting the same for each reactor type?
10
What Limits Graphite Lifetime in Nuclear Reactors?
Pebble Bed Prismatic HTGR
- Need to remove graphite HTGR MSR from core is most likely life limit for prismatic HTGR, since fuel is located in blocks
- Pebble bed HTGR may only need to replace graphite if strength cannot hold up the block stack in the reflector region
- Molten salt lifetime may be when flow channels cannot be maintained 11 Campbell, et al., Carbon, 109 (2016) 860-873 Concerns with Graphite in MSRs
- Salt wetting into pores
- Fission gas retention
- Localized changes in temperature, mismatch - 3H top concern with any salt with Li from of graphite and salt 6Li(n,3H)He and 7Li(n,3H)He+n reactions thermal/physical/mechanical properties - 135Xe - neutron poison, main concern in
- Fuel intrusion with salt - makes reactor breeder reactors simulations and source terms difficult to define
- Reduce gas permeability below 10-8 cm2/s of He at
- LiF-BeF2 (Capillary experiments (ORNL-4254)) STP (ORNL-4812)
- No intrusion in 0.8 mm hole
- 5HTXLUHVVXUIDFHSRUHVGLDPHWHU P
- <1 mm intrusion into 1.6 mm hole
- Multiple surface preparation techniques suggested to reduce permeability
- 1.1 cm intrusion in 3.2 mm hole
- 0.7 cm intrusion into 6.35 mm hole
- Salt intrusion may locally effect graphite
- Treatment of salt with HF or metallic Zr temperature, strength, and lifetime decreased wetting by 30% - Localized hot/cold spots reduces graphite lifetime
- Salt, graphite, and metal - Salt has different physical/mechanical/thermal reactions/corrosion properties than the graphite
- Irradiation effects on graphite size, strength, thermal properties 12
Understanding Graphite Behavior in Nuclear Reactor Environments for Lifetime Predictions Anne A. Campbell, Ph.D.
Oak Ridge National Laboratory Oak Ridge, TN campbellaa@ornl.gov US-NRC Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop, December 9-11 ORNL is managed by UT-Battelle, LLC for the US Department of Energy
The Chemistry of Graphite in FHRs and MSRs Raluca O. Scarlat scarlat@berkeley.edu Advanced Non-Light Water Reactors - Materials & Component Integrity U.S. NRC December 10, 2019 Acknowledgements:
Digby Macdonald (UC Berkeley). Cristian Contescu, Tim Burchell, Nidia Gallego, Anne Campbell (ORNL).
Will Windes (INL). David Carpenter, Lin wen Hu, Ron Balinger, Charles Forsberg (MIT). Craig Marshall Funding acknowledgements:
(University of Kansas). Martin Straka (Rez Institute, NEUP-15-8352 Czech Republic). Kumar Sridharan, Mark Anderson, NRC-HQ-84-15-G-0046 Huali Wu. Francesco Carotti (UW Madison). IRP-14-7476 FLiBe-Graphite l Wetting and Surface Tension
#
#
LJFGHK
R*+ DQP IE
& Q
B Q >@=6235 ?5:>8;:
& C Q 3;:?23? 2:795
# (1 Q <=5>>@=5 48665=5:?829 2? <;=5 5:?=2:35
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% W. R. Grimes, Reactor Chemistry Division Anuual Progres Report for Period Ending January 31, 1964, ORNL-3591. 1964.
W. R. Grimes, Molten Salt Reactor Program Semi-annual Progres Report for Period Ending July 31, 1964, ORNL-3708. 1964. p. 255.
FLiBe-Graphite l Contact Angle Measured contact angle: large scatter in measured data points.
Salt intrusion is not expected at 1.1 kPa-gauge:
LJFGHK DQP 95o contact angle, 0.2 N/m surface tension => Intrusion pore diameter > 60 um IE 120o contact angle, 0.25 N/m surface tension => Intrusion pore diameter > 400 um
[1] A. R. Delmore, W. Derdeyn, R. Gakhar, R. O. Scarlat. Wetting of Nuclear Graphite by Molten Fluoride Salts: Initial Experiments. American Nuclear Society Annual Meeting. Philadelphia, PA. June 17-21, 2018. [2] W. R. Grimes, Reactor Chemistry Division Anuual Progres Report for Period Ending January 31, 1964, ORNL-3591. 1964. [3] Briggs. ORNL-3529, p.125.(1965).
FLiBe-Graphite l Salt-Intrusion
" !
#
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We do not expect 2LiF-BeF2 (FLiBe) to intrude in the pores of graphite, at ambient pressure.
[1] H. Wu, F. Carotti, N. Patel, R. Gakhar, R. O. Scarlat. Fluorination of Nuclear Graphite IG-110 in Molten FLiBe salt at 700 oC. Journal of Fluorine Chemistry. 211 (2018) 159-170. [2] H. Wu et. al. Comparative analysis of microstructure and reactive sites for nuclear graphite IG-110 and graphite matrix A3. Journal of Nuclear Materials. 528 (2020) 151802.
FLiBe-Graphite l Salt-Exposure Experiment
[inches]
12h exposure at 700 oC
<1ppm O2, <1ppm H2O in Ar no observable weight change of salt-exposed graphite: < 0.016 % [2]
H. Wu, F. Carotti, N. Patel, R. Gakhar, R. O. Scarlat. Fluorination of Nuclear Graphite IG-110 in Molten FLiBe salt at 700 oC. Journal of Fluorine Chemistry. 211 (2018) 159-170.
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No Salt Intrusion Observed Fluorination was observed H. Wu, F. Carotti, N. Patel, R. Gakhar, R. O. Scarlat. Fluorination of Nuclear Graphite IG-110 in Molten FLiBe salt at 700 oC. Journal of Fluorine Chemistry. 211 (2018) 159-170.
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FLiBe-Graphite l Graphite Fluorination MFn = metal fluorides CS = control sample TS = test sample C* = reactive carbon sites in graphite M = reduced metal CFx = fluorinated carbon sites in graphite
- / - O :A .50M , :/0MN O A.5 H. Wu, F. Carotti, N. Patel, R. Gakhar, R. O. Scarlat. Fluorination of Nuclear Graphite IG-110 in Molten FLiBe salt at 700 oC. Journal of Fluorine Chemistry. 211 (2018) 159-170.
FLiBe-Graphite l Reactive Carbon Sites MFn = metal fluorides C* = reactive carbon sites in graphite M = reduced metal CFx = fluorinated carbon sites in graphite Reactive carbon sites (C*) in graphite, enable this reaction.
Otherwise, the following reaction does not proceed spontaneously:
FLiBe + C -> metal + fluorinated carbon
[1] H. Wu, F. Carotti, N. Patel, R. Gakhar, R. O. Scarlat. Fluorination of Nuclear Graphite IG-110 in Molten FLiBe salt at 700 oC. Journal of Fluorine Chemistry. 211 (2018) 159-170. [2] HSC 9.2.
Variability Among Graphite Grades l Reactive Carbon Sites
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J. D. Hunn and M. P. Trammell. Data Compilation for AGC-2 Matrix-only Compact Lot A3-H08. ORNL/TM-2010/304. Oak Ridge National Laboratory, Oak Ridge, TN. (2010).
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%" $*#" $ %" )%*#" $ " )% 9 Variability Among Graphite Grades l Reactive Carbon Sites Cryst. 1 Cryst. 2 Slightly higher A3 degree of graphitization would predict lower or similar H2 chemisorption.
[1] Huali Wu, Ruchi Gakhar, Allen Chen, Stephen Lam, Craig P. Marshall and Raluca O. Scarlat. Comparative Analysis of Microstructure and Reactive Sites for Nuclear Graphite IG-110 and Graphite Matrix A3. Journal of Nuclear Materials. 2019. [2] 657/,##$0$%((6$.?;'41*(0%(+$8,14,0&$4%10$0'
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Variability Among Graphite Grades l Reactive Carbon Sites Prior correlations (valid across 16 grades of isotropic graphite, 0-0.7 dpa neutron irradiation) assume uniform density of reactive carbon sites at crystallite edges.
Correlations do not extend to graphite matrix.
Huali Wu, Ruchi Gakhar, Allen Chen, Stephen Lam, Craig P. Marshall and Raluca O. Scarlat.
Comparative Analysis of Microstructure and Reactive Sites for Nuclear Graphite IG-110 and Graphite Matrix A3. Journal of Nuclear Materials. 528 (2020) 151802.
%" $*#" $ %" )%*#" $ " )% 11 The Chemistry of Graphite in FHRs and MSRs Future Questions to Investigate O2 solubulity in the melt accelerates graphite degradation?
W. Derdeyn, R. Scarlat. 2019 (unpublished)
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Graphite Electrode Behaviors and its Application for Salt Purification Jinsuo Zhang Nuclear Engineering Program, Mechanical Department Virginia Tech Outline Anodic Effects Non-metal element separation Metal element separation Galvanic Corrosion by graphite 2
Anode effect on g raphite electrode in molten FLiNaK Redox couple (C2/O2): WFx + e W + xF-Redox couple (C1/O1): K+ + e K(l)
Reaction O3-anode effect: C + xF- - xe = CFx Redox couple (C5/O5): Cl2 + 2e 2Cl-Redox couple (C4/O4): Mg2+ + 2e Mg (a) CV tests recorded from W and graphite WE in FLiNaK at 1023 K; (b) CV test recorded from graphite electrode in MgCl2-NaCl-KCl at 1023 K. Scan rate: 500 mV/s.
3 Anode effect on g raphite electrode in molten FLiNaK A suddenly drop shows the anodic effects (a) CV curve recorded from a graphite WE (area: 0.21 cm2) in FLiNaK melt at 973 K under argon atmosphere; (b) the enlargement of the curve in Figure 2a. Scan rate: 500 mV/s. CE and RE: Graphite and Pt; (c) CV curves recorded from graphite WE at various scan rates; (d) CA tests recorded in FLiNaK at 973 K. Electrolysis time interval: 150s. OCP interval: 300 s. OCP curves are not shown here.
4
F LiNaK--L a 2O3 : Separation of Oxygen Ip Er CO or CO2 Ip decreases with Er increasing, CF4 Or other C-F compounds which shows the anodic effects can limit the current, therefore, can influence the separation of oxygen WE: graphite, CE: graphite, RE: tungsten (a) CV scan in FLiNaK-La2O3 at 700°C, scan rate 100mV/s, electrode surface area: 0.9381cm2 (b) CV scan in FLiNaK-La2O3 at 700°C, scan rate 1000mV/s, electrode surface area: 0.8514cm2 5
Graphite electrode in molten FLiNaK with NaI-II odine Separation Potassium CV curves recorded from a graphite WE (area: 0.21 cm2) in FLiNaK melt with various NaI concentration at 973 K I2 in ethonal (left) and in starch solution (right) under argon atmosphere; Scan rate: 500 mV/s. CE and RE: Graphite and Pt; 6
F LiNaK--L a 2O3-SSeparation La3+/LaC2 (b) Black curve: CV scan in FLiNaK-La2O3 at 700°C, scan rate 200mV/s, electrode surface area:
1.2346cm2; Red curve: CV scan in La3+/LaC2 FLiNaK at 700°C, scan rate:
K+/K 200mv/s, electrode surface K+/K area:1.2574cm2; (c) CP scan in FLiNaK-La2O3 at (C) 700°C, applied constant current 9mA, electrode surface area:1.0236cm2; WE: graphite, CE: graphite, RE: tungsten XRD analysis of graphite cathode surface powder sample Overall Reaction:
xLa2O3 + (3+4x)C = 3COx + 2xLaC2, where x=1 or 2 7
L iCl--KCl--E uCl3-SSeparation Eu m etal Eu (a) SEM results of Eu deposition on graphite in LiCl-KCl-EuCl3 at 550° Eu Deposition Test:
- CP scan, applied current I = -40mA, t = 2h
- WE: graphite, CE: graphite, RE: Ag/AgCl
- WE surface area: 1.0342 cm2 8
Galvanic Corrosion by Graphite
= ln 9
Galvanic Corrosion-Ni Corrosion by Graphite Deep cavities were observed on the surface. Salt were embedded into these cavities.
Nickel deposits along with the salt on the graphite specimen surface, the Ni deposits also showed presence of S. These were introduced from the nickel wire degradation due to galvanic corrosion.
10
Acknowledgement Funds: DOE and Industrial Team member: Dr. Jianbang Ge, Qiufeng Yang, Brendan Dsouza, Dr.
Shaoqiang Guo 11
Metallurgical Challenges associated with using Grade 91 steels at Elevated Temperature Jonathan Parker Senior Technical Executive Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 10th 2019 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
EPRI History of Materials Research Includes key collaborations with Energy Sector Stakeholders and Global Technology Transfer including involvement with International Conferences. For example 0th Conference: Chicago, IL (1987) 1st-London, UK (1995) 2nd-San Sebastian, Spain (1998) 3rd-Swansea, Wales (2001) 4th-Hilton Head, SC (2004) 5th-Marco Island, FL (2007) 6th-Santa Fe, NM (2010) 7th-Waikoloa, HI (2013) 8th-Albufeira, Portugal (2016) 9th - Nagasaki , Japan (2019)
EPRIs extensive experience in high temperature materials performance offers benefit to Advanced Nuclear applications, in general, and long term service issues, in particular 2 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Introduction Design codes for alloys used in high energy applications typically specify that the components fabricated will exhibit homogeneous composition, microstructure and properties. Experience has shown that these assumptions may not be valid.
This presentation highlights known problems associated with heterogeneity in as manufactured components and welds with particular reference to Grade 91 steel.
There is growing recognition that further work is required to understand the factors affecting variability and then to use this knowledge to underpin solutions.
Solutions may involve use of non traditional fabrication methods 3 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
EPRI Library of Information on 9%Cr Steels initiated by collaboration on Grade 91 Life Management in Fossil Plants Type II Code Case Grade 91 Metallurgical ASME Risk Factors Damage Grade 91 Weld Tolerance Repair Project Grade 91 Life New repair methods NBIC WS11 Management Project Grade 91 NDE Project Damage Step Weld Tolerance Grade 92 Life DOE-sponsored R&D Management Project on Gr. 92 Optimization Now over 100 reports on 9%Cr steels - comprehensive understanding linking Fabrication to Microstructure and Performance 4 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Components can exhibit significant microstructural variability.
Proper documentation of microstructure is NOT straightforward.
Heterogeneity linked to manufacturing issues, including: Typical martensitic structure
- Steel Composition,
- Steel Making,
- Segregation
- Hot Working method and conditions,
- Degree of Hot Reduction, Non-Typical Ferritic structure
- Heat Treatment History, such as:
- Normalizing Temperature,
- Normalizing Time,
- Cooling Rate from Normalizing, and
- Tempering temperature, time and controls see The Effect of Metallurgical Factors & Stress State on the Performance of High Energy Components Manufactured from Creep Strength Enhanced Steels Parker and Siefert, ECCC Conference 2017 5 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Imaging Methods Critical to Meaningful Characterization Etching followed by optical metallography reveals a Ion Beam imaging reveals the true non martensitic microstructure which appears martensitic structure and the precipitate distribution.
Reference f Loughborough hb h University 6 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Detailed Characterization reveals heterogeneity of composition in Grade 91 components 7 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Variable Creep for Smooth Bar Tests in Grade 91 Steel Parent Reduction of Area, %
Focused ion beam (FIB) milling allows detailed 3D characterization of individual creep voids which are not previously exposed 8 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
In-service Creep Cracks in Grade 91 Welds Expected Performance > 200,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Cracking at stub & attachment welds after about 50 000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />; Replacement at about 79 000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Operational temperature ~570°C in line with Design Component Stresses in line with Design S. J. Brett and J. D. Parker. Creep Performance of a Grade 91 Header. International Journal of Pressure Vessels and Piping 111 (12), 2013. pp. 82 to 88.
9 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Metallurgy of Grade 91base steel influences creep of weld HAZs, Creep Life of Welds changes by up to 30 depending on cavity susceptibility of base steel Life 1,685h Life 13,201h 10 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Cavities in Grade 91 steel linked to Inclusions:
Cu Concentration around MnS Al 3
1 S Cu S Cu Al Cu S 2
In Cavity Susceptible Steels Cu was frequently found around inclusions 11 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Summary of Observations Grade 91 steel (Creep Life Finite)
Low (very) Creep Strength typically linked to bad (very) Heat treatment Low Creep Ductility linked to hard particles which are present after steel making -
this are difficult to change by heat treatment Complexity of Problems increased by segregation (heterogeneous)
Low ductility failures associated with Factors which promote cavity nucleation and growth
- Metallurgical AND Stress State Lower bound heat affected zone life linked to a high density of voids, Upper bound heat affected zone life noted in cavity resistant steel Lower bound poor damage tolerance 12 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
ASME Code Case 2864, 2016 13 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Comparison of Current and New Allowable Stress Values 91 & 92 In 2009, the specified Allowable Stress values for Grade 92 steel were reduced to be 15.7 ksi, 12.0 ksi, 8.6 ksi and 5.6 ksi for use temperatures of 1050 oF, 1100 oF, 1150 oF and 1200 oF respectively. Allowable stresses of Grade 91 reduced in 2018.
130 Code Case 2179-7 (Pipe) 110 Allowable Stress (MPa)
Foulds Assessment, Grade 91 Steel 90 SA-335 P91, t < 3 inches 600°C, 1112°F 70
-15.5%
50 30 SA-335 P22
-29.5%
10 525 550 575 600 625 650 Temperature (qC) 14 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Advanced Reactors (AR)
EPRIs Advanced Reactors (AR) technical focus area primarily addresses the next generation of nuclear reactors (often referred to as Generation IV)
- Includes R&D relevant to light water SMRs and fusion technologies Objective is to build the technical foundation needed to ensure advanced reactors are real, deployable generation options when and at scale needed Program is evolving four years after launch
- Formal incorporation into the EPRI Advanced Nuclear Technology (ANT) program portfolio
- New AR Supplemental project now available, providing low-cost access to and engagement in EPRI AR research collaborative 15 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Scouting + R&D Portfolio Requirements and Strategic Analysis and Guidance Technology
- Align design with Assessment customer needs
- Economic analysis
- Leverage existing nuclear industry and market insights experience
- Safety-in-design
- Assemble and
- Technology and Scouting and disseminate best manufacturing practicess readiness Engagement Technology l
- Innovation network Development and
- Awareness and Transfer understanding
- Address common needs and gaps through collaboration
- Leverage and extend Advanced Materials EPRI core competencies and Manufacturing
- Technology transfer 16 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Powder Metallurgy-Hot Isostatic Pressing Why PM-HIP?
Near-net shaped Subsea Manifold.
components Courtesy: Sandvik Homogenous microstructure
- Ease of inspection!
40 diameter HIP Vessel Elimination of welds Courtesy: Isostatic Forge International 4-6 months lead times typical Ideal for multiple penetration applications (RPV or CNV head) vs expensive forgings Large Bore Valve NNS Reactor Coolant Pump (courtesy Roll-Royce) Impeller (courtesy Framatome 3600lb (1630kg) BWR and Albert & Duval) Nozzle 17 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Small Modular Reactor Upper Head--Example Approximately 44% scale Single monolithic structure Photographs courtesy of EPRI and NuScale Power A508 Class 1, Grade 3 27 penetrations 1650kg (3650lbs); 1270mm (50 inches) diameter Next, 2/3-scale head Need larger HIP Vessel -- ATLAS DOE Project: DE-NE0008629 18 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Summary of Benefits of Clean Homogeneous Steels Improved Properties include Reduced variability (Uncertainty) and Lower FATT, higher fracture toughness, higher upper shelf energy Better creep strength and ductility Higher yield strength Higher low cycle fatigue strength Greater resistance to SCC initiation Uniform radial and longitudinal properties These improved Properties should offer performance benefits such as:
Increased life under conventional service conditions, Increased critical crack size ( greater duration of stable crack growth)
Greater opportunity for weld repair, Reduced damage initiation sites provides a lower risk of cracking during Flexible Operation.
A simple take away is well made, clean steel components reduce variability in properties and provide the margin which aids Damage Tolerance 19 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
TogetherShaping the Future of Electricity 20 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Recent EPRI Position Papers / Theses Loughborough Uni Life Management of 9Cr Steels - Damage Tolerance Assessment of Header End Cap Geometries , EPRI, Palo Alto, CA: 2018 3002011049 Life Management of 9%Cr Steels - Damage Tolerance Assessment of Novel Step Weld Geometry for Girth Welds in Thick-section Components, EPRI, Palo Alto, CA: 2018 3002011053 Life Management of 9%Cr Steels - Damage Tolerance Assessment of a Common Hot Reheat Lateral Geometry, EPRI, Palo Alto, CA: 2018 3002011051 Xu Xu https://repository.lboro.ac.uk/articles/Microstructural_evolution_and_creep_damage_accu mulation_in_Grade_92_steel_weld_for_steam_pipe_applications/9230171 Gu https://repository.lboro.ac.uk/articles/Microstructural_investigation_of_creep_behaviour_i n_Grade_92_power_plant_steels/9230141 Siefert https://repository.lboro.ac.uk/articles/The_influence_of_the_parent_metal_condition_on_t he_cross-weld_creep_performance_in_Grade_91_steel/8309882 21 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Advanced dvanced ced d Structural Materials ffor Nonon-onn-Light Water Reactors Steven J. Zinkle1,2 1Dept. Nuclear Engineering and Dept. Materials Science & Engineering, University of Tennessee, Knoxville, TN, USA 2Oak Ridge National Laboratory, Oak Ridge, TN, USA Workshop on Advanced Non-Light Water Reactors: Materials and Component Integrity Nuclear Regulatory Commission December 9-11, 2019 1
Overview
High sink strength has been a long-standing scientific tenet for superior radiation resistance in structural alloys
Cold-worked and Ti-modified SS alloys (e.g., D9) developed by LMFBR program in the 1970s
Improved structural materials are needed for nuclear power to fully achieve its promise
High burnup, accident tolerant LWRs
Fusion and Gen IV reactors
Nanostructured alloys enable simultaneous achievement of radiation resistance and high performance (strength)
Radiation resistance (sink strength): S~4RN
Dispersed barrier hardening: y~b(2NR)1/2
For a given precipitate volume f=4R /3, best radiation resistance 3 sista and mechanical anic strength simultaneously occurs for high N, small R
High density of uniformly distributed nanoscale precipitates preferred 2
Thermal creep and void swelling in sodium-cooled fast reactor cladding is problematic for conventional steels
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ODS steels provide improved void swelling resistance compared to standard ferritic/martensitic steels
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Conclusions
Nanostructured (high sink strength) alloys are promising options for the structural materials in next-generation fission reactors
Enables simultaneous superior radiation resistance and superior mechanical property performance
ASME code qualification is needed to enable their deployment
Currently in boutique materials stage Timeline of structural materials used in light water reactors
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Near-term Importance of Advanced Steam
13 Advanced Steam Conditions Push Metallurgical Limits of Alloys in Use Today
5400/1300/1325/1325 Improvement in Energy Efficiency 4000/1165/1200 4000/1100/1150 R&D Do on ongoing ngoing 4000/1085/1100 (Ni Ni-based)
Market introduction by 3600/1050/1085 3480 psi / 1005oF / 1050oF Current Market Japan R&D Japa & ongoing o go g 2400/1005/1005 introduction (COST) in Europe Mature technology
Illustration: EPRI Data: Alstom 14
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Objective: qualify/develop advanced alloys to enable reliable, high efficiency operation of A-USC plants
Key Deliverables:
Generate alloy properties database for U.S. boiler manufacturers to enable component design and fabrication
Qualification of fabrication/welding techniques for use with specific alloys
ASME Boiler and Pressure Vessel Code case for Inconel 740
Evaluate environmental compatibilityy Demonstration header fabricated from advanced alloys qualified by the U.S. A-USC Consortium 15
Objective: Determine capabilities of new commercially-produced ODS alloys for application at temperatures up to 1200C in advanced fossil combustion and conversion processes
Key Deliverables
Data on the full range of properties required to qualify the alloy for fossil applications
Evaluate joining technologies
Feasibility of employing a less costly route for producing ODS alloy powders 25 mm diam x 4 m long ODS FeCrAl tubes used in British Gas/COST 501 1100C air heater demonstrator nstrator 16
Historical development of improved high-temperature steels has exhibited slow and steady progress Based on 17
WE START WITH YES.
CHALLENGES IN QUALIFYING ADVANCED MANUFACTURING TECHNOLOGIES FOR HIGH TEMPERATURE NUCLEAR SERVICE MARK MESSNER NRC Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop 10 December 2019 WHY WOULD YOU WANT TO USE AM TECHNOLOGIES IN ADVANCED REACTORS?
Complicated, small, difficult to conventionally manufacture parts:
- Pump impellers and casings
- Core internals
- Compact heat exchangers Bimaterial cladded components Maier et al. 2018 Replacement parts High performance, functionally graded Inconel 625 Westinghouse components Unique, difficult to manufacture 304L SS materials 2
Carroll et al. 2016
WHAT ARE THE CHALLENGES QUALIFYING AM MATERIALS IN GENERAL?
Creep at same Variability in AM material properties is much Wrought 316L conditions AM 316L greater than for conventional wrought/cast 300 m material - more akin to welds
- Less understood processes
- Many processing parameters controllable by users
- Wide variety of technologies
- Manufacturing likely to occur at a number of smaller sites, rather than at large, central production facilities AM methods often result in significant material property variations within a single build We want a process that can take advantage of the AM material good, bad, or just different?
flexibility of AM processes - not trying to simply 3D print conventional material AM creep specimens courtesy UW Madison 3
WHAT ARE THE CHALLENGES QUALIFYING MATERIALS FOR HIGH TEMPERATURE SERVICE?
At high temperatures long-term, time-dependent material properties control design:
- Creep strength and ductility
- Creep-fatigue life
- Thermal aging characteristics Short-term tests might tell you very little about Creep cavitation important long-term properties (INL)
Statistical variation in mechanical properties tends to be high, even for well-controlled traditional wrought material processes Seam pipe failure at coal power station Weld resilience can be challenging (Viswanathan and Stringer, 2000)
Very little long-term mechanical test data on AM material for properties relevant to high temperature design HRSG tube failure 4 (EPRI, 2005)
THE CLASSICAL QUALIFICATION PROCESS Get a material specification from the metallurgists - this ensures repeatability Procure several heats of material -
this tests material variability Long term mechanical testing (including creep, creep-fatigue, and thermal aging)
Extrapolate properties (factor of 3 typical)
For design, qualification is a contract:
Establish time-dependent design I promise you will have material properties out to desired/available properties better than xxx design life 5 AM QUALIFICATION, BASIC STANDARD:
REPEATABILITY Process, machine, and operator qualification Dont worry about properties (yet) just make sure builds are repeatable with adequate geometric tolerances Example of an existing standard: ASTM-ISO/ASTM52902-19
- Use test artifacts to assess machine capabilities and build repeatability Could use other properties to assess repeatability:
- Mechanical properties
- Microstructural measurements ASTM-ISO/ASTM52902-19
- Indirect signals from in-situ monitoring 6
PROPERTIES QUALIFICATION OPTION 1: WITNESS TESTING Materials with adequate short-term properties will meet long-term property requirements (?)
- 1. Fabricate part along with Short-term witness test samples
- 2. Measure mechanical properties of witness samples, compare to minimum specified properties
- 3. Assume long-term material properties will be comparable to reference material having the same short term properties -
Long-term Example from ASTM requires test program F42.01 Work Item WK49229 7
PROPERTIES QUALIFICATION OPTION 2: QUALIFY BY MICROSTRUCTURE Two materials with the same microstructure will have the same long-term properties Microstructure Characterization 1. Define a material class with definite microstructural characteristics (i.e. a fingerprint)
- 2. Develop qualified properties for that class via testing or modeling/simulation
- 3. Ensure that production material falls into the material class via Design Category in-situ or ex-situ high properties definition Fingerprint throughput characterization 8
ACCELERATING QUALIFICATION Several options to accelerate the traditional high temperature qualification process Year 0 1 3 9 Test Qualified life Wrought 316L Can we make AM 316L (as printed) Test these look the same? Qualified life And so on Emulate traditional wrought material Staggered testing schedules Hayhurst Log stress Huddleston Max principal Mises 0 5 10 Log time Use models to replace (some) tests Partially rely on in-situ monitoring 9
CONCLUSIONS AND QUESTIONS TO ANSWER Qualifying AM materials for high temperature nuclear service will have some unique challenges, but options are available. It would be best to start now with likely technologies and materials, given the reliance of high temperature design on long-term material properties.
- 1. What is going to be the first use of AM material in high temperature reactors?
What time frame? Replacement parts or new construction?
- 2. Given the answer to #1, what basic research will be required before we need to start the qualification process?
- 3. What properties/measurements should we use to assess components?
- 4. How much trust do we have in physically-based models for relevant properties?
How much (if any) testing could we forgo and replace by simulation?
- 5. Can vendors live with the risk of a staggered testing schedule and/or in-situ monitoring? 10
Qualification of Materials for Elevated Temperature Nuclear Components Nuclear Regulatory Commission Workshop Richard Wright on Materials for Non-Lightwater Reactors Idaho National Laboratory December 10, 2019 ASME Section III, Rules for Construction of Nuclear Facility Components - Division 5, High Temperature Reactors Many of the proposed applications for advanced manufactured materials and components are for service in the time dependent material property range
- ASME Section III Division 5 Scope
- Division 5 rules govern the construction of vessels, storage tanks, piping, pumps, valves, supports, core support structures and nonmetallic core components for use in high temperature reactor systems and their supporting systems Construction, as used here, is an all-inclusive term that includes material, design, fabrication, installation, examination, testing, overpressure protection, inspection, stamping, and certification High temperature reactors include gas-cooled reactors, liquid metal reactors and molten salt reactors (liquid or solid fuel) 2 energy.gov/ne
ASME Code Qualification of a New Material or Process
- Division 5, Appendix HBB-Y, Guidelines for Design Data Needs for New Materials
- Recently exercised these rules through DOE ART base program on the Alloy 617 Code Case in support of HTGR/VHTR applications Required testing to introduce a new structural material into Section III, Division 5, or a Division 5 Code Case
- HBB-Y-2100 Requirement For Time-independent
- HBB-Y-3500 Data Requirement for Cyclic Stress-Data Strain Curves
- HBB-Y-2110 Data Requirement for Tensile
- HBB-Y-3600 Data Requirement for Inelastic Reduction Factors for Aging Constitutive Model
- HBB-Y-2200 Requirement for Time-Dependent
- HBB-Y-3700 Data requirement for Huddleston Data multiaxial failure criterion
- HBB-Y-2300 Data Requirement for Weldments
- HBB-Y-3800 Data Requirement for Time-
- HBB-Y-3100 Data Requirement for Isochronous Temperature Limits for External Pressure Charts Stress-Strain Curves
- HBB-Y-4100 Data Requirement for Cold Forming
- HBB-Y-3200 Data Requirement for Relaxation Limits Strength
- Validation of Elastic-Perfectly Plastic (EPP)
- HBB-Y-3300 Data Requirement for Creep-Fatigue Simplified Design Methods for the new alloy
- HBB-Y-3400 Data Requirement for Creep-Fatigue of Weldments 3 energy.gov/ne Additional Requirements Outside the Scope of Section III, Division 5
- Design procedures and materials data not contained in Division 5 may be required to ensure the integrity or the continued functioning of the structural part during the specified service life
- Rules do not provide methods to evaluate deterioration that may occur in service as a result of corrosion, mass transfer phenomena, radiation effects, or other material instabilities
- Owner/operator has the responsibility to demonstrate to NRC that these effects are accounted for in the design of the component 4 energy.gov/ne
Example: Creep-Fatigue Behavior
- ASME approach is to develop a damage diagram based on a time-fraction method 5 energy.gov/ne Creep-Fatigue Example Results 6 energy.gov/ne
7 Why the Interest in Advanced Manufacturing?
- Conventional light water reactor technology is seeking manufacturing methods to reduce lead times, material costs and improve the ability to inspect components
- Vendors of small modular reactors are considering manufacturing methods that can be applied to reduce fabrication cost and enhance the economics of factory fabrication
- Small modular reactor concepts have complex geometries and unique needs for compact systems
- ASME Code issues have arisen with conventionally processed materials (rolled or forged) that are difficult or impossible to inspect due to directional microstructures that are not homogenized with modern mill practices 7 energy.gov/ne 8
The Potential Consequences of Advanced Manufacturing are Being Assessed 8 energy.gov/ne
9 NRC Pathways to Regulatory Acceptance There are several regulatory paths available to a licensee for utilizing an AMT in a nuclear application.
The following four regulatory paths are possible:
- Develop a Code or standard that can be incorporated by reference (IBR) in 10 CFR 50.55a.
- Select an unregulated in-service application.
- Submit generic technical reports or plant-specific submittals for NRC approval.
- Implement the 10 CFR Part 50.59, 10 CFR Part 70.72, or 10 CFR Part 72.48 process.
9 energy.gov/ne Example: Room Temperature Tensile Properties of Diffusion Bonded Alloy 617 10 energy.gov/ne
Creep Rupture of Conventional and Diffusion Bonded Material 11 energy.gov/ne Creep Fatigue Behavior of Diffusion Bonded Alloy 617 12 energy.gov/ne
13 How Can Qualification be Accelerated?
- Characterizing time dependent properties cannot be avoided - but extrapolation by a factor of 3 to 5 in time is reasonable
- One approach is to limit the initial design life, particularly for technology demonstration, and continue testing while demonstration is ongoing
- Cladding concepts may be applicable - e.g., a corrosion resistant layer could be added on a Code qualified material
- Determination of mechanisms giving rise to time dependent properties through simulation validated by experiment could allow acceptable accelerated testing or greater extrapolation 13 energy.gov/ne
Challenges with ASME Code Qualifying Graphite Irradiation Effects in Test Reactors Anne A. Campbell, Ph.D.
Oak Ridge National Laboratory Oak Ridge, TN campbellaa@ornl.gov US-NRC Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop, December 9-11 ORNL is managed by UT-Battelle, LLC for the US Department of Energy What does the ASME B&PV Code Say (1/3)?
- Irradiated materials properties (HHA-2220 and HHA-III-3300):
- Dimensional change
- Strength
- Elastic modulus
- Creep coefficient
- Coefficient of thermal expansion (temperature dependent)
- Thermal conductivity (temperature dependent)
- Damage dose and irradiation temperature shall cover qualification envelope
- Maximum temperature increments of 200ºC shall be used (HHA-III-3300)
- Temperature-dependent properties will be measured with a maximum temperature increments of 200ºC shall be used (HHA-III-3300)
- Measurements at temperatures below the envelope temperature are permissible due to the necessity to retain rather than anneal radiation damage (HHA-III-3300)
ASME III-5, 2017, "ASME Boiler and Pressure Vessel Code An 2 International Code, ASME International, New York, NY.
What does the ASME B&PV Code Say (2/3)?
- Irradiated material properties data shall be obtained from material that is representative of the material used for the generation of the irradiated material design data (HHA-III-4200)
- Designer is responsible for determination and justification of the representative data (HHA-III-4200)
- The historical data shall meet the requirements of this Appendix and be on the same graphite grade. If credit is taken from previous qualification programs, then testing shall also demonstrate that the historic data is applicable to the current production material. (HHA-III-5000)
ASME III-5, 2017, "ASME Boiler and Pressure Vessel Code An 3 International Code, ASME International, New York, NY.
What does the ASME B&PV Code Say (3/3)?
- Irradiation Fluence Limits (HHA-3142.1)
(a) < 0.001 dpa the effects of neutron irradiation are negligible and may be ignored (b) > 0.001 dpa the effect of neutron irradiation on thermal conductivity shall be taken into account.
(c) > 0.25 dpa all effects of neutron irradiation (described in HHA-2200) shall be considered and a viscoelastic analysis applied.
- HHA-3142.2 Stored (Wigner) Energy. Should graphite be irradiated to significant fluence [> 0.25 dpa] at a temperature < 200°C, the effect of stored (Wigner) energy buildup shall be accounted for when evaluating reactor thermal transients.
ASME III-5, 2017, "ASME Boiler and Pressure Vessel Code An 4 International Code, ASME International, New York, NY.
Graphite Cohesive Life Limit (HHA-3142.4)
- A temperature-dependent cohesive life limit fluence is to be defined for the graphite grade used for Graphite Core Components. Material that exceeds this fluence limit is considered to provide no contribution to the structural performance (stiffness and strength) of the Graphite Core Component. This fluence limit shall be set to the fluence at which the material experiences a +10% linear dimensional change in the with-grain direction. For full assessment (HHA-3230), this material shall not be included in the volume of the Graphite Core Component assessed.
ASME III-5, 2017, "ASME Boiler and Pressure Vessel Code An 5 International Code, ASME International, New York, NY.
ARTICLE HHA-II-4000 DETAILED REQUIREMENTS FOR DERIVATION OF THE MATERIAL DATA SHEET IRRADIATED MATERIAL PROPERTIES
- (a) The changes in these properties with neutron irradiation shall be determined 900 at sufficient intervals over the design fluence and temperature range.
Temperature (ºC)
Temperature and fluence intervals shall be selected so as to provide adequate confidence in the accuracy of the interpolations. Limited extrapolations are 750 permitted, but such extrapolations shall be justified.
- (b) Property changes (such as strength, thermal conductivity, the linear 600 coefficient of thermal expansion, and the elastic modulus) are to be reported in the form of the relative change with regard to value before irradiation as a 450 function of the fluence over the relevant operating temperature range. When reporting fractional change of these properties, for near-isotropic graphite 300 grades, distinction with respect to the preferred grain orientation is not required.
Fluence
- (c) The creep coefficient (creep model parameters) is to be given for those temperatures that occur in the irradiated components, in accordance with HHA-3142.1(c).
- (d) The effect of creep strain on coefficient of thermal expansion and elastic modulus shall be determined and included in the Material Data Sheet.
- (e) The irradiation-induced relative linear dimensional changes are to be given for the design service temperature range for the irradiated components. The relative linear dimensional changes are to be given separately for both grain orientations. For isotropic graphite, where the difference between dimensional change curves is insignificant, only one curve is necessary.
ASME III-5, 2017, "ASME Boiler and Pressure Vessel Code An 6 International Code, ASME International, New York, NY.
MANDATORY APPENDIX HHA-III REQUIREMENTS FOR GENERATION OF DESIGN DATA FOR GRAPHITE GRADES
- It is the policy of the ASME Boiler and Pressure Vessel Committee to adopt for inclusion only such specifications adopted by the American Society for Testing and Materials (ASTM) and by other recognized national or international organizations. Mandatory Appendix HHA-I contains two ASTM material specifications for nuclear grade graphites. (HHA-III-1000)
- Material Specifications The following material specifications are accepted: (HHA-I-1110)
- (a) ASTM D7219-08: Standard Specification for Isotropic and Near-isotropic Nuclear Graphites
- (b) ASTM D7301-08: Standard Specification for Nuclear Graphite Suitable for Components Subjected to Low Neutron Irradiation Dose ASME III-5, 2017, "ASME Boiler and Pressure Vessel Code An 7 International Code, ASME International, New York, NY.
HHA-III-3000 Properties to be Determined
- The Designer shall record the product cutting and sampling plan from which the specimens and various properties were taken, as well as all supporting test records. The heterogeneity of the graphite and statistical nature of the data shall be considered. Where property means and standard deviations are reported, the size of the sample population shall be reported as well.
- (a) The test standards specified in ASTM C781 shall be used.
- (b) Where a test standard does not exist or has to be customized for application, the test procedure used for the completion of the tests shall be filed with the material test data.
ASME III-5, 2017, "ASME Boiler and Pressure Vessel Code An 8 International Code, ASME International, New York, NY.
ASTM Standards
- Bulk Density (ASTM C559-16) - mass of specimen >2000x
- Elastic properties (ASTM C747-93, ASTM C769-15, C1198-09 balance sensitivity, Volume>500 mm3 ceramics)
- Minimum dimension > 10 largest particle, and >2000x - Impulse/Resonance frequency - Obtain representative measurement device resolution specimens, slender rod geometry (5 < L/T < 20)
- Flexural Strength - Sonic Velocity (ASTM C769-15) Approximate value,
- 4-point flexural strength (ASTM C651-15), minimum dimension specimen size dependent on grain size and wavelength (t) > 5x particle size, length > 8t, width (w) t < w < 2t
- Wavelength longer than grain size (2.25 MHz frequency has =1.1 mm), diameter >5, length sufficiently long to include more than a
- Support span minimum length 40 mm (and 3x longer than load few grains span), load span > 2t
- Specimen overhang support span by at least 1t at each end
- Thermal conductivity (ASTM E1461-13), thin disks (6-30 (specimen length > 40 +2t) mm diameter), thickness to get thermal half-rise time 10-
- 3-point flexural strength (ASTM D7971-14) because of the 1000 ms (in graphite 2.8-11 mm pre-irradiation and 0.5-5 small volume under stress so four-point flexure is preferred mm post-irradiation) and recommended for most characterization purposes
- Coefficient of thermal expansion (ASTM E228-17), length
- Minimum dimension (t) > 5x grain size, L/t > 6, and 1 < W/t < 2 25-60 mm, 5-10 mm diameter (or equivalent)
- Compressive strength (ASTM C695-15), diameter (d) > 10x
- Cross section length > 5x grain size (no smaller than 4 mm),
particle size, length (l) 1.9d < l < 2.1d (standard calls out length > 25 mm (50-150 preferred) (ASTM C781-18) recommended minimum of 9.5 mm x 19 mm)
- Irradiation creep - no ASTM standard, safe to assume
- Tensile strength (ASTM C749-15), gauge diameter no specimen dimensions would have similar requirements as smaller than 3-5x maximum particle size, specifies compression or tensile/flexural specimens specimen shapes and dimensions, at least 50% of specimens should fracture in gauge region
- Older version of ASTM C781-08 had annex about using glued-end rods (removed in the newest version of standard, still in ASTM D7775-16), diameter 6.5 mm, length 26 mm 9
Minimum Specimen Dimensions per ASTM Standards Grade Grain Density 4F 3F C T ThD CTE EIR ESV Size (mm)
NBG-18 1.6 163 (4096 mm3) 8x8x64 8x8x48 16x32 4.8 6x6x3 8x25 8x40 12x16a PCEA 0.8 83 (512 mm3) 4x4x32 4x4x24 8x16 0.8 6x6x3 4x25 4x20 5.5x15 IG-110 0.05 23 (8 mm3) 0.25x0.25x2 0.25x0.25x2 0.5x1 0.15 6x6x3 0.25x25 0.25x1.25 5.5x15 a May require lower frequency since 2.25 MHz =1.1 mm (1.0 MHz =2.5 mm)
- Specimen sizes conforming to ASTM standards are typically larger than those that can fit in capsules for irradiation testing
- At ORNL, irradiations in the HFIR flux trap (~6-10 dpa per year) are usually limited to capsule irradiations, with internal space for specimens limited to 6 x 6 x 48 mm
- Longer irradiation time, higher costs (instrumented), still have size limitations 10
How to use specimens that fit in irradiation capsules?
- ASTM D7775-16 covers the measurement of properties on specimens that are smaller than standard recommendations
- responsibility of the user to demonstrate that the application of a standard outside any specified constraints is valid and reasonably provides properties of the bulk material from which the nonstandard specimen was extracted
- ORNL advises Size Effect Studies to quantify properties measured on specimens that conform at ASTM sizes and sub-sized specimens ASTM D7775-16, 2016, "Standard Guide for Measurements on Small Graphite Specimens", ASTM International, West Conshohocken, PA, DOI:
11 10.1520/D7775-16, www.astm.org.
Major Challenges for Qualification Programs
- Need multiple temperature/fluence combination (10-30)
- Need replicate measurements of properties on multiple specimens (is triplicate sufficient?)
- Representative specimen sizes limits number of specimens that can be irradiated in a single capsule/zone
- Multiple capsules/zones per temperature/fluence combination
- This results in 10s-100s of irradiation capsules in the ORNL HFIR to complete a full irradiation campaign
- Cant complete irradiation of all specimens at one time 12
Challenges with ASME Code Qualifying Graphite Irradiation Effects in Test Reactors Anne A. Campbell, Ph.D.
Oak Ridge National Laboratory Oak Ridge, TN campbellaa@ornl.gov US-NRC Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop, December 9-11 ORNL is managed by UT-Battelle, LLC for the US Department of Energy
POTENTIAL MATERIAL ISSUES FOR CNSC TO LICENSE ADVANCED REACTORS X. Wei and T. Nitheanandan e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019
1 Background
The Canadian Nuclear Safety Commission (CNSC) is performing a pre-licensing review of several advanced reactor designs with various reactor coolants, e.g. high temperature gas, liquid metals and molten salts The focus of the review is to provide early identification of:
- potential regulatory and technical issues in licensing a vendors design
- additional regulatory research that may be needed to inform regulatory requirements in specific cases This presentation discusses some potential material issues identified/anticipated based on available information to date e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 -2
2 Scope The majority of the vendors propose to use ASME Section III Division 5 High Temperature Reactors (Division 5 thereafter) for advanced reactor designs HAA-1130 of Division 5 states that the rules do not cover deterioration that may occur in service as a result of radiation effects, corrosion, erosion, thermal embrittlement, or instability of the material The following slides mainly focus on these material issues that are not currently covered by codes and standards e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 -3 3 Potential Material Issues 3.1 Neutron irradiation embrittlement 3.2 Effect of irradiation on safety margin of using code limits 3.3 Corrosion 3.4 Coating and cladding 3.5 Mechanical connection 3.6 Material development and qualification e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 -4
3.1 Neutron Irradiation Embrittlement Neutron irradiation embrittlement is a key issue for reactor safety operation.
The existing test data or operation experience may not be applicable for advanced reactors due to the differences in:
- material selection
- service temperature
- neutron flux / accumulated fluence
- interaction of corrosion and neutron irradiation Information/justification is needed to address this non-code issue in the future licence review.
e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 -5 3.2 Effect of Irradiation on Safety Margin of Using Code Limits Neutron irradiation affects code limits for load-controlled stresses. For example, the temperature & time-dependent stress intensity limit, St, is defined as the lesser of:
(a) 100% of the average stress required to obtain a total strain of 1%;
(b) 80% of the minimum stress to cause initiation of tertiary creep; (c) 67% of the minimum stress to cause rupture St values could be lower under neutron irradiation Neutron irradiation also affects other code limits, such as Smt, So, and limits/values for use-fraction sum analysis associated with the general primary membrane stresses during Level A, B and C Loading e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 -6
3.2 Effect of Irradiation on Safety Margin of Using Code Limits (continued)
Neutron irradiation also affects the code limits on deformation-controlled quantities Irradiation affects isochronous stress-strain curves, stress-to-rupture curves, and therefore affects some key limits for creep and fatigue analysis, for example:
- Sa in HBB-T-1322 Test No. A-1
- tid and i in HBB-T-1324 Test No. A-3
- in HBB-T-1333 Test No. B-3 (Equation 5)
Vendors will be requested to verify whether there is sufficient safety margin to use Division 5 limits for components under neutron irradiation e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 -7 3.3 Corrosion Corrosion degradations in advanced reactors could be very different depending on the combination of material, environment (coolant, temperature), and stress:
- general corrosion (molten salt reactors, liquid metal reactors)
- galvanic corrosion (molten salt reactors)
- wear and friction
- flow-accelerated corrosion and erosion corrosion (e.g.
heat exchange)
- environmental embrittlement (e.g. carburization and nitridation in gas-cooled reactor) e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 -8
3.3 Corrosion (continued)
Under each material/environment/stress combination, the degradation mechanism or corrosion rate could be different. The understanding and prediction of the corrosion effect on structure integrity will be reviewed with a consideration of specific reactor design and operating conditions.
e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 -9 3.3 Corrosion (continued)
Corrosion degradation could be accelerated significantly by other factors, such as:
- neutron irradiation
- impurities (e.g. H2O, O, S in molten salt; H2O, O in liquid metals; H2O, O2 and N2 in high temperature gas)
- redox potential (Maintaining mildly reducing condition is key to avoiding significant corrosion and tellurium cracking in molten salt)
The understanding on irradiation-enhanced corrosion will be reviewed. A Design Manual or Guide for chemistry or impurity control in each advanced reactor should be established. Technology for chemistry control should be demonstrated at an industrial scale.
e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 - 10
3.4 Coating and Cladding Vendors may use cladding or coating (such as Fe/Al, Ni, carbides, nitrides, borides, phosphides, and refractory coatings) to improve corrosion resistance However, coating affects heat transfer capability. Thermal cycling can cause coatings to crack or delaminate. Thin coatings reduce the tendency of cracking or delamination but are more vulnerable to imperfections; radiation-enhanced-intermixing could also be a significant issue for thin coatings Qualification or justification on the applicability of coating or cladding to advanced reactors will be assessed.
e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 - 11 3.5 Mechanical Connection Chemical compatibility of welds need to be investigated:
- class A materials can be sensitized during welding and become less resistant to corrosion
- brazing would cause less damage to the base metal microstructure, but is not addressed in nuclear portion of ASME BPVC Gaskets may be challenging due to the tendency to develop leaks over time:
- gasket degradation (e.g. nitriding/carburizing in high temperature gas, corrosion in molten salt or liquid metal)
- bolt creep
- sealing-surface deformation e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 - 12
3.6 Material Development & Qualification Material qualification and design data validation is important for advanced reactor design:
- new materials may be developed for meeting design requirements
- existing materials may be modified to improve component integrity
- materials with a lower code class (or not in code) may be used for fabrication of higher class components
- structural ceramic composites remain at a low technology readiness level e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 - 13 3.6 Material Development
& Qualification (continued)
Material properties and design data for long-term operation need to be verified and validated Question: should we allow design to use 100% of non-code stress/deformation limits?
e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 - 14
4 Conclusion Detailed information or justification may be required to address the non-code issues relevant to materials and component integrity in the stage of licence review of advanced reactors e-Docs# 6025322 Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 - 15 Thank you! Questions?
nuclearsafety.gc.ca
Jeffery Aguiar, Allen Roach, and Matthew Kerr Nuclear Science and Technology Division Idaho National Laboratory Email: Jeffery.Aguiar@inl.gov and Robert.Roach@inl.gov Work supported in part through the INL Laboratory Directed Research& Development (LDRD)
Program under DOE Idaho Operations Office Contract DE-AC07-05ID145142.
Apollo Moon Landing Supersonic X1 Flight USS Nautilus EBR-I The first step is to establish that something is possible; then probability will occur.
- Elon Musk EBR-I
Innovation equates more than the cumulative sum of the individual parts and spurs growth.
Data Analytics Advanced Manufacturing Nuclear clear Materials and Fu Fuels Physics Based Modeling Irradiation Testing Computational omputational Nuclea Nuclear High Throughput Materials Design Characterization 3
Where have others been successful?
- QuesTek Innovations, LLC. a successful outcome of the Materials Genome initiative.
- Develop and qualify a high strength, lightweight steel for the hook shank for the T-45 training aircraft where certification was required.
- Design to deployment in 8 years using the materials genome g pp approach.
Dave Da ve Snyder, Sny nyde der, r, C Chris hris hr is Kern, Ker ern, n, JJeff efff Gr ef Grab Grabowski abow owsk skii C. J. Kuehmann and G. B. Olson, Material Science and Technology, 2009, 25 (4). Aeromat, A t AAprilil 12 12, 2017 2017, Charleston, Ch l t South S th Carolina C li
Current nuclear materials (e.g. alloys) development cycle Design, Fabrication, and Post Irradiation Characterization Exam and ~10 (3 to 4 years)
Performance years Assessment per (3 to 5 years) cycle Irradiation (3 to 5+ years)
~20 to 30 year development cycle 1 + 2 + 3 = Example: TRISO, HPRR, Metal fuel, advanced claddings, Optimized Approach How fast can we go?
Physics Based Modeling Data Analytics Nuclear Materials Testing Advanced Manufacturing
Data Analytics & Modeling:
Enabling Materials Discovery and Optimization Adaptive Engineering Huang, Jen-Chiang, Scanning, 2012, 34, 325-31.
T. R. Munter et. al, Comput. Sci. Discov, 2009, 2, 015006. Z. Li, K. Pradeep, Y. Deng, D. Raabe, and C.C. Nature, 2016, 534, 227-230.
B. Gludovatz,..., and R. O. Ritchie, Science, 2014, 345, 1153-1158.
Data Analytics:
Capability for streaming higher throughput materials characterization INL is advancing adv ng the use of deep learning and open materials to characterize and predict crystal structure from datasets in milliseconds. milliseconds.
Aguiar et al. in press, Science Advances, 2019. Funded by INLs LDRD office.
NMDQi takes a Grand Challenge g approach to accelerate development and qualification of new nuclear materials ls and fuels for future advanced ad reactor technologies s.
Enabling Technologies and Capabilities:
- Physics-based M&S for materials discovery and optimization.
- Data analytics for machine and deep learning.
- High-throughput material fabrication/characterization applying advanced manufacturing principles.
- Nuclear material testing over a wide range of conditions, including accelerated irradiation testing.
Discovery:
Supports new materials and pipeline development for high-throughput.
- Targeting discovery of new alloy y classes that are high strength, low cross section, and stable claddings above 400 °C.
- Incorporating modeling and experimental results in the samee workflows to enable training, evaluation, and deployment.
- Establishing combined modeling and experimental frameworks to shorten the nuclear materials development For more information, see poster by Danielle Beatty and Marcus Parry and research cycle.
Discovery:
Alloy Composition - Property Relationships Regression to Predict Elastic Modulus
- Training & evaluation dataset: t:
30,000+ alloys containing varying concentrations
- Down select to 130 alloys for 5 Element System fabrication and testing 4 Element System Fe3x Ni Element Cr CoSystemCuv Fex Niyy Crrzz Cow Fex Niy Crz alloys w
Material Descriptors Ranked by Importance Shows above a 90% prediction rate Optimization:
Deploys state of the art design of experiment (DOE) modeling capability to structural materials and fuel development.
Hypothesis: Addition of minor metal additives to 316L stainless steel promotes sluggish diffusion kinetics mitigating irradiation assisted stress corrosion cracking (IASCC).
Model-based Design & Assessment. Materials Library Data & Validation. Testing Assessment.
Maximize sluggish diffusion rates over Compositionally-graded samples Proton irradiation irradiation flux and temperature. explore and validate hypothesis.
Stress Corrosion Test P. Tsai et al. Acta Materialia, 2016, 120, 426-
T. Schuler et al., Phys. Rev. B,, 2017, 95, 174102.
For more information: Daniel.Schwen@inl.gov 12
Utilization:
Makes use of new and existing sources of data to streamline model development in support of qualification. IFR Materials Information For more information:
Doug.Porter@inl.gov System (IMIS) database for the legacy Sodium Fast Reactor (SFR) U-xPu-Zr (0 < x < 28 wgt%),
stainless steel cladding.
Link to multi-physics modeling and simulation.
Data format must be easily usable by modelers and fuel qualification safety case composers.
INL L is advancing the use of data for model qualification to support port ortt accelerate accelerated licensing by utilizing unique experimental data from the EBR R-II
- reactor.
Qualification: Accelerated fuel testing platforms Fission Accelerated Steady-state Testing example Revised Capsule Design Objectives:
- 1) Increase power density to reduce time to achieve high burnup Fabrication trials for 1/2- and 1/3-scale fuel
- 2) Decrease pin diameter to keep peak fuel temperature constant and rodlets is
- 3) Reduce sensitivity to fabrication tolerances and capsule/pin eccentricity underway Semi-integral Irradiation Test Zone 2.5 mm (0.100-in)
Picture of HT9 D123 0.75 HT9 93 mm Fuel Alloy 1/2/3/
Liner Clad O thi d di One-third diameter t pins i couldld achieve
>5% burnup in ATR 55-day cycle and Helium achieve 30% burnup in less than 2 7.2 mm Sodium years.
(0.284-in)
Stainless Steel For more information: For more information, see poster by Geoffrey Beausoleil Geoffrey.Beausoleil@inl.gov
Collaboration Collaboration iis sEEssential ssential
- Materials Genome Initiative (MGI)-NIST:
Internal Working Groups (IWGs) - Pioneers in advancing the agenda on materials informatics for discovery, optimization, and qualification.
- Data analytics
- Physics cs-s-based modeling
- Center for Hierarchical Materials Design (CHiMaD)-
Northwestern University:
- Advanced manufacturing - Outlining processing, structure, and properties for nuclear materials.
- Nuclear Materials anand Fuels Irradiation Testing
- Materials Project (MP): Coordinate on prospective and needs for accessing nuclear materials relevant modeling data.
- University partnerships: Growing with the NSUF CINR call.
- National Laboratories.
15 Intersecting Opportunities Qualification 3/4 Requires a frameworks fra focus on comprehensive approaches to licensing using qualified model development, qualification by performance and/or code qualification for new suggested fabrication and testing methodologies.
Advanced Reactors 3/4 Demonstrate early on new advanced materials and supply chains that expand on U.S nuclear innovation leveraging advances in high-throughput materials and data frameworks.
Fleet Sustainment 3/4 Develop materials and applied technologies that enables sustainment and demonstrate new methods for materials deployment.
Thank You!
Further Questions?
Jeffery.Aguiar@inl.gov Robert.Roach@inl.gov
Development of Technical Basis for In-situ, Passive Surrogate Materials Surveillance for Advanced Non-LWRs Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop Sam Sham December 9-11, 2019 ART Technology Area Lead, Advanced Materials U.S. Nuclear Regulatory Commission Argonne National Laboratory Rockville, MD Acknowledgment Contributors from ANL
- Mark Messner, Yoichi Momozaki, Edward Boron 2 energy.gov/ne
=
Background===
- The technology maturity of Molten Salt Reactor (MSR) systems is substantially less than other advanced non-LWRs such as GCRs and SFRs
- Information on materials degradations due to irradiation, corrosion, elevated temperature exposure and stress (creep-fatigue loading) during MSR operations is limited
- The establishment of an in-situ, passive surrogate materials surveillance program that would allow the collection of information on these materials degradations would be an important pathway in support of timely licensing MSRs 3 energy.gov/ne Current Practice on Materials Surveillance for Advanced Non-LWRs (1/3)
- ASTM E531 - Standard Practice for Surveillance Testing of High-Temperature Nuclear Component Materials
- Originally approved in 1975; current edition approved in 2013; there is current effort for re-approval
- Practice is used when nuclear reactor component materials are monitored by specimen testing
- Covers procedures for periodic specimen testing performed through the service life of the components to assess changes in selected metallic material properties that are caused by neutron irradiation and thermal effects 4 energy.gov/ne
Current Practice on Materials Surveillance for Advanced Non-LWRs (2/3)
- Test specimens covered include
- Tensile, creep/stress rupture, low-cycle fatigue/creep-fatigue, swelling, and Charpy impact
- Provides guidance on how to place surveillance samples to obtain the desired irradiation conditions
- Temperature, neutron flux, and neutron spectrum
- Acceptance criteria are not provided in the ASTM E531 Standard Practice 5 energy.gov/ne Current Practice on Materials Surveillance for Advanced Non-LWRs (3/3)
- ASTM E531 standard practice is based on the following strategy
- Know initial structural design properties before reactor operations
- In-reactor test specimens exposed to neutron and temperature (and coolant) during reactor operations
- Test specimens removed from reactor per surveillance program schedule for mechanical properties testing
- Testing for time dependent properties of the exposed materials tends to be short 6 energy.gov/ne
Materials Degradations During Operations of Advanced Non-LWRs
- Effects of materials degradations on structural materials during Stress &
Temperature reactor operations are integral (Creep-Fatigue, effects Aging)
- Would be advantageous to have additional test specimens/articles Advanced Addvan dvannce n ed e
to capture these integral effects Non on-on n-LWRs to complement the current test Irradiation C Coolant specimens selection in ASTM Effects Effects E531 Standard Practice 7 energy.gov/ne In-situ, Passive Test Article to Capture Integral Effects of Materials Degradations
- DOE-NE ART Program has developed a test article, called SMT (Simplified Model Test), that models the key features of structural interaction in prototypical reactor components, the so-called elastic follow up effects
- The advancement in the SMT testing methodology allows the generation of structural data in air at high temperatures using standard specimen testing setup, without the expensive scaled component testing
- Goal is to develop an in-situ, passive test article based on the SMT concept to capture the integral effects of irradiation, temperature, coolant and stress 8 energy.gov/ne
Use Differential Thermal Expansion to Drive Test Article During Reactor Operations Cross section of axi- Fabricated a proof-of-concept test article using symmetric test article 316H (test material) and A617 (driver)
- 1) Stir- 3) E-beam friction weld weld outer inner bars casing to together inner bars
- 2) Machine 4) Attach inner bars strain gauges 9 energy.gov/ne 10 Strain-Time Results for Furnace-Tested 316H/A617 Proof-of-Concept Test Article from Temperature cycling Ch8, Tag=Reference gauge 1 3500 Ch9, Tag=Reference gauge 2 Slight 3000 Ch 4, Tag=Test sample 2B ratcheting Ch 5, Tag=Test sample 2A 2500 Ch 6, Tag=Test sample 1B strain Strain Strain (x10-6) 2000 Ch 7, Tag=Test sample 1A strain Total strain 1500 Mechanical strain 1000 Thermal strain 500 0
0 100 200 300 400 500 600 700 800
-500 Temperature (°C)
Temperature (ºC)
- 100ºC temperature change during steady-cycle
- Test article only has thermocouple and strain gauge - cannot measure stress
- Two measurements of strain (two strain gauges on reference bar, two strain gauges on sample) 10 energy.gov/ne
Summary of Progress to Date
- Developed and fabricated a proof-of-concept in-situ passively loaded test article
- Trial test by temperature cycling showed that:
- The fabrication process was adequate
- Tensile strain could be produced using a driver material with a lower CTE than the test material
- Strains could be predicted using simple structural models 11 energy.gov/ne Near Term Tasks
- Develop prediction tool to select test article geometry and driver material based on
- Plant temperature transient parameters
- Desired strain range and elastic follow-up factor
- Test and driver material thermal and mechanical properties
- Fabricate new test articles based on prediction tool results
- Conduct thermal cycling tests (with hold time)
- Explore if there are issues in fabricating smaller test articles that can conform to the space limitation of an operating reactor 12 energy.gov/ne
Future Issues to Be Addressed
- What methodologies can be developed to leverage the availability of these test articles that capture plant-specific materials degradations due to irradiation, temperature, corrosion, cyclic loading, and structural interaction effect
- What types of data can be extracted from these test articles?
- Can we use these test articles as canaries?
- How to set acceptance criteria
- Can we use them to gather material data in a test reactor, e.g., VTR?
- We dont have means to measure stress and strain at this time
- Other questions 13 energy.gov/ne Thank you Contact information Sam Sham Argonne National Laboratory Applied Materials Division Email: ssham@anl.gov Tel: 630-252-7873 14 energy.gov/ne
PNNL-SA-149883 Progress Toward Bridging Harsh Environment Online Monitoring Gaps for Advanced Reactors December 2019 S. W. Glass Presentation Outline
- Why advanced reactors?
- Designs, operating history, and gaps
- PNNL initiatives Sensors for High Temperature & Radiation Cold Spray Magnetostrictive High-Temperature Technology Optical Sensors for Imaging, Motion, Flow, Material Identification, High-Temperature On-line Heat Exchanger Tube Monitoring Sensor Website and Database Code Involvement
- Conclusions 2
Why Advanced Reactors? Its the MONEY!
From EIRP; WHAT WILL ADVANCED NUCLEAR POWER PLANTS COST? A Standardized Cost Analysis of Advanced Nuclear Technologies in Commercial Development (2018) 3 First Generation of MSRs Plan to Rely on Known Component Technology (from ORNL Module 2: Overview of MSR) Technology and Concepts
- Pumps
- Vertical shaft, cantilever-style similar to those used with sodium fast reactors
- May require pressurization of fuel system to avoid pump cavitation
- Could be coupled with spray ring to evolve fission gases and tritium
- Heat exchangers
- Tube and shell remains leading candidate technology
- Tube vibration and flow-accelerated corrosion present the most significant power density limits
- Double wall possible/considered for tritium release mitigation
- Many designs have dual-stage to isolate high-pressure steam from fuel
- Vessel and piping
- Either clad ASME BPVC code-qualified material or modified Alloy N used under a limited-term code case
- Interior shielding to minimize radiation damage is planned by multiple vendors 5
Operating Experience Related to Materials and Component Integrity for Advanced Non-LWR Rxs
- MSR experience is limited. ORNLs MSRE Summary of NRC TLR-RES/DE/CIB-1965-69 used Hastelloy N but concluded 2019-01 Advanced Non-Light-Water mechanical properties...are not sufficient Reactors Material and Operational for long-term operation Experience (SFRs only)
- The report focused on SFRs profiting from decades of OE. Distribution of events are summarized in chart.
- Recommendation: Accurate detection methods of corrosion and leaks are necessary The design phase should consider sensor placement and reliability under operating conditions.
6 Typical layout of SFR (Super Phenix [France], BN-600/800
[Russia], FFTR [US], Monju
[Japan], EFR [China], PRISM
[US];TWR [US]
- Na - Na Primary Heat-X loop inside shielding vessel - <2 atm pressure
- Pump on Heat-X cold leg
- Na-Water/Steam on secondary loop
- Pump on Heat-X cold leg
- 360-600oC 7
Hydromine Lead Reactor and Single-Stage Spiral Heat-X 8
Kairos Power TRISO Fuel-loaded MSR 0.5-0.7 mm dia.
9
Flibe Dissolved Fuel MSR with Hot Gas Heat-X and Turbine 10 Reactor Operating Parameter Comparison 11
PNNL Initiatives
- High-Temperature Cold Spray Magnetostrictive EMAT Guided-wave flaw detection Non-invasive flow meter
- Acoustic Emission
- High-Temperature Piezoelectrics for On-line Heat Exchanger Monitoring
- Multi-Modal Optical Sensors
- Multi-Modal SAW Sensors (NEET)
- Molten Salt Chemistry/Corrosion Science Focus
- Advanced Rx Sensor Website/Database
- ASME Code Involvement 12 Magnetostrictive Cold Spray Sensor Sends SH-0 Waves Perpendicular to Magnet N-S Axis and Lamb Waves Parallel to Magnet N-S Axis 13
Conceptual Representation of a Cold Spray System 14 Typical Field-portable Manual Application System Image courtesy of VRC Metal Systems 15
Candidate Materials for MS Sensors Material MS Coeff. Curie Temp. Notes (mostly related to cold spray)
(3/2) s u 10-6 Nickel (Ni) 50 354qC Corrosion-resistant; commonly cold-sprayed Cobalt (Co) 93 1120qC Can be cold-sprayed; may need environmental control Iron (Fe) 14 770qC Can be cold-sprayed; corrosion-susceptible Ferrous Cobalt 87 500qC Standard for adhesive strip alloyed 50%/50%
(FeCo)
Galfenol >200 670qC No cold-spray history Met-Glass 60 370qC No cold-spray history but similar to ceramics that have been sprayed 16 5RXQG/DPE:DYH5HVSRQVHRQó663ODWH
FeCo Strip Compared to CPNi Patch CPNi 1/4 in. S.S. plate with coil over FeCo 1/4 in. S.S. plate; 0.5 mm CPNi patch strip and stationary permanent magnet and stationary permanent magnet 17
SH-0 Response to Various Cold-Spray Patch Configurations Relative to FeCo Strip 18 The MS Sensor in the Lamb Wave Mode Could Be Extended to Measure Flow, Inclusions, and Temperature 19
Alternatively Piezoelectric Sensors Can Be Evaluated for Flow, Temperature, and Reflections of Inclusions 20 Acknowledgments Department of Energys
- Magnetostrictive EMAT is an established way to Technology perform guided-wave inspections. Commercialization Fund.
- BUT the current state of the art uses adhesives or pressure coupling that are difficult for many advanced reactor configurations.
- Magnetostrictive Ni tests indicate that the cold-spray patch sensors can be used for SH-0 and/or Lamb-wave inspections. Other materials may work even better.
- Thanks to a TCF award and a follow-up DOE SBIR, this technology has been licensed and is being exploited commercially.
21
Typical power plant tube bundle shell-and-tube heat exchanger configuration 22 Primary ID sensors would be subject to high flow forces.
Secondary OD sensors are protected from high flow forces and also are held at lower temperatures than the primary tube ID fluid.
23
System In-Situ In n-Situ 24 Progress Toward Feasibility Assessment of On-line Heat Exchanger Tube Monitor Torsional SH-0 Wave PZT Low-Temperature 50% Pit (6 mm Sensor dia.) @ 95 cm 50% Notch198 cm from Sensor from sensor 6 mm dia. Hole Tube End228 cm 150 cm from sensor from transducer
Initial Target for On-line Heat-X Monitoring in Molten Lead Reactor 26 Ultrasonic Power & Communication - Virtual Penetrations Sensors (temperature, radiation, UT, ?)
Electronics with Ultrasound UT/electrical transducer charging capacitor and communication Vessel wall Electronics with UT and communication Power only one side 27
Upper Regions of a Liquid-Cooled (left) and Gas-Cooled (right)
Reactor, Showing Optical Access Concepts.
- Optical access through the containment vessel also may be required.
28 Basic Optical Fiber Sensor, Showing How the Incident Light is Separated into Reflected and Transmitted Spectra 29
Heterodyne Laser Vibrometer 30 Chemical Characterization: U(III)
Optical Spectroscopy Pu(III)
Pu(IV) U(IV)
- Provides chemical information U(VI)
Identification and quantification Oxidation state 9 Essential information for control of systems Pu(III)
Molecular and elemental species 9 Essential information to understand/control Pu(IV) separation efficiency or general system 1.2 Pu4+ Pu(NO3)62-behavior 1 Np
- Fast 0.8 Pu(NO3)3+ Pu(NO3)4
- Robust absorbance 0.6 Pu(NO3)22+
Cr 0.4
- Versatile 0.2 Lns 0
450 500 550 600 650 700 750 800 850 wavelength, nm 31
Nuclear Energy Sensor Database
Purpose:
Collect, store, and maintain nuclear power plant sensor technology information for nuclear energy applications. Provide the nuclear industry with the ability to browse and search sensor data.
- Initial Content: ORNL/TM-2016 Assessment of Sensor Technologies for Advanced Reactors.
Nuclear energy sensors Sensor use cases Sensor needs and gaps Status: The website is currently going through a soft launch and looking for subject matter experts interested in providing or reviewing content.
32 ASME CODE Section XI Division 2 - Requirements for Reliability and Integrity Management (RIM)
- Published July 1, 2019, after 15 years of development
- Addresses the entire life cycle (design through decommissioning) of all types of NPPs
- Sets Reliability Targets for each passive structure, system & component (SSC)
- Because advanced NPP designs do not exist and there is no operational experience, RIM details a process to achieve Reliability Targets
- RIM utilizes expert panels to guide the RIM program development
- PNNL has supported the development of RIM and will continue as the code evolves considering both regulatory and industrial parties 33
Conclusions
- Advanced reactor technology promises lower cost, safe, and sustainable energy.
- BUT there are serious technology gaps that must be bridged before we can realize this goal.
High-temperature ultrasound sensors for structural integrity, flow, and temperature Magnetostrictive cold spray as a high-temperature sensor On-line heat exchanger tube monitoring Acoustic power and modem for virtual penetrations Fiber and optical guides for motion, temperature, and material identification An advanced reactor sensor website and database ASME code evolutions to guide the industry as new technology emerges 34 Thank you 35
Need for Advanced NDE Methods in Non-Light Water Reactors Pradeep Ramuhalli (ramuhallip@ornl.gov)
Yarom Polsky (polskyy@ornl.gov)
Workshop on Advanced Non-Light Water Reactors - Materials and Component Integrity December 9-11, 2019 ORNL is managed by UT-Battelle, LLC for the US Department of Energy Outline
Background:
Advanced Non-Light Water Reactors (ANLWR)
- Needs and challenges for NDE and SHM
- Examples of recent R&D efforts
- Open research questions
- Summary 2 ANLWR Monitoring Needs and Challenges
Advances in Nuclear Power Technology Advanced Non-Light Water Reactors http://www.gen-4.org 3 ANLWR Monitoring Needs and Challenges Nominal Operating Parameters of ANLWRs SFR LFR MSR SCWR GCR Core Inlet 290-610 550-650 350 250-587 Core outlet 704*** 465-780 700-1000 625 530-850 Maximum ~825+, 705++ 814+ 1300***, 947** 1900+ 1238+
Temperature (oC)
Primary loop 338/485 405/561 570-650 /700-(Inlet/outlet) 1000 Secondary Loop 282/443 392/541 450-600 /633-690 (Inlet/outlet)
Pressure Range Reactor Vessel ~0.1-0.2 ~0.1 ~0.1 - 0.5 26 ~5-9 (MPa)
Flow Rate (kg/s) Primary loop 174,128 (l/min) 2150-16200 1418 96-320 Peak fast fluence 6.8x 1012* 3.7 x 1023 0.33-1 x 1021*
n/cm2 4.0x 1023 (limit) 3 x 1023**
Neutron Flux Flux (average) 2.35 x 1015 n/cm2-s Power density Average 17-210 69 Varies from 67- 4-6.5 (MW/m3) 300
- Reactor vessel
- Graphite moderator
- Coolant maximum
+ Fuel
++ Reactor Vessel Wall 4 ANLWR Monitoring Needs and Challenges
Operating Experience for ANLWR Passive Components
- Sodium reactors
- Cause: Manufacturing defects, defective welds, fatigue cracking, erosion, sodium deposits, contamination,
- HTR
- Cause: Material incompatibility, moisture intrusion, manufacturing issues,
- Effect: cracking, chloride corrosion, failure of nut/bolts, plugging of pressurization lines, mechanical jamming,
- LFRs, MSR
- Cause: Material choice, high coolant velocity, coolant contamination
- Effect: irradiation hardening, cracking, corrosion/erosion,
- Issues identified from OE have been largely addressed through design modifications in current concepts and selection of improved materials
- Some ANLWR operational concepts also include periodic replacement of critical passive components 5 ANLWR Monitoring Needs and Challenges Anticipated Challenges in ANLWRs 3/4 Harsh operating environments 3/4 Potential for higher neutron damage levels for core structures and vessel components, and extended exposure to higher temperatures/corrosive coolant chemistry 3/4 Extended life operation of ANLWR 3/4 Limited information on performance of some materials in ANLWR environments over lifetime 3/4 Non-traditional operations possible J.G. Marques, Energy Conversion and Management, 2010 (on-demand power, multiple missions, advanced energy conversion cycles, etc.)
3/4 Potential for increased stress from non-traditional operating profiles 3/4 Possibility of extended operating intervals and online refueling between refueling outages 3/4 Potentially limited opportunities for periodic in-service inspection of primary system components blog.ngnpalliance.org 6 ANLWR Monitoring Needs and Challenges
Defense-in-Depth Philosophy
- All U.S. nuclear power plants designed, built, and operated so as to:
- maintain structural and leak-tight integrity of components important to safety,
- prevent or minimize accidents, and
- mitigate the effects of accidents, should they occur.
- Defense-in-depth: Multi-layered approach to maintaining safety and high reliability
- No one action, system, or component is depended upon to maintain safety
- An integrated number of actions, systems, and components with multiple backups
- In-service inspection (ISI) using nondestructive evaluation (NDE) is one component of defense-in-depth
- ISI of safety components required by 10CFR 50 through endorsement of ASME Boiler and Pressure Vessel (BPV) Code
- ASME BPV Code specifies type of NDE inspection, frequency of inspection, and sampling criteria based on system classification (Class-1, 2, or 3) 7 ANLWR Monitoring Needs and Challenges ISI Scope and Effectiveness
- Inspection sample sets selected to be representative
- Populations and methods vary by safety class
- Finite periods of operation, or inspection intervals, defined
- Current 10-year interval established by consensus and practical convenience
- Effectiveness affected by:
- Unknown degradation progression rates - failures may occur between planned inspections
- Variable crack initiation times
- Degradation outside of initial inspection sample
- ASME NDE methods applied may not always be targeted for appropriate degradation processes
- Unpredicted (or undiscovered) degradation leads to augmented ISI programs and mandated requirements exceeding those in ASME XI
- SCC in SS and Alloy 600/82/182 welds
- Boric acid corrosion
- Flow-accelerated corrosion
- Other forms 8 ANLWR Monitoring Needs and Challenges
Visual Examination NDE Methods in Nuclear Industry
- Ultrasound, Eddy current, Radiography, Visual, Liquid penetrant testing, Magnetic particle testing
- NDE Reliability
- For a given flaw, what is the repeatability of detection by a given technique?
Ultrasonic Examination
- What is the smallest flaw that can be detected by a given technique? 1.2
- NDE reliability influenced by many 1
factors including equipment, materials 0.8 PIRR MRR POD 0.6 PISC-AST and surface condition, flaw size and 0.4 PDI PDI PASS + FAILED orientation, procedures, 0.2 From Doctor, SMIRT 2009 0
0 2 4 6 8 10 12 14 16 TWD (m m )
9 ANLWR Monitoring Needs and Challenges Potential ISI Needs and Issues in ANLWR
- Wide variation in materials Components Materials Potential Degradation Modes Desired Measurements
- Stainless steel, F/M steel, ceramics, graphite,
- Reactor vessel
- Austenitic
- Thermal and
- Cracking and
- Core structure, stainless steel mechanical corrosion Ni-base superalloys, shields
- Ni-base fatigue cracking
- Creep
- Reflectors, superalloys
- Creep/irradiation
- Coolant
- Locations vary for potential degradation absorbers,
- F/M steels creep/creep- parameters moderators
- ODS F/M steels fatigue (temperature,
- Welds and joints
- Piping and tanks
- Ceramics,
- Oxidation/corrosio pressure, flow,
- Heat exchangers, composites, n chemistry, level)
- Bends/elbows
- steam generators Turbines,
- polymers Graphite Embrittlement Stress corrosion Neutron flux Contamination compressors
- Concrete cracking (coolant and
- Tubing
- Valves, pumps
- Void swelling cover gas)
- Loose-parts
- Data on material performance over ANLWR lifetime is limited
- Tests ongoing for material qualification, codes and standards development
- NDE measurement challenges
- Potential access limitations for ISI
- Measurement parameter sensitivity
- Deployment issues for in-situ measurements 10 ANLWR Monitoring Needs and Challenges
ISI for Reliable Degradation Detection
- Most effective technique - continuously monitor all plant components 100% of the time On-line monitoring (SHM) sensors provide data as a function of time at discrete locations
- Next best method would be to examine all components during each refueling outage (or periodically)
- Not economically viable - plants would spend more time being inspected than making power TIME NDE provides data as a function of discrete times at discrete
- Would require very large population of skilled NDE and locations crafts personnel; especially for many plants in simultaneous outages SPACE
- Structural health monitoring (SHM), which continuously Fundamental differences in monitors a subset of components, may need to be data structure between considered for ANLWRs that may have longer refueling Nondestructive Evaluation cycles (NDE) and Structural Health Monitoring (SHM))
- Component selection based on contribution to risk, and (After Thompson [2009])
perhaps limited accessibility 11 ANLWR Monitoring Needs and Challenges Simulation Studies Can Provide Insights Into Inspection Performance 3/4Potential for highest sensitivity from contact probes using bulk wave Immersion inspection Specular Amplitude Immersion systems limited in ability to detect and size cracking Guided waves: sensitivity may be a challenge Tip Amplitude 3/4Potential for detection and sizing Contact capability on crack growth in SFR environments using ultrasonic measurements Temperature effects on detection and prognostic ability appear to be limited Studies indicate other factors (orientation, grain structure, presence of Na, etc.) have greater effect Guided Wave Source: Dib et al, IWSHM (2018) 12 ANLWR Monitoring Needs and Challenges
SHM Systems for Nuclear Power May Require High Temperature, Rad-tolerant Sensors Probe-Array Connector 3/4Materials selection is key End End Example: Prior research has shown viability Under-sodium Viewing Ultrasonic Phased Array of AlN and BiT composites for high-temp, in- (Larche et al 2017) reactor ultrasonic measurements High temperature (>550oC) Monolithic Ultrasonic Transducer (Ramuhalli et al 2018)
Figure . Relationship between piezoelectric coefficient d33 and the maximum use Sol-gel High-Temperature Composite temperature (for most materials, TC) for piezoelectric ceramics. Reproduced from M.
Akiyama, et al, Advanced Materials 21 (5), 593-596 (2009). Transducers (courtesy C. Lissenden, B. Tittman) 13 ANLWR Monitoring Needs and Challenges Advanced NDE Methods are Being Investigated for Early Detection and Monitoring of Degradation Low Cycle Fatigue (A36 Steel)
(Walker et al, 2011)
(Jacobs 2015)
Irradiation (A533B Steel)
(Matlack et al, 2014)
Nonlinear Ultrasonics MBN (Tensile Strain, 410 Steel)
(Ramuhalli et al, 2015) High Cycle Fatigue (304 SS)
(Ramuhalli et al, 2014)
MBN Peak (Tensile Strain, 410 Steel)
(Ramuhalli et al, 2015) 3500 Perpendicular 3000 Parallel 2500 2000 mV1500 Magnetic Barkhausen Noise 1000 500 0
-2% 8% 18% 28% 38%
Strain 14 ANLWR Monitoring Needs and Challenges
SHM/NDE Measurements May Provide Necessary Sensitivity to Other Degradation Modes Elongation (mm)
Energy y Time of Flight ghtht High Temperature (600oC) Creep Test Setup with In-Situ Ultrasonic Guided Wave Monitoring Source: Dib et al, (IWSHM 2018) 15 ANLWR Monitoring Needs and Challenges NDE and SHM Methods for Non-Metallic ANLWR Components are Likely Needed CTRLRL UASR R CASR Test specimenss W. Glass et al (2017)
D. Ezell (2019), High Fidelity Ultrasonic Imaging of Concrete NDE techniques for characterizing polymers and concrete for LWRs can likely be adapted for use with ANLWRs 16 ANLWR Monitoring Needs and Challenges
NDE and SHM Methods for Non-Metallic ANLWR Components are Likely Needed Source: Y. Katoh (2013), Continuous Fiber Ceramic Composites for Fluoride Salt Systems New techniques will be needed for inspecting/monitoring and characterizing materials such as SiC/SiC composites and graphite 17 ANLWR Monitoring Needs and Challenges Measurements Are Also Usually Easier than Interpretation p 3/4Most NDE methods for microstructure characterization provide relative and not absolute information Classical inverse problem: non-uniqueness 3/4Correlative analyses provide vital insights into measurement change with MBN Peak Value Change with Tensile Strain in 304 SS degradation (Ramuhalli et al, 2015) 3/4Approaches for quantifying material state from NDE/SHM measurements and its remaining life are needed Density and Ultrasonic Velocity Change with Dose in Stainless Steel Blocks Garner et al, INL/CON-14-33001, 2015 18 ANLWR Monitoring Needs and Challenges
Characterizing Reliability of Advanced NDE and SHM methods Will be Necessary
- Diversity in materials and material microstructure; fabrication history
- Harsh environments seriously challenge available sensors and instrumentation
- Delineating effects of multiple stressors
- Insufficient information for inverse models and damage accumulation models
- Damage threshold for failure (especially if defining degradation with respect to precursors) 19 ANLWR Monitoring Needs and Challenges Ongoing Research
- Reliability of SHM and NDE, for mechanisms other than cracking
- Measures of degradation severity or component health are needed
- Measurement uncertainty quantification and its effect on reliability
- Concurrent damage mechanisms - detection sensitivity and selectivity
- NDE/SHM of AM manufactured materials for ANLWR reactor applications
- Sensors and instrumentation survivability
- Online calibration of aging sensors for in-situ monitoring to correct measurement drift
- Optimal sensor placement for in-situ monitoring
- Inverse methods for quantifying material condition from NDE/SHM, estimating remaining service life
- Models relating material changes to measured quantities, and models for assessing component health change over time
- Failure/Acceptance criteria - define acceptable and rejectable damage thresholds. Traditionally defined for cracks using fracture mechanics calculations to determine minimum acceptable flaw size that does not compromise structural integrity - how do we extend this concept to other mechanisms?
- Human factors
- Inspection and data analysis can vary with operator
- Resulting analysis information will need to be presented to operators 20 ANLWR Monitoring Needs and Challenges
Summary
- NDE plays (and will continue to play) a critical role in defense-in-depth for nuclear power plants
- As ANLWR concepts mature, new challenges to assessing degradation level and growth rates are foreseen; Research (nationally and internationally) is addressing many of these challenges
- Sensing: what, where, and how to measure; sensitivity & fidelity; applications to non-metals and AM materials
- Sensors and instrumentation for in-vessel/in-containment use
- Inverse models for rapid, robust data analysis
- Qualification of sensors and instrumentation, systems, methods, and procedures and personnel
- Data and testbeds for testing and qualifying methods and developing analysis methods are necessary 21 ANLWR Monitoring Needs and Challenges Acknowledgments 3/4 A number of collaborators have contributed to the work presented here, and include staff from ORNL, PNNL, ANL, Bettis, INL, Universities (UT-Knoxville, PSU, WSU, ISU, CSU-LB, WUSTL, Ajou University), and Industry (AMS Corp.)
3/4 A portion of the research presented here was supported by the USDOE Office of Nuclear Energy through the Advanced Reactor Technologies (ART), Nuclear Energy Enabling Technologies (NEET), and the National Scientific User Facility (ATR-NSUF) programs. A portion of the research was supported by the NNSA Office of Defense Nuclear Nonproliferation (NA22). Parts of this work were supported by Ajou University (S. Korea) and USNRC.
3/4 Oak Ridge National Laboratory is managed by UT-Battelle for the US Department of Energy.
22 ANLWR Monitoring Needs and Challenges
Questions?
23 ANLWR Monitoring Needs and Challenges Extras 24 ANLWR Monitoring Needs and Challenges
Nondestructive Evaluation (NDE) 25 25 ANLWR Monitoring Needs and Challenges Example: Magnetic Barkhausen Noise Measurement - Effects of Stressors 0% Strain A softest HT-9 410SS 9.5% Strain 410SS B middle HT-9 17.2% Strain C hardest HT-9 Magnetic Barkhausen measurements (Different Hardness Levels)
Magnetic Barkhausen 26 measurements (Different Strain ANLWR Monitoring Needs and Challenges Levels)
Even How We Make The Measurement Can Impact The Result!
27 27 ANLWR Monitoring Needs and Challenges Online Monitoring for Materials Degradation:
Structural Health Monitoring (SHM)
- In-situ online monitoring
- Monitoring hard-to-access or high-risk regions
- Flaw growth monitoring
- Component/system-scale monitoring
- Acoustic emission only currently sanctioned technique for online monitoring of materials degradation by the ASME BPV Code
- Flaw growth monitoring only (flaw must be characterized using other methods)
- Guided ultrasonic waves being discussed for inclusion in Code
- Many other methods being researched
- Guided ultrasonic waves
- Electromagnetic methods
- Vibration monitoring AE System Circa 2010
- Advances in technology are positively impacting AE System Circa 1993 the development of OLM for materials degradation!
28 28 ANLWR Monitoring Needs and Challenges
Coupled Materials Evolution-Measurement Models May Help With Mechanistic Understanding Simulation (at
( 0G))
0G 0G 600 G 400 G Neel walls Bloch walls 1400 1300 G G 1700 1750 G
[-110] G (Saturat ed)
[110]
29 ANLWR Monitoring Needs and Challenges Probabilistic Failure Analysis May Also Include Crack Initiation
- Crack initiation probabilities calculated using empirical data
- Models usually augmented using statistical analysis
- Example Examp ple (Simonen et al 2001, al 2001, NUREG/CR-NUREG/CR-6335) 6335)
Calculated Probabilities of Crack Initiation and Through-wall Crack F.A. Simonen et al. / Nuclear Engineering and Design 208 (2001) 143-165 30 30 ANLWR Monitoring Needs and Challenges
Integrating Measurements with Analytics for Prognostics & Decision Making Risk Models and Metrics; Predictive Risk Estimates Data/Physic Sensor s Driven Design and Prognostics Inverse Survivability Problems and Data Fusion Online Monitoring Systems Reliability and Resiliency 6 31 ANLWR Monitoring Needs and Challenges Harmonics in Ultrasound may Provide Additional Sensitivity to Earlier Stages of Degradation 3/4 Sensitive to microstructural changes due to degradation Precipitates size and number density Dislocation loop density 3/4 Interpretation of measured response challenging and is an open question 3/4 Use in remaining life assessment will require long-term monitoring Model Predictions 32 Measurement ANLWRData Monitoring Needs and Challenges
NRC Workshop Advanced Sodium cooled fast reactors (SFR)
Non-Light Water Reactors
- Materials and Component Integrity
- Japan s experience and future Dec. 11 2019 1. Fast reactor development in Japan (Experimental reactor Joyo, Prototype FBR NP Monju, Future )
- 2. Design future of Sodium cooled Fast Reactor(SFR) and deterioration of sodium retaining components.
- 3. Surveillance and monitoring experience of Joyo and Monju Authors: 4. Activities towards future T. Asayama and S. Nakai 5. closing Joyo and Monju toward Commercial Reactors
3 Outline of Joyo Containment vessel Secondary pump Control room Dump heat exchanger Major specs Type :Loop type (2loops)
Rated power :140 MWt Fuel :MOX IHX Primary pump Reactor vessel Core Structure :Stainless steel Coolant :Liquid sodium R/V inlet temp. :350 deg-c Start of construction :1970 exit temp. :500 deg-c Operation breeding coreMK-) :1978 Core dia. :80cm (irradiation coreMK-II :1983 Core height :50cm Upgrading core MK-III criticality :July 2003 Start of irradiation with MK-III core :May 2004 History of Joyo
- 2004.5~ Rated Power Operation 2003.7First Criticality of MK-III Core MK MK-III Renovation 2000.6 Carbide and Nitride Fuel Irradiation Collaboration with JAERI Power-to-Melt TestPTM Fuel Failure Simulation Test MK High Burn-up Test Peak Burn-up of 144 GWd/t, Collaboration with CEA France)
Served Mainly as an Irradiation Facility for FBR Fuel and Material 82.7First Criticality of MK-II Core Natural Circulation Test Confirm Breeding Ratio MK- Accumulate Technical Experience Through Planning, Construction and Operation
.4Attain First Criticality
5 Operating history of Joyo Year 1970 ~ 1977 1978 1979 1980 1981 1982 1983 1984 1985 1986 1987 1988 MK-I Breeder Core MK-II Irradiation Core Core Low Power Test Performance 75MW Performance 100MW Operation Key Items Performance Test Test Operation Test Construction 50MW Operation 100MW Replacement 50MW 75MW 0 1 2 3 4 5 6 7 8 9 1011 12 13 14 15 16 0 1 2 3 4 5 6 0 1 2 0~100kW Initial Criticality MK-II Initial Criticality 1977. 4. 24 1982. 11. 22 Reprocessed Joyo Fuel loaded in the Core Year 1989 1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 MK-II Irradiation Core Transition Core Key Items 100MW Operation MK-III Modification 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 Work Year 2002 2003 2004 2005 2006 2007 MK-III Core Performance Test 140MW Operation April 24 1975 :
Key Items MK-III Modification Work 0 1 2 3 4 5 6 The 1st operational safety program was approved.
MK-III Initial Criticality March 31 2003 :
2003. 7. 2 Operated time 60,725h Accumulated thermal power: 5,061GWh 1st PSR Review Period 6
Feature of SFR Containment vessel PWR Containment vessel SFR Steam Pressurize Reactor r vessel Secondary sodium Generator Steam Turbine CR System CR Water Turbine Generator Water SG SG Condenser Water Water To Discharge Water(moderator Reactor Water Condenser channel Fuel coolant Sea pump To Discharge water Circulation
)
Water channel pump Circulating Reactor Feed water pressure water Sea pump vessel pump Circulating water Circulation Feed water Purification water Primary sodium System pump pump System pump IHX
7 Aging phenomena identified in Sodium Damage Aging degradation Consideration in design Corrosion of steel in sodium at low oxygen concentration is very low Thickness and not considered as degradation to care.
Corrosion reduction Significant corrosion of inner surface of sodium retaining pipe of Joyo piping operated over 60,000hours was not observed.
Fatigue and creep fatigue are major degradation of SFR, because of SFR future ,i.e. elevated temperature and high temperature difference.
Fatigue and creep In the design, cumulative creep and fatigue damage were evaluated and fatigue satisfy with criteria of JSME.
On the other hand, cracks are reported many SFRs in the world at Crack earlier stage, fatigue and creep fatigue are to be cared.
No environmental factors for SCC in impurity controlled sodium exist SCC and no SCC damage were reported so far.
SCC is not degradation to care.
Damage of reactor vessel and internal structure increase with neutron irradiation.
In design, the degree of damage is evaluated using accelerated material Material Neutron irradiation irradiation test data by irradiation reactor.
property degradation To confirm material property change under the actual irradiation degradation condition, the test pieces are loaded.
As the property change of SFR irradiation shows ductility reduction, elongation at fracture of unloaded test piece is measured.
8 Specimen Irradiation Positions R=180cm Core Component Sodium Temp. Material Irradiation Spent Fuel Irradiation Position R=96cm (MK-II) Rack Storage R=160cm Rack Reflector 500deg-C Core Barrel (316SS)
CORE Ex-vessel Irradiation Hole R=207.0 Core Center Level Reactor Vessel (304SS)
Z=109cm Core Support Plate (316SS)
Dosimeter Set 370deg-C
9 Post Irradiation Examination (Tensile Test)
Before After Material : 304SS (Reactor vessel material)
Irradiation subassembly No. : II-04 (Irradiated at in-vessel spent fuel storage rack)
Test temp. 400 deg-C Radiation Deterioration: Neutron 10
- Core Support Plate (Base metal: 316SS) -
500 Irradiated at : Reflector region 0.2% Proof stress (MPa)
R.T.
400 In-vessel spent fuel 400 storage rack 500 300 550 200 100 0
End of design life (2.6E+21) 100 Integrity of core support R.T.
80 400 500 plate material 60 550 was confirmed 40 20 Criteria 0
1.0E+19 Unirradiated 1.0E+20 1.0E+21 1.0E+22 1.0E+23 Fast neutron fluence (n/cm2,E 0.1MeV)
Integrity of Core Structure Materials (Irradiation effects) 11 Reactor Vessel Specimen (304SS)
End of design life:3.481019n/cm2.
1000 Creep rupture strength unirradiated 19 2 I-03 (4.2x10 n/cm )
20 2 II-02 (9.1x10 n/cm )
21 2 II-04 (2.2x10 n/cm )
Stress (Mpa)
( ) : Fast neutron fluence Design criteria at 500 Fast neutron fluence at the end of design life : 3.5x1019n/cm2 100 100 1000 10000 Time to Rupture (h)
Measured results satisfied the design criteria Design criteria: JSME Codes for Nuclear Power Generation Facilities Rules on Design and Construction for Nuclear Power Plants Part 2: Fast Reactor Erosion for the removed Secondary Main pipe 12 (by MK-III modification)
Results:
9 No significant reduction in the wall thickness of straight pipe and elbow part of the secondary main cooling system.
9 No corrosion/erosion in the inner surface.
IHX outlet elbow inner and outer surface Material:2 1/4Cr-1Mo steel
13 Yield strength of removed secondary pipe)
Secondary hot leg pipe tensile characteristic (Base metal)
Virgin Virgin 6700 hr in sodiumNa 20000 hr in sodiumNa 0.2 Yield sterngth (MPa) 20000 hr in sodiumNa Virgin (MK-III)
Replaced MK-III Pipe (60000 hr in sodium)
BDSSy Temperature (deg-c) 14 Sodium Leak Continuous Monitoring Concept of the LBB (Leak before Break)
Leak Before Break (LBB) is satisfied for SFRs.
Ductile Structural Material : Austenitic SS Low Pressure Boundary : Sodium Coolant Continuous Monitoring is adopted as a main measure of the In-Service Inspection.
Sodium Leak Monitoring (SoLM)
Ar gas Leak Monitoring (ArLM)
Periodic tests such as Visual Test (ViT) and Material Surveillance are also adopted complementary.
15 Sodium Leak Detection Systems (example)
The Various kinds of sodium leak detectors in Monju are classified as shown in the table.
Object Detector Type Leak Scale Reactor Operation SID (Sodium Ionization Detector)
Very Small DPD (Differential Pressure Detector) Scale*1 CLD (Contact Type Sodium Leak Detector) Manually Scram Primary Process Data Changing System (Process instrumentation : O/T Sodium Level)
Middle Scale*2 Process Data Changing to (Safety protection system: R/V Sodium Level or G/V Automatically Large Scale*3 Sodium Level or C/V Under Floor Temp. or Scram Pressure or Radiation Dosage of C/V Above Floor)
RID (radioactive Ionization Detector), CLD Very Small Scale Smoke type Detector to Manually Scram Secondary (Fire Alarm System) Small Scale System Process Data Changing Automatically Large Scale (Sodium Level or Temp.) Scram
- 1: Sodium leak rate >1 X10-3 kg/h. R/V: Reactor Vessel O/T: Overflow Tank
- 2: Sodium leak rate 1 to 50 X103 kg/h. G/V: Guard Vessel C/V: Containment Vessel
- 3: Sodium leak rate >50 X103 kg/h.
16 Installation of Sodium Leak Detectors (R/V)
17 Gas Sampling Type Sodium Leak Detectors (Example) 3/4SID (Sodium Ionization Detector)
Na Aerosol Ion Collector Filament Ion Current Power for Heating Sodium Piping Heat Insulator S SID is best suitable to detect sodium Sampling Piping leak in the inert gas atmosphere (detector would burn if used in air).
Detector Pump Out Put Signal (Sensitivity: 110-10 g Na/cc (minimum))
<Gas Sampling Method>
SourceM. Sawada & G. Rodoriguez, pocketbook on Sodium Technology by JAEA & CEA, P36, July 2006 18 ISI Technology Development ISI device for Reactor Vessel Circumferential Welds of Reactor Vessel ISI Region Weld Joints of Inlet Nozzles and Vessel Weld Joints of Inlet Pipes and Nozzles VitFiber Scope CCD Camera Shielding Plug Methods VotUT by EMAT(Electro Magnetic Acoustic Transducer) (R&D)
Outlet Pipe EMAT Inlet Pipe (R&D)
Reactor Vessel ISI Robot CCD Inspection Conditions Temperature: 200 deg-c Guard Vessel Radio Activity: 10 Sv/hr Atmosphere: Nitrogen Gas
19 ISI device for Primary Pipes Controller Operation Floor Cable UT Device PHTS IHX Testing Condition Pipe Surface (Dose Rate 15mSv/h)
Allowable Working : 5min Narrow Space with Obstacles Temp. (Atmosphere 55deg-c, Piping Surface 80deg-c)
GV 20 Codes and standards Surveillance JSME Codes for Nuclear Power Generation Facilities
- Rules on Design and Construction for Nuclear Power Plants Part I: Light Water Reactors JSME S NC1 - 2008/2009 Article 12 Surveillance Testing Monitoring and inspection JSME Codes for Nuclear Power Generation Facilities
-Rules on Fitness-for-Service for Fast Reactor Nuclear Power Plants Under deliberation for class1 component in JSME
21 Class 1 components examination methods Parts Rules on Fitness-for-Service LWR Examined for Fast Reactor Primary Sodium Continuous monitoring 1 Volumetric examination, surface coolant retaining parts (CM1) examination boundary Small diameter pipe: Small diameter pipe:
welds System leak test and VT-2 Continuous monitoring 2 (CM2)
Cover gas Continuous monitoring 3 retaining parts (CM3)
Welded attachment VTM1 Surface examination z Leak detection sensitivity is Sodium and radioactive decided by LBB evaluation.
Retaining z JSME code Rules on Leak CM-1Leak detection Before Break Assessment for CM-2(CM-3) sensitivity is required Sodium Cooled Fast Reactor Plants that is under cover gas System leakage NDI and system leakage deliberation, may be applied.
Not retaining test test z As for Monju, primary main piping satisfies LBB assessment by using sodium leak detector Small Large that can detect leak rate of Influence of damage 1kg/h sodium leak.
22 JSME Fast Reactor Codes
System Based Code Concept Present System Based Code Margin accumulated but how Target reliability is determined much is not clear. first.of Expansion technical options LOAD LOAD Margin exchange MATERIAL MATERIAL Update of U
DESIGN DESIGN reliability r
re evaluation ev ev INSPECTION INSPECTION etc... etc...
APPROPRIATE APPROPRIATE EndTOTAL up Target TOTAL INTEGRITY with INTEGRITY EXCESSIVE Intermediary targets Design to required reliability Asada, Y., Japanese Activities Concerning Nuclear Codes and Standards - Part II, Journal of Pressure Vessel Technology, ASME 128 (2006) 64.
23 24 ASME Code Case N-875
- The SBC process consists of Stage I and II evaluations that have different objectives.
(a) Stage I is a structural reliability evaluation which considers the component level structural integrity and the probability of failure of the component at design basis conditions The contribution of inservice inspections is not taken into account at this stage.
(b) Stage II is a safety-related evaluation of detectability of a flaw which ensures that the plant can be safely shut down before the flaw reaches the maximum acceptable size. The evaluation is performed taking into account the component safety functions during plant operation and the events that have been postulated in the safety evaluation of the plant. Any flaw can be detected either directly or indirectly.
Indirect detection includes the detection of a leak or an unintentional discontinuity of plant parameters such as temperature and leak rate. If a flaw is not detectable, then additional margins in structural integrity will be required
25 Conclusions The experience of the experimental fast reactor Joyo and the prototype reactor Monju provides valuable information for the development of next generation liquid metal reactors.
The Joyo operation and maintenance (O&M) experiences over 30 years and integrity confirmation results of PSR will be reflected to future SFR not only O&M but also design and construction.
The R&D and design for Monju, the SSC developed for Monju such as leak detectors, ISI device are taken in codes and standards.
Codes and standards development is ongoing within JSME utilizing the experience. Collaboration with ASME is concurrently underway.
Nondestructive evaluation in advanced reactors Greg Selby Senior Technical Executive, Plant Support Electric Power Research Institute Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop December 9-11, 2019 Rockville, MD www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Topics Necessary areas of research focus Relevant EPRI research in process Highlight: NDE research in support of compact heat exchangers 2 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Necessary areas of research focus 3 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
CHALLENGE RESEARCH Understanding the degradation mechanisms Materials and corrosion research specific to the Knowing what to look for, and where, before plant operational temperatures, loads and coolant chemistry design is finalized Optimizing O&M costs Robotics, permanent sensors, data analytics The plants must be economically competitive Robust sensors, coupling and connections at Severe environment - during power operation temperature For permanently mounted sensors Ultrasonic, strain, optical Severe environment - during outages
- Robotics hardened for the anticipated radiation fields Conditions affecting human- and robotically-delivered
- Minimizing human entry NDE Plant design to minimize radiation fields and Access - human temperature, and to provide adequate space, where human entry will be required Plant design to enable robotic access and operations to the fullest extent possible Access - robotic The ideal: no human entry at all (ref. high-radiation hot cell facilities) 4 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Why Material Characterization?
Material Characterization Traditional NDE Inspect for Susceptibility Inspect for Effect Depth of Penetration Traditional NDE Detection Limit Slow Growth Crack Propagation Damage Precursor Still sub-millimeter Accelerated growth Accumulation (up to 10s - 100s Pm) rates Time 5 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Relevant EPRI research in process 6 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Void Swelling Tihange 1 Bolt Irradiation of austenitic stainless steel can lead to dimensional changes, distortion, and embrittlement
- Detrimental for components that do not have a lot of dimensional tolerance
- Small levels of differential swelling in large components could result in significant 330qC (626qF) strains 7.5dpa 0.24% Swelling Swelling could be an end-of-life determinant for a component 7 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Void Swelling: Approach Test samples Techniques EBR-II Hex Block Samples Nonlinear Ultrasonic Testing with known swelling (0-3%) - Nonlinear parameter E Linear Ultrasonic Testing Unirradiated blank coin
- Speed of Sound specimens
- Attenuation Four-Point Probe Resistivity 2019 UT measurements 2020 Resistivity measurements Identify and assess best NDE method Develop concepts for in-situ inspections 8 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Fracture Toughness Test samples Compact CT: SS304, SS316, SS347, A286 Cold work to simulate embrittlement
- Cold work 0%, 20%, 40%, 60%, 80%
Techniques
- Linear UT (velocity, attenuation, 2019 absorption NDE on surrogate CT samples
- Nonlinear UT Identify relevant irradiated materials
- Thermal Thermoelectric Power 2020
- Instrumented Indentation System Destructive testing of surrogate CT samples NDE on irradiated materials 9 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Microstructure Characterization A significant portion of failures in grade 91 Current NDE methods use and 92 steels have been attributed to poor hardness measurements fabrication - Difficult to obtain Damage susceptibility is linked to meaningful results operational conditions as well as variability - Ambiguous, insufficient of microstructure measure of quality P92 - 800oC P92 - 860oC P92 - 900oC P92 - 960oC 10 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Microstructure Characterization: Test Samples Quantity OD (in) Wall Thickness (in) Axial Length (in) Angular Section Tubes 16 2 0.165 12 360q Pipe Sections 8 18 1.5 9 a60q Median Condition Over-Hardness As-received Tempered HAZ Ferritic tempered As-received 220 Normalized 428 Normalized + Tempered 208 Tempered 207 Over-tempered 198 Fully Ferritic 147 HAZ 405 HAZ + Tempered 170 11 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Microstructure Characterization: NDE Techniques Magnetic Methods Schedule
- BH loops
- Incremental permeability
- 2019
- Barkhausen noise Magnetic & Ultrasonic NDE Linear Ultrasound testing
- Attenuation Identify additional NDE
- Backscatter techniques
- Absorption
- 2020 Additional proposed techniques under review Additional NDE testing
- Thermal properties
- Electric properties 12 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
NDE research in support of compact heat exchangers 13 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Compact heat exchanger (CHX)
Not a tube-and-shell design Solid metal block with many small hot and cold flow channels 1.5000 parting plates 6.0000 channel plates 6.0000 27.0 instrumented 0 plates 1.5000 27.0000 channel plates 7.5000 dimensions in mm parting plates 6.0000 14 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
NDE research is part of a DOE project directed at a Section III Code Case supporting CHX construction for advanced reactors Components Scenarios
- Diffusion-bonded block - Construction examination - highest priority for the Code Case Solid side walls
- Other examinations of interest to the Channeled core owner/operator - in support of run/repair/replace decisions
- Header attachments to the block Pre-service examination, after the CHX is installed In-service examination (during operation or during outages) 15 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
NDE results on a failed CHX provided by UW 16 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Comprex CHX 17 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Scanning setup 18 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Side A - entire surface Pressure boundary thickness ~18mm Full width ~ 500 mm Pressure boundary thickness Full length ~ 800 mm 19 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Side A - detail of largest indication Conservative measure 10 x 11 mm Actual size will be less 20 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Side B - entire surface Pressure boundary thickness ~18mm Full width ~ 500 mm Pressure boundary thickness Full length ~ 800 mm 21 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Side B - detail of largest indication Conservative measure 9 x 6 mm Actual size will be less 22 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Block interior, during service Strain gages, embedded Challenges:
- Sensor design and bonding method
- Accurate numerical modeling
- Sensor location optimization
- Algorithm design
- Sensor calibration 23 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
discussion 24 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
TogetherShaping the Future of Electricity 25 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
RESEARCH REACTOR EXPERIMENTS TO STUDY MATERIALS AND FUEL SALT PERFORMANCE NRC Advanced Reactor Materials Workshop Uazir Bezerra de Oliveira 11 December 2019 2
OUTLINE
- The Dutch Molten Salt Program
- Projects of the MSR Program
- Roadmap - Molten Salt Reactor Program
- Projects in a nutshelll S Salient 1
- Take away message S Saga
- Acknowledgements Enickma E
Salient 3 S
Waste W
3 THE DUTCH MOLTEN SALT PROGRAM
- Molten Salt Technology fits with the Dutch energy R&D program:
- Contribute to CO2-free energy market
- Reduce resource consumption / waste
- Improve safety Ensuring Nuclear Performance 4 THE DUTCH MOLTEN SALT PROGRAM
- NRG = Enabler of MSR Technology due to nuclear know-how, infrastructure, international network.
- Collaborations with competence centers:
JRCs, TUDelft, FUBerlin and CV Rez.
- Objectives:
- 1. Obtain operational experience
- 2. Safety
- Investigate FP management methods
- 3. Material investigation:
- Material properties of irradiated containment materials
- In-pile corrosion / deposition of metal alloys and SiC
- 4. Waste:
- Provide a waste route for spent molten salt fuel
- 5. Integral Demonstration:
- Feasibility of experimental Molten Salt loop for the HFR Petten n
5 PROJECTS OF MOLTEN SALT PROGRAM
- Focus on irradiation technology for generation of reliable data.
- Focus on generic topics
- Ambitious R&D program open for partnering Roadmap - Molten Salt Reactor Program Under review 6
2018 2019 2020 2021 2022 2023 IRRADIATION SALIENT-01 (H4, TRIO)
PIE SALIENT-02 TRANSPORT DESIGN / IRRADIATION ENICKMA BUILD (v.a. Q2, H2/H8, TRIO)
PIE MSR Waste DEVELOPMENT MSR WA W WASTE HANDLING METHOD MSR WASTE HANDLING BASIC MSR Loop DESIGN DESIGN / IRRADIATION SAGA BUILD (v.a. NOV-19, GIF)
PIE DESIGN / IRRADIATION SALIENT-03 BUILD (v.a. 2020-02, H4, REFA)
PIE STATIC WATER/GLYCOL ATE STATIC SALT He-Bubbling (TUD) COLUMN and an LOOP COLUMN Fission Product DB (TUD) DB DEVELOPMENT (COMP.)
PRE-DESIGN DESIGN BUILD IRRADIATION SALIENT-04 (in-pile, v.a. 2022)
Cold Tests Corrosion (out-of-pile) Electrochemical lec e tests IRRADIATION ENICKMA II DESIGN BUILD (v.a. 2021)
PIE
7 SALIENT-01 Idea: To build up experience with molten salt fuel irradiations.
Choices:
- Graphite was chosen for being corrosion free.
- Salt composition limited by JRC Karlsruhe capability at the time (no U nor Pu salts).
Scope:
- Irradiation of 78LiF-22ThF4 salt.
- In-pile temperature monitoring
- At JRC Karlsruhe: Knudsen cell effusion (determination of salt stability)
- Extensive PIE:
- 1. Gamma scan (ongoing): qualitative view of fission product distribution
- 2. Puncturing of 1st containment: fission gas characterization
- 3. Calibrated burn-up analysis based on activation monitor set results
- 4. Rinsing capsule + Analysing release of volatiles
- 5. Microscopy (Salt and Fission product penetration, surface characterization, etc)
- 6. Salt impregnation test of irradiated graphite 8
SALIENT-01 ASSEMBLY Synthesis and crucible loading at JRC Karlsruhe Assembly of sample holder at NRG Design TMA X-ray
9 SALIENT-01 EXPERIMENT
- Open capsules fabricated from nuclear-grade graphite
- Fuel power rises during irradiation due to production of U-233
- Fixed crucible temperature (~600 oC)) activelyy maintained 10 SALIENT 1: IN-PILE TEMPERATURE 700 L1 L2 L3 L4 L5 (dummy)
Salt power, through Temperature (oC) thermal 600 neutron flux, is relatively 500 sensitive to changes in the core 400 300 2017-07-25 2018-01-25 2018-07-25 2019-01-25 Date (yyyy-mm-dd)
11 SAGA: GAMMA IRRADIATION OF FUEL SALT AT LOW TEMPERATURE Idea: Simulate the formation of F2 gas when the salt cools down (range 50-150 oC).
Scope:
- HFR Spent fuel used as the gamma source
- 40-45 oC base irradiation
- Monitoring of pressure, dose rate and temperature
- 5 salt samples provided by CV Rez & FUBerlin:
- Powder: LiF, BeF2, ThF4, UF4, LiF-BeF2-UF4
- Fused: 1 Empty reference capsule as reference 12 SAGA: EXPERIMENT Spent fuel Spent fuel
- 5 salt Capsules + 1 Empty Gamma Irradiation started
- Instrumentation (on-line measurement) 27 November 2019
13 ENICKMA: EMBRITTLEMENT OF NICKEL-BASED ALLOYS IN HELIUM Idea: Material transformations of Nickel Grade Supplier based alloys during irradiation. 3166 L(N) CEA Scope: Hastelloy N Haynes GH3535 SINAP Irradiation parameters: HN80MTY COMTES FHT
- Temperature: 650 and 750 oC MONICR COMTES FHT
- Up to 1E21 n/cm2 thermal, 3E21 n/cm2 Hastelloy 242 Haynes fast (up to 50 appm helium, >1 dpa expected)
PIE:
- Microstructure analysis
- Tensile testing
- Low Cycle Fatigue
- Small Punch testing
- Oven anneal test at same temperatures as references Start of irradiation foreseen Q2/3 2020 14 SALIENT-03: IN-PILE CORROSION OF NICKEL ALLOYS Idea: Investigate in-pile corrosion of Nickel Electrodes Pressure alloys by fluoride fuel salt. Heaters added to sensor keep temp. > 150C. Heater wire Scope:
- Corrosion assessment >13.000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> in-pile NBG-18 graphite Gas tube
- Determine the influence of fission products and redox buffering on corrosion.
- Compare experimental mass transport in a Nickel alloy non-isothermal salt column to CFD capsule simulations.
(Fission product behavior)
- Determine in-pile fission gas release.
- Establish which fission products/species relocate to cold spots during irradiation.
- Determine post-irradiation fission product release temperatures (Knudsen Cell Effusion test at JRC Karlsruhe).
Start of irradiation foreseen Q2 2020
15 WASTE STRATEGY AND R&D
- Idea: Conversion of salt to recognizable, acceptable chemical forms, i.e.
- Actinide-bearing oxide high level waste
- Cemented intermediate level waste
- Fluoride intermediate level waste (CaF2 or fluorapatite)
- Discussion with national repository
- Route: direct oxidation, aqueous processing
- Can be performed at NRG hot cells with relatively little infrastructure changes
- No complicated gas streams
- Limited spreading of dust 16 TAKE AWAY MESSAGE
- NRG is an enabler of Molten Salt Reactor Technology by developing testing irradiation capabilities to produce reliable data and knowledge.
- R&D Projects are tailored aiming to understand mechanisms, such as: corrosion (Salient 3), alloy embrittlement (Enickma), radyolitic production of F2 gas (Saga), behavior of fission products (Sal. 1,3.)
- NRG is open for R&D collaboratons with MSR community.
- Projects can be set to support specific needs of commercial clients.
17 ACKNOWLEDGEMENTS EU DuC = N Goods labeled with an EU DuC (European Dual-use Codification) not equal to N are subject to European and national export authorization when exported from the EU and may be subject to national export authorization when exported to another EU country as well. Even without an EU DuC, or with EU DuC N, authorization may be required due to the final destination and purpose for which the goods are to be used. No rights may be derived from the specified EU DuC or absence of an EU DuC.
Molten Fluoride Salt Chemistry Raluca O. Scarlat scarlat@berkeley.edu Advanced Non-Light Water Reactors -
Materials & Component Integrity Workshop U.S. NRC Rockville, MD December 11, 2019 Funding acknowledgements:
NEUP-15-8352 NEUP-16-10647 NEUP-17-13232 NEUP-IRP-17-14541
/-0,&8/./2-"4*/.:8"-0,& */#*5- /,5#*,*49*.*&
Pseudo-Pourbaix Diagram Pourbaix diagram for niobium: solubilitization of an otherwise volatile species state (Nb5+)
Oxidation (electron transfer) reaction:
Nb4+ <-> Nb5+ + e-Association-dissociation (complexation) reaction:
Nb5+ + 2O2- <-> NbO2-Other examples:
Be2+ + 4F- -> BeF42-2BeF42- <-> Be2F7- + F-
&
/,9-&2*$ 425$452&3".%2*%(*.(,5/2*.&3 Ting, Baes, Bamberger, Mamantov. The Oxide Chemistry of Niobium in Molten LiF-BeF2 Mixtures. J. Inorg. Chem.
1977 (39) pp. 1803-1808.
Olander, Fukuda, Baes. Equilibrium Pressures over BeF2/LiF(Flibe) Molten Mixtures. Fusion Sci. Technol.
41, (2001).
",5$" $"2,"4:3$"2,"4.5$#&2+&,&9&%5:3$"2,"4#&2+&,&9&%5
/-0,&8/./2-"4*/.: 0&$*"4*/.*"(2"-3".%,5/2"$*%*49 0 : 0 Pseudo-Pourbaix Diagram Speciation Diagram Zinc in aqueous solution at 25oC Activity Coefficients LiF and BeF2 in LiF-BeF2 binary mixture Baes, Molten Salt Reactor Fuels. Journal of Nuclear Materials 51 (1974) 149-162.
",5$" $"2,"4:3$"2,"4.5$#&2+&,&9&%5:3$"2,"4#&2+&,&9&%5 R. A. Reichle et al., Can. J. Chem. 53 (24) 3841-3845 (1975). [
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F. Carotti, H. Wu, and R. O. Scarlat. Characterization of a Thermodynamic Reference Electrode for Molten LiF-BeF2 (FLiBe). Journal of The Electrochemical Society, 164 (12)
H854-H861 (2017).
-=:??5 -@019.-/4 )@!&?=-6-%#&/-=7-?4-=-/?1=5E-?5:9:2- 80&$4&%$)&-*$",*.4&2"$4*/.#&47&&.* ".%
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&'&2&.$&,&$42/%&3: )&2-/%9."-*$&'&2&.$&,&$42/%&'/2*&
",4,&-&.4",.",93*3 Standards and Standard Methods for Elemental Analysis are Needed
!-=:??5 -@019.-/4 )@!&?=-6-%#&/-=7-?4-=-/?1=5E-?5:9:2-'41=8:0D9-85/%121=19/171/?=:012:=!:7?19 51
51$-=? !-?1=5-7>9-7D>5> :@=9-7:2?4171/?=:/4185/-7&:/51?D ! J. E. Seifried, R. O. Scarlat, P. F.
Peterson, E. Greenspan. A General Approach for Determination of Acceptable FLiBe Impurity Concentrations in FHRs. Nuclear Engineering and Design. 343: 85-95 (Mar. 2019).
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Francesco Carotti, Leo L Liu, Digby Macdonald, Raluca Scarlat. Electrochemical
",5$" $"2,"4:3$"2,"4.5$#&2+&,&9&%5:3$"2,"4#&2+&,&9&%5 Studies of Hydrogen in Molten 2LiF-BeF2 (FLiBe). (in preparation) 7
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F. Carotti. Ph.D. H. Atsumi et al.
Thesis. UW Madison. Journal of Feb. 2019. Nuclear Materials 417 (2011) 633-636.
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,&$42/$)&-*$", 45%*&3/'8*%&3*.*&
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",5$" $"2,"4:3$"2,"4.5$#&2+&,&9&%5:3$"2,"4#&2+&,&9&%5 9 Molten Fluoride Salt Chemistry Future Questions to Investigate
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Molten Fluoride Salt Chemistry ADDITIONAL SLIDES
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Thermodynamic Drivers for Corrosion
- 1. Oxidating species: a balance among cation oxidation states.
Fission, activation, temperature gradients and addition of materials can have oxidative effects
- 1. Standard states can have SEVERAL definitions
- 2. Molecular weight of FLiBe (i.e. xi) can have (Zhang, 2018) several definitions
- 3. The activity coefficient provides conversion among standard states
- 4. In order of significance::
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More on Activity Coefficients:
Complex Ion Formation and Acidity
- 1. New complex ionic species can form in the melt
- 2. An equilibrium exists among many complexes, and it is temperature and composition-dependent.
(Quist, 1972) 3. Effect on activity coefficient:
The Gibbs energy of formation is captured in the activity coefficient.
The interaction energies of the new species with their surroundings are captured in the activity (Baes, 1970) coefficient.
- 4. Effect on other solutes:
Solubilities and Gibbs energy of dissolution are affected
Solvation mechanism can change, affecting reaction kinetics and diffusivity (Baes, 1969)
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CHARACTERIZATION OF THE HYDROGEN ELECTROCHEMICAL REACTION IN FLIBE
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Francesco Carotti, Leo Liu, Digby Macdonald, Raluca Scarlat. Electrochemical 1 Studies of Hydrogen in Molten 2LiF-BeF2 (FLiBe). (in preparation) 4
Fundamental properties and Removal of Impurities in Molten Salt systems Jinsuo Zhang Nuclear Engineering Program Mechanical Engineering Department Virginia Tech Outline Introduction New method for Red potential measurements New method for exchange current density measurements New method for lanthanides separation On line corrosion measurement Summary of what we have done Nuclear Materials and Fuel 2 Cycle center
Introduction Tritium:
Noble Gas:
Halogens (iodine)
Alkaline Metals Rare Earths Noble Metals:
Tellurium and Antimony Actinides Corrosion Products Oxygen and Moisture Nuclear Materials and Fuel Cycle center Redox potential measurement C M W Cl2/Cl- Standard WE RE CE Alumina Tube Cl-Salt + MCln Real RE Chloride salt (AgCl/Ag)
Function of Alumina tube
- Isolation of the metal ion of interest to prevent it from being oxidized. E.g.
Fe2+ and Cr2+
- Allowing Cl- exchange between RE Men+/Me and bulk salt.
- If the metal ion has only one stable state (such as Ni2+), the alumina tube can be removed.
Nuclear Materials and Fuel Cycle center
The Methodology of CP Data Analyze Potential change across the interface, can be neglected when I is small
=
+ ln + + ()
/
Solution When I=0 resistance
() = + ln
/
- Redox potential
- =1 pure substance =
/ + ln = ()
/
- =1 in Chloride salt
- Formation potential E = / + ln
/
() = /
Reverse the reference
() = /
= / + ln
/ /
Nuclear Materials and Fuel 5 Cycle center Result example if NiCl2 in MgCl2 / KCl / NaCl (a) Different current densities CP curve for 5 wt% NiCl2 - MgCl2 - KCl - NaCl solution at 873 K. (b) The plateau potential (zoomed-in (a)) with respect to current densities for 5 wt% NiCl2 - MgCl2 - KCl - NaCl at varies temperatures. WE: graphite CE & RE: nickel rod Nuclear Materials and Fuel 6 Cycle center
Result example if NiCl2 in MgCl2 / KCl / NaCl The summary results of NiCl2 redox potential (a) and formal potential (b) of three different concentrations at 773 K to 1073 K. WE: graphite CE & RE: nickel rod.
- E0*= 5.798E-04T - 1.363E+00 (Formal potential is using the average value at each temperature)
Where E0* is the formal potential of NiCl2, and T is the temperature in Kelvin Nuclear Materials and Fuel 7
Cycle center Exchange current Density- Optimization fitting Fitting results:
Difference between experimental and fitted
( )
data is typically 5 -10%
Potentiodynamic polarization curves obtained in LiCl-KCl-1-LaCl3 melts at 1 mv/s Nuclear Materials and Fuel 8 Cycle center
Examine concentration correlation
()
- = ()
= 9.00 .
= 6.90 .
= 4.76 .
Plots of i0 versus GdCl3 and LaCl3 concentrations Nuclear Materials and Fuel 9 Cycle center Conditions for Electrochemical Separation
- The Standard deposition potential of the impurity should be more positive than major metal ion redox potential but more negative then F/F-In LiF-NaF-KF, La and Ce can not be deposited Nuclear Materials and Fuel 10 Cycle center
Electrochemical Separation of LaF3 from FLiNaK Molten Salt on Inert Mo/W Electrode Nuclear Materials and Fuel La metal 11 Cycle center Why? Lanthanum Deposition Potential Shifts When KF or NaF Presents 3/4 Lanthanum is found to predominantly exists in the species of LaF63- instead of La3+ in molten fluoride salts when KF or NaF presents.
+ +
( )
=
+
= 3 ( )
Standard reduction Redox couple potential Eo (V vs. F2/F-)
LiF/Li -5.38 NaF/Na -4.85 KF/K -4.83 LaF3/La -4.99 K3LaF6/La -3.97 Nuclear Materials and Fuel 12 Cycle center
Does this happen for all the molten fluoride?
With the presence of the KF or NaF in molten fluoride salts, the deposition potential of lanthanum shifts to the positive direction greatly.
Nuclear uclear Materials M terialls and Fuel Ma Fuel Fuel 13 Cycle ycle center On-line corrosion Measurement-Eu effects on Alloy 709 and Inconel 718 Nuclear ccllear ea ear Materials Ma M atte erriia alls and and Fuel an Fu Fue ell 14 Cycle cclle center cce cent enntter er
What we have done/will do Redox Control Separation Kinetics Materials Transport in Primary Salt loop Materials Corrosion by Fission Products Flow induced corrosion (loop tests)
Compound formation and Plating out Available facilities in our lab Molten salt flow loop 4-High-temperature Electrochemical cell (Up to 1000 C)
DSC machine CALPHAD saltware Corrosion model, materials transport model in flow loop Static corrosion test autoclave (can work at high pressure)
Radioactive laboratory Nuclear Materials and Fuel 15 Cycle center Acknowledgement
- Funded by DOE
- Team Member: Mingyang Zhang, Yafei Wang, Shaoqiang Guo Nuclear Materials and Fuel 16 Cycle center
Fluoride Salt Properties and Chemistry Advanced Non-Light Water Reactors -
Materials and Component Integrity Workshop Dr. Matthew J Memmott - Brigham Young University December 11th 2019 2
Molten Salt Reactor
3 Licensing Source Term Thermophysical Properties Transport (System Models)
Thermodynamic Behavior 4
Fission Product Transport
5 Thermophysical Properties
- Salt is challenging!
- Anaerobic
- Anhydrous
- Wall creep
- Be toxicity
- High temperature
- Prone to impurity retention
- Need , k, Cp, , ,
etc.
6 Actinide Bearing Salts
- Purification/standards for clean salt not set
- Clean salt data exists, though limited
- Salt wont remain clean for long!
- Actinide salt
- Fission product salt
- Millions of experiments for all combinations
7 Thermodynamic Analysis
- Equilibrium potentials for actinides (Th, U) or fission products at low concentrations
- Analyze/evaluate thermodynamic data, assess trends/behaviors
- Phase prediction using modified quasi-chemical model 8
Atomistic Modeling T
Thermophsyical Property Modeling Ion-Ion Interactions
9 Salt Property Characterization
- Inductively coupled plasma
- Common method for composition analysis
- Cant see Hydrogen
- Cant see Oxygen
- Need alternative method to fully characterize salts Neutron Total Scattering Experiment on FLiNaK
- Experiment conducted at Oak Ridge National Laboratory Spallation Neutron Source (NOMAD instrument) November 17-19, 2019
- Objective: Obtain neutron scattering data to probe the structure of the salts in both the solid and molten states, enabling quantitative comparison with simulations
- Preliminary analysis is promising
Temperature Dependent Scattering Pattern FLiNaK
- Sharp peaks in the scattering SDWWHUQVIRU&&DQG
&LQGLFDWHORQJ-range structural correlations
&
- Diffuse features in the scattering
& SDWWHUQVIRU&DQGKLJKHU
& indicate short-range structural
&
& correlations (but no long-range
& structure)
&
&
- Consistent with the known
& PHOWLQJSRLQWRI&
&
- Note: Patterns are offset vertically for clarity Temperature Dependent Pair Distribution Function (PDF)
- The PDF is essentially the Fourier FLiNaK transform of the scattering pattern,
& yielding structural information in
& real space rather than reciprocal
& space
&
- Peaks in the PDF indicate the
& presence of well-defined pairs of
& atoms separated by that distance
&
- Long-lived peaks at low
& temperature reveal well-defined,
& long-range correlations in the solid
& state
- The peaks are much broader in the molten state due to the amorphous nature of the liquid
- Note: Patterns are offset vertically for clarity
Temperature Dependent Pair Distribution Function (PDF)
- Negative peak at 1.8 Å originates FLiNaK from Li-F nearest-neighbor pairs Longer-range correlations (negative due to negative
&
&
scattering length of Li)
&
- Na-F and K-F nearest neighbor
& pairs contribute to the peak
& centered around 3 Å
&
- Broad features persist to 8-9 Å,
&
indicated non-random correlations on this length scale
- These experimental patterns can Na-F, K-F correlations be compared quantitatively to MD Li-F correlations simulations (work in progress) to extract detailed structural information
- Note: Patterns are offset vertically for clarity 14 Conclusion
- Salt experiments needed to inform MSR design and licensing processes
- Actinide and FP bearing salts essential for system modeling, but massive in scope
- Thermodynamic assessment informs experiments, reveals correlations
- Trends in ion-ion interactions minimize experimental load
- Improved characterization capability needed: PDF analysis promising
DECEMBER 11, 2019 ONLINE MONITORING OF MOLTEN SALT REACTORS Laboratory Director Argonne National Laboratory NATHANIEL C. HOYT ELIZABETH A. STRICKER JICHENG GUO MARK A. WILLIAMSON Argonne National Laboratory NRC Advanced Reactor Workshop 2019 Paul Kearns INTRODUCTION ONLINE MONITORING OF MSR CHEMISTRY Chemical considerations within fuel salts and secondary salts demand high-fidelity process monitoring capabilities Concentrations of reactants in core (neutronics)
Fission product removal system (in situ processing)
Noble metal deposition monitoring Structural corrosion monitoring Safeguards DOE Gen4 Road Map (downloaded from:
http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf) 2
INTRODUCTION ONLINE MONITORING OF MSR CHEMISTRY Chemical considerations within fuel salts and secondary salts demand high-fidelity process monitoring capabilities Concentrations of reactants in core (neutronics)
Fission product removal system (in situ processing)
Noble metal deposition monitoring Structural corrosion monitoring Safeguards DOE Gen4 Road Map (downloaded from:
http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf) 3 INTRODUCTION ONLINE MONITORING OF MSR CHEMISTRY Chemical considerations within fuel salts and secondary salts demand high-fidelity process monitoring capabilities Concentrations of reactants in core (neutronics)
Fission product removal system (in situ processing)
Noble metal deposition monitoring Structural corrosion monitoring Safeguards DOE Gen4 Road Map (downloaded from:
http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf) 4
INTRODUCTION ONLINE MONITORING OF MSR CHEMISTRY Chemical considerations within fuel salts and secondary salts demand high-fidelity process monitoring capabilities Concentrations of reactants in core (neutronics)
Fission product removal system (in situ processing)
Noble metal deposition monitoring Structural corrosion monitoring Safeguards DOE Gen4 Road Map (downloaded from:
http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf) 5 INTRODUCTION ONLINE MONITORING OF MSR CHEMISTRY Chemical considerations within fuel salts and secondary salts demand high-fidelity process monitoring capabilities Concentrations of reactants in core (neutronics)
Fission product removal system (in situ processing)
Noble metal deposition monitoring Structural corrosion monitoring Safeguards DOE Gen4 Road Map (downloaded from:
http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf) 6
INTRODUCTION ONLINE MONITORING OF MSR CHEMISTRY Chemical considerations within fuel salts and secondary salts demand high-fidelity process monitoring capabilities Concentrations of reactants in core (neutronics)
Fission product removal system (in situ processing)
Noble metal deposition monitoring Structural corrosion monitoring Safeguards DOE Gen4 Road Map (downloaded from:
http tp::/
tp ://w
- //
//w
///
/ ww ne do http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf) 7 ELECTROANALYTICAL MEASUREMENTS VOLTAMMETRY FOR IN-SITU PROCESS MONITORING Voltammetric techniques can be used to monitor actinide concentrations in MSRs -0.02 UU03+/U 3+
/U0 Gd 3+ /Gd 0 3+
- Allows rapid, real-time measurements Gd /Gd0 species concentrations are
- Equipment not affected by high radiation -0.016 related to respective Run 1peak background heights Run 2 Current (A) current (A)
- Wetted materials are salt and temperature -0.012 Run 3 compatible
- Technique based on electrochemical properties and -0.008 3+
Np 0/Np Np 0
/Np 3+
does not require use of standards 4+ 3+
UU /U/U4+
3+
- Well-developed theory for voltammetric response for -0.004 a given redox reaction
- Analyze for multiple components with single 0
-2.4 -1.9 -1.4 -0.9 -0.4 0.1 indicator electrode +
voltage Voltage(V (V.vs. Ag/Ag vs Ag /Ag)+)
- Multiple voltage perturbation waveforms and methods of analyzing resultant current are available1-5 Square-wave voltammogram for molten salt Concentrations and other salt characteristics containing UCl3, NpCl3 and GdCl3 determined from current response to voltage waveforms 1Hill, Perano, Osteryoung, JES 107 (8) 698 (1960) 2Mamantov, Manning, Anal. Chem. 38 (11) 1494 (1966) 3Tylka, et al. JES 162 (9) H625 (2015) 8 4Tylka, et al. JES 162 (12) H852 (2015) 5Hoyt, et al. JES 164 (2) H134 (2017)
SENSOR DEVELOPMENT MULTIELECTRODE ARRAY SENSOR The multielectrode array provides a flexible platform for a wide range of voltammetry methods combined with quasi-reference and dynamic reference electrode measurements.
- Salt redox potential
- Species concentrations
- Salt level Advantages
- Fast measurement rates
- No moving parts
- Wide potential range
- Long electrode service life
- Tolerant of thermal cycling 9
SENSOR DEVELOPMENT FUNDAMENTAL SALT PROPERTIES Fundamental salt properties including diffusion coefficients and kinetics parameters must be experimentally determined for each species of interest in a given salt across a complete range of temperatures.
Typical measurements during Cr2+ Experimentally determined diffusion Parity plot comparing measured and and Fe2+ property measurements coefficients for Cr2+ and Fe2+ versus known Cr2+ concentrations in MgCl2-testing temperature for MgCl2-KCl-NaCl KCl-NaCl 10
SENSOR DEVELOPMENT FUNDAMENTAL SALT CHEMISTRY Electroanalytical techniques and supporting salt properties have been established for a variety of species in several key salts Actinides Fission Products Corrosion Impurities Corrosion Products Once suitable techniques have been developed, fundamental investigations of salt chemistry can be readily performed.
MgOHCl concentration versus time during decomposition reaction in MgCl2-KCl-NaCl 11 SENSOR DEVELOPMENT ELECTROANALYTICAL SENSOR CHALLENGES Challenges exist for the use of electroanalytical sensors in real molten salt systems Electrical Noise Highly Concentrated Salts Non-ideal Electrochemical Behavior Materials Stability Seals Automation of Acquisition Automation of Analysis Representative electroanalytical Representative electroanalytical measurement in molten salt loop with measurement in molten salt loop with significant interference from successful isolation of electrical noise resistance heaters 12
LONG-DURATION TESTING FLOW LOOP INTEGRATION Versions of the sensors suitable for loop Sensor Location operations were constructed and installed into thermal convection loops (TCLs) at ORNL Materials compatibility tests conducted at maximum specified hot leg temperature (750 °C)
Sensor provides long-term quantitative measurements of salt potential, metal ion corrosion products, and impurities Rendering of TCL loop with indicated location of sensor installation1 1HARP rendering courtesy 13 of Bruce Pint (ORNL)
LONG-DURATION TESTING FLOW LOOP INTEGRATION The sensor has been providing measurements throughout the >650 hour test. Principal measurements include salt potential, Cr2+ concentration, MgOHCl concentration, H+ concentration, and the salt depth.
Salt redox potential versus time Cr2+ concentration versus time MgOHCl concentration versus time during TCL loop operations during TCL loop operations during TCL loop operations 14
LONG-DURATION TESTING SENSOR PERFORMANCE ASSESSMENT A 100 day trial sequence was conducted to assess the long-term stability, accuracy, uncertainty, and longevity of the sensors in a molten salt electrorefiner.
The experimental sequence included electroanalytical measurements and in situ electrochemical procedures to maintain electrode Rendering of surfaces. electrorefining system for uranium and U/TRU co-deposition 15 LONG-DURATION TESTING VOLTAMMETRY DURING LONG-DURATION TESTING With proper procedures, long-duration service life tests demonstrated highly repeatable electroanalytical measurements over the 100 day testing period (and well beyond).
With proper electrochemical conditioning, in situ monitoring of the electrode health showed no film formation or material degradation. Electroanalytical measurements taken over EIS Nyquist plots for electrode 100 day service life test in the pilot-scale health monitoring (20 mV electrorefiner (500 mV/s, with indicated perturbation, 100 kHz to 100 Hz) immersion times) 16
LONG-DURATION TESTING CONCENTRATIONS DURING LONG-DURATION TESTING Measurements of the salt potential were consistent over the entire 100 day duration (mean absolute deviation less than 15 mV over entire test).
Measurements of the UCl3 and CeCl3 concentrations were also consistent over the entire test (relative standard deviation over the entire test of 1.08% and 1.10% for UCl3 and CeCl3 respectively; 0.13% and 0.60% on a daily basis)
Salt potential versus time during long- Species concentrations versus time during long-duration test as measured by quasireference duration test as measured by quasireference using Li+/Li0 as an internal standard using Li+/Li0 as an internal standard 17 LONG-DURATION TESTING CONCENTRATIONS DURING LONG-DURATION TESTING Measurements of the salt potential were consistent over the entire 100 day duration (mean absolute deviation less than 15 mV over entire test).
Measurements of the UCl3 and CeCl3 concentrations were also consistent over the entire test (relative standard deviation over the entire test of 1.08% and 1.10% for UCl3 and CeCl3 respectively; 0.13% and 0.60% on a daily basis)
Salt potential versus time during long- Species concentrations versus time during long-duration test as measured by quasireference duration test as measured by quasireference using Li+/Li0 as an internal standard using Li+/Li0 as an internal standard 18
LONG-DURATION TESTING CONCENTRATIONS DURING LONG-DURATION TESTING Measurements of the salt potential were consistent over the entire 100 day duration (mean absolute deviation less than 15 mV over entire test).
Measurements of the UCl3 and CeCl3 concentrations were also consistent over the entire test (relative standard deviation over the entire test of 1.08% and 1.10% for UCl3 and CeCl3 respectively; 0.13% and 0.60% on a daily basis)
Salt potential versus time during long- Species concentrations versus time during long-duration test as measured by quasireference duration test as measured by quasireference using Li+/Li0 as an internal standard using Li+/Li0 as an internal standard 19 LONG-DURATION TESTING AUTOMATED ACQUISITION AND ANALYSIS raw electroanalytical signals experiment duration and most recent update species time concentrations versus time system status histogram of daily concentration measurements electrode status salt potential histogram of daily versus time potential measurements 20
LONG-DURATION TESTING AIR INGRESSION MONITORING SALT REDOX POTENTIAL CORROSION RATES A variety of realistic scenarios including air ingression tests have been performed to demonstrate the ability of the sensor to indicate off-normal conditions
- Changes in the salt potential were immediately detected as impurities entered the salt
- Increased corrosion rates could be calculated from the measurements and were highly correlated to the condition of the atmosphere Salt potential and glovebox O2 Fe corrosion rate and glovebox O2 concentration versus time concentration versus time 21 LONG-DURATION TESTING CORROSION MONITORING AND CONTROL Operation of the salt monitoring potential too high capabilities can be used in tandem (structural metal corrosion) with corrosion control capabilities to maintain the salt in a healthy state control band The salt redox potential was properly potential too low controlled during a trial 20+ day trial (reactive metal flooding) that included air ingression testing
- The rate of actuation of the corrosion control system increased in response to impurities in the glovebox atmosphere
- No corrosion products were detected by the voltammetry sensor during the course of the test 22
ONLINE MONITORING OF SALT CHEMISTRY SENSOR FUSION Electrochemical sensors provide a wide array of information, but there are some chemical measurements they cannot observe.
Solids Off-gas Etc.
A variety of sensor technologies are under development across the national laboratory complex to close these gaps.
Fusion of all these sensing capabilities will provide a full picture of the MSR chemistry. DOE Gen4 Road Map (downloaded from:
http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf)
One example of a complementary technology for corrosion monitoring are the eddy current corrosion sensors under development at ORNL.
24
25 26
27 CONCLUSIONS AND FUTURE WORK Robust electroanalytical sensors have been demonstrated for durations longer than a year and a half. They provide high stability, longevity, and accuracy combined with very low uncertainty.
Future work will be directed toward:
Application of the sensor to additional nuclear-relevant fluoride and chloride salts Shakedown testing in noisy EMI environments Standardization of sensor construction Forced convection studies Sensor fusion will provide process monitoring capabilities for reactor control, safeguards, material accountancy, and corrosion control applications 28
GOVERNMENT LICENSE NOTICE The submitted manuscript has been created by UChicago Argonne, LLC, Operator of Argonne National Laboratory (Argonne). Argonne, a U.S. Department of Energy Office of Science laboratory, is operated under Contract No. DE-AC02-06CH11357. The U.S. Government retains for itself, and others acting on its behalf, a paid-up nonexclusive, irrevocable worldwide license in said article to reproduce, prepare derivative works, distribute copies to the public, and perform publicly and display publicly, by or on behalf of the Government.
Overview of Environmental Issues and Material Property Gaps for Commercial Viability of Advanced Reactors Mike Burke David Gandy Technical Executives EPRI Nuclear Sector International Workshop on Advanced Reactor Materials and Component Integrity Rockville Md.
Tuesday December 10th 2019 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Development and Validation for Advanced Reactor Applications In-reactor, In-environment Reactor Structures Mechanical Response, Life Prediction and IASCC, Rad. Fatigue Safety Validation Data Stress-Corrosion Cracking Behavior of Previously Irradiation Effects on Lifing Irradiated Material Parameters Environmentally Affected Cyclic Deformation Response Structure/Component
- Stress Amplitude vs Life, Crack Growth Rates etc. Irradiation Effects on Creep :
Lifing Design Data Creep Rates, Creep Life Mechanical Properties, Stress Corrosion in Environment Cracking Stress vs Life, Crack Growth Rates Irradiation Effects on Material Viability/
Mechanical Properties Selection for Cyclic Properties Fatigue Life Cyclic / Curves Components Irradiation Effects on Microstructure Time Dependent Corrosion Behavior Swelling, He Production, Matrix Mechanical Properties Weight Loss, IGA Attack Hardening, Phase Changes Licensing and Environmental Effects Irradiation Effects Qualification Mechanical Properties Construction Code Data New Materials Development: Composition Control, Processing, Microstructural Optimization Scale Up, Standardization and Commercial Availability 2 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Data Generation and Material Testing for Structures in Advanced Reactors In-reactor, In-environment Long Time Fleet Performance Life Prediction from Mechanical Response, From Monday : Comparison IASCC, Rad. Fatigue Assessment of Aged Materials with Light Water Reactors via representative lab Stress-Corrosion Cracking testing Irradiation Effects on Lifing Behavior of Previously Parameters Irradiated Material Moderate-time Multi-unit Environmentally Affected Cyclic Performance data
? Deformation Response : Stress Amplitude vs Life, Crack Growth Rates etc. ? Irradiation Effects on Creep :
Creep Rates, Creep Life Laboratory test data for interacting variables generating durability data for Time Dependent Cyclic Irradiation Effects on lifing Stress Corrosion Cracking Properties Mechanical Properties Stress vs Life, Crack Growth
? Rates Mechanical Properties,
? Prototype Performance
? in Environment Data Intermediate-time durability test data for Time Dependent Corrosion Behavior simple effects Irradiation Effects on Mechanical Properties Weight Loss, IGA Attack Microstructure Phase Changes Construction Code Data with margin of safety on Mechanical Properties Environmental Effects Irradiation Effects short time durability New Materials Development: Composition Control, Processing, Microstructural Optimization data 3 www.epri.com
©Properties 2019 Electric Power Research Institute, Inc. All rights reserved.
How Well Can We Follow The LWR Paradigm To Support The Development Of Advanced Reactors? (2)
Lots of white space compared to the data that supports LWR reactors
- Immediate need for new materials/properties to be developed, validated and standardized High temperatures, irradiation and corrosion data
- Simple properties are available to support concepts - but there are many gaps in supporting data - long time data, time dependent properties, severe environment service
- Progression from properties under combined conditions-Stress + Environment, Cyclic at Temperature, Strain +
Irradiation Without these data, life estimates cannot be performed without life estimates for structures, assessments of economic viability cannot be performed 4 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Gaps identified for Advanced Reactors (MSR, LFR, VHTR/GFR, SFR) in EPRI Material Gap Studied - 1 of 2 GEN IV REACTOR COMPONENT MATERIAL NEEDED R&D Molten Salt Reactors Core Support/ 316 Proof of resistance to long time corrosion in properly controlled salt environment. Time dependent Structural Materials properties for ASME code Sec III Div 5 certification. Demonstration of performance (resistance to EAC) in salt under loading.
Stainless Steels Ferritic-Martensitic & LAS Development and demonstration of cladding (Mo rich) for protection Nickel-based Alloys Hastelloy N and Demonstration of radiation tolerance of Hast N variants (Proper understanding of chemistry Other (Graphite, Ceramics) variants microstructure properties Development of properties for ASME code Sec III Div 5 certification Corrosion Cladding Coolant Salt Development of salt chemistry (and impurity) control. Demonstration of Te control Moderator Graphite Development of long time properties in salt etc. for the specific type of graphite to be employed High temperature Core Support/ 316 and Code approval of time dependent properties - creep, creep-fatigue Gas Reactor Structural Austenitic Alloys Materials 316FR Code qualification properties for ASME code Sec III Div 5 for 316FR including time dependent properties Vessel LAS Time dependent and fatigue properties for ASME code Sec III Div 5 Moderator Graphite Development of long time properties etc. for the specific type of graphite to be employed Na SFR Vessel and Core 316 Stainless Extend code properties to include time dependent behavior (Creep. Creep fatigue)
Support Structure D9 Development of for ASME code Sec III Div 5 properties (including time dependent properties) for D9 Development of swelling behavior at long times under realistic conditions - demonstrate adequacy Core Support Ferritic Prove adequacy of swelling resistance at high fluence Structure and Martensitic Development of fabrication technology and proof pf performance of welds Cladding 5 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Gaps identified for Advanced Reactors (MSR, LFR, VHTR/GFR, SFR) in EPRI Material Gap Studied - 2 of 2 GEN IV COMPONENT MATERIAL NEEDED R&D REACTOR GFR Core Ferritic Martensitics Demonstration of adequate resistance to swelling at high dpa. Time dependent properties for support ASME code Sec III Div 5. (include development of fabrication technologies - and demonstrate properties of joints)
Cladding Ceramics For advanced GFR - SiC-SiC, Zr3Si need materials endurance data for these materials and reflector Lead Fast Structural 316 (code qualified already) but need creep and creep fatigue data to be added into code.
Reactor Materials/ Need corrosion data/demonstration of resistance to lead corrosion Stainless Steels Vessel Type 15-15Ti stainless Verification of swelling resistance Ferritic-Martensitic &
Development of code properties for 15-15Ti material design LAS Nickel-based Alloys Near core Ferritic Martensitics Demonstration of adequate resistance to swelling at high dpa.
Other (Graphite, structures Time dependent properties for ASME code Sec III Div 5. (include development of fabrication Ceramics)
Corrosion and technologies - and demonstrate properties of joints)
Cladding cladding Demonstration of resistance to lead corrosion/development of corrosion data Development of fabrication and effective joining methods High Structural Alumina Forming Austenitic Demonstration of resistance to lead corrosion Temperature Materials/ Stainless Steels Demonstration of adequate resistance to irradiation/swelling at expected high dpa Lead Reactors Vessel Development of processing and joining of alumina forming austenitic stainless steels Cladding SiC-SiC Development of SiC-SiC structures Demonstration of resistance to lead corrosion Development of properties and support to code qualification 6 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Metallic Materials Environmentally Related Properties Needed for Design and Development of Advanced Reactors Austenitic Stainless Steels Long time, high dose performance in sodium environment Long time high dose performance in He environment Gas & Na Ferritic-Martensitic Long time, high dose performance in sodium environment Long time high dose performance in He environment Reactors Demonstration of long time resistance to molten salt Demonstration of sustained mechanical behavior in molten salt (Salt equivalent of SCC)
Demonstration of sustained mechanical behavior of CW/Irrad. material mechanical behavior in molten salt 316 Data/demonstration of resistance to lead corrosion Sustained mechanical behavior in molten lead Proof of resistance to long time corrosion in properly controlled salt environment.
316H Demonstration of performance (resistance to EAC) in salt under loading.
Alumina Forming SS Demonstration of resistance to Lead corrosion Demonstration of resistance to molten salt corrosion Molten Determination of the effects of potential EAC Salt Extended corrosion data/demonstration of resistance to molten salt Reactors Hast N & Variants Demonstration of sustained mechanical behavior in molten salt (Salt equivalent of SCC)
Demonstration of sustained mechanical behavior of CW/Irradiated material mechanical behavior in molten salt Proof of resistance to long time corrosion in properly controlled salt environment.
Alloys 800H and 617 Demonstration of performance (resistance to EAC) in salt under loading.
Comparison of behavior vs 316 and Hastelloy N Ferritic-Martensitic 9Cr Demonstration of resistance to lead corrosion/development of corrosion data Lead Cooled Demonstration of sustained mechanical properties in lead Ferritic-Martensitic 12Cr Demonstration of resistance to lead corrosion/development of corrosion data Reactors Demonstration of sustained mechanical properties in lead 7 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Materials Properties Data needed to Support Reactor Design Commitments Commitment to build ?
Time dependent mechanical properties of non-irradiated material in environment Time dependent mechanical properties of non-irradiated material in environment Corrosion behavior in Confidence neutron environment Mechanical properties of in reactor non-irradiated material design in environment process Basic design(code)
Corrosion behavior of properties non-irradiated material 8 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
Development of Materials Technology for Advanced Reactors Design and Development Three categories of development needed:
- Materials properties to support initial design and to attain ASME code acceptance for constructions - including high temperature and time dependent properties
- Response of candidate materials to neutron irradiation - effects of realistic levels of irradiation on microstructural and property stability
- Materials response in environment - effects measured in realistic environments: stand alone & effects on mechanical behavior (equivalent to IASCC)
Initial Focus on development of code required material properties
- Support for design and build of prototype(s) Å Simple properties
- Initial design for short life (predict prototype performance) Å Analysis incorporates some time dependent property data
- Development of extended time dependent properties & properties of appropriately aged and exposed materials (T,t + n,t)
Future Needs 3/4 Need for better knowledge of the chemical composition of coolants after significant operations of the advanced reactors provide a basis for environmental testing 3/4 More knowledge of the critical variables in the most significant degradation mechanisms of materials 3/4 Rig and loop testing in simulated environments - on previously neutron exposed materials 3/4 Value of prototyping and post-mortem analyses to identify key assessment variables A clear need for investment in materials testing facilities Specialized test rigs, dedicated loops and reactors for irradiation exposures 9 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
TogetherShaping the Future of Electricity 10 www.epri.com © 2019 Electric Power Research Institute, Inc. All rights reserved.
LIQUID METAL SYSTEMS SODIUM Outlet temp ~550°C SODIUM Operationally oxygen controlled by a cold trap at 2-3 ppm (up to 10 ppm during maintenance)
- Cr has little solubility as a metal and the transport of chromium from the metal surface to form an oxide in the sodium is a stronger driving force.
- Molybdenum is unlikely to form an oxide at typical sodium reactor operating temperatures, and with limited solubility, is likely to stay in the steel.
- Nickel oxides are not thermodynamically stable so Ni loss from the steel surface is driven by the difference in the solubility of nickel in sodium and the amount of nickel in the alloy.
- Iron does not form a thermodynamically favorable oxide yet corrosion data for pure iron indicates that the iron dissolution increases with the square of the oxygen concentration, indicating a kinetic effect in which oxygen promotes the local dissolution of iron into sodium SODIUM MONJU adopted corrosion rate
SODIUM An important observation after the sodium was drained from the primary tank was that the condition of the tank and the components submerged in sodium was pristine. There was absolutely no corrosion of the stainless steel after 35 years in contact with hot sodium.
Dr. John Sackett Testimony to the NRC in December 2008 Transport of carbon and nitrogen from the reactor vessel (proposed to be constructed from 316 stainless steel), and associated decrements in strength was noted as an open question for General Electric as part of the Nuclear Regulatory Commission review of the proposed PRISM sodium-cooled reactor design LEAD ALLOYS Outlet temp up to ~800°C
LEAD ALLOYS Pb and Pb-Bi are both in consideration Si-containing alloys provide a more protective oxide Al-containing coatings are being studied in Europe for higher temperature application LEAD ALLOYS
- At very low oxygen, both austenitic and ferritic-martensitic steels are subject to dissolution, even at low temperature.
- From 300°C to 470°C, with sufficient oxygen (>10 4 ppm), protective oxide films can be formed on both austenitic and ferritic-martensitic steels.
- For temperatures above 550°C, austenitic stainless steels undergo heavy dissolution and ferritic-martensitic steels form a very thick and potentially unstable oxide. This thick oxide may be susceptible to erosion at high flow rates.
- Between 470°C and 550°C, the corrosion behavior in structural steels appears to make transition from oxidation to dissolution. Furukawa et al. determined that at these higher temperatures, the iron oxide form changed from magnetite to wustite which is less adherent and thus more prone to detachment
LEAD ALLOYS
- Tensile properties appear to be unaffected for both T91 and 316 L steels when exposed to oxidizing LBE but does lose some ductility when exposed to an LBE under reducing conditions.
- Low cycle fatigue in 316 L contacted with LBE shows only a weak damaging effect but T91 shows a decrease in low cycle fatigue resistance. If the T91 is brought to reducing conditions, the decrease in low cycle fatigue growth is greater, supporting the idea that reducing conditions appear to be detrimental to mechanical properties.
- Creep-rupture tests in flowing LBE at 550°C showed a marked acceleration of creep rate at stresses >180 MPa.
Overall, the mechanical properties of austenitic steels appear to be little affected but those of ferritic-martensitic steels can be affected and must be chosen carefully.
COMPARISON The U.S. has operated multiple sodium-cooled demonstration plants and environmental effects from the coolant was not limiting. Oxygen control was critical The U.S. has only operated test loops to understand liquid lead alloy systems but the Russians did operate submarines using lead alloys for the coolant.
REFERENCES BACKUPS
WHAT DOES CORROSION LOOK LIKE Corrosion Effects in Materials in High Temperature Gas-Cooled Reactor (HTGR)
Environments Kumar Sridharan Department of Engineering Physics Department of Materials Science & Engineering University of Wisconsin, Madison e-mail: kumar.sridharan@wisc.edu Advanced Non-Light Water Reactors - Materials and Component Integrity Workshop Nuclear Regulatory Commission, Rockville, MD December 9th to 11th, 2019 1 Highh-Temperature h -Tem mperature Gas mp s-Cooled s - Reactor (HTGR)) - Background Uses helium gas as primary coolant Operating temperature ~ 750oC to 950oC Graphite moderator TRISO fuel particles in pebble bed or prismatic configuration Byproduct heat for, e.g., chemical plants, hydrogen production, desalination Five decades of building and operating experience (Dragon, UK; Ft. St. Vrain, US; AVR, Germany) 2
HTGR R - ASME Code Certified Materials ASME Section III Division 5 Framework for Nuclear Components (Part II) allows for the use of only the following six alloys Alloy 800H codified for use up to 750oC for 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Alloy 617s code case just completed for operation up to 954oC for 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 3 Dr. Richard Wright, INL Environmental Effects in Alloys in HTGR Helium Helium is inert - should not corrode alloys However, trace levels (few ppm) impurities of H2O, CH4, CO,CO2, H2, O2 in helium can induce corrosion in alloys Effect of impurities in helium on corrosion of structural alloys studied since the 1970s Ground-breaking papers in the 1980s (and subsequently) that provide useful guidance for the path forward 4
Impurity Regimes Regim of Corrosion in HTGR Helium Zone 1: Reducing environment (may be decarburization)
Zone II: Strongly oxidizing Zone III: Stable oxide and stable internal carbides in the alloy (most desirable)
Zone IV: Strongly carburizing Zone IVa: Carburizing and oxidizing (mixed phases) ppm levels of impurity shifts can alter the corrosion mechanism Cr is the most important participant in the corrosion reactions (present in 19% to 24% in 800H and 617) 5 Dr. Richard Wright, INL, INL/EXT-06-11494 Where do impurities come from?
Small amounts of H2O and O2 in the He entering the core O2 reacts with hot graphite to form CO Some H2O reacts with the hot graphite to form CO and H2 CO2 degassing from graphite converts to CO Corrosion reactions with alloys may also produce H2 and CO CH4 can come from leakages of oils Radiolytic reaction of H2 with graphite can lead to CH4 formation 6
L. Graham, UK and K. Natesan et al, ANL, NUREG/CR-6824
Primary Mechanisms of Corrosion and Impurities Involved Oxidation: H2O and CO Carburization: CO and CH4 Decarburization: H2O Need to form a thin, dense, adherent, thermodynamically and mechanically stable oxide layer which prevents further oxidation, and acts as a barrier for carburization and decarburization 7 K. Brenner and L. Graham, UK, 1984 Primary Mechanisms Me of Corrosion n-Oxidation Oxidation: H2O and CO 2/3Cr + H2O = 1/3 Cr2O3 + H2 Possibly protective Cr2O3 CO + 3Cr = 1/3 Cr2O3 + 1/3 Cr7C3 Alloy 617 exposed to oxidizing conditions Not be protective due to co- (1000oC/1000 hours);
internal oxidation formation of Cr7C3 due to Al Micrograph Dr. Richard Wright, INL, INL/EXT-06-11494 8
Primary Mechanisms Mecha of Corrosion n-Carburization Carburization: CO and CH4 7/3Cr + CH4 = + 1/7 Cr7C3 + 2H2 Carburized layer in Alloy 800H -2500 hrs, 1000oC (H. Inouye, ORNL, 1983)
CO + 3Cr = 1/3 Cr2O3 + 1/3 Cr7C3 Carburization can lead to lead to loss of ductility and toughness Mixed carbide/oxide layer in Alloy 617 -10,000 hrs, 900oC (Brenner, Germany, 9
1983)
Primary Mechanisms Mechanis of Corrosion n-Decarburization Normal Decarburized microstructure with layer carbides at GBs Decarburization: H2O Strengthening carbide phases in alloy (typically M23C6 and M6C) can get decarburized M-carbide + H2O = M + CO + H2 Alloy 800H subjected to decarburizing Decarburization leads to loss of environment 870oC, 2500hrs (Cappelaere, creep-rupture strength France, 1984) 10
Conceptual Ternary Env Environmental Attack (TEA) Diagram Decarburizing Protective Oxide Regimes of corrosion in various regimes of the three most Carburizing significant impurity species 11 Work on Nimonic alloys (Ni-based, similar to Alloy 617) K. G. Brenner, UK Consequenc Consequences of Corrosion n - Mechanical Properties No instances of mechanical failures due to corrosion in HTGRs; most experimental studies performed by using accelerated carburization or decarburization conditions CH4 injected in helium 900oC for 3000hs, post-tensile test cracks (Li, US)
Creep-rupture tests (617);
Cracks initiate in carburization has no effect but carburized layer (SCC could decarburization does at 1000oC 12 also be a possibility) (Ennis, Germany)
Consequ Consequences equ uences e of Corrosion n - Wear W Behavior (NEUP P - University of Wisconsin)
Wear and friction behavior between rubbing components (e.g, valves) can vary dramatically between the various corrosion regimes 13 K. Sridharan PI, University of Wisconsin NEUP; D. Singh ANL (collaborator); Dr. Sam Sham TPOC Consequences of Corro Corrosion orro ro ossionn-W Wear Behavior: Allo Alloys 800H and 617 (NEUP P - University of Wisconsin) 800H 617 Once through He flow loop (impurity High temperature pin-on-control at ppm level, UW disk wear test (ANL)
Wear volume Friction measurements measurements Formation of a nanocrystalline glaze oxide layer lowers friction and wear 14 University of Wisconsin - ANL NEUP
Conseque Consequences que ue ences of Corrosion & Environmental Environ ron nmen m
Effects s-E Emissivity Changes Cha (NEUPP - U.
Wisconsin/U. Missouri) 1 In-617 in 0.8
( , n,T=300C) 1000oC He 0.6 Oxidizing He Env.
for 500 0.4 Carburizing hours He Env.
0.2 Radiation heat (samples Ref.
0 transfer(Q) from INL 3 6 9 12 15 18 21 important in helium loop, Wavelength ( m)
HTGR Dr. Wright)
Q = K.HT4 Stefan-Graphite Boltzmann deposition equation, His on metallic emissivity and T surfaces is temperature can change emissivity 15 University of Wisconsin - University of Missouri NEUPs; Dr. Sam Sham TPOC Concluding Remarks Controlling total impurity content in helium to below 10ppm should be targeted*
Molecular sieves are effective, but they cannot capture CO and H2; gas flowed over CuO to convert CO to CO2 and H2 to H2O*
Back streaming of oils must be minimized*
Graphite in HTGR core plays a dominant role in corrosion*
Effect of flow-velocity and system pressure should be considered in corrosion prediction Data mining and analysis of the large amount of literature on HTGR-He corrosion of 800H and 617 will be very valuable Since the alloy composition cannot be altered, surface modification approaches to promote passive oxide layer formation (making corrosion immune to the regime) is an attractive option 16 *Dr. Richard Wright, INL, INL/EXT-06-11494
Chemical Reactions in Gas Phase Cont Controlling Carbon Activity and Oxygen Potential Carbon activity results from reactions such as:
CO + H2 = [C] + H2O ac= k (pCO . pH2)/pH2O 2CO = [C] + CO2 CH4 = [C] + 2H2 CH4 + CO = [2C] +H2O + H2 CO = [C] + 1/2O2 Oxygen potential is determined by reactions such as:
H2O = 1/2O2 + H2 pO2= [k pH2O / pH2)]2 CO2 = 1/2O2 + CO 17 K. Natesan et al, ANL, NUREG/CR-6824
Corrosion in molten salts: What matters, what doesnt, and what we can do about it.
Stephen Raiman R&D Associate Materials Science and Technology Division Supported by:
- DOE Molten Salt Reactor Campaign
- US NRC
- ORNL Laboratory Directed R&D ORNL is managed by UT-Battelle, LLC for the US Department of Energy This is corrosion (most of the time)
Environment Outer oxide Metal Kinetic Oxygen Ions Oxide Metal Barrier Metal 2
Materials in molten salts degrade by chromium depletion S.S. Raiman, Solar Energy Materials and Solar Cells 201 (2019) 3 Stability of Metals in Chloride Salts Redox condition 4
Chromium reacts at salt-metal interface, diffuses outward from bulk, and leaves porous layer Porous Bulk Region Salt
- Salt impurities react with structural alloys Interfacial Cr Diffusion reactions 2HCl(d) + Cr(s) CrCl2(d) + H2(d)
- Metallic halides Cr Cr(s) + FeCl2(d) CrCl2(d) + Fe(s)
- Fuel Cr(s) + 2UCl4(d) CrCl2(d) + 2UCl3(d)
Similar reactions in F salt, just replace Cl with F 5
Our objective is to understand degradation of structural materials in molten salts, and turn that understanding into actionable engineering knowledge
- Optimized chemistry-material choices
- Accurate models of in-reactor behavior
- Development of advanced materials
- Longer lifetimes & improved economics
- Prudent and informed regulatory decisions 6
What matters?
7 This data is a mess Raiman, Lee, J. Nucl. Mater. 511 (2018) 8
Container material doesnt seem to matter Raiman, Lee, J. Nucl. Mater. 511 (2018) 9 Salt purity seems to matter Raiman, Lee, J. Nucl. Mater. 511 (2018) 10
Experimental Techniques Providing the evidence basis for good decisions 11 Capsule testing Capsules filled with KCl-MgCl2 and model Ni-Cr alloys for 700°C tests 12
Mass change in MgCl2 is generally higher than in KCl-MgCl2 Cr samples Mo capsules Raiman (unpublished) 13 Flowing salt with a thermal gradient is required to accurately reproduce MSR conditions
- Temperature-dependent equilibrium constants drive mass Hot Hot Cold transfer Leg Leg Mass transport Leg
+
Hot leg: reaction moves to right (depletion) Dissolution Deposition Liquid Liquid Cold leg: reaction moves to left (deposition)
Containment Containment 14
Why a thermal gradient matters Computed equilibrium concentration of CrCl2 in KCl-MgCl2 based on Cr + MgCl2 (liq &U&O2 (liq) + Mg (liq)
Pint et al. Mater and Corros (2018) 15 15 Thermal convection loops at ORNL Fill tank
- Salt is introduced at top
- Flow is driven by thermal Overflow gradient tank
- Coupons are hung in both the hot leg and cold Hot leg Cold leg leg Drain tank 16
Recently completed TCL experiment
- 650°C max temp, 540°C minimum
- FLiNaK salt
- 316 tubing
- 316 samples 17 TCLs have been run at Oak Ridge since the Aircraft Reactor Experiment Manly et al. ORNL-2349 (1957) 18
Mass transfer is visible on TCL samples 316 TCL with purified FLiNaK salt Cold Leg Coupon 1.0 Tensile Spacer Specimen Mass Change (mg/cm2)
Hot Leg 0.5 Coupon Tensile 0.0 Spacer
-0.5
-1.0
-1.5
-2.0 540 560 580 600 620 640 660 T (°C)
T 19 Preparing good salt 20
As-received anhydrous MgCl2 is a mixture of mono- and di-hydrate Phases identified:
MgCl2
- 2(H2O) - 78.3%
Intensity (r.a.u.)
MgCl2 * (H2O) - 21.8%
2 D. Sulejmanovic (unpublished) 21 This is a big problem (because salt chemistry matters) 22
Chemical chlorination successfully dehydrates MgCl2 Phases identified:
MgCl2 Intensity (r.a.u.)
23 2 D. Sulejmanovic (unpublished)
Chloride salts purified by chemical dehydration and chemical chlorination
- Dehydration with NH4Cl:
- () + ()
- Chlorination:
- + + + +
- + +
Salt purity must be maintained under all handling and operating conditions 24
Fluoride salt preparation
- Hydrofluorination process
- Capacity of ~ 3.3 L (6.6 kg) per batch
- Ability to produce a variety of salts including fuel salts and FLiBe 25 Source: K. Robb So now that weve got that right 26
Salt purification matters Stainless Steel Hastelloy N 700°C 700°C 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Treatment 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Treatment before before fusing fusing KCl+MgCl2 KCl+MgCl2 No treatment prior to fusing KCl+MgCl2 No treatment prior to fusing KCl+MgCl2 27 Capsule material? Matters 700°C Cl salt Cl salt Cl salt Cl salt 500h Purified Cl salt aggressive FLiNaK Sulejmanovic et al. (unpublished) 28
Moisture? Redox control additives?
They matter too Additives Additives 316 Stainless Chloride salt with:
- Metallic additions
- 0.1% H2O
- Both
- Neither Raiman et al. (unpublished) 29 Conclusions (and a commercial)
- Consistent experiments are important
- Control variables
- Salt chemistry matters most. When its good, other data emerges
- DOE can provide characterized salts for research under certain circumstances 30