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Fluoride-Salt-Cooled High Temperature Reactor (FHR) Materials, Fuels and Components White Paper

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Fluoride-Salt-Cooled High Temperature Reactor (FHR) Materials, Fuels and Components White Paper FHR Materials, Fuels and Components White Paper Integrated Research Project Workshop 3 Fluoride-Salt-Cooled High Temperature Reactor (FHR) Materials, Fuels and Components White Paper Panel Members Todd Allen (University of Wisconsin, Madison) Gordon Kohse (Massachusetts Institute of Technology) Mark Anderson (University of Wisconsin, Madison) Regis Matzie (ret. Westinghouse Electric Company) Ron Ballinger (Massachusetts Institute of Technology) Matthew Memmott (Westinghouse Electric Company) Tim Burchell (Oak Ridge National Laboratory) Jim Nestell (MPR Associates) Edward Blandford (University of New Mexico) Michael Patterson (Idaho National Laboratory) Denis Clark (Idaho National Laboratory) Scott Penfield (Technology Insights) William Corwin (Department of Energy) Per Peterson (University of California, Berkeley) George Flanagan (Oak Ridge National Laboratory) Sam Sham (Oak Ridge National Laboratory) Charles Forsberg (Massachusetts Institute of Technology) Lance Snead (Oak Ridge National Laboratory) Peter Hosemann (University of California, Berkeley) Kumar Sridharan (University of Wisconsin, Madison) David Holcomb (Oak Ridge National Laboratory) Carl Stoots (Idaho National Laboratory) Lin-Wen Hu (Massachusetts Institute of Technology) Richard Vollmer (Westinghouse Electric Company) John Hunn (Oak Ridge National Laboratory) David Williams (Oak Ridge National Laboratory) Victor Ignatiev (Kurchatov Institute) Dane Wilson (Oak Ridge National Laboratory) UW Madison Facilitators MIT Facilitators UC Berkeley Facilitators Guoping Cao David Carpenter Anselmo Cisneros Tyler Gerczak Becky Romatoski Lakshana Huddar Brian Kelleher John Dennis Stempien Raluca Scarlat Guiqiu Zheng Nicolas Zweibaum UCBTH-12-003 Final July 2013 Department of Nuclear Engineering and Engineering Physics University of Wisconsin, Madison This research is being performed using funding received from the U.S. Department of Energy Office of Nuclear Energy’s Nuclear Energy University Programs. Preamble The University of California, Berkeley; Massachusetts Institute of Technology; and University of Wisconsin, Madison, hosted a series of four workshops during 2012 under a U.S. Department of Energy-sponsored Integrated Research Project (IRP) to review technical and licensing issues for fluoride-salt-cooled, high-temperature reactors (FHRs). The focus of the third workshop was to discuss key fuel and materials needs unique to FHRs, as well as methods for tritium and beryllium control. FHRs deliver heat at temperatures in the range of 600°C to 700°C, and thus structures and components must be designed for high-temperature service where time-dependent phenomena are important. Recently, closely related progress has been made in high-temperature design and materials as a part of U.S. research for the Next Generation Nuclear Plant and for liquid metal reactors. This white paper reviews the current status of this work, and identifies additional work needed to support FHR development. The four workshops are a central element of developing an FHR preliminary conceptual design report to be completed in 2014. This third white paper focuses on material covered by the third workshop and is divided into six chapters. The first chapter provides an overview and discusses the unique environmental conditions that FHR structures and components experience. The second, third, and fourth chapters review FHR fuels, ceramic materials, and metallic materials respectively. The fifth chapter reviews FHR salt corrosion and chemistry control. The sixth chapter reviews tritium and beryllium control issues for FHRs. The comments of the experts attending the workshop were also integrated into this white paper. The IRP team sincerely appreciates the input of all of the experts who attended and contributed to the third FHR workshop, as well as the hard work of the graduate students and postdoctoral scholars who organized it, facilitated the sessions, and drafted the major sections of this white paper based on their research and the review and input of the workshop experts. FHR Materials, Fuels and Components White Paper 2 | 163 Executive Summary The fluoride-salt-cooled, high-temperature reactor (FHR) is a new reactor concept that uses a novel combination of fuel, coolant and materials: graphite-matrix, coated particle fuel; fluoride salt coolant; graphite moderator materials; and high-temperature metallic structural materials. This white paper provides a review of the results from a two-day expert workshop held in Madison, Wisconsin, in August 2012. It reviews the state-of-the-art for these materials and fuels, and the issues that arise from their application to FHRs. FHRs are graphite-moderated, thermal-spectrum nuclear power systems. The neutron spectrum in these systems is selected to enable negative coolant temperature reactivity feedback and maximize burnup. The neutron spectrums in FHRs are much softer than those in pressurized water reactors and slightly harder than those in high-temperature, gas-cooled reactors (HTGRs). FHR fuel utilization is quite similar to that of HTGRs, but components in FHRs are exposed to larger neutron fluxes than in HTGRs due to the much higher power density in FHRs. To deal with the radiation damages to graphite reflectors in FHRs, the inner graphite reflector is designed to be replaced once it reaches its radiation damage limit and the outer graphite reflector is protected from radiation damage by graphite pebble reflectors. The baseline pebble fuel uses an annular fuel layer of graphite-matrix-coated particles, with a center, inert kernel of lower-density graphite. The graphite matrix is the primary reactor neutron moderator and the structural form of the fuel. These coated particle fuels have the ability to maintain their integrity up to temperatures of 1600°C or higher, which provides highly robust safety characteristics to FHRs. The fuel can be in many geometric forms including pebbles, prismatic blocks, plates, and stringers. The pebble fuel was selected as the baseline FHR fuel form because of the lowest development risk and lower fabrication costs. This white paper discusses the performance of FHR fuel, quality verification requirements and development needs. The baseline FHR design assumes limited use of carbon fiber reinforced composites (CFRC), for example for the core barrel assembly, and silicon carbide fiber reinforced composite (SiC/SiC) for structures in high neutron rate regions of the core, for example for liners for shutdown rod channels. Experts agreed that these composite materials are likely to perform as expected in nuclear reactor environments. However, since there is no precedent for using ceramic composites within a nuclear reactor, American Society for Testing and Materials standard test procedures will be established to qualify ceramic composites for nuclear reactor applications. Since information on the compatibility of CFRC and SiC/SiC in fluoride salts is also limited, corrosion data to validate lifetime predictions of composite materials in FHRs is needed. FHRs operate at significantly higher temperature than light water reactors and even liquid metal reactors. To ensure safe and reliable operation of FHRs for thirty or sixty years, time- dependent creep deformation of reactor structural materials must be limited. Type 316 stainless steel (SS) and Alloy N are two candidate alloys to make the reactor vessel and intermediate heater exchangers. Experts emphasized that only a single metallic structural material should be used in contact with the FHR coolant salts to prevent electro-chemical interactions. 316 SS is an FHR Materials, Fuels and Components White Paper 3 | 163 attractive candidate material for use in FHRs due to the extensive experience for nuclear applications, its good tolerance for neutron irradiation and the well developed ASME Section III code case for high temperature use. Because the code for 316 SS does not address corrosion and neutron embrittlement, some further evaluation is needed to address these effects. Alloy N is a reasonably well proven alloy for structural components that operate at temperatures up to 704°C in low neutron flux regions (up to 1 dpa). It has excellent corrosion resistance in fluoride salts. However, Alloy N is not code qualified for Subsection NH – Class 1 Components in Elevated Temperature Service. To qualify Alloy N for Subsection NH, extensive data, especially creep rupture and creep fatigue, in relevant FHR fluoride salt environment is required. Optimizing the method to control corrosion of structural materials in fluoride salts is a key research goal for FHRs. Presence of graphite and the production of tritium fluoride (TF) due to neutron irradiation will affect the corrosion of structural materials. This white paper presents studies performed at the University of Wisconsin in recent years on corrosion of structural alloys (316 SS and Alloy N) in different salt systems, including flinak and KCl-MgCl2 systems, and corrosion control by redox control. It was found that graphite can accelerate corrosion in molten salt systems and metallic redoxagents (Zr or Na) can improve the corrosion resistance of 316 SS. However, graphite damage is also observed when Zr or Na redox is added into the salt. This shows optimal redox control is important to minimize corrosion of metallic materials and also avoid graphite damage. Further study is needed on the compatibility of structural alloys, graphite and composite materials in the flibe primary coolant. Impurities in the salt, particularly moisture and
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