Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

Field, Kevin G.¹  Howard, Richard H.¹ 

[1] Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

关键词: FUEL-CLADDING INTERACTIONS;    FERRITIC STEELS;    FUEL CANS;    WATER MODERATED REACTORS;    TEMPERATURE RANGE 0400-1000 K;    WATER COOLED REACTORS;    PROCESSING;    FABRICATION;    ALLOY SYSTEMS;    MECHANICAL PROPERTIES;    IRON BASE ALLOYS;    CHROMIUM ALLOYS;    ALUMINIUM ALLOYS;    ACCIDENT-TOLERANT NUCLEAR FUELS;    URANIUM DIOXIDE;   

DOI  :  10.2172/1239760
RP-ID  :  ORNL/TM--2015/573
PID  :  OSTI ID: 1239760
Others  :  Other: AF5810000
Others  :  NEAF278
Others  :  MSFT-15OR0202271
Others  :  TRN: US1600568

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来源: SciTech Connect

【 摘 要 】

FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO₂ clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO₂ fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys, hence promoting FCCI between the fuel-clad systems. The other factor was to develop a test bed where multiple candidate alloys could be evaluated within a single irradiation test train, thereby reducing overall costs and increasing efficiency in alloy screening efforts. A collaboration between ORNL and INL was developed to facilitate the completion of the test bed for FCCI testing. The report highlights the activities related to the development of the ATF-1 ORNL FCCI rodlets for irradiation in INL???s ATR as part of the on-going ATF-1 experiments.