Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

Kontogeorgakos, D.¹  Derstine, K.¹  Wright, A.¹  Bauer, T.² 

[1] Argonne National Lab. (ANL), Argonne, IL (United States);Argonne National Lab.

关键词: SLIGHTLY ENRICHED URANIUM;    HIGHLY ENRICHED URANIUM;    TREAT REACTOR;    URANIUM DIOXIDE;    GRAPHITE;    CONVERSION;    MONTE CARLO METHOD;    FUEL ASSEMBLIES;    NUCLEAR FUELS;    PARTICLES;    DESIGN;    HEATING;    REACTOR KINETICS;    FEASIBILITY STUDIES;   

DOI  :  10.2172/1224971
RP-ID  :  ANL/GTRI/TM--13/4
PID  :  OSTI ID: 1224971
Others  :  Other: 121484
Others  :  TRN: US1500895

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来源: SciTech Connect

【 摘 要 】

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO₂ particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N???Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.