Analysis of the TREAT LEU Conceptual Design

Connaway, H. M.¹  Kontogeorgakos, D. C.¹  Papadias, D. D.¹  Brunett, A. J.¹  Mo, K.¹  Strons, P. S.¹  Fei, T.¹  Wright, A. E.¹ 

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

关键词: REACTOR CORES;    HIGHLY ENRICHED URANIUM;    NUCLEAR FUELS;    TREAT REACTOR;    SLIGHTLY ENRICHED URANIUM;    DESIGN;    THERMAL HYDRAULICS;    URANIUM DIOXIDE;    GRAPHITE;    CONVERSION;    PERFORMANCE;    TRANSIENTS;    STEADY-STATE CONDITIONS;    S;   

DOI  :  10.2172/1245183
RP-ID  :  ANL/RTR/TM--16/1
PID  :  OSTI ID: 1245183
Others  :  Other: 126340
Others  :  TRN: US1601179

美国|英语

来源: SciTech Connect

【 摘 要 】

Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy???s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration???s Material Management and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO₂-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.