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

Technical Report · Sat Jun 01 04:00:00 EDT 2013

DOI:https://doi.org/10.2172/1224971· OSTI ID:1224971

Kontogeorgakos, D. ^[1]; Derstine, K. ^[1]; Wright, A. ^[1]; Bauer, T. ^[1]; Stevens, J. ^[1]

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

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.

• Research Organization: Argonne National Lab. (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation (NA-20). Office of Material Management and Minimization (M3)
• DOE Contract Number: AC02-06CH11357
• OSTI ID: 1224971
• Report Number(s): ANL/GTRI/TM--13/4; 121484
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
CONVERSION
DESIGN
FEASIBILITY STUDIES
FUEL ASSEMBLIES
GRAPHITE
HEATING
HIGHLY ENRICHED URANIUM
MONTE CARLO METHOD
NUCLEAR FUELS
PARTICLES
REACTOR KINETICS
SLIGHTLY ENRICHED URANIUM
TREAT REACTOR
URANIUM DIOXIDE