Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

Wilson, E.H. ; Horelik, N.E. ; Dunn, F.E. ; Newton, T.H., Jr. ; Hu, L. ; Stevens, J.G. (Nuclear Engineering Division) ; Department), (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering

Argonne National Laboratory

关键词: Research And Test Reactors;    Alloys;    Power Distribution;    Uranium-Molybdenum Fuels;    Slightly Enriched Uranium;   

DOI  :  10.2172/1037973
RP-ID  :  ANL/RERTR/TM-12-3 REVISION 0
RP-ID  :  DE-AC02-06CH11357
RP-ID  :  1037973

美国|其它

来源: UNT Digital Library

【 摘 要 】

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.