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Title: Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion

Journal Article · · Progress in Nuclear Energy
ORCiD logo [1]; ORCiD logo [1];  [1];  [2];  [2];  [2];  [2]
  1. Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
  2. Argonne National Lab. (ANL), Argonne, IL (United States)
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The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to power light water reactors (LWRs) in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion of U.S. high performance research reactors (USHPRRs) like the MITR. The Preliminary Safety Analysis Report (PSAR) for the MITR LEU conversion has been submitted and currently under review by the U.S. Nuclear Regulatory Commission (NRC). A transition core plan, from 22 fresh LEU fuel elements (i.e., beginning-of-life) gradually to 24 of them arranged in an equilibrium configuration, is expected to serve as an appendix chapter in the final SAR. The current study reported here presents the fuel cycle development, which eventually leads to the transition core plan. The results confirm the equilibrium state, where both the shim bank movement (i.e., core reactivity) as well as the content of fissile materials stabilize, can be achieved by fixed-pattern fuel management. Fission density has been evaluated for a number of fully discharged LEU fuel elements, using a best-estimate approach. The fuel cycle calculations also generate power profiles at each core state. A steady-state thermal-hydraulic safety analysis has thus been performed, where onset of nucleate boiling (ONB) is considered as the safety criterion. The results confirm significant margin to ONB at all analyzed transition and equilibrium fuel cycle states.

Research Organization:
Argonne National Laboratory (ANL), Argonne, IL (United States)
Sponsoring Organization:
USDOE National Nuclear Security Administration (NNSA)
Grant/Contract Number:
AC02-06CH11357; 2J-30101
OSTI ID:
1606513
Alternate ID(s):
OSTI ID: 1580309
Journal Information:
Progress in Nuclear Energy, Vol. 119, Issue C; ISSN 0149-1970
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English
Citation Metrics:
Cited by: 3 works
Citation information provided by
Web of Science

References (4)

Thermal-hydraulic analyses of MIT reactor LEU transition cycles journal January 2020
Experimental Study of Incipient Nucleate Boiling in Narrow Vertical Rectangular Channel Simulating Subchannel of Upgraded JRR-3 journal January 1986
Validation of a fuel management code MCODE-FM against fission product poisoning and flux wire measurements of the MIT reactor journal August 2014
A correlation for the minima in the pressure drop versus flow-rate curves for sub-cooled water flowing in narrow heated channels journal August 1967

Figures / Tables (15)

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Table V(p. 15)table Table V
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Table VI(p. 17)table Table VI
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11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
LEU conversion
MITR
fission density
fuel cycle
transition core