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Title: Thermal-hydraulic analyses of MIT reactor LEU transition cycles

Journal Article · · Progress in Nuclear Energy
ORCiD logo [1]; ORCiD logo [1];  [1];  [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. In the framework of non-proliferation policy, research and test reactors have started a program to convert highly enriched uranium (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. A transition core plan, from 22 fresh LEU fuel elements to 24 in an equilibrium state, is proposed. This study evaluates the thermal-hydraulic safety margins of each transition cycle and state (i.e. beginning, middle and end). The STAT7 and RELAP5 codes are used in the study. STAT7 provides an integrated platform for statistical propagation of uncertainties in determining steady-state thermal-hydraulic margins. A RELAP5 model was created to verify MITR's STAT7 model, and provide transient analyses. The steady-state analyses determined that, during the entire transition from a fresh to equilibrium core, the Limiting System Safety Settings (LSSS) has a 10.2% margin to the minimum power at which onset of nucleate boiling (ONB) is precluded at 3σ confidence. The loss of flow (LOF) transient, which is the most limiting anticipated transient for the MITR, was simulated. With a conservative initial state of the reactor (operating at μ + 3σ power level and LSSS mass flow rate), the maximum calculated centerline fuel temperature was 115 degrees C, which is a significantly lower value than the blistering limit of 350 degrees C. The cladding wall temperatures did not exceed the ONB temperature throughout the transient. Therefore, during LOF, no nucleate boiling is expected, precluding any critical heat flux trigger. Overall, this study indicates that the proposed LEU transition core specifications have significant margin to thermal-hydraulic limits, during steady-state operation and the LOF transient.

Research Organization:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Organization:
USDOE National Nuclear Security Administration (NNSA)
Grant/Contract Number:
AC02-06CH11357; 2J-30101
OSTI ID:
1606512
Alternate ID(s):
OSTI ID: 1703404
Journal Information:
Progress in Nuclear Energy, Vol. 118, Issue C; ISSN 0149-1970
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English
Citation Metrics:
Cited by: 2 works
Citation information provided by
Web of Science

References (5)

The Determination of Forced-Convection Surface-Boiling Heat Transfer journal August 1964
Heat Transfer Performance of Internally Finned Tubes in Turbulent Flow journal April 1980
Development of a Low-Enriched-Uranium Core for the MIT Reactor journal November 2007
Experimental Study of Incipient Nucleate Boiling in Narrow Vertical Rectangular Channel Simulating Subchannel of Upgraded JRR-3 journal January 1986
If the Dittus and Boelter equation is really the McAdams equation, then should not the McAdams equation really be the Koo equation? journal March 2011

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