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

Abstract

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 objectivemore » 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.« less

Authors:
; ; ; ; ;  [1];  [2]
  1. (Nuclear Engineering Division)
  2. (
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1037973
Report Number(s):
ANL/RERTR/TM-12-3 REVISION 0
TRN: US1201780
DOE Contract Number:
DE-AC02-06CH11357
Resource Type:
Technical Report
Country of Publication:
United States
Language:
ENGLISH
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ALLOYS; GEOMETRY; HIGHLY ENRICHED URANIUM; MITR REACTOR; NUCLEAR WEAPONS; PEAK LOAD; POWER DISTRIBUTION; REACTOR CORES; RESEARCH AND TEST REACTORS; RESEARCH REACTORS; SLIGHTLY ENRICHED URANIUM; THERMAL HYDRAULICS; URANIUM-MOLYBDENUM FUELS

Citation Formats

Wilson, E.H., Horelik, N.E., Dunn, F.E., Newton, T.H., Jr., Hu, L., Stevens, J.G., and 2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department). Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.. United States: N. p., 2012. Web. doi:10.2172/1037973.
Wilson, E.H., Horelik, N.E., Dunn, F.E., Newton, T.H., Jr., Hu, L., Stevens, J.G., & 2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department). Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.. United States. doi:10.2172/1037973.
Wilson, E.H., Horelik, N.E., Dunn, F.E., Newton, T.H., Jr., Hu, L., Stevens, J.G., and 2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department). Wed . "Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.". United States. doi:10.2172/1037973. https://www.osti.gov/servlets/purl/1037973.
@article{osti_1037973,
title = {Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.},
author = {Wilson, E.H. and Horelik, N.E. and Dunn, F.E. and Newton, T.H., Jr. and Hu, L. and Stevens, J.G. and 2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)},
abstractNote = {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.},
doi = {10.2172/1037973},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Wed Apr 04 00:00:00 EDT 2012},
month = {Wed Apr 04 00:00:00 EDT 2012}
}

Technical Report:

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  • 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 Enrichedmore » 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. Towards this goal, comparisons of MCNP5 Monte Carlo neutronic modeling results for HEU and LEU cores have been performed. Validation of the model has been based upon comparison to HEU experimental benchmark data for the MITR-II. The objective of this work was to demonstrate a model which could represent the experimental HEU data, and therefore could provide a basis to demonstrate LEU core performance. This report presents an overview of MITR-II model geometry and material definitions which have been verified, and updated as required during the course of validation to represent the specifications of the MITR-II reactor. Results of calculations are presented for comparisons to historical HEU start-up data from 1975-1976, and to other experimental benchmark data available for the MITR-II Reactor through 2009. This report also presents results of steady state neutronic analysis of an all-fresh LEU fueled core. Where possible, HEU and LEU calculations were performed for conditions equivalent to HEU experiments, which serves as a starting point for safety analyses for conversion of MITR-II from the use of HEU fuel to the use of UMo LEU fuel.« less
  • The WWR-SM reactor in Uzbekistan is preparing for the conversion from HEU (36%) fuel to LEU (19.8%) fuel. During this conversion, the HEU fuel assemblies (IRT-3M FA) being discharged at the end of each cycle will be replaced by LEU fuel assemblies (IRT-4M FA); this gradual conversion requires 9 cycles. The safety analysis report for this conversion process has been prepared. This paper presents selected results for postulated transient/accidents during this conversion process; results for transient analysis for the HEU core, the 1st mixed (HEU-LEU) core, and for the first full LEU core are presented for the following initiators: controlmore » rod motion (2 cases), loss of power, and FA blockage. These results show that safety is maintained for all transients analyzed and that the behavior of all the analyzed cores is essentially the same. (author)« less
  • The WWR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences (INP AS) in Uzbekistan is preparing for the conversion from HEU (36%) fuel to LEU (19.8%) fuel. During this conversion, the HEU fuel assemblies (IRT-3M FA) being discharged at the end of each cycle will be replaced by LEU fuel assemblies (IRT-4M FA); this gradual conversion requires 9 cycles. Conversion to LEU without loss of performance for the present experimental program requires the size of the core to increase from 18 to 20 fuel assemblies and the power of the reactor to increase from 10 tomore » 11 MW. The safety analysis report for this conversion process has been prepared. This paper presents the methods and results for the neutronics analysis (burnup, power distributions and shutdown margin), the steady-state thermal hydraulics analysis and the kinetics parameters for the HEU, all mixed and the first full LEU cores. (author)« less
  • This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy`s Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elementsmore » in January 1996, 18,800 in January 2001, and 22,700 in January 2006.« less
  • Calculations have been performed for postulated transients in the VVR-SM Reactor at the Institute of Nuclear Physics (INP) of the Academy of Sciences in the Republic of Uzbekistan. (The reactor designation in Cyrillic is BBP-CM; transliterating characters to English gives VVRSM but translating words gives WWR-SM.) These calculations have been performed at the request of staff of the INP who are performing similar calculations. The transients considered were established during working meetings between Argonne National Laboratory (ANL) and INP staff during summer 2006 [Ref. 1], subsequent email correspondence, and subsequent staff visits. Calculations were performed for the current high-enriched uraniummore » (HEU) core, the proposed low-enriched uranium (LEU) core, and one mixed HEU-LEU core during the transition. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.« less