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

Abstract

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 anticipatedmore » 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.« less

Authors:
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)
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1606512
Alternate Identifier(s):
OSTI ID: 1703404
Grant/Contract Number:  
AC02-06CH11357; 2J-30101
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Progress in Nuclear Energy
Additional Journal Information:
Journal Volume: 118; Journal Issue: C; Journal ID: ISSN 0149-1970
Publisher:
Elsevier
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; LEU conversion; natural circulation; safety analysis; uncertainty propagation

Citation Formats

Dave, Akshay J., Sun, Kaichao, Hu, Lin-wen, Pham, Son Hong, Wilson, Erik H., and Jaluvka, David. Thermal-hydraulic analyses of MIT reactor LEU transition cycles. United States: N. p., 2019. Web. doi:10.1016/j.pnucene.2019.103117.
Dave, Akshay J., Sun, Kaichao, Hu, Lin-wen, Pham, Son Hong, Wilson, Erik H., & Jaluvka, David. Thermal-hydraulic analyses of MIT reactor LEU transition cycles. United States. https://doi.org/10.1016/j.pnucene.2019.103117
Dave, Akshay J., Sun, Kaichao, Hu, Lin-wen, Pham, Son Hong, Wilson, Erik H., and Jaluvka, David. 2019. "Thermal-hydraulic analyses of MIT reactor LEU transition cycles". United States. https://doi.org/10.1016/j.pnucene.2019.103117. https://www.osti.gov/servlets/purl/1606512.
@article{osti_1606512,
title = {Thermal-hydraulic analyses of MIT reactor LEU transition cycles},
author = {Dave, Akshay J. and Sun, Kaichao and Hu, Lin-wen and Pham, Son Hong and Wilson, Erik H. and Jaluvka, David},
abstractNote = {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.},
doi = {10.1016/j.pnucene.2019.103117},
url = {https://www.osti.gov/biblio/1606512}, journal = {Progress in Nuclear Energy},
issn = {0149-1970},
number = C,
volume = 118,
place = {United States},
year = {2019},
month = {8}
}

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Figures / Tables:

Fig. 1 Fig. 1: Top view of the MITR core (left) and horizontal cross-section (right).

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Works referenced in this record:

The Determination of Forced-Convection Surface-Boiling Heat Transfer
journal, August 1964


Experimental Study of Incipient Nucleate Boiling in Narrow Vertical Rectangular Channel Simulating Subchannel of Upgraded JRR-3
journal, January 1986