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Title: Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10 MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.
Authors:
 [1] ;  [1] ;  [1] ;  [2] ;  [2] ;  [2]
  1. Argonne National Lab. (ANL), Argonne, IL (United States)
  2. Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.
Publication Date:
Grant/Contract Number:
AC02-06CH11357
Type:
Accepted Manuscript
Journal Name:
Nuclear Engineering and Design
Additional Journal Information:
Journal Volume: 317; Journal ID: ISSN 0029-5493
Publisher:
Elsevier
Research Org:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org:
USDOE National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation (NA-20); USDOE National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation (NA-20), Office of Material Management and Minimization (M3)
Contributing Orgs:
Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; MITR; TRIGA; Low-enriched uranium; Reactor conversion; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; Low‐enriched Uranium; Reactor Conversion
OSTI Identifier:
1349058
Alternate Identifier(s):
OSTI ID: 1366484; OSTI ID: 1415517

Dunn, F. E., Wilson, E. H., Feldman, E. E., Sun, K., Wang, C., and Hu, L. -W.. Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium. United States: N. p., Web. doi:10.1016/j.nucengdes.2017.02.034.
Dunn, F. E., Wilson, E. H., Feldman, E. E., Sun, K., Wang, C., & Hu, L. -W.. Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium. United States. doi:10.1016/j.nucengdes.2017.02.034.
Dunn, F. E., Wilson, E. H., Feldman, E. E., Sun, K., Wang, C., and Hu, L. -W.. 2017. "Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium". United States. doi:10.1016/j.nucengdes.2017.02.034. https://www.osti.gov/servlets/purl/1349058.
@article{osti_1349058,
title = {Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium},
author = {Dunn, F. E. and Wilson, E. H. and Feldman, E. E. and Sun, K. and Wang, C. and Hu, L. -W.},
abstractNote = {The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10 MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.},
doi = {10.1016/j.nucengdes.2017.02.034},
journal = {Nuclear Engineering and Design},
number = ,
volume = 317,
place = {United States},
year = {2017},
month = {3}
}