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Title: Neutronics Modeling of the High Flux Isotope Reactor using COMSOL

Abstract

The High Flux Isotope Reactor located at the Oak Ridge National Laboratory is a versatile 85 MWth research reactor with cold and thermal neutron scattering, materials irradiation, isotope production, and neutron activation analysis capabilities. HFIR staff members are currently in the process of updating the thermal hydraulic and reactor transient modeling methodologies. COMSOL Multiphysics has been adopted for the thermal hydraulic analyses and has proven to be a powerful finite-element-based simulation tool for solving multiple physics-based systems of partial and ordinary differential equations. Modeling reactor transients is a challenging task because of the coupling of neutronics, heat transfer, and hydrodynamics. This paper presents a preliminary COMSOL-based neutronics study performed by creating a two-dimensional, two-group, diffusion neutronics model of HFIR to study the spatially-dependent, beginning-of-cycle fast and thermal neutron fluxes. The 238-group ENDF/B-VII neutron cross section library and NEWT, a two-dimensional, discrete-ordinates neutron transport code within the SCALE 6 code package, were used to calculate the two-group neutron cross sections required to solve the diffusion equations. The two-group diffusion equations were implemented in the COMSOL coefficient form PDE application mode and were solved via eigenvalue analysis using a direct (PARDISO) linear system solver. A COMSOL-provided adaptive mesh refinement algorithm was usedmore » to increase the number of elements in areas of largest numerical error to increase the accuracy of the solution. The flux distributions calculated by means of COMSOL/SCALE compare well with those calculated with benchmarked three-dimensional MCNP and KENO models, a necessary first step along the path to implementing two- and three-dimensional models of HFIR in COMSOL for the purpose of studying the spatial dependence of transient-induced behavior in the reactor core.« less

Authors:
 [1];  [1];  [1];  [1]
  1. ORNL
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); High Flux Isotope Reactor
Sponsoring Org.:
USDOE Office of Science (SC)
OSTI Identifier:
1042798
DOE Contract Number:  
DE-AC05-00OR22725
Resource Type:
Journal Article
Journal Name:
Annals of Nuclear Energy
Additional Journal Information:
Journal Volume: 38; Journal Issue: 11; Journal ID: ISSN 0306-4549
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ALGORITHMS; CROSS SECTIONS; DIFFERENTIAL EQUATIONS; DIFFUSION; DIFFUSION EQUATIONS; DISCRETE ORDINATE METHOD; EIGENVALUES; HEAT TRANSFER; HFIR REACTOR; HYDRODYNAMICS; IRRADIATION; ISOTOPE PRODUCTION; NEUTRON ACTIVATION ANALYSIS; NEUTRON TRANSPORT; REACTOR CORES; RESEARCH REACTORS; SCATTERING; THERMAL HYDRAULICS; THERMAL NEUTRONS; TRANSIENTS; HFIR; COMSOL; Multiphysics; diffusion theory; neutronics; SCALE; NEWT

Citation Formats

Chandler, David, Primm, Trent, Freels, James D, and Maldonado, G Ivan. Neutronics Modeling of the High Flux Isotope Reactor using COMSOL. United States: N. p., 2011. Web. doi:10.1016/j.anucene.2011.06.002.
Chandler, David, Primm, Trent, Freels, James D, & Maldonado, G Ivan. Neutronics Modeling of the High Flux Isotope Reactor using COMSOL. United States. doi:10.1016/j.anucene.2011.06.002.
Chandler, David, Primm, Trent, Freels, James D, and Maldonado, G Ivan. Sat . "Neutronics Modeling of the High Flux Isotope Reactor using COMSOL". United States. doi:10.1016/j.anucene.2011.06.002.
@article{osti_1042798,
title = {Neutronics Modeling of the High Flux Isotope Reactor using COMSOL},
author = {Chandler, David and Primm, Trent and Freels, James D and Maldonado, G Ivan},
abstractNote = {The High Flux Isotope Reactor located at the Oak Ridge National Laboratory is a versatile 85 MWth research reactor with cold and thermal neutron scattering, materials irradiation, isotope production, and neutron activation analysis capabilities. HFIR staff members are currently in the process of updating the thermal hydraulic and reactor transient modeling methodologies. COMSOL Multiphysics has been adopted for the thermal hydraulic analyses and has proven to be a powerful finite-element-based simulation tool for solving multiple physics-based systems of partial and ordinary differential equations. Modeling reactor transients is a challenging task because of the coupling of neutronics, heat transfer, and hydrodynamics. This paper presents a preliminary COMSOL-based neutronics study performed by creating a two-dimensional, two-group, diffusion neutronics model of HFIR to study the spatially-dependent, beginning-of-cycle fast and thermal neutron fluxes. The 238-group ENDF/B-VII neutron cross section library and NEWT, a two-dimensional, discrete-ordinates neutron transport code within the SCALE 6 code package, were used to calculate the two-group neutron cross sections required to solve the diffusion equations. The two-group diffusion equations were implemented in the COMSOL coefficient form PDE application mode and were solved via eigenvalue analysis using a direct (PARDISO) linear system solver. A COMSOL-provided adaptive mesh refinement algorithm was used to increase the number of elements in areas of largest numerical error to increase the accuracy of the solution. The flux distributions calculated by means of COMSOL/SCALE compare well with those calculated with benchmarked three-dimensional MCNP and KENO models, a necessary first step along the path to implementing two- and three-dimensional models of HFIR in COMSOL for the purpose of studying the spatial dependence of transient-induced behavior in the reactor core.},
doi = {10.1016/j.anucene.2011.06.002},
journal = {Annals of Nuclear Energy},
issn = {0306-4549},
number = 11,
volume = 38,
place = {United States},
year = {2011},
month = {1}
}