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Title: Benchmarking of software and methods for use in transient multidimensional fuel performance with spatial reactor kinetics

Abstract

The key physics involved in accurate prediction of reactor-fuel-element behavior includes neutron transport and thermal hydraulics. The thermal hydraulic feedback mechanism is primarily provided through cross sections to the neutron transport that are temperature and density dependent. Historically, this coupling was primarily seen only in reactor simulators, which are well suited to model the reactor core, giving only a coarse treatment to individual fuel pins as well as simple models for thermal distribution calculations. This poor resolution on the primary coupling mechanisms can lead to conservatisms that should be removed to improve fuel design and performance. This work seeks to address the resolution of space-time-dependent neutron kinetics with thermal feedback within the fuel pin scale in the multi-physics framework. The specific application of this new capability is transient performance analysis of space-time-dependent temperature distribution of fuel elements. The coupling between the neutron transport and the thermal feedback is extremely important in this highly coupled problem, primarily applicable to reactivity-initiated- accidents (RIAs) and loss-of-coolant-accidents (LOCAs). The capability developed will include the coupling of the time-dependent neutron transport with the time-dependent thermal diffusion capability. An improvement in resolution and coupling is proposed by developing neutron transport models that are internally coupled withmore » high fidelity within fuel pin thermal calculations in a multi-physics framework. Good agreement is shown with benchmarks and problems from the literature of RIAs and LOCAs for the tools used. (authors)« less

Authors:
 [1]; ;  [2];  [1]; ;  [2]
  1. Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)
  2. Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)
Publication Date:
Research Org.:
American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
OSTI Identifier:
22107781
Resource Type:
Conference
Resource Relation:
Conference: ICAPP '12: 2012 International Congress on Advances in Nuclear Power Plants, Chicago, IL (United States), 24-28 Jun 2012; Other Information: Country of input: France; 14 refs.; Related Information: In: Proceedings of the 2012 International Congress on Advances in Nuclear Power Plants - ICAPP '12| 2799 p.
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; COMPUTER CODES; CROSS SECTIONS; DESIGN; FEEDBACK; FUEL PINS; LOSS OF COOLANT; NEUTRON FLUX; NEUTRON TRANSPORT; NEUTRONS; NUCLEAR FUELS; NUCLEAR POWER PLANTS; REACTOR CORES; REACTOR FUELING; REACTOR KINETICS; REACTOR SIMULATORS; TEMPERATURE DISTRIBUTION; THERMAL DIFFUSION; THERMAL HYDRAULICS; TIME DEPENDENCE

Citation Formats

Banfield, J. E., Clarno, K. T., Hamilton, S. P., Maldonado, G. I., Philip, B., and Baird, M. L. Benchmarking of software and methods for use in transient multidimensional fuel performance with spatial reactor kinetics. United States: N. p., 2012. Web.
Banfield, J. E., Clarno, K. T., Hamilton, S. P., Maldonado, G. I., Philip, B., & Baird, M. L. Benchmarking of software and methods for use in transient multidimensional fuel performance with spatial reactor kinetics. United States.
Banfield, J. E., Clarno, K. T., Hamilton, S. P., Maldonado, G. I., Philip, B., and Baird, M. L. Sun . "Benchmarking of software and methods for use in transient multidimensional fuel performance with spatial reactor kinetics". United States.
@article{osti_22107781,
title = {Benchmarking of software and methods for use in transient multidimensional fuel performance with spatial reactor kinetics},
author = {Banfield, J. E. and Clarno, K. T. and Hamilton, S. P. and Maldonado, G. I. and Philip, B. and Baird, M. L.},
abstractNote = {The key physics involved in accurate prediction of reactor-fuel-element behavior includes neutron transport and thermal hydraulics. The thermal hydraulic feedback mechanism is primarily provided through cross sections to the neutron transport that are temperature and density dependent. Historically, this coupling was primarily seen only in reactor simulators, which are well suited to model the reactor core, giving only a coarse treatment to individual fuel pins as well as simple models for thermal distribution calculations. This poor resolution on the primary coupling mechanisms can lead to conservatisms that should be removed to improve fuel design and performance. This work seeks to address the resolution of space-time-dependent neutron kinetics with thermal feedback within the fuel pin scale in the multi-physics framework. The specific application of this new capability is transient performance analysis of space-time-dependent temperature distribution of fuel elements. The coupling between the neutron transport and the thermal feedback is extremely important in this highly coupled problem, primarily applicable to reactivity-initiated- accidents (RIAs) and loss-of-coolant-accidents (LOCAs). The capability developed will include the coupling of the time-dependent neutron transport with the time-dependent thermal diffusion capability. An improvement in resolution and coupling is proposed by developing neutron transport models that are internally coupled with high fidelity within fuel pin thermal calculations in a multi-physics framework. Good agreement is shown with benchmarks and problems from the literature of RIAs and LOCAs for the tools used. (authors)},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2012},
month = {7}
}

Conference:
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