Whole–core neutron transport calculations without fuel-coolant homogenization

M. A. Smith; N. Tsoulfanidis; E. E. Lewis; G. Palmiotti; T. A. Taiwo

Whole–core neutron transport calculations without fuel-coolant homogenization

M. A. Smith, N. Tsoulfanidis, E. E. Lewis, G. Palmiotti, T. A. Taiwo

Mechanical Engineering

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution

4 Scopus citations

Abstract

The variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the full spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. Comparisons are made of the accuracy of various space-angle approximations and of the corresponding CPU times.

Original language	English (US)
Title of host publication	Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium
Publisher	American Nuclear Society
ISBN (Electronic)	0894486551, 9780894486555
State	Published - 2000
Event	2000 ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium, PHYSOR 2000 - Pittsburgh, United States Duration: May 7 2020 → May 12 2020

Publication series

Name	Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium

Conference

Conference	2000 ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium, PHYSOR 2000
Country/Territory	United States
City	Pittsburgh
Period	5/7/20 → 5/12/20

Funding

This work was supported by the U.S. Department of Energy under Contract No. ENG-38 and by U.S. Department of Energy Contract No. DE-FG07-98ID13632.

ASJC Scopus subject areas

Nuclear Energy and Engineering
Nuclear and High Energy Physics
Radiation
Safety, Risk, Reliability and Quality

Cite this

Smith, M. A., Tsoulfanidis, N., Lewis, E. E., Palmiotti, G., & Taiwo, T. A. (2000). Whole–core neutron transport calculations without fuel-coolant homogenization. In Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium (Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium). American Nuclear Society.

Smith, M. A. ; Tsoulfanidis, N. ; Lewis, E. E. et al. / Whole–core neutron transport calculations without fuel-coolant homogenization. Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium. American Nuclear Society, 2000. (Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium).

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title = "Whole–core neutron transport calculations without fuel-coolant homogenization",

abstract = "The variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the full spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. Comparisons are made of the accuracy of various space-angle approximations and of the corresponding CPU times.",

author = "Smith, {M. A.} and N. Tsoulfanidis and Lewis, {E. E.} and G. Palmiotti and Taiwo, {T. A.}",

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Smith, MA, Tsoulfanidis, N, Lewis, EE, Palmiotti, G & Taiwo, TA 2000, Whole–core neutron transport calculations without fuel-coolant homogenization. in Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium. Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium, American Nuclear Society, 2000 ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium, PHYSOR 2000, Pittsburgh, United States, 5/7/20.

Whole–core neutron transport calculations without fuel-coolant homogenization. / Smith, M. A.; Tsoulfanidis, N.; Lewis, E. E. et al.
Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium. American Nuclear Society, 2000. (Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium).

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution

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N1 - Publisher Copyright: © 2000 Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium. All rights reserved.

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N2 - The variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the full spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. Comparisons are made of the accuracy of various space-angle approximations and of the corresponding CPU times.

AB - The variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the full spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. Comparisons are made of the accuracy of various space-angle approximations and of the corresponding CPU times.

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Smith MA, Tsoulfanidis N, Lewis EE, Palmiotti G, Taiwo TA. Whole–core neutron transport calculations without fuel-coolant homogenization. In Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium. American Nuclear Society. 2000. (Proceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium).

Whole–core neutron transport calculations without fuel-coolant homogenization

Abstract

Publication series

Conference

Funding

ASJC Scopus subject areas

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