Whole–core neutron transport calculations without fuel-coolant homogenization

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

Research output: Chapter in Book/Report/Conference proceedingConference contribution

3 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 languageEnglish (US)
Title of host publicationProceedings of the PHYSOR 2000 - ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium
PublisherAmerican Nuclear Society
ISBN (Electronic)0894486551, 9780894486555
StatePublished - 2000
Event2000 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 2020May 12 2020

Publication series

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

Conference

Conference2000 ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium, PHYSOR 2000
CountryUnited States
CityPittsburgh
Period5/7/205/12/20

ASJC Scopus subject areas

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

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