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.

Department(s)

Nuclear Engineering and Radiation Science

Comments

U.S. Department of Energy, Grant DE-FG07-98ID13632

International Standard Book Number (ISBN)

978-089448655-5

Document Type

Article - Conference proceedings

Document Version

Citation

File Type

text

Language(s)

English

Rights

© 2024 American Nuclear Society, All rights reserved.

Publication Date

01 Jan 2000

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