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Title: A 2-D/1-D transverse leakage approximation based on azimuthal, Fourier moments

Journal Article · · Nuclear Science and Engineering
 [1];  [2];  [3]
  1. Univ. of Michigan, Ann Arbor, MI (United States); Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  2. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  3. Univ. of Michigan, Ann Arbor, MI (United States)
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Here, the MPACT code being developed collaboratively by Oak Ridge National Laboratory and the University of Michigan is the primary deterministic neutron transport solver within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). In MPACT, the two-dimensional (2-D)/one-dimensional (1-D) scheme is the most commonly used method for solving neutron transport-based three-dimensional nuclear reactor core physics problems. Several axial solvers in this scheme assume isotropic transverse leakages, but work with the axial SN solver has extended these leakages to include both polar and azimuthal dependence. However, explicit angular representation can be burdensome for run-time and memory requirements. The work here alleviates this burden by assuming that the azimuthal dependence of the angular flux and transverse leakages are represented by a Fourier series expansion. At the heart of this is a new axial SN solver that takes in a Fourier expanded radial transverse leakage and generates the angular fluxes used to construct the axial transverse leakages used in the 2-D-Method of Characteristics calculations.

Research Organization:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility (OLCF)
Sponsoring Organization:
USDOE Office of Science (SC)
Grant/Contract Number:
AC05-00OR22725
OSTI ID:
1344985
Journal Information:
Nuclear Science and Engineering, Vol. 185, Issue 2; ISSN 0029-5639
Publisher:
American Nuclear Society - Taylor & FrancisCopyright Statement
Country of Publication:
United States
Language:
English
Citation Metrics:
Cited by: 16 works
Citation information provided by
Web of Science

References (6)

2D/1D fusion method solutions of the three-dimensional transport OECD benchmark problem C5G7 MOX journal July 2006
Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers journal December 2013
The Development and Implementation of a One-Dimensional S n Method in the 2D-1D Integral Transport Solution journal February 2014
Solution of the C5G7MOX benchmark three-dimensional extension problems by the DeCART direct whole core calculation code journal July 2006
Axial SP N and Radial MOC Coupled Whole Core Transport Calculation journal September 2007
Benchmark on deterministic 3-D MOX fuel assembly transport calculations without spatial homogenization journal July 2006

Cited By (2)

Improved Accuracy in the 2-D/1-D Method with Anisotropic Transverse Leakage and Cross-Section Homogenization journal September 2018
Polar Parity for Efficient Evaluation of Anisotropic Transverse Leakage in the 2D/1D Transport Method journal July 2019

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Related Subjects

97 MATHEMATICS AND COMPUTING
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS