Performance Improvements to the Cross Section Calculation in MPACT

Journal Article · Fri Jul 01 04:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:23042652

Liu, Yuxuan; Kochunas, Brendan ^[1]; Stimpson, Shane; Kim, Kang-Seog; Collins, Benjamin ^[2]

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1. Department of Nuclear Science and Engineering, University of Michigan: 2355 Bonisteel Blvd. Ann Arbor, MI, 48105 (United States)
2. Oak Ridge National Laboratory: One Bethel Valley Road, P.O. Box 2008, Oak Ridge, TN 37831-6172 (United States)

The MPACT code is the primary deterministic neutron transport solver available in the Virtual Environment for Reactor Applications (VERA) as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL). In contrast to the conventional two-step method using homogenization techniques, MPACT employs a direct transport methodology, where the transport calculation is performed with the explicit reactor core geometry, material compositions, and temperature distribution. As the direct whole-core transport calculation requires significant computational resources, performance improvements have been recently conducted in various components of MPACT. The initial MPACT profiling was performed on the EOS cluster at the Oak Ridge Leadership Computing Facility. Table I shows that the cross section processing accounted for 74% and 33% of the total runtime for VERA Progression Problems 5 and 7. Starting with a problem-independent multigroup (MG) library, two primary functionalities are performed in MPACT to obtain the problem-dependent MG cross sections: (1) the resonance self-shielding calculation, which currently MPACT is approaching with the subgroup method; (2) use of equivalence cross section from resonance calculation to compute the microscopic cross sections, and thus, the macroscopic cross sections of material regions. As shown in Table I, the time distribution in cross section processing varies from case to case. For Problem 5, a 2-D quarter core depletion case, the macroscopic XS dominates due to the increased number of isotopes in the depleted fuel. For Problem 7, a 3-D quarter core PWR at HFP and BOL, more subgroup calculations are performed to obtain updated self-shielded cross sections during the iteration. In this summary, the approaches to reduce the computing time in cross section processing are presented. (authors)

• OSTI ID: 23042652
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 115; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
97 MATHEMATICS AND COMPUTING
COMPUTERIZED SIMULATION
CROSS SECTIONS
GEOMETRY
ISOTOPES
NEUTRON TRANSPORT
PERFORMANCE
PWR TYPE REACTORS
REACTOR CORES
SELF-SHIELDING
TEMPERATURE DISTRIBUTION