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Title: Reactor physics verification of the MCNP6 unstructured mesh capability

Abstract

The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

Authors:
 [1]; ;  [2];  [1]
  1. Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States)
  2. X-Computational Physics Division, Monte Carlo Codes Group, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)
Publication Date:
Research Org.:
American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
OSTI Identifier:
22212779
Resource Type:
Conference
Resource Relation:
Conference: M and C 2013: 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Sun Valley, ID (United States), 5-9 May 2013; Other Information: Country of input: France; 6 refs.; Related Information: In: Proceedings of the 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering - M and C 2013| 3016 p.
Country of Publication:
United States
Language:
English
Subject:
97 MATHEMATICAL METHODS AND COMPUTING; 22 GENERAL STUDIES OF NUCLEAR REACTORS; BENCHMARKS; COMPARATIVE EVALUATIONS; COMPUTER CODES; GEOMETRY; MATHEMATICAL SOLUTIONS; MESH GENERATION; MONTE CARLO METHOD; REACTOR PHYSICS; VERIFICATION

Citation Formats

Burke, T. P., Kiedrowski, B. C., Martz, R. L., and Martin, W. R.. Reactor physics verification of the MCNP6 unstructured mesh capability. United States: N. p., 2013. Web.
Burke, T. P., Kiedrowski, B. C., Martz, R. L., & Martin, W. R.. Reactor physics verification of the MCNP6 unstructured mesh capability. United States.
Burke, T. P., Kiedrowski, B. C., Martz, R. L., and Martin, W. R.. 2013. "Reactor physics verification of the MCNP6 unstructured mesh capability". United States. doi:.
@article{osti_22212779,
title = {Reactor physics verification of the MCNP6 unstructured mesh capability},
author = {Burke, T. P. and Kiedrowski, B. C. and Martz, R. L. and Martin, W. R.},
abstractNote = {The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2013,
month = 7
}

Conference:
Other availability
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  • New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructivemore » Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.« less
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