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Title: Verification of Unstructured Mesh Capabilities in MCNP6 for Reactor Physics Problems

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 Constructive 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.
Authors:
 [1] ;  [1] ;  [1] ;  [1]
  1. Los Alamos National Laboratory
Publication Date:
OSTI Identifier:
1049351
Report Number(s):
LA-UR-12-24260
TRN: US1204492
DOE Contract Number:
AC52-06NA25396
Resource Type:
Technical Report
Research Org:
Los Alamos National Laboratory (LANL)
Sponsoring Org:
DOE/LANL
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 97 MATHEMATICAL METHODS AND COMPUTING; 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; BENCHMARKS; CRITICALITY; EIGENVALUES; GEOMETRY; NUCLEAR FUELS; OXIDES; REACTOR PHYSICS; VERIFICATION