Solution of the C5G7 Benchmark using the MOCUM Transport Code with ANSYS Unstructured Mesh
Journal Article
·
· Transactions of the American Nuclear Society
OSTI ID:23042828
- Texas A and M University-Kingsville: 700 University Blvd, MSC 191, Kingsville, TX, 78363 (United States)
The objective of this research is to use finite element analysis (FEA) software to perform the unstructured meshing for reactor geometries and port the mesh information to MOCUM code. This workflow takes the advantage of the mature meshing module of the mainstream FEA; reduces the development effort for nuclear codes; and increases the nuclear code accuracy by using high quality meshes. In this work, ANSYS is used to build the reactor model and perform the unstructured meshing. The mesh information is then imported to the MOCUM code, a method of characteristic (MOC) transport program supporting unstructured meshes, using a developed MATLAB script. This procedure is verified by modeling the C5G7 benchmark problem. (authors)
- OSTI ID:
- 23042828
- 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
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