VVER-1000 Fuel Assembly Model in CAD-Based Unstructured Mesh for MCNP6

Martin Lovecký, Jiří Závorka, Jan Vimpel (ŠKODA JS)

28th Symposium of AER on VVER Reactor Physics and Reactor Safety (2018, Olomouc, Czechia)
Reactor physics experiments and code validation (benchmarks)


Geometry models for Monte Carlo transport codes have been using standard constructive solid geometry (CSG). The standard approach is using analytical equations for defining surfaces from which spatial cells are constructed. Both union and intersection operators are available, therefore, arbitrary 3-D geometry can be modeled with CSG. However, this approach can be quite time consuming and possibly error prone for complex models. Monte Carlo transport codes are continuously developed, one of the paths is using CAD-based mesh geometry. MCNP6 features unstructured meshes (UM) created with Abaqus/CAE as geometry description. Attila4MC package for creation of UM geometry from CAD model can be used for MCNP6 models.

VVER-1000 fuel assembly model in UM geometry was created for TVSA-T.mod.2 fuel type. This TVSA fuel type is exclusively operated at Temelin NPP in Czechia. Creation of the model consists of deleting small assembly parts that can be neglected for transport calculations, choosing the size of tetrahedron mesh cells and verification of the model. Basic validation of the model was performed, initially for criticality calculations. In the future, the model will be used for criticality safety analyses, preparation of boundary conditions for diffusion codes and radiation shielding analyzes of spent fuel transport and storage facilities.

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