VVER-1000 fuel assembly model in CAD-based unstructured mesh for MCNP6

Abstract

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. 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. 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.

Czech national programme and disposal canister concept

“…will have to be fulfilled (criticality, mechanical resistance, radiation protection), the temperature applied to the surface of the canister will be <95°C, the canister will be required to endure an external pressure of 20 MPa, and the minimum lifetime will be set at 10 000 years. Calculations of the inventory [5–7], thermal output [8] and criticality [9–11], and mechanical [12] and radio-shielding [13–15] have already been conducted employing data on the SNF released by existing power plants. Indeed, the calculation of the inventory of planned new nuclear units was based on available data from EPR1600-type and AP1000-type plants.…”