Calculation of radiation fields in VVER shielding using approximations supporting local balance of materials mass and fission-source neutrons

Abstract

The possibility of using approximations of geometry, which support the balance of materials mass on each spatial cell of a grid covering the computational region, to calculate radiation fields in VVER shielding by introducing additional mixtures of materials for cells where several initial materials are present is analyzed. The methods of combinatorial geometry, which are implemented in the MCU computer program for solving transfer equations by the Monte Carlo method, are used to specify the 3-D geometry of the shielding and a fuel-element and (or) fuel assembly fission source and the method of ray tracing is used to convert the combinatorial specification into a grid specification. A numerical experiment in which the VVER-1500 radiation shielding is calculated showed that the conservation of the materials mass/number of source neutrons in the grid cells during the conversion ensures rapid convergence of the results when the density of the spatial grid increases. Good agreement is attained between MCU Monte Carlo and KATRIN S n calculations of VVER-1000 shielding.Determining with the required accuracy the radiation fields in VVER shielding to substantiate anticipated regimes and the service life of the vessel and in-vessel apparatus as well as the standard operating regime of sensors for thermal neutrons in the channels of ionization chambers is a topical problem. This problem can be solved by performing a three-dimensional calculation of the real composition by the S n method using, for example, the VITAMIN-B6 multigoup cross-sections library obtained from the BGL440 and BGL1000 problem-oriented libraries [1] of group cross-sections with 47 groups of neutrons and 20 groups of photons, taking account of neutron thermalization, as well as data on fuel burnup (which are given for a fuel elements and/or fuel assemblies), whose profile repeats the time-integrated profile of the fission-neutron source. The difficulties encountered in such calculations are due to the necessity of prescribing the three-dimensional geometry of the computational region and converting the geometry and the source to a difference grid using spatial grids with several million cells and effective methods for accelerating the inner and outer (with respect to the region of neutron thermalization) iterations.