Authors & Affiliations

Seleznev E.F.1, Belov A.A.1, Belousov V.I.1, Chernova I.S.1, Drobyishev Yu.Yu.2
1. Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
2. All-Russian Research Institute for Nuclear Power Plants Operation, Moscow, Russia

Seleznev E.F. – Head of Laboratory, Dr. Sci. (Tech.), Nuclear Safety Institute of the Russian Academy of Sciences. Contacts: 52, Bolshaya Tulskaya st., Moscow, Russia, 115191. Tel.: +7(495) 955-23-11; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Belov A.A. – Researcher, Cand. Sci. (Tech.), Nuclear Safety Institute of the Russian Academy of Sciences.
Belousov V.I. – Researcher, Cand. Sci. (Tech.), Nuclear Safety Institute of the Russian Academy of Sciences.
Chernova I.S. – Junior Researcher, Nuclear Safety Institute of the Russian Academy of Sciences.
Drobyishev Yu.Yu. – Engineer, All-Russian Research Institute for Nuclear Power Plants Operation.

Abstract

This paper considers the statement of the problem of designing of a combined code taking into account the heterogeneity of the reactor environment properties in the diffusion approximation. The advantages (a promptitude, developed and an approved base of constant) and the weakness of the diffusion approximation (associated with homogenization of computational cells) are known. Implementing the code in the diffusion approximation, taking into account the heterogeneity of the environment, will eliminate a known weakness, while retaining all its advantages.
The code assumes an implementation of the homogeneous, adjoint, inhomogeneous and transient neutron transport problems in the cores of fast reactors. All tasks have been tested by the authors in the creation of previous codes, so the problems in the implementation of these solutions in new code are not expected. But due to the high level of detailing of the task it is possible to use new algorithms, for example, for computation of areas with cavities.
Great opportunities of the code in the detailing of calculated core will certainly lead to the complication of the algorithm of initial data input. Using the constructor for creating a computational model of the core providing a visualization of the model including spatial grid is scheduled in the code to facilitate the user's work. With the constructor all possible types of modeling FA (calculated channels) with selectable fill each calculated cell by any material available in the database of the code will be prompted to the user.
For user's convenience it is planned to implement the code's postprocessor with the possibility of visualization of calculated neutronic characteristics of a core.

Keywords
neutronic calculation, heterogeneity, diffusion approximation, code, core constructor, postprocessor, visualization

Article Text (PDF, in Russian)

References

UDC 621.039.5

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2018, issue 1, 1:16