Russian Codes for Safety Analysis of Sodium-Cooled Fast Reactors

Abstract

An approach to validating the safety of sodium-cooled fast reactors which was devised in Russia as these reactors were developed is presented. The collection of computational codes used for validating safety is described. Attention is focused on a group of codes which are intended for systematic analysis of all phases of a serious beyond design-basis accident resulting in the destruction of the core. The domestic approach to the analysis of the consequences of serious accidents is compared to the methodology developed abroad. Up to now the code system under development has made it possible to perform the main tasks of design validation, but the chosen path of development, where a separate computational code was developed for each individual problem, is not exhausted. On the one hand the complexity of the mathematical models drawn into the analysis has increased significantly, while on the other hand significant advances have been made in the development of computational technologies. The answer to the present situation is the tendency observed in the world nuclear community toward developing integrated codes and multipurpose computational systems. In Russia, this tendency is realized in the program for the development of a new generation of code systems within the framework of a special federal program.