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.The approach to the safety analysis of sodium-cooled fast reactors was formed on the basis of domestic and foreign experience accumulated in the process of designing and operating these reactors. One of the main tools for solving this general problem is a set of computational codes used for validating the characteristics of fast reactors and analyzing their changes under different initial events, including accidents. As numerical modeling methods develop and the computational hardware improves, computer codes play an increasingly important role in the validation process. In practice, both the code developers and code users rejected, first and foremost, the list of applied problems whose solution required complete and comprehensive validation of the reactor parameters and safety. Such lists are, on the one hand, universal and applicable for definite types of reactors, for example, sodium-cooled fast reactors, while on the other hand the characteristics of each particular design must be taken into account in them. The lists are compiled by the joint efforts of design organizations and oversight agencies in application to each design being developed and include a list of disruptions of normal operation, design-basis and beyond design-basis accidents.The task of safety analysis is to show that the operational limits and safety criteria formulated in the regulatory documents are not violated in all initial failures and accidents included on the list.Since the lists of initial disruptions and accidents cover a large number of phenomena on different levels, a set of different codes differing by the spectrum of the phenomena being modeled, the degree of detail of the modeling and, in consequence, the computational accura...