Safety of sodium-cooled fast reactors

In the paper some physical and technical aspects of sodium-cooled fast reactor safety are discussed. An approach to the solution of safety problems in the designs of the first fast-reactor generation in the USSR was based (as a result of thermal reactor operating experience) on the logical analysis of the tree of failures; detailed investigation of really predictable emergency conditions; careful working out of safety systems and devices; duplication and echelonment of the most important of them. There are shown general features and some peculiarities of this approach as compared to that used for ensuring safety of other reactor types. A review of main works carried out for the last two decades is given and the most important results of design studies, research and operating experience on fast reactors for accumulation of data and characteristics in the area of safety are presented. The most valuable of the above results have been obtained in the process of operation and investigations on the BR-10 and BOR-60 reactors, from design studies of the BN-350 and BN-600, and from experiments and operation of the BN-350 reactor. A contemplated extensive program of fast reactor development calls for a consideration, in the future, of a wider range of problems including definite attention to very low probability accidents. In the paper some advantages and disadvantages of the concepts of nuclear power plant safety evaluation using the maximum credible accident analysis and quantitative probability methods are discussed. The necessity of all possible development of the quantitative methods and expediency of their application already at the present stage for evaluating reliability of individual safety systems and devices is shown. Taking into account lack of statistical data and initial information about the ways of hypothetical accident development, a possibility is considered for introducing intermediate qualitative and quantitative criteria whose application allows to accelerate significantly the promotion of probability methods for safety evaluation and initiates more active work on accumulation of required statistical material. Introduction of these criteria will permit in the nearest years to use a combined approach to the evaluation of fast reactor safety