Sodium as a coolant for fast reactors

Abstract

The results of a comparative analysis and choice of sodium as the coolant for fast reactors are presented. The facilities developed for removing impurities present in the sodium coolant and monitoring their content are described. The modeling of the mass transfer of impurities in coolants and the development of new liquid-metal coolants are examined. The results of an analysis of the anomalous situations in fast reactors, and methods for removing coolant residues from equipment and salvaging wastes are presented. It is shown that the technical solutions adopted provide reliable protection from accidents. New problems of sodium technology are formulated in application to the development of a new generation of fast reactors.Work on studying and mastering liquid metals as coolants began in the 1950s at the Physics and Power-Engineering Institute under the direction of A. I. Leipunskii as one of the avenues for developing nuclear power facilities with liquid-metal cooling. Metals with satisfactory nuclear, thermophysical, and physicochemical properties as well as low vapor pressure at high temperature (see Table 1) were considered as candidates.The requirements which coolants must satisfy to be used in different types of nuclear power facilities were formulated during this period. The effect of a coolant on the physical, technological, corrosion, and thermohydraulic characteristics of a reactor, the toxicity, and the costs were taken into account. These questions were closely tied to safety (physical, fire, toxic, and technological) [1][2][3][4][5].In our country, sodium was chosen as the coolant for nuclear power plants with fast reactors because of its thermophysical properties and the simplicity of the technological operations during repair work [5][6][7]. The disadvantages inherent to sodium are high chemical activity in an oxidative medium and intense interaction with water (with formation of hydrogen gas), induced radioactivity of 24 Na and 22 Na with half-life 15 h and 2.6 yr, respectively -resulted in a second loop, also with sodium, in the reactor facility. The second loop prevents the products of the interaction of water with sodium from entering the core and the effect of high pressure on the first loop.The experimental and computational-theoretical studies led to an understanding of the liquid-metal system of nuclear power facilities as a complex heterogeneous multicomponent system whose components react with the structural materials and with one another [1][2][3][4][5][6][7][8][9][10][11][12][13][14][15][16]. Physicochemical Properties and Fundamentals of Liquid-Metal Coolant.In studies of the coolant-structural (technological) materials-protective gas system, a great distance has been traversed from obtaining the solubility constants of impurities in liquid metals to constructing models of mass transfer in liquid-meal loops taking account of their thermohy-