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The subject of this paper involves monitoring the temperature of the heat transfer agent above the core and fluctuations in the temperature of the sodium, affecting the strength of the constructions under steady-state reactor operating conditions. Sources of temperature differences may be local superheating with deformations of the fuel element lattices, blocking of parts of the cross sections of the fuel assemblies, and other such factors. Some information about this is presented in [1][2][3].The experimental assembly consisted of seven models of fuel assemblies (Fig. 1). Nineteen fuel element simulators separated by wire coils were mounted in each model. Electrical heating of the simulators was provided only in the central model. The thermocouple probe (Fig. 2) consisted of 30 chromel-alumel thermocouples in stainless-steel capillaries of diameter 0.8 mm. The probe could be placed along the height from the middle of the caps (h = 0) to 200 mm. The placement of the thermocouples is shown in Fig. la. There were also thermocouples in the caps of the fuel assemblies in the cross section of the lower edges of their windows, at the outlet from the model and at the inlet to the "cold" fuel assemblies.There were two independent heat-transfer-agent feeds: into the central model and simultaneously into the six lateral models. We considered the case of superheating of the heat transfer agent in the central model, accomplished either by feeding the heat transfer agent at a higher temperature or by heating it with the electric heaters of the fuel element simulators. The basic parameters for the two series of experiments are given in Table 1. When feeding hotter heat-transfer agent, the possibility of generating additional thermal noise from the electric heaters is eliminated.The average temperature was measured by means of a scanning system. The temperature pulsations were measured and statistically treated using a system including an amplifier circuit, a magnetograph, and a computer. We entered 8100 numbers for each statistical treatment.The temperature distribution over the radius of the model, averaged over the thermocouple readings, is shown in Fig. 3. From this figure it follows that the "hot spot" from the central fuel assembly spreads out with an increase in height. There is some anomaly for low velocity of the "hot" flow (see Fig. 3a, b, c): at some distances from the end of the cap of the fuel assembly, a temperature minimum appears. The effect is possibly connected with cold sodium from adjacent cells entering this region.The flow of heat with an increase in distance from the caps of the models occurs differently for the central sodium flow and the flows from the lateral windows (at the radius rl) (Fig. 4). At a height of 200 mm, the temperature difference between the central flow (t 0) and the cold sodium tr4 (at the radius r4) is only 20% of the initial temperature difference (t h -tc) (the indices h, c mean hot, cold at the inlet to the model). Within the cap, the temperature of this flow decreases by 10%:tO-tr...

Investigations of processes associated with the disruption of cooling of the reactor core and possible subsequent destruction and melting of the core have become of paramount importance in connection with the stricter safety requirements for operational and planned nuclear power plants. It is necessary to know how to correctly describe the scenario of the development of accident situations, to characterize quantitatively accident processes, to predict the probable consequences of accidents, and to develop methods and technical means for decreasing undesirable consequences.Boiling of liquid-metal coolant in the fuel assemblies of a fast reactor is one of the uncalculated operating regimes of a reactor and can arise as a result of a decrease of the coolant flow rate, increase of power production, and other reasons. Overheating of the coolant in the fuel assembly (and, as a consequence, boiling of the coolant) is also possible after accidental reactor shutdown, when the core is cooled. In the process, the power released corresponds to the residual energy release, and the coolant either stops moving or it moves as a result of natural convection.Boiling of the coolant as an accident situation, characterized by a large difference of the specific volume of the vapor and the liquid, is a strong perturbative factor, which gives rise to a sharp restructuring of the hydrodynamics of the entire flow. as a result of which the flow becomes unstable. The character of boiling of a liquid metal (sodium) in the core of a fast reactor and. primarily, the propagation velocity of the steam phase determine the rate of increase of the positive reactivity in the reactor core. Further development of the accident process depends on the amount of energy released as a result of the increase of reactivity because of the boiling of the hquid metal.In some countries (Japan, Germany. USA, France, and Great Britain) long-term programs of experimental and computational thermal-hydraulic investigations of the development and consequence of accident situations in a liquid-metalcooled reactor core have been conducted since the 1970s. Attention was focused mainly on the investigation of the dynamics of the development of the vapor phase m a fuel assembly, the character of the cooling during boiling of the coolant in the region behind the blockage in the fuel assembly, and so on. For example, staged investigations of accident processes accompanied by boiling of a liquid-metal coolant in ring-shaped channels and multirod assemblies with a duration of not more than 10 sec (power increase, decrease of the flow rate, blockage) were conducted in Japan [1][2][3][4][5]. In Germany nonstationary processes associated with the study of a loss of coolant accident and blocking of part of the flow section of the fuel assembly were investigated [6, 7]. In France investigations of boiling of the coolant in a fuel assembly under stationary conditions, during slow transient processes with a decrease of the flow rate, natural convection, fast processes with an increase...