The Challenging Thermophysical Issues of Fast Reactors

The article presents and analyzes the results from topical thermophysical investigations aimed at substantiating the characteristics and assessing the safety of liquid metal cooled fast reactors: an advanced largecapacity fast sodium-cooled reactor and a fast lead-cooled reactor. We also outline the results from experimental investigations into hydrodynamics and heat transfer for stratified flow of coolant, into thermal hydraulics of a large-modular “sodium–water” steam generator in different modes of reactor operation, and into the fuel assembly degradation process in the course of a severe accident involving loss of sodium flowrate in a fast reactor. The basic mechanisms governing degradation of dummy fuel rod claddings are identified; the distribution of marker materials over the fuel assembly height in its ultimate state is estimated, and the phenomena of blocking the fuel assembly flow pass section and the marker materials being thrown to beyond the assembly boundaries are studied. The article demonstrates the advisability of using a combined sodium purification system built into the reactor vessel that contains cold traps as a mandatory element along with hot traps, which serve to perform accelerated removal of oxygen during the NPP operation at its nominal parameters. The results from thermal–hydraulic tests of the lead-cooled reactor core carried out on the 6B experimental setup and of the steam generator carried out on the SPRUT experimental setup installed at the SSC RF–IPPE are presented.1 It has been found from the experiments on the thermal–hydraulic model of the lead-cooled reactor’s steam generator that the steam temperatures at the outlets from both headers are identical with each other and so are the lead temperatures at the downcomer section outlet and in the main lead path. The experiments did not show pulsations of feedwater flowrate or pressure in the loops, which points to the stable nature of operation at partial load. The state of the technology of heavy metal coolants and the prospects for its future development are considered. It is shown that there is a possibility in principle to realize the required parameters of a high-temperature sodium-cooled fast reactor for producing large quantities of hydrogen, e.g., based on one of the thermochemical cycles or high-temperature electrolysis with a high electricity thermal utilization factor. The problems that must be solved during the further thermophysical investigations are analyzed.