Vol 125, No 5 (2019)

It is proposed that a specialized molten-salt reactor facility for burning long-lived actinides be located in direct proximity to a pilot-demonstration center. A full-scale ZhSR-S with thermal power 2400 MW will make it possible to burn all transuranium elements released during the reprocessing of spent VVER-1000 nuclear fuel and part of the regenerated plutonium and in the process produce 1 GW of electricity and radioisotope products. The inclusion of a specialized reactor facility for burning transuranium elements from the spent nuclear fuel of VVER-1000 reactors into the nuclear energy system as an integral element will make it possible to solve the problem of closing the nuclear fuel cycle for actinides and reduce the volume and lower the cost of the long-term storage and subsequent disposal of high-level wastes produced in reprocessing.

Several variants of intensifier grids which have a positive influence on the equalization of the coolant parameters over a section of an assembly as well as on the enhancement of critical heat-emission are proposed for intensifying heat-and mass-transfer in connection with the increase in the power of operating power-generating units with VVER. A large volume of experimental studies of the efficiency of different types of intensifier grids has been conducted at the National Research Center Kurchatov Institute. Among the intensifier grids studied is the developed universal intensifier spacer-grid. The investigations were performed on 37-pin models of VVER-1000 fuel assemblies. The influence of the grid spacing on the critical heat flux was studied. The dependence of the hydraulic resistance of a grid on the size of the deflector was studied. The experimental studies can be used to develop intensifier grids with new designs and to verify computational models and codes.

An analysis is presented of different methods of controlling the useful capacity of reactor units with hightemperature gas-cooled reactor and a gas-turbine system for converting energy on the basis of the Brayton cycle: by changing the pressure (mass) of the gas in the loop, the power of the reactor, and the temperature of the gas at the entry into the turbine by means of internal transfers of gas in the loop. The power dependences of the change in the basic parameters are obtained in the range from minimum to 100% in application to a reactor unit with thermal power 600 MW with a single-shaft turbomachine, which qualitatively, taking account of the lag of the processes in the reactor, can be extended to low-capacity units.

The complex program of computational and experimental validation of dense fuel for fast reactors in Project Breakthrough provides for testing of fuel rods with mixed uranium-plutonium nitride fuel in the BOR-60 and BN-600 reactors. The irradiation parameters and post-reactor studies of the fuel rods of a BREST reactor are presented. Testing of fuel rods in BOR-60 to maximum burnup 5% h.a. and damaging dose 74.8 dpa and in BN-600 to maximum burnup 4.5% h.a. and damaging dose 53 dpa did not result in depressurization of the rods. The results of post-reactor tests of irradiated fuel rods – swelling of fuel, gas release from the fuel, mechanical and corrosion properties of the cladding – are presented.

The results of studies of the physical and chemical properties of samples obtained during solidification of simulators of spent ion-exchange resin pulps with the use of polymer binders are presented. Investigations were conducted for pulps simulating heterogeneous radioactive wastes from the Kursk NPP. It is shown that the physical and chemical properties of the final compounds and the technological processes used to obtain them correspond to the requirements of solidified radioactive waste sent to storage or disposal. The surface of the samples is uniform and the components are distributed uniformly. Water resistance and radiation resistance tests showed that the mechanical strength of the tested samples is almost identical to the initial strength. It is shown that the mechanical load at which plastic deformation of the polymer compound occurs is a factor of two larger than and the leach rate does not exceed the established value.

Data on the radiation loads of a radiotherapy system developed at NIITFA are presented. The main problems of the calculations are to establish the region in the mobile part of the system that is best suited for arrangement of blocks with electronics and to calculate the parameters of local radiation shielding that ensures fault-free operation of the electronics for the life of the system. The calculations were performed by the Monte Carlo method using the FLUKA software package.