A method for spontaneous formation of a protective coating, consisting of zirconium nitride in a eutectic based on lead with 2.25 mass % magnesium and 0.2 mass % zirconium, on a steel surface is proposed for decreasing the interaction between the sublayer in a fuel element and the fuel-element cladding. It is confirmed on the basis of gravimetric analysis, analysis with a deuteron beam from an EG-2.5 accelerator, x-ray diffraction analysis performed with a DRON-1 diffractometer, and x-ray spectral wave microanalysis that at temperatures in the range 813-1023 K a zirconium nitride coating on the surface of 16Kh12VMSFBR steel will effectively increase the service life of the fuel assemblies in the BREST-OD-300 system (35000 h). The maximum mass loss will be 0.12 kg/m 2 and the maximum thinning a fuel-element wall will be 15 µm.BREST-OD-300 is an experimental sample of a new-generation nuclear power facility with inherent safety [1]. Lead, which has a high boiling point, is chemically inactive, and exhibits low radiation activity, was chosen to remove heat. The low neutron moderation by heavy lead makes it possible to discharge a lattice of fuel elements, compensating the low pump-through rate and high melting temperature of lead.The fuel elements developed for BREST-OD-300 fuel elements consist of pellets fabricated from uranium and plutonium nitride, 16Kh12VMSFBR steel cladding, and a liquid-metal filled gap between the fuel and cladding. Fuel elements with nitride fuel have certain advantages over models used in other systems. The liquid-metal sublayer inside a fuel element makes it possible to decrease the temperature of the nitride fuel to a level that ensures reliable operation of the fuel element with deep burnup and low swelling and low gas-release. The corresponding technology has been developed. The fuel elements have been tested in BOR-60 [2]. The results of the tests performed on a channel with a 43.4 kW fuel assembly, coolant temperature 813 K at the exit into the pump and 888 K at the exit from the fuel assembly in the BOR-60 reactor are presented in [3]. The damaging dose was 6.5 displ./atom and the fuel burnup was 0.44% h.a. No corrosion damage to or unsealing of fuel elements were visible. However, the compatibility of the liquid-metal sublayer inside a fuel element with the fuel-element cladding in the presence of nitride fuel needs further investigation.In the present work, the method of spontaneous formation of a protective coating of zirconium nitride in a lead-based eutectic with 2.25 mass % magnesium and 0.2 mass % zirconium on a steel surface is proposed for decreasing the interac-
The sources of impurities entering the sodium in fast reactors were investigated. The analysis showed that oxygen and hydrogen can be removed from the sodium by using cold traps in all operating regimes of a nuclear power plant as well as hot traps. An operating regime preventing hydrogen accumulation in the first-loop cold trap is proposed for a system purifying the first two loops. A computer code for calculating the impurity mass transfer is perfected. Test calculations showed that the procedure developed and the code are both serviceable. The deviation of the computational results from the experimental data is about 30% on average. For a built-in purification system, it is essential to develop a cold trap with a large impurity capacity. It is shown on the basis of experiments that such cold traps can in principle be developed. Thermohydraulic and mass-transfer codes must be developed in order to realize this possibility.Sodium-coolant purification systems in nuclear power plants with fast reactors must provide the required purity in all operating regimes taking account of all sources of impurities and must have the capacity required to handle the impurities accumulating in the purification system (it is permissible to replace the elements of the system during purification but the number of such replacements must be minimal). When significant contamination is present (routine maintenance, refueling, and accidental contamination), their capacity must ensure that impurities will be removed from the coolant as quickly as possible before the power reaches the operating level and the accumulation of suspensions in the first loop must be prevented.At the present stage of development of nuclear power, considering that safety, cost-effectiveness, and environmental compatibility must be improved, the requirements for the equipment in a nuclear power facility have been raised. Specifically, it has been decided that all systems with radioactive sodium be placed inside the reactor tank. This limits the dimensions of the first-loop systems. Therefore, the positive experience gained in placing purification systems outside the reactor tank cannot be fully utilized.For optimization, the purification characteristics were analyzed for the first loop. The analysis is based on evaluating the possible sources of impurities: their composition, amount, and rate of flow into the coolant in all possible operating regimes of a nuclear power plant. The relationship between the first-and second-loop purification systems and their effect on the distribution of impurities and the possibility and desirability of using not only cold traps but also other methods of purification, for example, hot traps, the structural implementation of the purification system, and the operation regime were all examined.The sources of impurities in systems with sodium coolant are oxygen, hydrogen (determining amount by mass and volume), products of corrosion of the structural materials, tritium, cesium if fuel-elements become depressurized, gaseous fission prod...
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