Uranium-zirconium carbonitride based fuel (CNF) is a consequential modification of UN retaining almost all the advantages of UN. This fuel boasts rather high uranium content and high heat conductivity. Uranium-zirconium carbonitride fuel is safer due to its higher tolerance (inertia) to accident processes development. Thanks to its properties, the CNF is an attractive candidate material for using in reactors of various types. The main drawback of UZr(CN) is insufficient amount of data regarding its performance and behavior under irradiation, especially at high burn-ups. To address this problem, preparations are currently underway to perform a reactor experiment with the goal to study the properties of the CNF after reaching ≈ 7% burn-up. The parameters of the reactor experiment are as follows: cladding temperature not exceeding 800 K, power density not exceeding 750 W/cm3. For the purposes of reactor testing of the CNF pellets at high burn-up, an experimental capsule installed into the irradiating device has been developed. To verify the choice of designs of both the experimental capsule and the irradiating device thermophysical calculations were made and a programme of pre-irradiation experiments was completed. This paper elaborates on the results of the calculations demonstrating that the reactor tests fit the set goals and objectives.
The results of reflooding investigations of WWER-type and PWR-type model fuel assemblies (FA) carried out at largescale facilities (RBHT, SVD, PARAMETER) under Design Basis Accident (DBA) and Beyond Design Accident (BDA) conditions are presented. The regime parameters ranges are: the initial fuel rod temperature -T clad = 460÷1600 о С , water mass rate -ρw = 47÷160 kg/(m 2 •s), FA inlet subcooling -∆T sub = 11 -118 o C, pressure -p=0,14÷0,42 MPa and the linear heat flux density simulating decay heat -q l = 0÷2,3 kW/m. The generalizing correlation on the quench front velocity of model FA's describing the experimental data with 30% error is proposed. It allows to predict the latency of model fuel assemblies reflooding of various design and to evaluate the time of WWER and PWR reflooding.
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