Post-reactor studies of experimental fuel pins with mixed nitride uranium-plutonium fuel and EP823-Sh, ChS68-ID kh.d., and EK164-ID kh.d. steel cladding, which have been irradiated in BOR-60 and BN-600 reactors, indicated the nitriding and carburization of the cladding inner layer, which leads to an increase in its microhardness and a reduction in ductility characteristics. In fuel pin claddings, made of ChS68-ID kh.d. and EK164-ID kh.d. austenitic steel, the processes of carburization and nitriding correlate with sample destruction during mechanical tests. The present article generalizes and analyzes the results of post-reactor studies concerning the carburization and nitriding of the fuel pin claddings made of ChS68-ID kh.d., EK164-ID kh.d., and EP823-Sh steel. The carburization and nitriding of the cladding inner surface was studied by electron microprobe analysis using the cross-sections of fuel pins.

A total number of 18 experimental fuel assemblies (FAs), containing mixed nitride fuel rods with claddings of various geometry, were irradiated in an BN-600 core. By the end of 2022, JSC SSC NIIAR had completed the post-reactor studies of 16 ETVS BN-600 nitride fuel rods, 4 fuel rods with a gas sublayer irradiated in the composition of OU‑1 and OU‑2 collapsible irradiation devices, and 3 OU‑4 BOR-60 fuel rods with a liquid metal sublayer. Post-reactor studies of nitride fuel rods were preceded by pre-test calculations of their stress-strain and temperature states, taking into account the actual irradiation parameters. The results of the studies were compared with the calculated data. Fuel radiation swelling represents one of the most important factors that determine the degree of thermomechanical interaction between the fuel and the cladding, thus limiting the performance of nitride fuel rods. The present article summarizes and analyzes the currently obtained data on the swelling and its rate for mixed uranium-plutonium nitride fuel in fuel rods with a gas sublayer.

Irradiation of experimental fuel assemblies with uranium-plutonium nitride fuel and shells made of ferritic- martensitic steel EP823-Sh is conducted in order to ensure normal operation of BREST-OD-300 fuel rods in the BOR-60 and BN-600 reactors. This article presents the results of an investigation of the short-time mechanical properties of fuel rod cladding made of EP823-Sh steel after irradiation in BOR-60 and BN-600 reactors at temperatures 315–610°C up to maximum damaging dose 94 dpa, obtained for circular and longitudinal segmented samples. Empirical temperature dependences are proposed for the ultimate strength and total relative elongation of the irradiated EP823-Sh steel.

The purpose of the program was to validate the resource dependability of fuel rods with mixed uranium-plutonium nitride fuel, the consistency of the reproduction of characteristics and quality required for this pilot technology for manufacturing pellets, fuel rods and experimental fuel assemblies for testing in the BOR-60 and BN-600 reactors, as well as optimization, in terms of the results of reactor testing, of the design and manufacturing technology of fuel rods for BN-1200 and BREST-OD-300. The article presents key results of the investigations: methods, codes, and criteria have been developed to validate the performance of fuel rods with nitride fuel, the necessary set of pre-reactor properties of mixed nitride fuel and cladding materials has been obtained, a cycle of reactor tests of experimental fuel rods in BOR-60 and BN-600 has been completed, the maximum fuel burnup 9% h.a., damaging dose 107.6 dpa, technical designs of fuel rods for initial loading have been developed, and the manufacturing technology for fuel and fuel rods has been developed and optimized.

The results of an investigation of fission gas release from mixed nitride fuel after irradiation to maximum burnup 7.5% h. a. in 10 experimental FA in BN-600 are reported. It is shown that gas release commences at burnup >2% h.a. and increases with increasing burnup. The burnup dependence of gas release for mixed and uranium nitride fuel coincide quite well. The fission gas beneath the fuel-rod cladding consists of three basic components: xenon, helium, and krypton. The content of the remaining gases (N 2 , Ar, O 2 , CO 2 ) does not exceed tenths of a percent. Xenon has the highest specific release, followed by helium and krypton. Helium has the highest relative release (>40%). Xenon and krypton have almost identical release not exceeding 20% for most of the investigated fuel rods. A higher release of fission gas is associated with either an increase of the fuel temperature (>1500°C) or features of fuel microstructure.

Reactor tests in BN-600 and post-reactor studies of experimental FA of fuel rods with mixed nitride fuel and different cladding are being conducted in Project Breakthrough. In the present article, the irradiation parameters of fuel rods with mixed nitride fuel and cladding made from ChS68-ID c.d. austenitic steel are presented: maximum linear power density 38.3 kW/m, maximum fuel burnup 7.5% h.a., and maximum damage dose 73.6 dpa. No depressurization of the fuel elements was recorded. The results of post-reactor examination (PIE) of the irradiated fuel rods are presented – swelling, gas release from fuel, and mechanical and corrosion characteristics of the cladding.

The results of an investigation of gas release and swelling of uranium mononitride fuel after irradiation in standard and experimental fuel assemblies in the BR-10 reactor in the burnup range 3.4–8.4% h.a. are presented and analyzed. The results obtained were used to verify the fuel code DRAKON.