This paper uses the results of material science studies whose purpose was to investigate contact compatibility of uranium-zirconium carbonitride-based fuel pellets with the chosen refractory structural materials under the conditions of high-temperature pre-irradiation tests. Monocrystalline W-3Ta spacers with [110], [123], [112] orientations and polycrystalline W and Mo spacers served as contact pairs for the fuel pellets. The fuel and structural materials underwent annealing at various temperatures significantly exceeding 1200 °C. Duration of experiments for contact pairs with monocrystalline spacers was 80 and 300 hours, for polycrys-talline materials the exposure was 50, 100 and 150 hours. High stability of microstructure and chemistry of the U,Zr(C,N)-based composition was observed throughout the entire temperature-time diapason of the tests. No formation of intermediate phases capable of impairing air-tightness and operability of the cladding materials have been found in the interface zone. As a result of the studies performed, it has been found that orientation of W-3Ta monocrystals has no impact on the diffusion rate. Depth of uranium atoms penetration into monocrystalline samples didn’t exceed 15μm in the entire temperature-time diapason. For polycrystalline materials this value was 40 and 350 for W and Mo respectively.
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 paper presents an alternative design of the micro fuel element with the use of a core made of uranium-zirconium carbonitride fuel instead of traditional uranium dioxide which has a detrimental effect on the micro fuel element operation due to oxygen. This fuel as part of the four-layer TRISO-micro fuel element creates significantly lower gas pressure under the coating, it is compatible well with a pyrocarbon coating and does not crack it. The fuel has high thermal properties and it is much less subject to the “amoeba”-effect.
The particulars of the azimuthal deformation and the reasons for the formation of faceting on single-crystal molybdenum cladding with the <111> axis oriented longitudinally in tests for structural creep at 1650°C under internal argon gas pressure are investigated. It is shown that the periodic local nonuniformity of the azimuthal deformation of <111> cladding and the appearance of faceting are associated with the difference of the maximum shear stress in the active slip planes for single crystals with a bcc lattice in the range of orientations of the stretch axis <110>-<112>. Cladding sections located at the center of the interval <110>-<112> possess weak resistance to high-temperature creep and sites of deformation localization.Heat-resistant single-crystal alloys based on molybdenum and tungsten are the most promising material for cladding of high-temperature fuel elements for nuclear thermionic converters and nuclear power-propulsion plants for use in space [1][2][3]. The long service life at high temperatures under stress and the stringent requirements for preserving geometric dimensions of the cladding require materials with high resistance to creep and an adequate long-time strength limit. Heat-resistant single-crystal materials best satisfy the complex and contradictory set of requirements for the cladding material of fuel elements for nuclear facilities of this kind.A characteristic feature of single-crystal materials is that their mechanical properties depend on the crystallographic direction of the acting load. For uniform azimuthal deformation of single-crystal cladding under the internal pressure of the nuclear fuel and gaseous fission products, a crystallographic orientation of the longitudinal axis of the cladding with maximum possible symmetry in the transverse section is preferred.Of the three main crystallographic directions of the longitudinal axis of the cladding for a bcc lattice <100>, <110>, and <111>, the latter possesses the highest symmetry and therefore highest isotropy of mechanical properties. In the transverse section, the crystallographic directions <110> and <112> appear six times on the cylindrical surface, <123> 12 times, and <110>-<112> segments 12 times (Fig. 1). For single crystals of molybdenum, tungsten and their alloys, the short-time mechanical properties and creep have been studied in detail in a wide range of temperatures in uniaxial tension in the <111> and <110> directions and it has been shown that the <110> direction is least resistant to creep [4,5]. The experimental creep of samples with stretching axis <110> is usually used to determine the maximum deformation of the cladding when validating its service life. However, it is not obvious that this orientation during high-temperature creep is weakest in the segment <110>-<112> of the stretching axis. As one can see in Fig. 1, in cladding with longitudinal axis <111> the possible directions of the axis of stretching occupy an angle range of 30°in the segment <110>-<112>. High-temperature creep and longand short-term st...
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