[en] Published experimental data relating to the enthalpy and specific heat of uranium metal, uranium alloys, uranium compounds and some elements that form alloys and compounds with uranium were reviewed and recommendation put forward for enthalpy and specific heat of these materials

[en] Mechanism and variable affecting on the thermal conductivity of UO₂ are studied and published experimental data relating to the thermal conductivity of UO₂, UO_2+x and mixed (U, Pu) oxides are reviewed and recommendation put forward for the thermal conductivity of these materials. Heat in UO₂ is transferred by phonon and heat transfer rate is decreased with increasing temperature in the low temperature region. However, heat in UO₂ is transferred by polaron and heat transfer rate is increased with increasing temperature in the high temperature region. The model of the thermal conductivity of UO₂ consists of hyperbolic term by phonon and exponential term by polaron as follows: κ=C₁/C₂+C₃t+C₄t² + C₅/t^5/2 * e^(C₆/t) where t = T/1000

[en] It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties, fission gas release, grain growth and et al. of the DUPIC fuel is different from the commercial UO₂ fuel. Published experimental data relating to the thermal expansion of UO₂, UO_2+x and mixed (U, Pu) oxides is reviewed and recommendation put forward for the lattice parameter and the thermal expansion of these materials. The theoretical density at room temperature is calculated using the lattice parameter. The density variation calculated using the thermal expansion model is recommended

[en] Published experimental data relating to the enthalpy and specific heat of UO₂, UO_2+x and mixed (U, Pu) oxides is reviewed and recommendation put forward for enthalpy and specific heat of these materials. We recommended Fink's eq. for sepcific heat of UO₂, Matzke's eq. for UO_2+x and SIMFUEL and Harding's eq. for PuO₂

[en] It is necessary to develop the technology to treat spent ceramic filters safely which have trapped emitted gases from the pre-treatment processes of pyroprocess. Investigations were made on the characteristics of representative immobilising materials and processes widely used, and the current situation of immobilising technology domestically and internationally. Based on this report, we intend to develop immobilising materials and processes appropriate to spent filter in the future

[en] There are three main categories in this report : 1) Basement construction of measurement standardization for nuclear material thermal property, 2) Reliability evaluation of measurement instrument, and 3) Standardization and industrial propagation

[en] It is required increasingly to demand a nuclear fuel for higher burnup to extend a fuel cycle and to reduce a amount of spent fuels. Various types of fission products were created in the matrix of fuel pellet during irradiation in the reactor. Especially, gaseous fission products are not almost solved in the matrix of the fuel, but released to the fuel pellet outside. With an increasing of the released gaseous fission products, the internal pressure of a fuel rod increases. The PCI(Pellet Cladding Interaction) and SCC(Stress Corrosion Cracking) will be resulted from the cladding tube inside due to the increasing internal pressure of released gaseous fission products. The release rate of gaseous fission products depends on the diffusion length or diffusion velocity of gaseous atoms in the fuel matrix. The former and the latter is accomplished by having the larger grain size and by decreasing the fuel temperature in the fuel, respectively. In this report, several approaches for nuclear fuel fabrication process, especially decreasing a fuel temperature and having a larger grain size in the matrix of fuel, are introduced and analyzed to solve the problems of these approaches

[en] Fission products, which were created in a UO₂ matrix during irradiation, depend on burnup, power, average neutron energy, irradiation time and cooling period, etc. According to their physical and chemical characteristics, fission products divide three groups, that is, soluble in UO₂(solid formation), insoluble in UO₂(precipitate) and volatile fission gases. Simfuels(simulated fuel) are fabricated to duplicate an irradiated nuclear fuel. There are abundant studies on some characteristics and fabrication of simfuel due to the non-active material. In this report, three different contents of (U,SS)O₂ pellet were fabricated. SS is a mixture of 7 soluble oxides -SrO, Y₂O₃, ZrO₂, TeO₂, La₂O₃, CeO₂ and Nd₂O₃- in UO₂ matrix. SS-1 have a 1.63wt%, SS-2 have a 3.25wt% and SS-4 have a 6.51wt% of soluble oxides in UO₂ matrix. These contents of (U,SS)O₂ pellets were investigated green densities, sintered densities, and microstructures change of (U,SS)O₂ matrix as a function of milling time and compaction pressure

[en] Fission products, which were created in a UO₂ matrix during irradiation, depend on burnup, power, average neutron energy, irradiation time and cooling period, etc. According to their physical and chemical characteristics, fission products divide three groups, that is, soluble in UO₂(solid formation), insoluble in UO₂(precipitate) and volatile fission gases. Simfuels(simulated fuel) are fabricated to duplicate an irradiated nuclear fuel. There are abundant studies on some characteristics and fabrication of simfuel due to the non-active material. In this report, among fission products, effect of precipitates on the fabrication of UO₂ sintered pellet was investigated. Three different contents of precipitates - that is, M-1(0.754wt%), M-2 (1.508wt%) and M-4(3.016 wt%)-were prepared as a function of burnup. And two parameters - milling time and heating step- were tried to fabricate sound sintered pellets. And then heating step showed a major factor to make defect-free sintered pellet, because heating steps was composed decomposition temperatures of precipitates

[en] Fission products, which were created in a UO₂ matrix during irradiation, depend on burnup, power, average neutron energy, irradiation time and cooling period, etc. According to their physical and chemical characteristics, fission products divide three groups, that is, soluble in UO₂(solid formation), insoluble in UO₂(precipitate) and volatile fission gases. Simfuels(simulated fuel) are fabricated to duplicate an irradiated nuclear fuel. There are abundant studies on some characteristics and fabrication of simfuel due to the non-active material. In this report, three different contents of (U,SS)O₂ pellet were fabricated. SS is a mixture of 7 soluble oxides -SrO, Y₂O₃, ZrO₂, TeO₂, La₂O₃, CeO₂ and Nd₂O₃- in UO₂ matrix. SS-1 have a 1.63wt%, SS-2 have a 3.25wt% and SS-4 have a 6.51wt% of soluble oxides in UO₂ matrix. Steady-state reep strain rate, transition stress, stress exponent and creep activation energy of these contents of (U,SS)O₂ pellets were investigated