Abstract: A method for manufacturing an annular nuclear fuel pellet is provided. In the method, an annular nuclear fuel green compact whose lateral cross-section is a trapezoid is prepared. The thickness of the annular nuclear fuel green compact reduces along one direction of the central axis, and a green density of the nuclear fuel green compact increases along one direction of the central axis. The annular nuclear fuel green compact is sintered under a reducing gas atmosphere so that the annular nuclear fuel pellet is obtained. According to this method, the annular pellet which has uniform inner and outer diameters and small diametric tolerances along the pellet height is fabricated without grinding the pellet surfaces.

Abstract: In a method of producing large-grained nuclear fuel pellet, Cr-compound contained in an uranium oxide green pellet is reduced to Cr phase at 1,470° C. or below and maintained to the Cr phase, and the uranium oxide green pellet containing the Cr-compound is then sintered at 1,650° C.-1,800° C. in a gas atmosphere of oxygen potential at which Cr element in the uranium oxide green pellet becomes liquid phase.

Abstract: A high-frequency inductive heating apparatus of ceramic material, whereby the nonconductive ceramic specimen in which induced current is not generated at room temperature is rapidly heated in a preheating housing, and a pressure-less sintering method using the same, are disclosed. The high-frequency inductive heating apparatus includes a preheating housing placed in a chamber to preheat a ceramic material; an induction coil installed around the preheating housing for supplying induced current so that the preheating housing is heated; and a high-frequency current generator for supplying high-frequency current to the induction coil. According to the present invention, inductive heating is made possible of nonconductive ceramic material for which inductive heating has thus far been impossible because induced current is not generated at room temperature, so that rapid heating by the self-heating of the specimen of ceramic material is possible.

Abstract: A method for fabricating a sintered duplex nuclear fuel pellet includes the steps of: preparing a first powder composed of a material selected from the group consisting of UO2 and UO2—Er2O3, and a second powder composed of UO2—Gd2O3 and a sintering additive; producing a duplex compact consisting of an annular outer portion composed of the first powder and a cylindrical inner portion composed of the second powder; and sintering the duplex compact under a reducing gas atmosphere, wherein the sintering additive contains manganese of 0.001% to 2% by weight based on the total weight of the cylindrical inner portion.

Abstract: There is provided a method of producing U3O8 powder having large surface area and small particle size by oxidizing defective UO2 pellets and manufacturing nuclear fuel pellets which are stable in a pore structure and high in density through the use of a mixture comprising UO2 powder and U3O8 powder. The method includes producing an U308 powder having a surface area of at least 1 m2/g by oxidizing defective UO2 pellets at a temperature of 300 to 370? in such a way that a maximum weight increase rate per 1 g of the UO2 pellets is up to 0.06 wt %/min; producing a mixed powder by mixing the U3O8 powder with an UO2 powder by 2 to 15 wt %; producing a compact by compression molding the mixed powder; and sintering the compact in a reducing gas atmosphere at a temperature of 1600 to 1800?. In addition, a small amount of an Al-compound may be added to the oxidized U3O8 powder before the U3O8 powder is mixed with the UO2 powder.

Abstract: The present invention relates to a method for measuring the content of Lanthanides dissolved in uranium oxide, wherein the Lanthanides content in the nuclear fuel pellet is measured using the thermo gravimetric analysis which measures the weight variation caused by the oxidation and heat treatment of the nuclear fuel pellet. This method provides an advantage in that the Lanthanide content can be measured using relatively simple equipments such as an electric furnace and a balance.

Abstract: The present invention relates to a nuclear fuel body comprising tungsten network and a method for manufacturing the same, more particularly, a nuclear fuel body in which tungsten network is continuously formed over the entire or some parts of a sintered body and a method for manufacturing the same. The sintered body in the nuclear fuel body of the present invention contains tungsten network having excellent heat conductivity, leading to the enhancement of heat conductivity of the nuclear fuel sintered body itself with decreasing the temperature of the nuclear fuel, so that it can contribute to the improvement in safety and performance of nuclear fuel.

Abstract: The present invention relates to a nuclear fuel body comprising tungsten network and a method for manufacturing the same, more particularly, a nuclear fuel body in which tungsten network is continuously formed over the entire or some parts of a sintered body and a method for manufacturing the same. The sintered body in the nuclear fuel body of the present invention contains tungsten network having excellent heat conductivity, leading to the enhancement of heat conductivity of the nuclear fuel sintered body itself with decreasing the temperature of the nuclear fuel, so that it can contribute to the improvement in safety and performance of nuclear fuel.

Abstract: The present invention relates to a method for measuring the content of Lanthanides dissolved in uranium oxide, wherein the Lanthanides content in the nuclear fuel pellet is measured using the thermo gravimetric analysis which measures the weight variation caused by the oxidation and heat treatment of the nuclear fuel pellet. This method provides an advantage in that the Lanthanide content can be measured using relatively simple equipments such as an electric furnace and a balance.

Abstract: Disclosed is a method of manufacturing large-grained uranium dioxide pellets through the use of U3O8 single crystals. The method consists of two main steps; producing U3O8 single crystals and manufacturing large-grained UO2 pellets. The U3O8 single crystals are produced by annealing U3O8 powder to make U3O8 polycrystalline aggregates having a large crystal size, and by dividing the U3O8 polycrystalline aggregate into its constituent U3O8 single crystals. Large-grained UO2 pellets are produced by forming a mixture comprising UO2 powder and the U3O8 single crystals, making granules of the mixture, pressing the granules into green pellets, and sintering the green pellets above 1600° C. for more than one hour in a reducing gas. The U3O8 powder is produced preferably by heating defective UO2 pellets in air to oxidize UO2 to U3O8. The invention provides a method of manufacturing a large-grained UO2 pellet with defective pellets reused.

Abstract: A method of manufacturing new UO2-based fuel pellets by recycling irradiated UO2-based fuel pellets. Irradiated UO2-based fuel pellets are oxidized so as to make U3O8-based powder, and then the U3O8 -based powder is mixed with an additive which contains at least one oxide of an element selected from the group consisting of niobium, titanium, vanadium, aluminum, magnesium, chromium, silicon and lithium. Green pellets are formed from the mixed powder, and then sintered, preferably at 1500° C. or higher, in a reducing gas atmosphere to produce UO2-based fuel pellets with high densities.

Abstract: A method is disclosed in which fuel scrap of UO.sub.2 alone or UO.sub.2 containing an oxide of plutonium, gadolinium or erbium is recycled into the manufacture of nuclear fuel pellets. The fuel scrap consisting of defective fuel pellets is comminuted through oxidation to fuel particles of U.sub.3 O.sub.8 alone or U.sub.3 O.sub.8 containing an oxide of plutonium, gadolinium or erbium, and a sintering aid containing an element selected from the group consisting of aluminum, magnesium, niobium, titanium, vanadium, chromium, lithium, silicon, tin and mixtures thereof is added in a quantity of about 0.02% to about 2% by weight to the sintering powder which consists of said recycled fuel particles and fresh fuel powder having a composition of UO.sub.2 alone or UO.sub.2 in a mixture of PuO.sub.2, Gd.sub.2 O.sub.3 or Er.sub.2 O.sub.3. The sintering powder is then mixed uniformly, in which the amount of the recycled fuel particles is in the range of about 10% to about 100% by weight.