Abstract: Proposed is a nuclear fuel pellet manufactured with UO2 powder and being in a cylindrical shape, the nuclear fuel pellet including: a dish (10) provided in a shape of a spherical groove having a predetermined curvature and a diameter of 4.8 to 5.2 mm at a center of each of top and bottom surfaces of the nuclear fuel pellet; a shoulder (20) provided in an annular plane along a rim of the dish (10); a first chamfer (310) provided along a rim of the shoulder (20) while being adjacent to the shoulder (20); and a second chamfer (320) provided along a rim of the first chamfer (310), wherein a width (SW) of the shoulder (20) is 0.4565 mm to 0.6565 mm, an angle between the first chamfer (310) and a horizontal plane is 2.0°, and an angle between the second chamfer (320) and the horizontal plane is 18.0°.

Abstract: Provided is a nuclear-fuel sintered pellet based on oxide in which a plate-type fine precipitate material in a base of a sintered pellet of uranium dioxide, used as nuclear fuel in nuclear power plants, is uniformly dispersed in a matrix of uranium dioxide fuel thereof so as to form a donut-shaped precipitate cluster, and to a method of manufacturing the same. The plate-type fine precipitate material is uniformly precipitated in a tissue thereof or forms a donut-shaped precipitate cluster having a two-dimensional structure through dispersion to improve thermal and physical performance of the nuclear-fuel sintered pellet of uranium dioxide, whereby the creep deformation rate and thermal conductivity of the sintered pellet are improved. The nuclear-fuel sintered pellet based on oxide can reduce the Pellet-Clad Interaction (PCI) failure and the core temperature of nuclear fuel when an accident occurs, thereby significantly improving the safety of a nuclear reactor.

Abstract: A method of manufacturing a zirconium alloy for a nuclear fuel cladding tube includes melting a mixture of 0.5 wt % of Nb, 0.4 wt % of Mo, 0.1 to 0.15 wt % of Cu, 0.15 to 0.2 wt % of Fe, and a balance of zirconium to prepare a melted ingot; heat treating the melted ingot at 1,000 to 1,050° C. for 30 to 40 min. followed by quenching in water to prepare a heat-treated ingot; preheating the heat-treated ingot at 630 to 650° C. for 20 to 30 min. to prepare a preheated ingot followed by hot rolling the preheated ingot at a reduction ratio of 60 to 65% to provide a hot-rolled material; thrice performing vacuum annealing followed by cold-rolling; and vacuum annealing a third cold-rolled material in a final vacuum annealing at 510 to 520° C. for 7 to 9 hrs. to provide the zirconium alloy as a cold-rolled material.

Abstract: Provided is a nuclear-fuel sintered pellet based on oxide in which a plate-type fine precipitate material in a base of a sintered pellet of uranium dioxide, used as nuclear fuel in nuclear power plants, is uniformly dispersed in a matrix of uranium dioxide fuel thereof so as to form a donut-shaped precipitate cluster, and to a method of manufacturing the same. The plate-type fine precipitate material is uniformly precipitated in a tissue thereof or forms a donut-shaped precipitate cluster having a two-dimensional structure through dispersion to improve thermal and physical performance of the nuclear-fuel sintered pellet of uranium dioxide, whereby the creep deformation rate and thermal conductivity of the sintered pellet are improved. The nuclear-fuel sintered pellet based on oxide can reduce the Pellet-Clad Interaction (PCI) failure and the core temperature of nuclear fuel when an accident occurs, thereby significantly improving the safety of a nuclear reactor.

Abstract: Provided is a nuclear fuel pellet having excellent compression resistance and a reduced missing pellet surface, the pellet being a cylindrical shape and including, on each of a top surface and a bottom surface thereof: a dish configured as a spherical groove shape having a predetermined curved surface at a center; a land configured as a horizontal annular shape along a rim of the dish; and a chamfer configured as a shape in which a corner is chamfered at a predetermined angle along a rim of the land, wherein the chamfer includes: a primary chamfer adjacent to the rim of the land; and a secondary chamfer configured as a shape additionally chamfered along a rim of the primary chamfer.

Abstract: Provided is a nuclear fuel pellet having excellent impact resistance, the pellet being prepared with UO2 powder and having a cylindrical shape with a height of 9 to 13 mm and a horizontal cross-section diameter of 8 to 8.5 mm, and including: at each of a top surface and a bottom surface thereof, a dish configured as a spherical groove shape having a predetermined curvature and a groove diameter of 4.8 to 5.2 mm on a center; a shoulder configured as an annular plane along a rim of the dish; and a chamfer configured as a shape in which a corner is chamfered along a rim of the shoulder, wherein a width of the shoulder is 0.20 mm to 0.80 mm, and an angle between the chamfer and a horizontal plane is a 14-degree angle to 18-degree angle.

Abstract: A zirconium alloy is manufactured through melting; solution heat treatment at 1,000 to 1,050° C. (?) for 30 to 40 min and ?-quenching using water; preheating at 630 to 650° C. for 20 to 30 min and hot rolling at a reduction ratio of 60 to 65%; primary intermediate vacuum annealing at 570 to 590° C. for 3 to 4 hr and primarily cold-rolled at a reduction ratio of 30 to 40%; secondary intermediate vacuum annealing at 560 to 580° C. for 2 to 3 hr and secondarily cold-rolled at a reduction ratio of 50 to 60%; tertiary intermediate vacuum annealing at 560 to 580° C. for 2 to 3 hr and tertiarily cold-rolled at a reduction ratio of 30 to 40%; and final vacuum annealing at 460 to 590° C. for 7 to 9 hr.

Abstract: Provided is a silicon-carbide-sintered body in which plural crystal grains including silicon carbide are densely formed so as to be adjacent to each other. Sc and Y elements are present in a rich phase at a triple point at which interfaces of the crystal grains forming the sintered body meet each other without solid-solution of the elements in the crystal grains. Accordingly, sintering is feasible at a temperature of 1950° C. or lower, and an EB layer including a rare-earth-Si oxide containing the Sc and Y elements is formed on a surface thereof without an EB coating process, and is also formed up to the inner region of a silicon carbide base, resulting in strong three-dimensional bonding, so that the possibility of peeling of the EB layer is reduced and a new EB layer is formed even when peeling occurs, increasing the resistance to corrosion of the silicon carbide material.

Abstract: Disclosed is a method of manufacturing a zirconium alloy plate, wherein fine precipitates having an average size of 35 nm or less are uniformly distributed in a matrix through multi-pass hot rolling, thus increasing corrosion resistance and fatigue failure resistance, the method including forming a zirconium alloy ingot (step 1); subjecting the ingot of step 1 to beta annealing and rapid cooling (step 2); preheating the ingot of step 2 (step 3); forming a multi-pass hot-rolled plate through primary hot rolling and then air cooling during which secondary hot rolling is subsequently conducted (step 4); subjecting the multi-pass hot-rolled plate of step 4 to primary intermediate annealing and primary cold rolling (step 5); subjecting the rolled plate of step 5 to secondary intermediate annealing and secondary cold rolling (step 6); subjecting the rolled plate of step 6 to tertiary intermediate annealing and tertiary cold rolling (step 7); and finally annealing the rolled plate of step 7 (step 8).

Abstract: This invention relates to a composition and method for manufacturing a large-grained uranium oxide nuclear fuel pellet containing an additive. The nuclear fuel pellet is configured such that a uranium oxide powder and an additive powder composed of an Mg compound and a Si compound or Ca compound and a Al compound are mixed together, thus increasing a grain size to thus suppress the release of fission products, thereby increasing the stability of nuclear fuel, preventing cladding tubes from breaking, and contributing to the stable operation of nuclear power plants, ultimately increasing the overall stability of nuclear power plants including nuclear fuel.

Abstract: Disclosed herein are zirconium alloy compositions having a low hydrogen pick-up rate and high hydrogen embrittlement resistance. This zirconium alloy composition can be usefully used as a nuclear fuel components in a nuclear power plant because it has a very low hydrogen pick-up rate and high hydrogen embrittlement resistance under operation environments of nuclear power plant.

Abstract: A zirconium alloy is manufactured through melting; solution heat treatment at 1,000 to 1,050° C. for 30 to 40 min and ?-quenching using water; preheating at 630 to 650° C. for 20 to 30 min and hot rolling at a reduction ratio of 60 to 65%; primary intermediate vacuum annealing at 570 to 590° C. for 3 to 4 hr and primarily cold-rolled at a reduction ratio of 30 to 40%; secondary intermediate vacuum annealing at 560 to 580° C. for 2 to 3 hr and secondarily cold-rolled at a reduction ratio of 50 to 60%; tertiary intermediate vacuum annealing at 560 to 580° C. for 2 to 3 hr and tertiarily cold-rolled at a reduction ratio of 30 to 40%; and final vacuum annealing at 440 to 650° C. for 7 to 9 hr.

Abstract: A zirconium alloy is manufactured through melting; solution heat treatment at 1,000 to 1,050° C. (?) for 30 to 40 min and ?-quenching using water; preheating at 630 to 650° C. for 20 to 30 min and hot rolling at a reduction ratio of 60 to 65%; primary intermediate vacuum annealing at 570 to 590° C. for 3 to 4 hr and primarily cold-rolled at a reduction ratio of 30 to 40%; secondary intermediate vacuum annealing at 560 to 580° C. for 2 to 3 hr and secondarily cold-rolled at a reduction ratio of 50 to 60%; tertiary intermediate vacuum annealing at 560 to 580° C. for 2 to 3 hr and tertiarily cold-rolled at a reduction ratio of 30 to 40%; and final vacuum annealing at 460 to 590° C. for 7 to 9 hr.

Abstract: Disclosed herein are zirconium alloy compositions having a low hydrogen pick-up rate and high hydrogen embrittlement resistance. This zirconium alloy composition can be usefully used as a nuclear fuel components in a nuclear power plant because it has a very low hydrogen pick-up rate and high hydrogen embrittlement resistance under operation environments of nuclear power plant.

Abstract: Disclosed herein is a free drop tester wherein an angle of a test piece can be freely adjusted, a weight can fall on the exact spot, and an impulse of the weight can be controlled, and provides a free drop tester comprising a vertical guide disposed perpendicularly to the ground, a support body for supporting the vertical guide against the floor, a weight of which falling path is guided by the vertical guide, a test piece holder disposed under the bottom surface of the vertical guide, a horizontal rotation shaft of which one end is connected to a test piece holder, a power transmission unit connected to the other end of the horizontal rotation shaft, a test piece angle adjuster including a driving unit for supplying power to the power transmission unit, and a means for adjusting the height of the weight.