Abstract: Proposed is a device for assembling and disassembling a top nozzle of a nuclear fuel assembly wherein the top nozzle of the nuclear fuel assembly is assembled and disassembled with improved convenience, and the load value is measured to ensure stable operations without damaging components. To this end, the device includes a pedestal including a threaded hole through both sides, a screw bar screw-coupled to the threaded hole to reciprocate in a straight line, a top nozzle plate configured to be detachably coupled to the screw bar and to be able to be hooked by and coupled to the top nozzle of the nuclear fuel assembly, and a load cell provided between the top nozzle plate and the screw bar to measure a load value applied to the top nozzle plate hooked on the top nozzle.

Abstract: Proposed is a method of synthesizing reactor core power distribution using an in-core instrument in a reactor core protection system, that is, a method of synthesizing reactor core power distribution for a reactor core protection system based on an in-core instrument signal using an ordinary kriging method. According to the present disclosure, a power of all fuel assemblies in a reactor core is calculated from powers of fuel assemblies where in-core instruments are located using the ordinary kriging methodology, and a hot-pin power distribution of each fuel assembly is synthesized from the power of all fuel assemblies calculated, whereby there is an effect that more accurate hot-pin axial power distribution, rather than pseudo hot-pin axial power distribution, may be synthesized.

Abstract: A light-water reactor nuclear fuel rod loading apparatus, which comprises a spacer grid holder and a loading power device installed with a traction module configured to reciprocally move toward the spacer grid holder. The loading apparatus further comprises a hollow cladding tube sample and a pulling bar comprising one end part and an opposite end part. The one end part passing in use through a cell of a spacer grid held by the spacer grid holder and configured to couple to the hollow cladding tube sample. The opposite end part is coupled to the traction module. The loading apparatus further comprises: a measuring device installed on the pulling bar and configured to measure loading force of the traction module loading the pulling bar; a holder cap; and a holder comprising an end part configured to in use pass through the cladding tube sample and couple to the holder cap.

Abstract: This invention relates to a solidifying agent for solidifying radioactive waste, and more particularly to a solidifying-agent composition for solidifying radioactive waste, including alumina cement and a gypsum powder. The solidifying-agent composition including alumina cement and a gypsum powder is capable of effectively minimizing an increase in the volume of a solidified radioactive waste product to a level satisfying physical and chemical safety regulations upon the solidification of radioactive waste.

Abstract: The present invention relates to an ultrasonic apparatus for decontaminating solid radioactive waste contaminated by radiation. The ultrasound apparatus for decontaminating solid radioactive waste includes: a container having a cylindrical shape; and a plurality of ultrasonic generators attached to an outer surface of a lower portion of the container and apart from each other at predetermined intervals, wherein in a state in which the container is filled with water, and the ultrasound apparatus decontaminates the solid radioactive waste by using ultrasonic waves generated from the plurality of ultrasonic generators, wherein the lower portion of the container has a downwardly-convex hemispherical shape, and the plurality of ultrasonic generators are attached to the hemispherical lower portion of the container such that ultrasonic waves generated from the plurality of ultrasonic generators may be focused on one point in the container.

Abstract: A method for preparing a UO2 mixture powder for nuclear fuel manufacturing, comprises the steps of: (a) weighing and sieving, by means of an automatic injection device, a UO2 powder, a porogen and a lubricant, and injecting same into a UC container; and (b) mixing the UO2 powder, the porogen and the lubricant by means of an IBC blender. According to a method for preparing a UO2 mixture powder for nuclear fuel manufacturing, mixing time is short, and degrees of mixing and homogeneity of a prepared UO2 mixture powder are excellent.

Abstract: A device for supplementing structural integrity of a spent nuclear fuel assembly during handling is provided. The spent nuclear fuel assembly includes spacer grids for supporting fuel rods, guide tubes, and top and bottom nozzles for supporting upper and lower ends of the guide tubes. The device includes a housing, a rod member and fixing members. The housing includes a plurality of free end portions, stopper protrusions formed at an end of each of the free end portions, a first male thread formed on an outer surface, and a female thread formed on an inner surface. The rod member includes a pressurization protrusion at one end and a second male thread for engaging with the female thread of the housing. The fixing members are configured engage with the first male thread to secure the device to the top nozzle of the spent nuclear fuel assembly.

Abstract: The present disclosure relates to a pellet containing an oxide additive to improve a nuclear-fission-gas-adsorption ability of a uranium-dioxide pellet used as nuclear fuel and increase the grain size thereof, and to a method of manufacturing the same. A La2O3—Al2O3—SiO2 sintering additive is added to uranium dioxide so that mass movement is accelerated due to the liquid phase generated during sintering of the uranium-dioxide pellet, which promotes the growth of grains thereof. Further, since less volatilization occurs during sintering due to the low vapor pressure of the liquid phase, efficient additive performance is exhibited, so the liquid phase surrounding the grain boundary effectively adsorbs cesium, which is a nuclear fission gas.

Abstract: Provided is an apparatus for fatigue testing a bulge tool having a WH-type skeleton, the apparatus including: a fixing bracket having tool holes penetrated through opposite sides thereof; a tool housing coupled to the tool hole of the fixing bracket and having the bulge tool inserted and installed therein; a moving rail installed at one side of the fixing bracket in a lengthwise direction of the tool housing and providing a reciprocating movement path facing the tool housing; a moving bracket reciprocating along the moving rail; a pusher protrudingly installed from the moving bracket toward the tool housing and moving in and out of the bulge tool; a measurement means installed between the pusher and the moving bracket, measuring a load applied to the bulge tool; and a drive means for generating power reciprocating the moving bracket on the moving rail.

Abstract: Disclosed is a system for sensing a UF6 gas leak in a nuclear fuel manufacturing process. The system is configured to sense whether or not there is a UF6 gas leak by optically detecting UO2F2 in a solid state generated due to a reaction with outside air. This allows prevention of damage to a detection apparatus by means of sensing in a non-contact manner whether or not there is a UF6 gas leak. Further, the system extends the mechanical life of and reduces the maintenance and repair costs for the detection apparatus.

Abstract: The present invention relates to a method for preparing a fully ceramic capsulated nuclear fuel material containing three-layer-structured isotropic nuclear fuel particles coated with a ceramic having a composition which has a higher shrinkage than a matrix in order to prevent cracking of ceramic nuclear fuel, wherein the three-layer-structured nuclear fuel particles before coating is included in the range of between 5 and 40 fractions by volume based on after sintering.

Abstract: Proposed is an internal electrode insertion device for transferring the internal electrode to a welding position in order to weld a skeleton of a nuclear fuel assembly, the device including: a table; a plurality of internal electrode bars disposed in parallel in a plurality of rows on a top portion of the table; guide members provided on the table and configured to guide a horizontal movement of the internal electrode bars; guide rails provided in parallel with an alignment direction of the internal electrode bars on the top portion of the table; a driving unit provided with a servomotor thereon, thereby performing horizontal movement driving along the guide rails; and a control unit for controlling driving of the servomotor based on a preset position of each of the internal electrodes.

Abstract: Embodiments of the disclosure relate to a ferritic alloy having excellent ability to withstand nuclear power plant accidents and a method of manufacturing a nuclear fuel cladding tube using the same. The alloy includes iron (Fe), aluminum (Al), chromium (Cr), and nickel (Ni). The nickel (Ni) may be included 0.5 to 10 wt % based on a total amount of the alloy. The chromium may be included 13 to 18 wt % based on the total amount of the alloy. The aluminum may be included 5 to 7 wt % based on the total amount of the alloy.

Abstract: The present invention relates to a method for preparing a fully ceramic capsulated nuclear fuel material containing three-layer-structured isotropic nuclear fuel particles coated with a ceramic having a composition which has a higher shrinkage than a matrix in order to prevent cracking of ceramic nuclear fuel, wherein the three-layer-structured nuclear fuel particles before coating is included in the range of between 5 and 40 fractions by volume based on after sintering.

Abstract: Proposed is a light-water reactor nuclear fuel rod loading force measuring apparatus provided on a nuclear fuel rod loading apparatus, which comprises a spacer grid holder and a loading power device installed with a traction module configured to reciprocally move toward the spacer grid holder, the measuring apparatus including: a hollow cladding tube sample positioned on one side of a spacer grid and having the same diameter as a nuclear fuel rod cladding tube; a pulling bar having one end part passing through a cell of the spacer grid and coupled to the hollow cladding tube sample and an opposite end part coupled to the traction module; and a measuring means installed on the pulling bar and configured to measure force of the traction module loading the pulling bar.

Abstract: System for separating and coupling a nuclear fuel assembly from/to a top nozzle which has a flow channel plate with guide holes. The system includes a lock insert and a separation member. The lock insert includes an insertion part provided on a top portion of a hollow body. The separation member is configured to separate the insertion part from a guide hole. The insertion part is variable in size. The insertion part comprises a first latching member and a second latching member, each having a step which contacts the flow channel plate. The first latching member includes a latching groove which is inserted into a member protruding from the top surface of the flow channel plate. The second latching member contacts a bottom surface of the flow channel plate. The separation member provides a space accommodating an outer circumferential surface of the first latching member.

Abstract: Disclosed is a composition of a lubricant for cold pilgering of zirconium alloy tubes. More particularly, disclosed is a composition of an external lubricant for cold pilgering of a zirconium alloy cladding tube, the composition exhibiting excellent lubricity and decomposition against microbes.

Abstract: Disclosed is a composition of a lubricant for cold pilgering of zirconium alloy tubes. More particularly, disclosed is a composition of an internal lubricant for cold pilgering of a zirconium alloy cladding tube, the composition exhibiting excellent lubricity and decomposition against microbes.

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: Proposed are nuclear fuel pellets showing high oxidation resistance in a steam atmosphere and a method for manufacturing same. The method includes: preparing a powder mixture by mixing a sintering additive powder including Cr2O3, MnO, and SiO2 with a uranium dioxide powder; forming a molded body by subjecting the powder mixture to compression molding; and sintering the molded body in a weak oxidative atmosphere in which an oxygen potential is ?581.9 kJ/mol to ?218.2 kJ/mol. The nuclear fuel pellets contain 0.05% to 0.16% by weight of the sintering additive composed of Cr2O3, MnO, and SiO2. A liquid phase generated during the sintering accelerates grain growth and inhibits reaction between uranium dioxide with steam by forming a film at the grain boundary of the uranium dioxide. This reduces leakage of a fission material by improving high-temperature water vapor oxidation resistance at around 1204° C. in a loss-of-coolant accident condition.