| KOREA HYDRO AND NUCLEAR POWER CO., LTD. Patent applications |
| Patent application number | Title | Published |
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| 20120135157 | Coating and Ion Beam Mixing Apparatus and Method to Enhance the Corrosion Resistance of the Materials at the Elevated Temperature Using the Same - The present invention relates to a ceramic coating and ion beam mixing apparatus for improving corrosion resistance, and a method of reforming an interface between a coating material and a base material. In samples fabricated using the coating and ion beam mixing apparatus, adhesiveness is improved, and the base material is reinforced, thereby improving resistance to thermal stress at high temperatures and high-temperature corrosion resistance of a material to be used in a sulfuric acid decomposition apparatus for producing hydrogen. | 05-31-2012 |
| 20120123721 | ELECTRONIC SCALE HAVING FUNCTION OF COMPENSATING FOR AIR PRESSURE CHANGES IN GLOVE BOX - The present invention relates to an electronic scale having an air pressure change compensation function, which can precisely measure the weight of an object to be weighed even in an environment in which air pressure changes moment by moment in an airtight space such as in a glove box. According to the present invention, there is provided an electronic scale having an internal air pressure measurement sensor, by which the function of correcting the weight of an object to be weighed is provided even in an environment in which air pressure changes such as in the glove box is provided, thus enabling the weight of a sample to be precisely measured even in an environment in which air pressure changes moment by moment. | 05-17-2012 |
| 20120106693 | HIGH Cr FERRITIC/MARTENSITIC STEELS HAVING AN IMPROVED CREEP RESISTANCE FOR IN-CORE COMPONENT MATERIALS IN NUCLEAR REACTOR, AND PREPARATION METHOD THEREOF - Disclosed herein is a high Cr Ferritic/Martensitic steel comprising 0.04 to 0.13% by weight of carbon, 0.03 to 0.07% by weight of silicon, 0.40 to 0.50% by weight of manganese, 0.40 to 0.50% by weight of nickel, 8.5 to 9.5% by weight of chromium, 0.45 to 0.55% by weight of molybdenum, 0.10 to 0.25% by weight of vanadium, 0.02 to 0.10% by weight of tantalum, 0.21 to 0.25% by weight of niobium, 1.5 to 3.0% by weight of tungsten, 0.015 to 0.025% by weight of nitrogen, 0.01 to 0.02% by weight of boron and iron balance. By regulating the contents of alloying elements such as nitrogen, born, the high Cr Ferritic/Martensitic steel with to superior tensile strength and creep resistance is provided, and can be effectively used as an in-core component material for sodium-cooled fast reactor (SFR). | 05-03-2012 |
| 20120063562 | DUAL-COOLED FUEL ROD'S SPACER GRIDS WITH UPPER AND LOWER CROSS-WAVY-SHAPE DIMPLE - A dual-cooled fuel rod's spacer grid with upper and lower dimples, including a blocking area of a flow passage that coolant flows through is reduced and dual-cooled fuel rods are supported, and reduces a turbulent flow of the coolant as well as vibrations of the dual-cooled fuel rods, thereby lessening fretting damage done to the rods. The spacer grid includes a plurality of unit grid straps, each of which includes a body disposed in a vertical direction, an upper dimple protruding from an upper portion of the body, and a lower dimple spaced apart from the upper dimple in a downward direction and protruding from a lower portion of the body. The unit grid straps form a grid structure that have inner grid holes into which the dual-cooled fuel rods are held, and the held dual-cooled fuel rods are each supported in four directions by the upper and lower dimples. | 03-15-2012 |
| 20120056108 | SURFACE MODIFICATION METHOD OF FLUOROPOLYMERS BY ELECTRON BEAM IRRADIATION AND THE FABRICATION OF SUPERHYDROPHOBIC SURFACES USING THE SAME - A method for the surface modification of fluoropolymer films using electron beam irradiation to generate superhydrophobic surfaces is provided. This surface modification method can cause simultaneously both a physical modification roughening the fluoropolymer surfaces and a chemical modification changing the surface composition of the fluoropolymers, and therefore fabricating the superhydrophobicity on a fluoropolymer surface by controlling the dose of electron beam irradiation. Therefore, this method for the surface modification of fluoropolymers by electron beam irradiation can be used in the generation of superhydrophobic surfaces required in various industries such as paint, glue, fine chemistry, electrical and electronics, cars, and display manufacturing. | 03-08-2012 |
| 20120043469 | RADIATION DETECTING DEVICE TO MEASURE GAMMA-RAY AND NEUTRON DISCRIMINATELY - A radiation detecting device is provided, according to which it is possible use only one radiation detecting device to measure radiation and measure gamma ray and neutron at once and discriminately in a restricted space. The radiation detecting device includes a radiation detecting unit to measure gamma ray and neutron discriminately at once, and a signal processing circuit which applies voltage to the neutron detecting unit and indicates measured gamma ray and neutron discriminately. | 02-23-2012 |
| 20110162764 | HIGH-CR FERRITIC/MARTENSITIC STEEL HAVING IMPROVED CREEP RESISTANCE AND PREPARATION METHOD THEREOF - High-Cr ferritic/martensitic steels having an improved tensile strength and creep resistance are provided, which includes 0.04˜0.13 weight % of carbon, 0.03˜0.07 weight % of silicon, 0.40˜0.50 weight % of manganese, 0.40˜0.50 weight % of nickel, 8.5˜9.5 weight % of chromium, 0.45˜0.55 weight % of molybdenum, 0.10˜0.25 weight % of vanadium, 0.02˜0.10 weight % of tantalum, 0.15˜0.25 weight % of niobium, 1.5˜3.0 weight % of tungsten, 0.05˜0.12 weight % of nitrogen, 0.004˜0.008 weight % of boron, and optionally, 0.002˜0.010 weight % of phosphorus or 0.01˜0.08 weight % of zirconium, and iron balance. By regulating the contents of alloying elements such as niobium, tantalum, tungsten, nitrogen, boron, zirconium, carbon, the high-Cr ferritic/martensitic steels with superior tensile strength and creep resistance are provided, and can be effectively used as an in-core structural material for Generation IV sodium-cooled fast reactor (SFR) which is used under high temperature and high irradiation conditions. | 07-07-2011 |
| 20110052461 | FILTER TYPE TRAPPING AGENT FOR VOLATILE CESIUM COMPOUND AND TRAPPING METHOD FOR VOLATILE CESIUM COMPOUND THEREOF - A filter type trapping agent for volatile cesium compound and trapping method for volatile cesium compound thereof are provided. More particularly, a filter type trapping agent for volatile cesium compound including silica 40˜65% by weight of silica, 15˜30% by weight of alumina, 5˜15% by weight of iron oxide, 1˜15% by weight of molybdenum oxide, 1˜10% by weight of chromium oxide, and 1˜10% by weight of vanadium oxide and trapping method for volatile cesium compound thereof are provided. Through a filter type trapping agent for volatile cesium compound and a trapping method, only cesium can be selectively separated among the nuclear fission gases. Accordingly, by disposing only the filter where cesium is trapped, the efficiency of an off-gas process improves, expense for disposing filter wastes decreases, and a cesium isotope of the waste filter can be recycled. Therefore, many forms of cesium compound gas are made insoluble efficiently. | 03-03-2011 |
| 20110051882 | TRUSS-REINFORCED SPACER GRID AND METHOD OF MANUFACTURING THE SAME - A truss-reinforced spacer grid and a method of manufacturing the same are provided, in which truss members having a small diameter are woven to form a truss structure surrounded by an external plate, and the truss structure is joined to the external plate to thereby improve the strength of the mechanical structure. The truss-reinforced spacer grid includes a truss structure in which horizontal trusses formed by horizontally weaving a plurality of truss members are vertically disposed at regular intervals, and an external plate is joined with ends of the horizontal trusses and surrounds the truss structure. | 03-03-2011 |
| 20100322370 | PROCESS OF MANUFACTURING ZIRCONIUM ALLOY FOR FUEL GUIDE TUBE AND MEASURING TUBE HAVING HIGH STRENGTH AND EXCELLENT CORROSION RESISTANCE - A process of manufacturing zirconium alloy. The process may be used to make a nuclear fuel guide tube and/or a measuring tube which are main components of a nuclear fuel assembly structure. While a nuclear fuel guide tube and a measuring tube are manufactured by performing three-step cold working, and intermediate and final thermal annealing from a semi-finished TREX shell in the conventional method, the present invention relates to zirconium alloy undergoing two-step cold working, and intermediate and final thermal annealing from a TREX shell, resulting in enhanced strength and corrosion resistance. The present invention may be applied to a nuclear fuel guide tube and a measuring tube used for a nuclear fuel assembly in a light water nuclear reactor because, by the shortened process, high percentage reduction in thickness between processes and an decrease in thermal annealing time may sustain high strength and excellent corrosion resistance, and achieve economy of manufacture by reducing the number of processes. | 12-23-2010 |
| 20100266094 | DUAL-COOLED NUCLEAR FUEL ROD HAVING ANNULAR PLUGS AND METHOD OF MANUFACTURING THE SAME - A dual-cooled nuclear fuel rod and a method of manufacturing the same are provided. The nuclear fuel rod includes an outer cladding tube having a circular cross section, an inner cladding tube having an outer diameter smaller than an inner diameter of the outer cladding tube, and a length longer than the outer cladding tube, and located in parallel in the outer cladding tube, a pellet charged in a space between the outer and inner cladding tubes and generating energy by nuclear fission, and first and second end plugs coupling opposite ends of the outer cladding tube to stepped outer joints formed on outer circumferences of first ends thereof and coupling opposite ends of the inner cladding tube to stepped inner joints formed on inner circumferences of the first ends thereof. | 10-21-2010 |
| 20100177860 | FULLY PASSIVE DECAY HEAT REMOVAL SYSTEM FOR SODIUM-COOLED FAST REACTORS THAT UTILIZES PARTIALLY IMMERSED DECAY HEAT EXCHANGER - Disclosed herein is a fully passive decay heat removal system utilizing a partially immersed heat exchanger, the system comprising: a hot pool; an intermediate heat exchanger which heat-exchanges with the sodium of the hot pool; a cold pool; a support barrel extending vertically through the boundary between the hot pool and the cold pool; a sodium-sodium decay heat exchanger received in the support barrel; a sodium-air heat exchanger provided at a position higher than the sodium-sodium decay heat exchanger; an intermediate sodium loop connecting the sodium-sodium decay heat exchanger with the sodium-air heat exchanger; and a primary pump, wherein a portion of the effective heat transfer tube of the sodium-sodium decay heat exchanger is immersed in the cold pool, particularly in a normal operating state, and the surface of the lower end of a shroud for the sodium-sodium decay heat exchanger, the lower end being immersed in the sodium of the cold pool, has perforated holes. | 07-15-2010 |
| 20100172460 | PERFORATED PLATE SUPPORT FOR DUAL-COOLED SEGMENTED FUEL ROD - A perforated plate support supports dual-cooled fuel rods, each of which has concentric outer and inner tubes and is coupled with upper and lower end plugs at upper and lower ends thereof, and guide thimbles, each of which is used as a passage for a control rod. The perforated plate support is formed as a support plate having the shape of a flat plate, which includes internal channel holes, each of which has a diameter corresponding to an outer diameter of the inner tube, guide thimble holes, each of which has a diameter corresponding to an outer diameter of the guide thimble, and sub-channel holes around each internal channel hole. The upper or lower end of the dual-cooled fuel rod is coupled to the support plate such that the outer diameter of the inner tube is matched with the diameter of the internal channel hole. | 07-08-2010 |
| 20100142668 | POROUS PLENUM SPACER FOR DUAL-COOLED FUEL ROD - A porous plenum spacer is inserted into the plenum of a dual-cooled fuel rod having concentric inner and outer cladding tubes. The porous plenum spacer includes a hollow cylindrical body inserted into the annular space between the inner and outer cladding tubes. The hollow cylindrical body includes a plurality of through-holes formed in an outer circumference thereof or at least one groove formed in one of outer and inner circumferences thereof in a lengthwise direction. Pores formed by the through-holes or the grooves of the hollow cylindrical body of the porous plenum spacer are allowed to secure a space containing fission gas inevitably generated by a nuclear reaction. | 06-10-2010 |
| 20100135452 | LIQUID-METAL-COOLED FAST REACTOR CORE COMPRISING NUCLEAR FUEL ASSEMBLY WITH NUCLEAR FUEL RODS WITH VARYING FUEL CLADDING THICKNESS IN EACH OF THE REACTOR CORE REGIONS - A liquid-metal cooled fast reactor core having a nuclear fuel assembly constituted of nuclear fuel rods with varying cladding thicknesses in reactor core regions, in which: the nuclear fuel assembly ( | 06-03-2010 |
| 20100128835 | INTERMEDIATE END PLUG ASSEMBLY FOR SEGMENTED FUEL ROD AND SEGMENTED FUEL ROD HAVING THE SAME - An intermediate end plug assembly for a segmented fuel rod can stably support the fuel rod to the end of its cycle even if an interval between the fuel rods becomes narrow due to application of a dual-cooled fuel rod, and reduce excess vibration induced by flows of interior and exterior channels of the dual-cooled fuel rod for obtaining high burnup and output. To this end, the fuel rod has a segmented structure so as to make its length short. A lower intermediate end plug includes at least one channel hole, through which a coolant flows into an internal channel of the fuel rod, so that a possibility of causing departure from nuclear boiling ratio (DNBR) of the dual-cooled fuel rod is reduced. | 05-27-2010 |
| 20100124669 | JOINING METHOD BETWEEN Fe-BASED STEELS AND Ti/Ti-BASED ALLOYS HAVING JOINT STRENGTH HIGHER THAN THOSE OF BASE METALS BY USING INTERLAYERS AND THE JOINTS PRODUCED USING THE METHOD - A joining method between Fe-based steel and Ti/Ti-based alloys having a joint strength higher than those of base metals by using interlayers. The production of intermetallic compounds at a joint portion between Fe-based steel and Ti/Ti-based alloys can be prevented using interlayers, and strong interface diffusion bonding can be formed at interfaces between interlayers, thereby producing a high-strength joint. Accordingly, the present disclosure can be used to develop high-strength, high-functional advanced composite materials. | 05-20-2010 |
| 20100108204 | ZIRCONIUM ALLOY COMPOSITION FOR NUCLEAR FUEL CLADDING TUBE FORMING PROTECTIVE OXIDE FILM, ZIRCONIUM ALLOY NUCLEAR FUEL CLADDING TUBE MANUFACTURED USING THE COMPOSITION, AND METHOD OF MANUFACTURING THE ZIRCONIUM ALLOY NUCLEAR FUEL CLADDING TUBE - Disclosed herein is a zirconium alloy composition for nuclear fuel cladding tubes, comprising: 1.6˜2.0 wt % of Nb; 0.05˜0.14 wt % of Sn; 0.02˜0.2 wt % of one or more elements selected from the group consisting of Fe, Cr and Cu; 0.09˜0.15 wt % of O; 0.008˜0.012 wt % of Si; and a balance of Zr, a nuclear fuel cladding tube comprising the zirconium alloy composition, and a method of manufacturing the nuclear fuel cladding tube. Since the nuclear fuel cladding tube made of the zirconium alloy composition can maintain excellent corrosion resistance by forming a protective oxide film thereon under the conditions of high-temperature and high-pressure cooling water and water vapor, it can be usefully used as a nuclear fuel cladding tube for light water reactors or heavy water reactors, thus improving the economical efficiency and safety of the use of nuclear fuel. | 05-06-2010 |
| 20100071408 | METHOD OF RECYCLING LiCl SALT WASTES BY USING LAYER CRYSTALLIZATION AND APPARATUS FOR THE SAME - Disclosed herein are a method of recycling LiCl salt wastes comprising radionuclides and an apparatus using the same. The method includes a) solidifying a LiCl salt contained in the LiCl salt wastes and contacting the outer wall of a housing, by charging a crystallizer comprising the housing having an internal accommodating space and an air cooler in the internal accommodating space into a crystallizing furnace accommodating the LiCl salt wastes comprising the radionuclides, and by cooling the housing to a temperature of two-phase region where the liquid state and the solid state of the LiCl salt waste coexist, b) separating the crystallizer where the LiCl salt is solidified from the crystallizing furnace, c) recycling the LiCl salt by heating the separated crystallizer to melt the solidified LiCl salt. | 03-25-2010 |
| 20100051246 | HIGH TEMPERATURE AND HIGH PRESSURE CORROSION RESISTANT PROCESS HEAT EXCHANGER FOR A NUCLEAR HYDROGEN PRODUCTION SYSTEM - A high-temperature and high-pressure corrosion-resistant process heat exchanger for a nuclear hydrogen production system decomposes sulfite (SO | 03-04-2010 |
| 20100032288 | COATING AND ION BEAM MIXING APPARATUS AND METHOD TO ENHANCE THE CORROSION RESISTANCE OF THE MATERIALS AT THE ELEVATED TEMPERATURE USING THE SAME - The present invention relates, in general, to shoes for measuring the quantity of motion and a method of measuring the quantity of motion using the shoes and, more particularly, to artificial intelligence shoes, in which various numerical values (calorie consumption, body fat, and a pulse), measured by a walking sensor ( | 02-11-2010 |
| 20090288949 | REFERENCE ELECTRODE INCLUDING ELECTROLYTE CONTAINING OPTICALLY-ACTIVE MATERIAL AND AUTOMATIC ELECTROCHEMICAL POTENTIAL CORRECTION APPARATUS USING THE SAME - Disclosed herein is a reference electrode including an electrolyte containing an optically-active material, including: an electrode body provided at an end thereof with an electrolyte separation membrane and charged therein with an optically-active material and an electrolyte solution; an inner electrode disposed in the electrode body to be immersed in the electrolyte solution; and an absorbance measurement probe for transmitting light to the electrolyte solution and collecting reflected light waves, which is disposed in the electrode body to be immersed in the electrolyte solution. Since the concentration of an electrode reaction material, such as Cl | 11-26-2009 |
| 20090279657 | SAFETY INJECTION TANK WITH GRAVITY DRIVEN FLUIDIC DEVICE - A safety injection tank, used for quickly injecting emergency core cooling water (ECCW) to a reactor vessel in the case of a cold leg large break accident (CLLBA) in a pressurized water reactor (PWR), is disclosed. The safety injection tank has a gravity-driven fluidic device configured to efficiently change the ECCW injection mode from a high flow injection mode to a low flow injection mode. The gravity-driven fluidic device includes a spring placed in the upper end of the vertical pipe, and a vertically movable water tub placed on the spring so as to be movable in a vertical direction. When ECCW contained in the pressure vessel is discharged and the water level is reduced lower than the height of the tub, the tub is moved downwards such that the lower surface thereof comes into contact with the vertical pipe and closes the high flow inlet port. | 11-12-2009 |
| 20090252883 | METHOD OF PREVENTING CORROSION DEGRADATION USING NI OR NI-ALLOY PLATING - Disclosed herein is a method of preventing corrosion degradation in a defective region including an expansion transition region and/or an expansion region of a heat transfer tube of a steam generator in a nuclear power plant by using nickel (Ni) plating or nickel (Ni) alloy plating. The method can prevent various types of corrosion damage, such as pitting corrosion, abrasion, stress corrosion cracking, lead-induced stress corrosion cracking and the like, occurring during the operation of the steam generator, and particularly, pitting corrosion or primary and secondary stress corrosion cracking, so that the life span of the steam generator is increased, maintenance costs are reduced, and the operation rate of a nuclear power plant is increased, with the result that the unit cost of the production of electric power can be decreased, thereby improving economic efficiency. Further, the method can be usefully used to prevent the corrosion damage of parts and equipment of nuclear, hydroelectric or thermoelectric power plants or of petrochemical plants, and that of industrial and machine parts and equipment, and parts and equipment in a defense industry. | 10-08-2009 |
| 20090245453 | DECAY HEAT REMOVAL SYSTEM COMPRISING HEAT PIPE HEAT EXCHANGER - Disclosed herein is a decay heat removal system, including: a decay heat exchanger that absorbs decay heat generated by a nuclear reactor; a heat pipe heat exchanger that receives the decay heat from the decay heat exchanger through a sodium loop for heat removal and then discharges the decay heat to the outside; and a sodium-air heat exchanger that is connected to the heat pipe heat exchanger through the sodium loop and discharges the decay heat transferred thereto through the sodium loop to the outside. According to the decay heat removal system, a heat removal capability can be realized by the heat pipe heat exchanger at such a high temperature at which the safety of a nuclear reactor is under threat, and a cooling effect can be obtained through the sodium-air heat exchanger at a temperature lower than that temperature. | 10-01-2009 |
| 20090232267 | EMERGENCY CORE COOLING SYSTEM HAVING CORE BARREL INJECTION EXTENSION DUCTS - An emergency core cooling system directly injects emergency core cooling water, which is supplied from a high-pressure safety injection pump or a safety injection tank for a pressurized light water reactor, into a reactor vessel downcomer. A pipe connector is completely removed from between each direct vessel injection nozzle and each injection extension duct installed on an outer surface of the core barrel, which are opposite to each other. An emergency core cooling water intake port, through which the water is injected from each direct vessel injection nozzle, is formed on the surface of each injection extension duct facing an axis of each direct vessel injection nozzle. Thereby, a structure in which a jet of the emergency core cooling water flows into the injection extension ducts is adopted in a hydraulic connection fashion. | 09-17-2009 |
| 20090141851 | NUCLEAR FUEL ROD FOR FAST REACTORS WITH OXIDE COATING LAYER ON INNER SURFACE OF CLADDING, AND MANUFACTURING METHOD THEREOF - Disclosed herein are a nuclear fuel rod for fast reactors, which includes an oxide coating layer formed on the inner surface of a cladding, and a manufacturing method thereof. The nuclear fuel rod for fast reactors, which includes the oxide coating layer formed on the inner surface of the cladding, can increase the maximum permissible burnup and maximum permissible temperature of the metallic fuel slug for fast reactors so as to prolong the its lifecycle in the fast reactors, thus increasing economic efficiency. Also, the fuel rod is manufactured in a simpler manner compared to the existing method, in which a metal liner is formed, and the disclosed method enables the cladding of the fuel rod to be manufactured in an easy and cost-effective way. | 06-04-2009 |
| 20090116606 | JOINT TONG APPARATUS FOR RADIATION SHIELDING FACILITY - A joint tong apparatus for a radiation shielding facility. A spherical ball has a through-hole. An inner spherical socket and an outer spherical socket are installed in a hole formed in a partition between a radiation-shielded room and a control room so as to enclose the spherical ball on inner and outer sides. A bar is inserted and coupled into and to the through-hole of the spherical ball. An inner joint assembly has a first housing coupled to a shielded room-side end of the bar and a first pivot member pivotably mounted on a free end of the first housing. An outer joint assembly having a second housing coupled to a control room-side end of the bar and a second pivot member pivotably mounted on a free end of the second housing. The apparatus further includes a tong assembly, a handle assembly, a tong manipulation cable, and a pair of pivot manipulation cables. | 05-07-2009 |
| 20090060117 | Guide thimble of dual tube type structure nuclear fuel assembly - Disclosed herein is a guide thimble of a nuclear fuel assembly, which is capable of improving the cooling performance and the stability of a nuclear fuel, preventing a flow split in dual-cooling nuclear fuel rod and guide thimble sub channels for obtaining high combustion degree and high power, and minimizing a neutron absorption section in a reaction degree region. Since the guide thimble having the dual tube type structure is adopted, a flow split in the fuel rod and guide thimble sub channels can be reduced, and the degradation in performance of nuclear fuel due to increase of a neutron absorption section can be prevented. In order for compatibility with an existing control rod, a typical guide tube is used as an inner guide thimble, and an outer guide thimble is provided outside the inner guide thimble. Thus, the guide thimble has the dual tube type structure as a whole, and is coupled to the upper and lower end fittings so that it can prevent a flow unbalance due to the flow split in the fuel rod and guide thimble sub channels. | 03-05-2009 |
