| KOREA HYDRO & NUCLEAR POWER CO., LTD Patent applications |
| Patent application number | Title | Published |
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| 20120037572 | STRAINER WALL STRUCTURE INCLUDING CURVED SECTIONS, METHOD OF MANUFACTURING THE SAME, AND FILTERING METHOD USING THE SAME - A strainer wall structure includes curved sections, a method of manufacturing the same, and a filtering method using the strainer wall structure to provide a substantially larger effective filtering area in the same length and width, substantially reducing foreign substances covering a suction surface and flow resistance of the foreign substances, and reducing pressure drop at a cooling water pass corresponding thereto. The strainer wall structure includes an inlet side through which cooling water is introduced and an outlet side through which the filtered cooling water is discharged, includes a body having openings in directions of the inlet side and the outlet side, and a first filter plate inserted into the body and including curved sections formed by alternately bending a first punched plate having filtering holes in opposite directions and at a predetermined interval. | 02-16-2012 |
| 20120037559 | STRAINER FILTERING APPARATUS INCLUDING FILTERING TUBE - A strainer filtering apparatus including a filtering tube providing a substantially larger effective filtering area for its length and width, substantially reducing foreign substances covering a suction surface and flow resistance of the foreign substances, and reducing pressure drop at a cooling water passage. The strainer filtering apparatus includes at least one inlet side into which cooling water is introduced and an outlet side through which the cooling water is discharged, hollow filtering tubes with filtering holes along their lengths, an upper plate having first grooves located at a lower surface and coupled to upper ends of the filtering tubes and an inlet part into which the cooling water is introduced, and a lower plate having punched holes to which lower ends of the filtering tubes are coupled. The cooling water in the filtering tubes is introduced through the punched holes and discharged at the outlet side. | 02-16-2012 |
| 20120026495 | Device for On-Site Measurement of Concentration of Uranium in High Temperature Molten Salts - A device for on-site measurement of concentration of uranium in high temperature molten salts is provided. More particularly, to a device for on-site measurement of concentration of uranium in high temperature molten salts that can be directly applied to a pyroprocess for reusing spent nuclear fuel and determine concentration of uranium 3+ and 4+ chemical species using ultraviolet-visible light absorption spectrometry. The device includes first and second optical waveguides submerged in molten salts including uranium through a port formed at an upper side of a pyrochemical process apparatus; a lengthwise driver installed at the port to be operated to adjust a distance between the optical waveguides; a light source for supplying light to the second optical waveguide as any one of the optical waveguides; and a spectrometer connected to the first optical waveguide as the other one of the optical waveguides to analyze the light emitted from the second optical waveguide and introduced through the first optical waveguide via the molten salts. | 02-02-2012 |
| 20110299645 | Breeding Nuclear Fuel Mixture Using Metallic Thorium - Disclosed is a breeding nuclear fuel mixture including metallic thorium useable in a nuclear power plant, prepared by mixing uranium dioxide (UO | 12-08-2011 |
| 20110297627 | STRAINER WALL STRUCTURE, FILTRATION METHOD USING THE SAME, AND METHOD OF FABRICATING THE SAME - A strainer wall structure that removes foreign substances from a fluid suctioned into a pipe and a re-circulation pump that is part of an emergency core cooling system (ECCS). The strainer wall structure has an inlet side and an outlet side through which cooling water is introduced and discharged, respectively, and includes a body having an opening in a direction of the inlet side, closed side surfaces, and an outlet port disposed at one of the closed side surfaces. The strainer includes a punched plate filter screen inserted into the opening. A modular cassette apparatus including grooved first filter plates is inserted into the body, and second filter plates having second grooves is inserted into the first grooves. | 12-08-2011 |
| 20110268238 | APPARATUS AND METHOD FOR AUTOMATICALLY AND REMOTELY MEASURING THE INTERNAL GAP OF A REACTOR - A remote, precise gap-measuring apparatus for automatically measuring the gap between reactor internals, including a nuclear reactor vessel, a core support barrel, a core shroud, and a lower support structure. The remote, precise gap-measuring apparatus includes digital probes measuring a gap between a nuclear reactor vessel protrusion and a core support barrel protrusion. The nuclear reactor vessel protrusions are disposed on an inner surface of the nuclear reactor vessel, and the core support barrel protrusions are disposed on an outer surface of the core support barrel, engaged with the nuclear reactor vessel protrusions. The apparatus includes a computer coupled to the digital probes to display and store measured values of the gap. A solenoid valve, controlled by the computer, controls compressed air supplied through an air hose to operate the digital probe. The digital probes are used to measure the gap after the core support barrel, the core shroud, and the lower support structure are welded to each other, decreasing construction time and simplifying construction. | 11-03-2011 |
| 20110240817 | TELESCOPIC TUBE SET FOR BRIDGE TRANSPORT SYSTEM - A telescopic tube set for a bridge transport system is disclosed. The telescopic tube set includes cylindrical tubes each having an upper end, at which an upper ring is mounted, and a lower end, at which a lower ring is mounted, the cylindrical tubes having different diameters, pulleys mounted to respective inner wall surfaces of a part of the tubes, and extension/retraction lines each having an end connected to the upper ring of a corresponding one of the tubes, and an opposite end connected to the upper ring of another corresponding one of the tubes while extending around the pulley mounted to the tube interposed between the corresponding tubes. A plurality of bearings are installed at the lower ring to guide the tube, at which the lower ring is mounted, such that the tube extends or retracts straight in a longitudinal direction. | 10-06-2011 |
| 20110196530 | APPARATUS AND METHOD FOR AUTOMATICALLY LOADING DRUMS INTO DRUM CONTAINER - An apparatus and method for automatically loading drums into a drum container, in which, when the drums filled with radioactive waste are loaded into the drum container, the drums are fed to a designated position, and placed at a fixed position, reducing loading time and manpower as well as minimizing radiation exposure risk associated with radioactive waste treatment. The apparatus includes a drum feeder having conveyor modules transferring the drums and a turntable rotating a drum transferred to an orientation where the drum can be gripped, for sequentially loading the drums, a support frame on which a lid of the drum container is placed when the drums are being loaded, and a crane that grips and transfers the drum or the drum container lid. | 08-11-2011 |
| 20110194922 | APPARATUS AND METHOD FOR LOADING DRUMS INTO DRUM CONTAINER - An apparatus and method for loading drums into a drum container, in which a function of gripping a lid of the drum container is added to a gripper used when a drum filled with radioactive waste is loaded into the drum container, thereby considerably reducing loading time of the drums, risk of dropping the drum, and preventing a radiation exposure risk. The apparatus includes a drum feeder transferring a drum, a drum container into which the drums transferred through the drum feeder are sequentially loaded, a support frame on which a lid of the drum container is placed when the drums are being loaded, and a crane having a gripper that selectively grips and transfers the drum or the drum container lid. The gripper includes radial gripper arms, an arm hydraulic unit reciprocating the gripper arms, jaws on inner sides of the gripper arms and gripping an outer surface of the drum, and latches for latching the drum container lid. | 08-11-2011 |
| 20110185790 | Leakage Detection Method and System Using Camera Image - Provided is a leakage detection system and method using an image that may detect a leakage of high temperature high pressure steam using an image. The leakage detection method may include: obtaining an image of a target where a leakage of high temperature high pressure steam occurs; detecting, in the obtained image, an edge image before the leakage occurs and an edge image after the leakage occurs; detecting only a leakage edge image by comparing the edge image before the leakage occurs and the edge image after the leakage occurs; removing noise from the leakage edge image; and displaying the leakage edge image in which the noise is removed. | 08-04-2011 |
| 20110121473 | Method for Fabricating Sintered Annular Nuclear Fuel Pellet Through Rod-Inserted Sintering - A method for fabricating a sintered annular nuclear fuel pellet includes: molding nuclear fuel powder or granules to fabricate an annular nuclear fuel green body; inserting a rod-like shaped structure into the annular nuclear fuel green body; sintering the rod-like shaped structure-inserted annular nuclear fuel green body in a reductive gas atmosphere; and separating the sintered annular nuclear fuel pellet from the rod-like shaped structure. | 05-26-2011 |
| 20110056647 | INGOT-CASTING APPARATUS USING URANIUM DEPOSITS - A casting apparatus includes a hermetically sealed chamber, wherein the chamber includes: a melt-heating portion to melt external uranium deposits, the melt-heating portion being tiltable to allow the resulting liquid molten material to be discharged to the outside, an ingot-forming portion to receive the molten material discharged from the melt-heating portion and to cool-solidify the molten material, and a salt-capturing portion provided on the melt-heating portion, to solidify a salt gas produced when the uranium deposits are molten, and thereby recover the salt gas, wherein the ingot-forming portion includes: a horizontal turn-table rotated by an operating means, and a plurality of molds mounted on the edge of the turn-table, each containing the molten material, which is then cooled. | 03-10-2011 |
| 20110050890 | METHOD AND SYSTEM FOR MEASURING STRUCTURAL VIBRATION USING CURVE FITTING FROM VIDEO SIGNAL - A method and system for measuring structural vibration using curve fitting from a video signal, which can reduce an error in vibration measurement displacement, is provided. A method for measuring structural vibration using curve fitting from a video signal, the method includes the steps of: obtaining a video signal of the object; converting the video signal of the object into a gray video signal; adjusting the brightness of the converted video signal; separating an area to be measured from the brightness-adjusted video signal; selecting an edge area of the object from a video signal of the separated area; removing noises from the edge regions; and performing curve fitting with respect to the noise-removed video signal of the edge area. Accordingly, a displacement error is reduced, so that vibration can be more exactly measured. | 03-03-2011 |
| 20110049734 | Method for Preparing Sintered Annular Nuclear Fuel Pellet - A method for fabricating a sintered annular nuclear fuel pellet includes molding nuclear fuel powder or granule, an oxide of a fissile element (M), to fabricate an annular nuclear fuel green body. A rod-like shaped structure is inserted into the annular nuclear fuel green body and sintered in a slight oxidizing gas atmosphere such that the oxide of the fissile element has a balanced O/M ratio higher than a desired O/M ratio (oxygen/fissile element) of a final sintered annular nuclear fuel pellet, while being maintained in a cubic phase. The sintered annular nuclear fuel pellet is then reduced in a reductive gas atmosphere so as to have the desired O/M ratio in the state that the rod-like shaped structure is inserted. | 03-03-2011 |
| 20110017601 | Method for Recovery of Residual Actinide Elements from Chloride Molten Salt - A method for recovery of residual actinide element from chloride molten salts that are formed after electro-refining and/or electro-winning of a spent nuclear fuel and include actinide elements and rare-earth elements is provided. The method comprises conducting electrolysis using a liquid cadmium cathode (LCC) in the chloride molten salt that is formed after electro-refining and/or electro-winning of a spent nuclear fuel and contains rare-earth elements and actinide elements; electro-depositing the actinide elements contained in the chloride molten salt on the LCC in order to reduce a concentration of the actinide elements; and adding a CdCl | 01-27-2011 |
| 20100317911 | METHOD FOR PREPARING CERAMIC WASTE FORM CONTAINING RADIOACTIVE RARE-EARTH AND TRANSURANIC OXIDE, AND CERAMIC WASTE FORM WITH ENHANCED DENSITY, HEAT-STABILITY, AND LEACH RESISTANCE PREPARED BY THE SAME - Disclosed herein is a method for preparing a ceramic waste form containing radioactive rare-earth and transuranic oxide, and the ceramic waste form with enhanced density, heat-stability, and leach resistance prepared by the same. | 12-16-2010 |
| 20100281954 | TEST APPARATUS AND METHOD FOR SAFETY VALVE - An apparatus and method for testing the performance of a safety valve. The apparatus includes an accumulator including at least one electric heater for heating a predetermined amount of water in the accumulator, and storing steam produced by the electric heater, a test vessel storing the steam supplied from the accumulator, and providing steam at a test pressure to a safety valve, the steam being supplied after controlling flow rate and pressure, a condensing tank storing demineralized water to be supplied to the accumulator, and condensing and collecting the steam discharged from the safety valve, and a water-supply pump supplying the water stored in the condensing tank to the accumulator. | 11-11-2010 |
| 20100278294 | EMERGENCY CORE COOLING DUCT FOR EMERGENCY CORE COOLING WATER INJECTION OF A NUCLEAR REACTOR - The present invention relates to a longitudinally divided emergency core cooling (ECC) duct in order to efficiently inject safety water to core of a pressurized light-water nuclear reactor. The ECC duct includes side supports for preventing the flow-induced vibration in the annular downcomer, and has structural stability while thermally expanding and contracting. A longitudinally divided ECC duct for emergency core cooling water injection of a nuclear reactor is provided on the periphery of a core barrel of a nuclear reactor, includes an emergency core cooling water inlet facing a direct vessel injection nozzle, and extends in a longitudinal direction of the core barrel. The longitudinally divided ECC duct is divided into a plurality of longitudinally-divided ducts in the longitudinal direction of the longitudinally divided ECC duct. | 11-04-2010 |
| 20100272616 | SYSTEM AND APPARATUS OF SEPARATING REMAINING POWER OF HULL - An apparatus and system for separating remaining powder of hulls includes a first remaining powder separating unit to be supplied with hulls of a spent nuclear fuel subjected to a high-temperature oxidation, and to include a first brush for separating remaining powder of the hulls; a hull alignment unit to be supplied with the hulls from the first remaining powder separating unit, and to align the hulls; a second remaining powder separating unit to be supplied with the aligned hulls from the hull alignment unit, and to include a second brush for separating remaining powder adhered on an inner peripheral surface of the hulls; and a third remaining powder separating unit to be supplied with the hulls from the second remaining powder separating unit, and to separate the remaining powder remaining on the inner/outer peripheral surface of the hulls by using air. | 10-28-2010 |
| 20100254847 | METHOD FOR MANUFACTURING SINTERED ANNULAR NUCLEAR FUEL PELLET WITHOUT SURFACE GRINDING - 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. | 10-07-2010 |
| 20100163746 | Irradiation device for material test using gamma ray from spent nuclear fuel assembly - The present invention relates to an irradiation device for material test using a gamma ray radiated from a spent nuclear fuel assembly and provides the irradiation device for material test using a gamma ray radiated from a spent nuclear fuel assembly wherein an irradiation device for material test to achieve a radiation effect evaluation is manufactured to be movable upward, downward and horizontally in order to study the hardening phenomenon of the frail materials to the radiation among the atomic power facilities using a gamma ray radiated from a spent nuclear fuel assembly, thereby it is possible to adjust a position of the spent nuclear fuel used for material test using a gamma ray radiated from a spent nuclear fuel and a test material, identify a distance between the spent nuclear fuel and the test material easily with a scale and evaluate the radiation effects on the materials used at facilities handling a spent nuclear fuel under the same situation as they are really exposed. Further the present invention comprises a support constructed vertically; a vertical moving table which is capable of moving upward and downward connected with a proper position of the support; a moving device for moving said vertical moving table upward and downward; a horizontal moving table which is capable of moving horizontally placed on the vertical moving table; a horizontal moving bar which is capable of moving horizontally placed on the vertical moving table; and a driving device for driving the horizontal moving table and the horizontal moving bar horizontally. | 07-01-2010 |
| 20100140522 | Separable Ball Valve Apparatus and Ball Valve Assembly - A separable ball valve apparatus and ball valve assembly capable of selectively opening/closing a separable flow path are provided. The separable ball valve apparatus comprises a separable pipe member having a flow path through which fluid flows, a ball member having at least two balls disposed in the pipe member, opening and closing the flow path, and adjacent to each other, and an opening/closing member configured to control the ball member to open/close the flow path, wherein the balls have through-holes, respectively, and the opening/closing member includes a gear train configured to control the opening/closing of the at least two balls. Therefore, it is possible to simply open/close pipes having the flow path and detachably attach the pipes, increase space utilization due to removal of a separate coupling space for attachment/detachment of the pipes, and simultaneously open/close the two pipes, which are separated from each other through one operation. | 06-10-2010 |
| 20100132291 | STEEL PLATE STRUCTURE AND STEEL PLATE CONCRETE WALL - A steel plate structure and a steel plate concrete wall are disclosed. A steel plate structure, which includes: a pair of steel plates, which are separated to provide a predetermined space; a structural member, which is positioned in the predetermined space, and which is structurally rigidly joined to one side of the steel plate in the direction of gravity; and a strut, which maintains a separation distance between the pair of steel plates, can be utilized to reduce the overall thickness of a steel plate concrete wall for efficient use of space, and to reduce the thickness of the steel plates for better welding properties and larger unit module sizes. Also, the axial forces or lateral forces applied on the steel plate concrete wall may be effectively resisted. | 06-03-2010 |
| 20100126713 | Fluid Collecting Apparatus - A fluid collecting apparatus inserted in a borehole for collecting fluid in the borehole. The fluid collecting apparatus may include a first packer and a second packer which are selectively adhered to an inside of the borehole, a first supply pipe and a second supply pipe which supply expansion fluid to each of the first packer and the second packer, and a guide tube which guides fluid in a collecting space between the first packer and the second packer to an outside of the borehole. Also, the second supply pipe moves with the second packer in a longitudinal direction of the borehole, and a passage which does not communicate with the first supply pipe is formed. Accordingly, the fluid collecting apparatus may independently control the first and second packers, and thus fluid in a desired depth and region within the borehole may be easily collected. | 05-27-2010 |
| 20100118294 | Apparatus for Detecting the Leakage of Heavy Water in Nuclear Reactor System and Detection Method Using the Same - An apparatus and a method for detecting a leakage of heavy water in a nuclear reactor system includes: a diode laser that generates a laser beam and injects the generated laser beam to a test sample placed in a light absorption cell; the light absorption cell accommodates a gathered test sample therein; a vacuum pump adjusts a degree of vacuum in the light absorption cell; a test sample introduction unit gathers an air test sample from a location with a high possibility of a leakage of heavy water or a light water test sample from a secondary side of a steam generator and transfers the sample to the light absorption cell; an optical detector detects the laser beam which has passed through the light absorption cell; and a microprocessor controls the operation of the diode laser, the light absorption cell, the vacuum pump, the test sample introduction unit, and the optical detector, receives a detect signal from the optical detector, and analyzes an absorption spectrum signal. | 05-13-2010 |
| 20100091933 | METHOD OF PRODUCING LARGE-GRAINED NUCLEAR FUEL PELLET BY CONTROLLING CHROME CATION SOLUBILITY IN UO2 LATTICE - 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. | 04-15-2010 |
| 20100084265 | CONTINUOUS ELECTROREFINING DEVICE FOR RECOVERING METAL URANIUM - Disclosed is a continuous electrorefining device for recovering metal uranium. The electrorefining device comprises an electrolytic cell | 04-08-2010 |
| 20100082318 | SINGLE QUANTIFICATION METHOD OF EXTERNAL EVENT PSA MODEL CONTAINING MULTI-COMPARTMENT PROPAGATION SCENARIOS - Provided is a single quantification method of an external event PSA model containing multi-compartment scenarios, including: loading an internal event PSA logic model having core damage as a top event; constituting a mapping table comprising external events containing the multi-compartment scenarios in consideration of information regarding external event occurrence frequencies, external event-induced initiators, and equipments damaged by external events; reflecting the mapping table in the internal event PSA logic model to establish an external event PSA model; calculating a final minimum cut set (MCS) based on the external event PSA model; and calculating a core damage frequency (CDF) value according to the final MCS. | 04-01-2010 |
| 20100051607 | High-Frequency Inductive Heating Apparatus and Pressure-Less Sintering Method Using the Same - 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. | 03-04-2010 |
| 20090294113 | Heat exchanger - Disclosed is a heat exchanger. The heat exchanger includes a plurality of plates superimposed on one another, and a plurality of heat transfer fins formed on the plurality of plates, and shaped into an airfoil, wherein a channel of a fluid between the superimposed plates is formed to perform a heat exchange. | 12-03-2009 |
| 20090252277 | Upper Plenum Structure of Cooled Pressure Vessel for Prismatic Very High Temperature Reactor - An upper plenum structure of a cooled pressure vessel for a prismatic very high temperature reactor which secures a space for coolant to supply to a core and also supports an upper reflector located inside a graphite structure on top of the core. The upper plenum structure includes a cavity structure where the coolant goes down in the upper plenum structure, a plurality of upper reflector supports formed with the cavity and supporting the upper reflector located on top thereof, and a plurality of coolant distributing blocks. Each of the coolant distributing blocks is coupled with a bottom portion of a respective one of the upper reflector supports and is located on top of the core in order to distribute the coolant collected in a cavity, formed by the upper reflector support, to the core. The coolant distributing blocks cooperate with the upper reflector supports to define the cavity structure. | 10-08-2009 |
| 20090116607 | Passive cooling and arresting device for molten core material - An apparatus for passively cooling and retaining molten core material discharged from a damaged reactor vessel during a severe accident in the nuclear plant including: a molten core material retention tank to retain molten core material; a compressed gas tank storing high-pressure inert gas; a cooling water storage tank being installed higher than the molten core material retention tank; and a mixing means. The molten core material retention tank includes an outer retention vessel having at least one coolant hole, a porous protection vessel formed at an inside of the outer retention vessel, and a gravel layer formed between the outer retention vessel and the porous protection vessel. The apparatus can be installed in a reactor cavity without changing the compartment structure of a containment building, and makes it possible to prevent a steam explosion during the cooling process for the ultrahigh-temperature molten core material and to secure the reliability of the cooling process. | 05-07-2009 |
| 20090010377 | INHIBITOR OF CORROSION AND STRESS CORROSION CRACKING CONTAINING NICKEL BORIDE (NiB) IN THE SECONDARY SIDE OF STEAM GENERATOR TUBES IN A NUCLEAR POWER PLANT AND INHIBITING METHOD USING THE SAME - A method of inhibiting corrosion and stress corrosion cracking of a steam generator tube in a nuclear power plant, includes the steps of providing a nuclear power plant having a secondary side feedwater system including a secondary side feedwater of a steam generator tube; and supplying nickel boride to said secondary side feedwater to inhibit corrosion and stress corrosion cracking. | 01-08-2009 |
| 20090001294 | Neutron Coincidence Counter for Non-Destructive Accounting for Nuclear Material and the Handling Thereof - A neutron coincidence counter for non-destructive accounting for a nuclear material according to the present invention comprises an outer case, neutron detectors mounted in the outer case while being surrounded by a moderator, and a basket horizontally movable in the outer case so as to be exposed outside the outer case and having a cavity for receiving a sample container therein. Further, a neutron coincidence counter for non-destructive accounting for a nuclear material according to the present invention comprises an outer case, neutron detectors mounted in the outer case while being surrounded by a moderator, a basket movable in the outer case so as to be exposed outside the outer case and having a cavity for receiving a sample container therein, and an external signal analyzer connected to the detectors through an electrically conductive path. Moreover, at least one facile connector of one-touch connection type is mounted on the electrically conductive path for connecting the detectors to the external signal analyzer, resulting in free removal and replacement of wires connected to the connector. | 01-01-2009 |
| 20080292041 | Method and system for early sensing of water leakage, through chemical concentration monitoring, in nuclear reactor system using liquid metal and molten salt - A method of early sensing of water leakage, through chemical concentration monitoring, in a nuclear reactor system using a liquid metal and molten salt, the method including: measuring an electrical conductivity and a mass spectrum of a first channel and a second channel of a heat-related device included in the nuclear reactor system using the liquid metal and the molten salt; calculating a first identification value associated with the water leakage in the heat-related device using the measured electrical conductivity; calculating a second identification value associated with the water leakage in the heat-related device using the measured mass spectrum; and sensing the water leakage by comparing a predetermined reference value and a summed identification value, the summed identification value being the sum of the first identification value and the second identification value. | 11-27-2008 |
| 20080234950 | WATER LEAKAGE-ACOUSTIC SENSING METHOD AND APPARATUS IN STEAM GENERATOR OF SODIUM-COOLED FAST REACTOR USING STANDARD DEVIATION BY OCTAVE BAND ANALYSIS - A water leakage-acoustic sensing method in a steam generator of a sodium-cooled fast reactor, the method including: calculating a standard deviation and an average of an octave band by octave band analysis of an input signal sound received from at least one predetermined acoustic sensor; comparing the calculated standard deviation and the calculated average of the octave band, and determining a size of the octave band based on a comparison result; calculating an average of standard deviations of the octave band recomposed by the determined size and normalizing the average of standard deviations; applying a predetermined weight, established by a predetermined neural network learning algorithm, to the normalized average of standard deviations; and generating leakage determination data based on the average of standard deviations to which the weight is applied. | 09-25-2008 |
| 20080219397 | Decay Heat Removal System for Liquid Metal Reactor - A decay heat removal system for a liquid metal reactor in which a decay heat exchanger (DHX) is installed concentrically with an intermediate heat exchanger (IHX) in the same cylinder which separates the DHX and IHX from the reactor pool fluid, and serves to remove the reactor core decay heat. The cylinder surrounds the IHX and DHX, and has an open top portion protruding out of the level of the fluid in a hot pool, a bottom portion connected to a cold pool and a guide pipe for allowing the passage of the fluid from the hot pool into the IHX. The decay heat removal system can remove decay heat immediately after occurrence of an accident, thereby improving the safety of a nuclear plant. | 09-11-2008 |
