Abstract: 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 (1) of a liquid-metal cooled fast reactor includes nuclear fuel assemblies (1a, 1b and 1c) in inner, middle and outer reactor core regions, respectively, and is installed in a hexagonal duct (3) with nuclear fuel materials (2-2a, 2-2b and 2-2c) surrounded by respective claddings (2-1a, 2-1b and 2-1c), and the claddings (2-1a, 2-1b and 2-1c) of a nuclear fuel rod (2a) in the inner reactor core region, a nuclear fuel rod (2b) in the middle reactor core region and a nuclear fuel rod (2c) in the outer reactor core region are formed at different thicknesses. The reactor core can flatten power distribution using a single-enrichment nuclear fuel in the liquid-metal cooled fast reactor.

Abstract: Provided is a method of manufacturing a high strength ferritic/martensitic steel. The method includes melting a ferritic/martensitic steel, hot-working the melted ferritic/martensitic steel, normalizing the hot-worked ferritic/martensitic steel at a temperature of about 1050° C. to about 1200° C., tempering the ferritic/martensitic steel at a temperature of about 600° C. or less, and leaving MX precipitates while preventing a M23C6 precipitate from being precipitated, and cold-working and thermal-treating the ferritic/martensitic steel in a multistage fashion, and precipitating M23C6 precipitates. Through the above described configuration, the high strength ferritic/martensitic steel that prevents a ductility from being deteriorated even in a high-temperature environment may be manufactured.

Abstract: 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.

Abstract: 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 (1) of a liquid-metal cooled fast reactor includes nuclear fuel assemblies (1a, 1b and 1c) in inner, middle and outer reactor core regions, respectively, and is installed in a hexagonal duct (3) with nuclear fuel materials (2-2a, 2-2b and 2-2c) surrounded by respective claddings (2-1a, 2-1b and 2-1c), and the claddings (2-1a, 2-1b and 2-1c) of a nuclear fuel rod (2a) in the inner reactor core region, a nuclear fuel rod (2b) in the middle reactor core region and a nuclear fuel rod (2c) in the outer reactor core region are formed at different thicknesses. The reactor core can flatten power distribution using a single-enrichment nuclear fuel in the liquid-metal cooled fast reactor.

Abstract: Provided is a method of manufacturing a high strength ferritic/martensitic steel. The method includes melting a ferritic/martensitic steel, hot-working the melted ferritic/martensitic steel, normalizing the hot-worked ferritic/martensitic steel at a temperature of about 1050° C. to about 1200° C., tempering the ferritic/martensitic steel at a temperature of about 600° C. or less, and leaving MX precipitates while preventing a M23C6 precipitate from being precipitated, and cold-working and thermal-treating the ferritic/martensitic steel in a multistage fashion, and precipitating M23C6 precipitates. Through the above described configuration, the high strength ferritic/martensitic steel that prevents a ductility from being deteriorated even in a high-temperature environment may be manufactured.

Abstract: 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.

Abstract: Disclosed herein is a steam generator for a liquid metal reactor and a heat transfer method thereof, and more particularly a steam generator for a liquid metal reactor having intermediate fluid forcibly circulated therein and structurally improved coiled heat transfer tubes installed therein, and a heat transfer method of such a steam generator, whereby an accident of the sodium-water reaction is practically prevented and heat transfer efficiency is improved. The heat transfer method of the steam generator for the liquid metal reactor including a reactor core, a steam generator, and a turbine, utilizes forced circulation for the intermediate fluid in the steam generator. The forced circulation is carried out through a flow channel defined in the steam generator, which forms a closed flow circuit for the intermediate fluid, by means of operation of a circulation pump installed in the flow channel.

Abstract: Disclosed herein is a steam generator for a liquid metal reactor and a heat transfer method thereof, and more particularly a steam generator for a liquid metal reactor having intermediate fluid forcibly circulated therein and structurally improved coiled heat transfer tubes installed therein, and a heat transfer method of such a steam generator, whereby an accident of the sodium-water reaction is practically prevented and heat transfer efficiency is improved.