Abstract: The present invention relates to a coated insulated wire having improved heat dissipation properties, insulation properties, flame retardancy, and lightweight properties; and a method of manufacturing the same.

Abstract: The present invention relates to a zirconium alloy cladding with improved oxidation resistance at a high temperature and a method of manufacturing the same. More particularly, the zirconium alloy cladding includes a zirconium alloy cladding; and a Cr—Al thin film coated on the cladding, wherein the thin film is deposited through arc ion plating and the content of Al in the thin film is 5% by weight to 20% by weight.

Abstract: The present invention relates to a coated insulated wire having improved heat dissipation properties, insulation properties, flame retardancy, and lightweight properties; and a method of manufacturing the same.

Abstract: The present invention relates to a zirconium alloy cladding with improved oxidation resistance at a high temperature and a method of manufacturing the same. More particularly, the zirconium alloy cladding includes a zirconium alloy cladding; and a Cr—Al thin film coated on the cladding, wherein the thin film is deposited through arc ion plating and the content of Al in the thin film is 5% by weight to 20% by weight.

Abstract: Provided are a ferritic/martensitic oxide dispersion strengthened steel with increased high temperature creep resistance, including 0.02 to 0.2 wt % of carbon (C), 8 to 12 wt % of chromium (Cr), 0.1 to 0.5 wt % of yttria (Y2O3), 0.2 to 2 wt % of molybdenum (Mo), 0.01 to 0.5 wt % of titanium (Ti), 0.01 to 1 wt % of manganese (Mn), 0.01 to 0.3 wt % of vanadium (V), 0 to 0.3 wt % of zirconium (Zr), 0 to 0.5 wt % of nickel (Ni), and the remaining content of iron (Fe), and a method of manufacturing the same. The ferritic/martensitic oxide dispersion strengthened steel may be useful as a material for core structural components of a nuclear power system, ultra supercritical pressure steam generator components of a thermal power plant, or engine components of an airplane due to a high tensile strength at 700° C. and excellent creep resistance.

Abstract: Provided are a metal foam stack and a manufacturing method thereof. The metal foam stack includes one or more stack units. The stack unit includes: a first metal foam sheet including an open cell, in which a plurality of internal cells is connected with one another; a first bonding member positioned on the first metal foam sheet; and a second metal foam sheet positioned on the first bonding member, and including an open cell, in which a plurality of internal cells is connected with one another. Materials of an interface between the first metal foam sheet and the first bonding member and an interface between the second metal foam sheet and the first bonding member are atomically diffused.

Abstract: An exemplary embodiment of the present invention provides an electrostatic metal porous body forming apparatus including: a transfer module transferring a porous body substrate; and a coating module coating a metal powder on the porous body substrate, wherein the transfer module includes a substrate supporter fixing the porous body substrate while the porous body substrate is transferred, and wherein the coating module includes: an electrifier including a first electrode electrifying the metal powder, a second electrode facing the first electrode, a first power supplier connected with the first electrode supplying electricity to the first electrode, and a second power supplier connected with the second electrode supplying electricity electrified with an opposite charge to a charge caused by the electrification of the first electrode to the second electrode, and generating a pulse type of voltage; and a metal powder supplier including a metal powder vessel storing the metal powder therein and supplying the met

Abstract: A steam trap system is provided which can improve efficiency of sensor assembly and air-tightness. The steam trap system includes: a housing that includes a feeding hole for feeding steam, a discharging hole for discharging condensate water, and an opening; at least one sensor that is disposed inside the housing via the opening; an insulation member that is disposed in the opening to maintain air-tightness and to which the sensor is fixed to expose a terminal; and a cover that is disposed in the insulation member to cover the terminal.

Abstract: A zirconium alloy for use in nuclear fuel assemblies is provided, which provides increased resistance against oxidation and corrosion and also improved bonding with parent material, because pure metallic material such as silicon (Si) or chromium (Cr) is evenly coated on the surface of the parent material by plasma spraying. Because the plasma spray coating used to coat the pure metallic material on the zirconium alloy does not require vacuum equipment and also is not limited due to the shape of the coated product, this is particularly useful when evenly treating the surface of the component such as 4 m-long tube or spacer grip arrangement which is very complicated in shape. Furthermore, because the coated zirconium alloy confers excellent resistance to oxidation and corrosion under emergency such as accident as well as normal service condition, both the economic and safety aspects of nuclear fuel are improved.

Abstract: An oxide dispersion strengthening (ODS) method of a metallic material using a laser is provided, which includes melting a surface of a metallic matrix placed on a movable stage by irradiating a laser onto the surface (step 1), supplying an oxide dispersion strengthening (ODS) powder at a site of the matrix surface which is melt at step 1 (step 2), and cooling the matrix in which the ODS powder is supplied at step 2 (step 3). Because oxide particles are directly supplied into previously-made sheet or tube matrix, fabrication process is simplified, fabrication cost is reduced, and end product is fabricated efficiently.

Abstract: Disclosed are a zirconium alloy for a nuclear fuel cladding having a good corrosion resistance by reducing an amount of alloying elements and a method of preparing a zirconium alloy nuclear fuel cladding using thereof. The zirconium alloy includes 0.2 to 0.5 wt % of niobium (Nb); 0.2 to 0.6 wt % of iron (Fe); 0.3 to 0.5 wt % of chromium (Cr); 0.1 to 0.15 wt % of oxygen (O); 0.008 to 0.012 wt % of silicon (Si) and a remaining amount of zirconium (Zr). The total amount of the niobium, the iron and the chromium is 1.1 to 1.2 wt %. A good oxidation resistance of the nuclear fuel cladding may be confirmed under accident conditions as well as normal operating conditions of a reactor, thereby improving economic feasibility and safety.

Abstract: A steam trap system is provided which can improve efficiency of sensor assembly and air-tightness. The steam trap system includes: a housing that includes a feeding hole for feeding steam, a discharging hole for discharging condensate water, and an opening; at least one sensor that is disposed inside the housing via the opening; an insulation member that is disposed in the opening to maintain air-tightness and to which the sensor is fixed to expose a terminal; and a cover that is disposed in the insulation member to cover the terminal.

Abstract: Disclosed are a zirconium alloy for a nuclear fuel cladding having a good corrosion resistance by reducing an amount of alloying elements and a method of preparing a zirconium alloy nuclear fuel cladding using thereof. The zirconium alloy includes 0.2 to 0.5 wt % of niobium (Nb); 0.2 to 0.6 wt % of iron (Fe); 0.3 to 0.5 wt % of chromium (Cr); 0.1 to 0.15 wt % of oxygen (O); 0.008 to 0.012 wt % of silicon (Si) and a remaining amount of zirconium (Zr). The total amount of the niobium, the iron and the chromium is 1.1 to 1.2 wt %. A good oxidation resistance of the nuclear fuel cladding may be confirmed under accident conditions as well as normal operating conditions of a reactor, thereby improving economic feasibility and safety.

Abstract: Provided are a ferritic/martensitic oxide dispersion strengthened steel with increased high temperature creep resistance, including 0.02 to 0.2 wt % of carbon (C), 8 to 12 wt % of chromium (Cr), 0.1 to 0.5 wt % of yttria (Y2O3), 0.2 to 2 wt % of molybdenum (Mo), 0.01 to 0.5 wt % of titanium (Ti), 0.01 to 1 wt % of manganese (Mn), 0.01 to 0.3 wt % of vanadium (V), 0 to 0.3 wt % of zirconium (Zr), 0 to 0.5 wt % of nickel (Ni), and the remaining content of iron (Fe), and a method of manufacturing the same. The ferritic/martensitic oxide dispersion strengthened steel may be useful as a material for core structural components of a nuclear power system, ultra supercritical pressure steam generator components of a thermal power plant, or engine components of an airplane due to a high tensile strength at 700° C. and excellent creep resistance.

Abstract: Disclosed are a zirconium alloy for a nuclear fuel cladding having a good oxidation resistance in a severe reactor operation condition and a method of preparing zirconium alloy nuclear fuel claddings by using thereof. The zirconium alloy includes 1.8 to 2.0 wt % of niobium (Nb); at least one element selected from iron (Fe), chromium (Cr) and copper (Cu); 0.1 to 0.15 wt % of oxygen (O); 0.008 to 0.012 wt % of silicon (Si) and a remaining amount of zirconium (Zr). The amount of Fe is 0.1 to 0.4 wt %, the amount of Cr is 0.05 to 0.2 wt %, and the amount of Cu is 0.03 to 0.2 wt %. A good oxidation resistance of the nuclear fuel cladding may be confirmed under a severe reactor operation condition at an accident condition as well as a normal operating condition of a reactor, thereby improving economic efficiency and safety.

Abstract: Disclosed are a zirconium alloy for a nuclear fuel cladding having a good oxidation resistance in reactor accident conditions, a zirconium alloy nuclear fuel cladding prepared by using thereof and a method of preparing the same. The zirconium alloy includes 1.0 to 1.2 wt % of niobium (Nb); at least one element selected from tin (Sn), iron (Fe) and chromium (Cr); 0.02 to 0.1 wt % of copper (Cu); 0.1 to 0.15 wt % of oxygen (O); 0.008 to 0.012 wt % of silicon (Si) and a remaining amount of zirconium (Zr). The amount of Sn is 0.1 to 0.3 wt %, the amount of Fe is 0.3 to 0.8 wt %, and the amount of Cr is 0.1 to 0.3 wt %. A good oxidation resistance of the nuclear fuel cladding may be confirmed under accident conditions as well as normal operating conditions of a reactor, thereby improving economic efficiency and safety.

Abstract: The present application discloses a martensitic oxide dispersion-strengthened alloy having enhanced high-temperature strength and creep properties. The alloy includes chromium (Cr) of 8 to 12% by weight, yttria (Y2O3) of 0.1 to 0.5% by weight, carbon (C) of 0.02 to 0.2% by weight, molybdenum (Mo) of 0.2 to 2% by weight, titanium (Ti) of 0.01 to 0.3% by weight, zirconium (Zr) of 0.01 to 0.2% by weight, nickel (Ni) of 0.05 to 0.2% by weight and the balance of iron (Fe). The application also discloses a method of making the alloy.

Abstract: A zirconium alloy with a coating layer formed on a surface comprising a mixed layer, the mixed layer comprises one or more very high temperature oxidation resistant material and zirconium alloy parent material selected from the group consisting of Y2O3, SiO2, ZrO2, Cr2O3, Al2O3, Cr3C2, SiC, ZrC, ZrN, Si and Cr, and in a vertical direction on a boundary between the mixed layer and the zirconium alloy parent material is formed a gradient of compositions between the very high temperature oxidation resistance material and the zirconium alloy parent material.

Abstract: An oxide dispersion strengthening (ODS) method of a metallic material using a laser is provided, which includes melting a surface of a metallic matrix placed on a movable stage by irradiating a laser onto the surface (step 1), supplying an oxide dispersion strengthening (ODS) powder at a site of the matrix surface which is melt at step 1 (step 2), and cooling the matrix in which the ODS powder is supplied at step 2 (step 3). Because oxide particles are directly supplied into previously-made sheet or tube matrix, fabrication process is simplified, fabrication cost is reduced, and end product is fabricated efficiently.

Abstract: Disclosed herein are a zirconium alloy composition, which exhibits excellent corrosion resistance by varying the kinds of metal oxides and controlling the size of precipitates of the composition, including: 1.05˜1.45 wt % of Nb; one or more selected from the group consisting of 0.1˜0.7 wt % of Fe and 0.05˜0.6 wt % of Cr; and residual Zr, and a method of preparing the same. The zirconium alloy composition exhibits excellent corrosion resistance by controlling the kinds and amounts of the elements included in the zirconium alloy composition and the heat-treatment temperature and thus varying the kinds of metal oxides formed during an oxidation process and controlling the size of precipitates of the zirconium alloy, so that it can be usefully used as a raw material for nuclear fuel cladding tubes, spacer grids, nuclear reactor internals and the like of a light-water reactor or a heavy-water reactor in a nuclear power plant.