Korea Atomic Energy Research Institute

Korea Hydro Nuclear Power Co., Ltd.

Byoung-Kwon Choi

Dong Jun Park

Jeong-Yong Park

Yang-Il Jung

Sang Yoon Park

Yang-Hyun Koo

Hyun Gil Kim

Tae Kyu Kim

Kun-Woo Song

Il Hyun Kim

Yong-Hwan Jeong

Il-Hyun Kim

Sanghoon Noh

Hyun-Gil Kim

Ki-Baik Kim

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 (YO), 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&#xb0; C. and excellent creep resistance.

Ferritic/martensitic oxide dispersion strengthened steel with enhanced creep resistance and method of manufacturing the same

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.

Zirconium alloy for improving resistance to oxidation at very high temperature and fabrication method thereof

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.

Method for oxide dispersion strengthening of metallic material using laser

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.

Zirconium alloys for a nuclear fuel cladding having a superior corrosion resistance by reducing an amount of alloying elements and methods of preparing a zirconium alloy nuclear fuel cladding using thereof

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.

Zirconium alloys for a nuclear fuel cladding having a superior oxidation resistance in a severe reactor operation condition and methods of preparing a zirconium alloy nuclear cladding by using thereof

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.

Zirconium alloys for a nuclear fuel cladding having a superior oxidation resistance in a reactor accident condition, zirconium alloy nuclear fuel claddings prepared by using thereof and methods of preparing the same

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Daejeon, South Korea

Byoung-Kwon Choi