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1.发明授权Zirconium alloy for improving resistance to oxidation at very high temperature and fabrication method thereof 有权在非常高的温度下提高耐氧化性的锆合金及其制造方法公开(公告)号：US09421740B2
公开(公告)日：2016-08-23
申请号：US13670515
申请日：2012-11-07
发明人： Kun-Woo Song , Yang-Hyun Koo , Byoung-Kwon Choi , Il-Hyun Kim , Hyun-Gil Kim , Yang-Il Jung , Dong Jun Park , Jeong-Yong Park
IPC分类号： C23C4/04 , C23C4/08 , C23C4/12 , B32B15/01 , C23C4/18 , C23C4/02 , C22C16/00 , C22F1/18 , G21C3/07 , G21C3/30
CPC分类号： B32B15/01 , C22C16/00 , C22F1/186 , C23C4/02 , C23C4/08 , C23C4/134 , C23C4/18 , G21C3/07 , G21C3/30 , Y02E30/40 , Y10T428/12674 , Y10T428/12812
摘要： 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.
摘要翻译： 提供了用于核燃料组件的锆合金，其提供增加的抗氧化和腐蚀性，并且还改善了与母体材料的接合，因为诸如硅（Si）或铬（Cr）的纯金属材料被均匀地涂覆在 母材通过等离子喷涂。 由于用于在锆合金上涂覆纯金属材料的等离子体喷涂不需要真空设备，并且由于涂覆产品的形状也不受限制，因此当将组分的表面均匀地处理如4 m形管子或间隔手柄装置，其形状非常复杂。 此外，由于涂覆的锆合金在事故和正常使用条件下在紧急情况下具有优异的抗氧化和耐腐蚀性，因此提高了核燃料的经济性和安全性。
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2.发明授权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 有权用于核燃料包层的锆合金通过减少合金元素的量而具有优异的耐腐蚀性和使用其制备锆合金核燃料包层的方法公开(公告)号：US09202597B2
公开(公告)日：2015-12-01
申请号：US13747614
申请日：2013-01-23
发明人： Jeong-Yong Park , Byoung-Kwon Choi , Hyun Gil Kim , Sang Yoon Park , Yang-Il Jung , Dong Jun Park , Yang-Hyun Koo
摘要： 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.
摘要翻译： 公开了一种通过减少合金元素量的具有良好耐腐蚀性的核燃料包层锆合金，以及使用该锆合金制备锆合金核燃料包层的方法。 锆合金含有0.2〜0.5重量％的铌（Nb）; 0.2〜0.6重量％的铁（Fe）; 0.3〜0.5重量％的铬（Cr）; 0.1〜0.15重量％的氧（O）; 硅（Si）0.008〜0.012重量％，锆（Zr）余量。 铌，铁和铬的总量为1.1〜1.2重量％。 可以在事故条件和反应堆的正常运行条件下确认核燃料包层的良好的抗氧化性，从而提高经济可行性和安全性。
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3.发明授权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 有权用于核燃料包层的锆合金在严格的反应器操作条件下具有优异的抗氧化性，以及通过使用它制备锆合金核包层的方法公开(公告)号：US09111650B2
公开(公告)日：2015-08-18
申请号：US13747763
申请日：2013-01-23
发明人： Jeong-Yong Park , Yang-Il Jung , Hyun Gil Kim , Byoung-Kwon Choi , Sang Yoon Park , Dong Jun Park , Yang-Hyun Koo
摘要： 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.
摘要翻译： 公开了一种在严格的反应器运行条件下具有良好的抗氧化性的核燃料包覆用锆合金，以及通过使用它们制备锆合金核燃料包层的方法。 锆合金含有1.8〜2.0重量％的铌（Nb）; 选自铁（Fe），铬（Cr）和铜（Cu）中的至少一种元素; 0.1〜0.15重量％的氧（O）; 硅（Si）0.008〜0.012重量％，锆（Zr）余量。 Fe的量为0.1〜0.4重量％，Cr的含量为0.05〜0.2重量％，Cu的含量为0.03〜0.2重量％。 在反应堆的事故状态以及反应堆的正常运行状态下，在严格的反应堆操作条件下，可以确认核燃料包层的良好的抗氧化性，从而提高经济效率和安全性。
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4.发明申请ZIRCONIUM ALLOY FOR IMPROVING RESISTANCE TO OXIDATION AT VERY HIGH TEMPERATURE AND FABRICATION METHOD THEREOF 有权用于在非常高的温度下提高氧化电阻的氧化锌合金及其制造方法公开(公告)号：US20130302639A1
公开(公告)日：2013-11-14
申请号：US13670515
申请日：2012-11-07
发明人： Kun-Woo Song , Yang-Hyun Koo , Byoung-Kwon Choi , Il-Hyun Kim , Hyun-Gil Kim , Yang-Il Jung , Dong Jun Park , Jeong-Yong Park
CPC分类号： B32B15/01 , C22C16/00 , C22F1/186 , C23C4/02 , C23C4/08 , C23C4/134 , C23C4/18 , G21C3/07 , G21C3/30 , Y02E30/40 , Y10T428/12674 , Y10T428/12812
摘要： 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.
摘要翻译： 提供了用于核燃料组件的锆合金，其提供增加的抗氧化和腐蚀性，并且还改善了与母体材料的接合，因为诸如硅（Si）或铬（Cr）的纯金属材料被均匀地涂覆在 母材通过等离子喷涂。 由于用于在锆合金上涂覆纯金属材料的等离子体喷涂不需要真空设备，并且由于涂覆产品的形状也不受限制，因此当将组分的表面均匀地处理如4 m形管子或间隔手柄装置，其形状非常复杂。 此外，由于涂覆的锆合金在事故和正常使用条件下在紧急情况下具有优异的抗氧化和耐腐蚀性，因此提高了核燃料的经济性和安全性。
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公开(公告)号：US09136023B2
公开(公告)日：2015-09-15
申请号：US13747614
申请日：2013-01-23
发明人： Jeong-Yong Park , Byoung-Kwon Choi , Hyun Gil Kim , Sang Yoon Park , Yang-Il Jung , Dong Jun Park , Yang-Hyun Koo
摘要： 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.
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6.发明授权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 有权在反应堆事故状态下具有优异的抗氧化性的核燃料包层的锆合金，使用它们制备的锆合金核燃料包层及其制备方法公开(公告)号：US09099205B2
公开(公告)日：2015-08-04
申请号：US13747042
申请日：2013-01-22
发明人： Jeong-Yong Park , Hyun Gil Kim , Byoung-Kwon Choi , Sang Yoon Park , Yang-Il Jung , Dong Jun Park , Yang-Hyun Koo
摘要： 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.
摘要翻译： 公开了一种用于核燃料包层的锆合金，其在反应堆事故条件下具有良好的抗氧化性，通过使用它制备的锆合金核燃料包覆层及其制备方法。 锆合金含有1.0〜1.2重量％的铌（Nb）; 选自锡（Sn），铁（Fe）和铬（Cr）中的至少一种元素; 0.02〜0.1重量％的铜（Cu）; 0.1〜0.15重量％的氧（O）; 硅（Si）0.008〜0.012重量％，锆（Zr）余量。 Sn的量为0.1〜0.3重量％，Fe的量为0.3〜0.8重量％，Cr的含量为0.1〜0.3重量％。 可以在事故条件以及反应堆的正常运行条件下确认核燃料包层的良好的抗氧化性，从而提高经济效率和安全性。
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7.发明申请MARTENSITIC OXIDE DISPERSION STRENGTHENED ALLOY WITH ENHANCED HIGH-TEMPERATURE STRENGTH AND CREEP PROPERTY, AND METHOD OF MANUFACTURING THE SAME 审中-公开具有增强的高温强度和侵蚀性的氧化镁分散强化合金及其制造方法公开(公告)号：US20140294653A1
公开(公告)日：2014-10-02
申请号：US14191963
申请日：2014-02-27
发明人： Tae Kyu Kim , Sanghoon Noh , Byoung-Kwon Choi , Chang-Hee Han , Ki-Baik Kim , Suk Hoon Kang , Young-Bum Chun , Jinsung Jang , Yong-Hwan Jeong
CPC分类号： C22C49/08 , B22F3/12 , B22F3/15 , B22F3/17 , B22F3/18 , B22F3/20 , B22F2302/25 , B22F2302/40 , B22F2998/10 , C21D6/002 , C21D7/13 , C21D2211/004 , C22C33/0285 , C22C38/002 , C22C38/44 , C22C38/50
摘要： 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.
摘要翻译： 本申请公开了具有提高的高温强度和蠕变性能的马氏体分散强化合金。 该合金包括8〜12重量％的铬（Cr），0.1〜0.5重量％的氧化钇（Y 2 O 3），0.02〜0.2重量％的碳（C），0.2〜2重量％的钼（Mo） ，钛（Ti）为0.01〜0.3重量％，锆（Zr）为0.01〜0.2重量％，镍（Ni）为0.05〜0.2重量％，余量为铁（Fe）。 该申请还公开了一种制造该合金的方法。
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公开(公告)号：US10011893B2
公开(公告)日：2018-07-03
申请号：US14485409
申请日：2014-09-12
发明人： Tae Kyu Kim , Sanghoon Noh , Byoung-Kwon Choi , Ki-Baik Kim , Yong-Hwan Jeong
IPC分类号： C22C38/18 , C22C38/50 , B22F3/12 , B22F3/15 , B22F3/17 , B22F5/00 , B22F5/10 , B22F9/04 , C22C33/02 , C22C38/04 , C22C38/22 , C22C38/28 , C22C38/44 , C22C38/46 , B22F5/12 , B22F3/18 , B22F3/20
CPC分类号： C22C38/50 , B22F3/12 , B22F3/1208 , B22F3/15 , B22F3/17 , B22F3/18 , B22F3/20 , B22F5/00 , B22F5/009 , B22F5/10 , B22F5/12 , B22F9/04 , B22F2003/185 , B22F2003/208 , B22F2009/041 , B22F2998/10 , C22C33/0285 , C22C38/04 , C22C38/22 , C22C38/28 , C22C38/44 , C22C38/46 , C22C1/1084
摘要： 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.
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9.发明申请公开(公告)号：US20150252458A1
公开(公告)日：2015-09-10
申请号：US14485409
申请日：2014-09-12
发明人： Tae Kyu Kim , Sanghoon Noh , Byoung-Kwon Choi , Ki-Baik Kim , Yong-Hwan Jeong
IPC分类号： C22C38/50 , B22F3/15 , B22F3/17 , B22F3/18 , B22F3/20 , B22F5/00 , B22F5/10 , B22F9/04 , C22C38/46 , C22C38/44 , C22C38/28 , C22C38/22 , C22C38/04 , C22C33/02 , B22F3/12
CPC分类号： C22C38/50 , B22F3/12 , B22F3/1208 , B22F3/15 , B22F3/17 , B22F3/18 , B22F3/20 , B22F5/00 , B22F5/009 , B22F5/10 , B22F5/12 , B22F9/04 , B22F2003/185 , B22F2003/208 , B22F2009/041 , B22F2998/10 , C22C33/0285 , C22C38/04 , C22C38/22 , C22C38/28 , C22C38/44 , C22C38/46 , C22C1/1084
摘要： 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.
摘要翻译： 本发明提供耐高温蠕变性提高的铁素体/马氏体分散强化钢，其中，碳（C）为0.02〜0.2重量％，铬（Cr）为8〜12重量％，氧化钇（Y 2 O 3）为0.1〜0.5重量％ ，钼（Mo）0.2〜2重量％，钛（Ti）0.01〜0.5重量％，锰（Mn）0.01〜1重量％，钒（V）：0.01〜0.3重量％，0：0.3重量％ 的锆（Zr），0〜0.5重量％的镍（Ni）和余量的铁（Fe）及其制造方法。 铁素体/马氏体分散强化钢可用作核动力系统的核心结构部件，火力发电厂的超超临界压力蒸汽发生器部件或飞机的发动机部件，由于700的高抗拉强度 °C和优异的抗蠕变性。
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