公开(公告)号：US20140185732A1

公开(公告)日：2014-07-03

申请号：US13729704

申请日：2012-12-28

申请人： Kevin Ledford ,  Yang-Pi Lin ,  Paul Cantonwine

发明人： Kevin Ledford ,  Yang-Pi Lin ,  Paul Cantonwine

IPC分类号： G21C3/07

CPC分类号： G21C3/07 ,  G21C3/324 ,  G21Y2002/101 ,  Y02E30/40

摘要： A method and apparatus for a fret resistant fuel rod for a Boiling Water Reactor (BWR) nuclear fuel bundle. An applied material entrained with fret resistant particles is melted or otherwise fused to a melted, thin layer of the fuel rod cladding. The applied material is made of a material that is chemically compatible with the fuel rod cladding, allowing the fret resistant particles to be captured in the thin layer of re-solidified cladding material to produce an effective and resilient fret resistant layer on an outer layer of the cladding.

摘要翻译： 一种用于沸水反应堆（BWR）核燃料束的耐燃燃料棒的方法和装置。 夹带有耐磨颗粒的涂布材料被熔化或以其它方式熔合到燃料棒包层的熔化的薄层。 施加的材料由与燃料棒包层化学相容的材料制成，允许耐磨颗粒被捕获在再固化的包层材料的薄层中，以在外层的外层上产生有效且有弹性的抗磨层 包层

公开(公告)号：US09646722B2

公开(公告)日：2017-05-09

申请号：US13729704

申请日：2012-12-28

申请人： Kevin Ledford ,  Yang-Pi Lin ,  Paul Cantonwine

发明人： Kevin Ledford ,  Yang-Pi Lin ,  Paul Cantonwine

IPC分类号： G21C9/00 ,  G21C3/07 ,  G21C3/324

CPC分类号： G21C3/07 ,  G21C3/324 ,  G21Y2002/101 ,  Y02E30/40

摘要： A method and apparatus for a fret resistant fuel rod for a Boiling Water Reactor (BWR) nuclear fuel bundle. An applied material entrained with fret resistant particles is melted or otherwise fused to a melted, thin layer of the fuel rod cladding. The applied material is made of a material that is chemically compatible with the fuel rod cladding, allowing the fret resistant particles to be captured in the thin layer of re-solidified cladding material to produce an effective and resilient fret resistant layer on an outer layer of the cladding.

公开(公告)号：US09637809B2

公开(公告)日：2017-05-02

申请号：US12624845

申请日：2009-11-24

申请人： Yang-Pi Lin ,  David W. White ,  Daniel R. Lutz

发明人： Yang-Pi Lin ,  David W. White ,  Daniel R. Lutz

IPC分类号： C22C16/00

CPC分类号： C22C16/00

摘要： An alloy according to example embodiments of the present invention may include zirconium, tin, iron, chromium, and nickel, with a majority of the alloy being zirconium. The composition of the alloy may be about 0.85-2.00% tin by weight, about 0.15-0.30% iron by weight, about 0.40-0.75% chromium by weight, and less than 0.01% nickel by weight. The alloy may further include 0.004-0.020% silicon by weight, 0.004-0.020% carbon by weight, and/or 0.05-0.20% oxygen by weight. Accordingly, the alloy exhibits reduced hydrogen absorption and improved corrosion resistance and may be used to form a fuel assembly component.

公开(公告)号：US09139895B2

公开(公告)日：2015-09-22

申请号：US10935157

申请日：2004-09-08

申请人： David White ,  Daniel R. Lutz ,  Yang-Pi Lin ,  John Schardt ,  Gerald Potts

发明人： David White ,  Daniel R. Lutz ,  Yang-Pi Lin ,  John Schardt ,  Gerald Potts

IPC分类号： C22F1/18 ,  C22C16/00

CPC分类号： C22F1/186 ,  C22C16/00

摘要： Disclosed herein are zirconium-based alloys and methods of fabricating nuclear reactor components, particularly fuel cladding tubes, from such alloys that exhibit improved corrosion resistance in aggressive coolant compositions. The fabrication steps include a late-stage β-treatment on the outer region of the tubes. The zirconium-based alloys will include between about 1.30 and 1.60 wt % tin; between about 0.06 and 0.15 wt % chromium; between about 0.16 and 0.24 wt % iron, and between 0.05 and 0.08 wt % nickel, with the total content of the iron, chromium and nickel comprising above about 0.31 wt % of the alloy and will be characterized by second phase precipitates having an average size typically less than about 40 nm. The final finished cladding will have a surface roughness of less than about 0.50 μm Ra and preferably less then about 0.10 μm Ra.

摘要翻译： 本文公开了锆基合金和从这种在侵蚀性冷却剂组合物中表现出改进的耐腐蚀性的合金制造核反应堆部件，特别是燃料包壳管的方法。 制造步骤包括在管的外部区域的后期和后处理。 锆基合金将包括约1.30至1.60重量％的锡; 约0.06至0.15重量％的铬; 约0.16至0.24重量％的铁，以及0.05至0.08重量％的镍，其中铁，铬和镍的总含量高于合金的约0.31重量％，其特征在于具有平均尺寸的第二相沉淀 通常小于约40nm。 最终完成的包层将具有小于约0.50μmRa的表面粗糙度，优选小于约0.10μmRa的表面粗糙度。

公开(公告)号：US20120033779A1

公开(公告)日：2012-02-09

申请号：US12850244

申请日：2010-08-04

申请人： Daniel Reese Lutz ,  Young Jin Kim ,  Yang-Pi Lin

发明人： Daniel Reese Lutz ,  Young Jin Kim ,  Yang-Pi Lin

IPC分类号： G21C17/00

CPC分类号： G21C17/06 ,  G01N17/04 ,  G21C17/00

摘要： A method of determining in-reactor susceptibility of a zirconium-based alloy to shadow corrosion according to a non-limiting embodiment of the present invention may include immersing a first electrode and a second electrode in an electrolytic solution. The first electrode may be formed of the zirconium-based alloy, while the second electrode may be formed of a metallic material suitable for use in a nuclear reactor and having a higher electrochemical corrosion potential than the zirconium-based alloy. The method may additionally include irradiating the immersed first and second electrodes with electromagnetic radiation. A galvanic current may then be measured between the first electrode and the second electrode to ascertain the relative in-reactor susceptibility of the zirconium-based alloy to shadow corrosion. The present invention allows a simplified and more rapid method of developing solutions that mitigate shadow corrosion, thereby potentially saving years of expensive in-reactor testing.

摘要翻译： 根据本发明的非限制性实施例，确定锆基合金对阴影腐蚀的反应器内易感性的方法可以包括将第一电极和第二电极浸入电解液中。 第一电极可以由锆基合金形成，而第二电极可以由适于在核反应堆中使用并且具有比锆基合金更高的电化学腐蚀电位的金属材料形成。 该方法可以另外包括用电磁辐射照射浸没的第一和第二电极。 然后可以在第一电极和第二电极之间测量电流，以确定锆基合金相对于反应器内的易受影响的腐蚀。 本发明允许简化和更快速的开发解决方案的方法，其减轻阴影腐蚀，从而潜在地节省了数年的昂贵的反应堆内测试。

公开(公告)号：US20100014624A1

公开(公告)日：2010-01-21

申请号：US12219212

申请日：2008-07-17

申请人： Daniel R. Lutz ,  Young Jin Kim ,  Yang-Pi Lin

发明人： Daniel R. Lutz ,  Young Jin Kim ,  Yang-Pi Lin

IPC分类号： G21C9/00 ,  G21F1/08

CPC分类号： G21C17/0225 ,  G21C3/356 ,  G21C21/02

摘要： Example embodiments are directed to providing a thin, adherent coating on the surfaces of nuclear reactor components, which are known to cause increased corrosion on adjacent zirconium alloy structures, and methods of reducing the increased corrosion. Example embodiments include coatings being structurally bonded to components such that the difference in the corrosion potential between a coated component and a zirconium alloy component is less than that between a component without the coating and the zirconium alloy component.

摘要翻译： 示例性实施例涉及在核反应堆组件的表面上提供薄的粘附涂层，这已知会对相邻的锆合金结构造成增加的腐蚀，以及减少增加的腐蚀的方法。 示例性实施方案包括在结构上结合到组分的涂层，使得涂覆组分和锆合金组分之间的腐蚀电位差异小于没有涂层的组分和锆合金组分之间的腐蚀电位差。

公开(公告)号：US20060048869A1

公开(公告)日：2006-03-09

申请号：US10935156

申请日：2004-09-08

申请人： David White ,  Daniel Lutz ,  Yang-Pi Lin ,  John Schardt ,  Gerald Potts ,  Robert Elkins ,  Hiroaki Kagami ,  Hideyuki Mukai

发明人： David White ,  Daniel Lutz ,  Yang-Pi Lin ,  John Schardt ,  Gerald Potts ,  Robert Elkins ,  Hiroaki Kagami ,  Hideyuki Mukai

IPC分类号： C22F1/18

CPC分类号： C22F1/186 ,  C22C16/00 ,  G21C3/07 ,  G21C21/00 ,  Y02E30/40

摘要： Disclosed herein are zirconium-based alloys that may be fabricated to form nuclear reactor components, particularly fuel cladding tubes, that exhibit sufficient corrosion resistance and hydrogen absorption characteristics, without requiring a late stage α+β or β-quenching processes. The zirconium-base alloys will include between about 1.30-1.60 wt % tin; 0.0975-0.15 wt % chromium; 0.16-0.24 wt % iron; and up to about 0.08 wt % nickel, with the total content of the iron, chromium and nickel comprising at least about 0.3175 wt % of the alloy. The resulting components will exhibt a surface region having a mean precipitate sizing of between about 50 and 100 nm and a Sigma A of less than about 2×10−19 hour with the workpiece processing generally being limited to temperatures below 680° C. for extrusion and below 625° C. for all other operations, thereby simplifying the fabrication of the nuclear reactor components while providing corrosion resistance comparable with conventional alloys.

摘要翻译： 本文公开的是锆基合金，其可以被制造成形成核反应堆组分，特别是燃料包覆管，其表现出足够的耐腐蚀性和吸氢特性，而不需要晚期α+β或β-猝灭过程。 锆基合金将包括约1.30-1.60重量％的锡; 0.0975-0.15重量％铬; 0.16-0.24重量％铁; 和至多约0.08重量％的镍，其中铁，铬和镍的总含量占合金的至少约0.3175重量％。 所得到的组分将显示平均沉淀尺寸为约50至100nm的表面区域和小于约2×10 -19小时的Sigma A，其中工件加工通常限于低于680℃的温度 ℃，对于所有其他操作，低于625℃，从而简化核反应堆组件的制造，同时提供与常规合金相当的耐腐蚀性。

公开(公告)号：US20070153963A1

公开(公告)日：2007-07-05

申请号：US11320477

申请日：2005-12-29

申请人： Daniel Lutz ,  Gerald Potts ,  Yang-Pi Lin ,  Sheikh Mahmood ,  Mark Dubecky ,  David White ,  John Schardt

发明人： Daniel Lutz ,  Gerald Potts ,  Yang-Pi Lin ,  Sheikh Mahmood ,  Mark Dubecky ,  David White ,  John Schardt

IPC分类号： G21C1/04

CPC分类号： G21C3/07 ,  G21C3/324 ,  G21Y2002/203 ,  G21Y2002/207 ,  G21Y2004/10 ,  G21Y2004/201 ,  Y02E30/40

摘要： A zirconium alloy suitable for forming reactor components that exhibit reduced irradiation growth and improved corrosion resistance during operation of a light water reactor (LWR), for example, a boiling water reactor (BWR). During operation of the reactor, the reactor components will be exposed to a strong, and frequently asymmetrical, radiation fields sufficient to induce or accelerate corrosion of the irradiated alloy surfaces within the reactor core. Reactor components fabricated from the disclosed zirconium alloy will also tend to exhibit an improved tolerance for cold-working during fabrication of the component, thereby simplifying the fabrication of such components by reducing or eliminating subsequent thermal processing, for example, anneals, without unduly degrading the performance of the finished component.

摘要翻译： 适用于形成反应器组分的锆合金，其在轻水反应器（LWR），例如沸水反应器（BWR）的操作期间表现出减少的照射生长和改善的耐腐蚀性。 在反应器的操作期间，反应器部件将暴露于足以引起或加速反应堆堆芯内辐射的合金表面的腐蚀的强烈且常常不对称的辐射场。 由公开的锆合金制造的反应器部件也将倾向于在部件制造期间显示出对冷加工的改善的公差，从而通过减少或消除随后的热处理（例如退火）来简化这些部件的制造，而不会不适当地降低 成品组件的性能。

公开(公告)号：US08043448B2

公开(公告)日：2011-10-25

申请号：US10935156

申请日：2004-09-08

申请人： David White ,  Daniel R. Lutz ,  Yang-Pi Lin ,  John Schardt ,  Gerald Potts ,  Robert Elkins ,  Hiroaki Kagami ,  Hideyuki Mukai

发明人： David White ,  Daniel R. Lutz ,  Yang-Pi Lin ,  John Schardt ,  Gerald Potts ,  Robert Elkins ,  Hiroaki Kagami ,  Hideyuki Mukai

IPC分类号： C22F1/18

CPC分类号： C22F1/186 ,  C22C16/00 ,  G21C3/07 ,  G21C21/00 ,  Y02E30/40

摘要： Disclosed herein are zirconium-based alloys that may be fabricated to form nuclear reactor components, particularly fuel cladding tubes, that exhibit sufficient corrosion resistance and hydrogen absorption characteristics, without requiring a late stage α+β or β-quenching processes. The zirconium-base alloys will include between about 1.30-1.60 wt % tin; 0.0975-0.15 wt % chromium; 0.16-0.24 wt % iron; and up to about 0.08 wt % nickel, with the total content of the iron, chromium and nickel comprising at least about 0.3175 wt % of the alloy. The resulting components will exhibit a surface region having a mean precipitate sizing of between about 50 and 100 nm and a Sigma A of less than about 2×10−19 hour with the workpiece processing generally being limited to temperatures below 680° C. for extrusion and below 625° C. for all other operations, thereby simplifying the fabrication of the nuclear reactor components while providing corrosion resistance comparable with conventional alloys.

摘要翻译： 这里公开的是锆基合金，其可以被制造成形成核反应堆组分，特别是燃料包覆管，其表现出足够的耐腐蚀性和氢吸收特性，而不需要晚期α+和bgr; 或＆bgr; - 修复过程。 锆基合金将包括约1.30-1.60重量％的锡; 0.0975-0.15重量％铬; 0.16-0.24重量％铁; 和至多约0.08重量％的镍，其中铁，铬和镍的总含量占合金的至少约0.3175重量％。 所得组分将表现出具有约50-100nm的平均沉淀尺寸的表面区域和小于约2×10-19小时的Sigma A，其中工件加工通常限于低于680℃的温度用于挤出 并且对于所有其他操作而言低于625℃，从而简化了核反应堆部件的制造，同时提供了与常规合金相当的耐腐蚀性。

公开(公告)号：US20110180184A1

公开(公告)日：2011-07-28

申请号：US11611348

申请日：2006-12-15

申请人： Daniel Reese Lutz ,  Yang-Pi Lin ,  Gerald Potts ,  William H. Jackson

发明人： Daniel Reese Lutz ,  Yang-Pi Lin ,  Gerald Potts ,  William H. Jackson

IPC分类号： C23C8/00 ,  G21C3/07 ,  B23K26/00

CPC分类号： G21C21/02 ,  C22F1/186 ,  G21C3/32

摘要： A method for treating a Zr-alloy fuel bundle material in a nuclear reactor includes treating a surface of the Zr-alloy fuel bundle material with a laser beam generated by a solid-state laser, and a nuclear reactor including a treated Zr-alloy fuel bundle material. This may reduce the generation of shadow corrosion and/or reduce the propensity for interference between control blade and fuel channel during operation of the nuclear reactor.

摘要翻译： 一种用于处理核反应堆中的Zr合金燃料束材料的方法，包括用固体激光器产生的激光束对Zr合金燃料束材料的表面进行处理，以及包括经处理的Zr合金燃料 束材料。 这可以减少在核反应堆运行期间阴影腐蚀的产生和/或降低控制叶片与燃料通道之间干扰的倾向。