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1. 发明申请

US20140112846A1 USE OF A KMGF3 COMPOUND FOR TRAPPING METALS IN THE FORM OF FLUORIDES AND/OR OXYFLUORIDES IN A GASEOUS OR A LIQUID PHASE 有权
标题翻译：使用KMGF3化合物在气相或液相中的氟化物和/或氧化物形式中捕获金属
公开(公告)号：US20140112846A1
公开(公告)日：2014-04-24
申请号：US14124545
申请日：2012-06-08
申请人： Bertrand Morel , Lorraine Leite , Laurent Moch , Alain Demourgues , Romain Clarenc , Alain Tressaud , Etienne Durand
发明人： Bertrand Morel , Lorraine Leite , Laurent Moch , Alain Demourgues , Romain Clarenc , Alain Tressaud , Etienne Durand
IPC分类号： C01D3/02 , C22B60/02
CPC分类号： C22B60/0278 , B01D53/02 , B01D53/685 , B01D2251/306 , B01D2251/402 , B01J20/046 , B01J20/28004 , B01J20/28057 , B01J20/3028 , B01J20/3078 , B01J20/3085 , C01D3/02 , C01F5/28 , C01G43/063 , C01P2002/72 , C01P2002/82 , C01P2006/12 , C22B60/0213 , G21C19/46 , G21F9/04 , G21F9/06 , Y02C20/30 , Y02W30/883
摘要： The invention relates to the use of a compound of formula KMgF3 to trap metals present in the form of fluorides and/or of oxyfluorides in a gaseous or liquid phase.It also relates to a compound of formula KMgF3 which has a surface specific area at least equal to 30 m2/g and at most equal to 150 m2/g and also to its methods of preparation.The invention notably finds application in the nuclear industry, in which it can advantageously be used to purify uranium hexafluoride (UF6) present in a gaseous or liquid stream, with regard to metal impurities which are also present in this stream.
摘要翻译：本发明涉及式KMgF 3化合物在气相或液相中捕集以氟化物和/或氟氧化物形式存在的金属的用途。它还涉及表面比面积至少等于30m 2 / g，最多等于150m 2 / g的式KMgF 3的化合物，还涉及其制备方法。本发明特别适用于核工业，其中可以有利地用于净化存在于气态或液态物流中的六氟化铀（UF 6），该金属杂质也存在于该物流中。

2. 发明授权

US06299748B1 Method and apparatus of treating waste from nuclear fuel handling facility 有权
标题翻译：处理核燃料处理设备废物的方法和装置
公开(公告)号：US06299748B1
公开(公告)日：2001-10-09
申请号：US09393317
申请日：1999-09-10
申请人： Naruhito Kondo , Reiko Fujita
发明人： Naruhito Kondo , Reiko Fujita
IPC分类号： C25C122
CPC分类号： G21F9/30 , C22B60/0213 , G21C19/48 , Y02W30/884
摘要： A waste treatment apparatus treats radioactive contaminated waste from a nuclear fuel material handling facility to decontaminate the radioactive contaminated waste by using an electrolytic molten salt, and reuses the electrolytic molten salt so that any effluents are not produced. Radioactive contaminated waste (10) from a nuclear fuel material handling facility is subjected to electrolysis by a molten salt electrolysis unit (20) to decontaminate the waste (10). The used salt (16) used for decontaminating the waste (10) is filtered to separate nuclear fuel materials (19) from the used salt (16). The filtered salt (18) is reused by the molten salt electrolysis unit (20). The salt adhering to the decontaminated waste (12) is recovered by an evaporating unit (59), and the recovered salt (15) is reused by the molten salt electrolysis unit (20).
摘要翻译：废物处理装置用核燃料处理设备处理放射性污染废物，通过使用电解熔融盐对放射性污染废物进行净化，并重新使用电解熔融盐，使得不产生任何污水。来自核燃料处理设备的放射性污染废物（10）通过熔融盐电解装置（20）进行电解以净化废物（10）。用于净化废物（10）的用过的盐（16）被过滤以将核燃料材料（19）与使用的盐（16）分离。过滤的盐（18）由熔盐电解装置（20）重复使用。附着在去污废物（12）上的盐通过蒸发单元（59）回收，回收盐（15）由熔融盐电解单元（20）再利用。

3. 发明授权

US06241800B1 Acid fluxes for metal reclamation from contaminated solids 有权
标题翻译：用于污染固体的金属回收的酸通量
公开(公告)号：US06241800B1
公开(公告)日：2001-06-05
申请号：US09389673
申请日：1999-09-02
申请人： Edward J. Lahoda , Khoa T. Ha
发明人： Edward J. Lahoda , Khoa T. Ha
IPC分类号： C22B304
CPC分类号： C22B60/0221 , C22B7/002 , C22B7/04 , C22B60/0213 , C22B60/0226 , G21F9/30 , Y02P10/212
摘要： A method of recovering contaminating or valuable components from a solid feed material (10), includes feeding the material (10) into an optional grinder as a pretreatment (12), then into a heated melter (14) along with a material (16) that provides fluorine, to provide a molten or semi-molten material, where the molten material is then reacted with water or an acid solution (22) in vessel (20), to form a dissolved molten or semi-molten salt in solution, which can be passed to extractor (26) or the like and provide a concentrated stream of the valuable or contaminating components (30).
摘要翻译：从固体进料（10）回收污染或有价值的组分的方法包括将材料（10）作为预处理（12）进料到可选的研磨机中，然后与材料（16）一起进入加热的熔化器（14）提供氟以提供熔融或半熔融材料，其中熔融材料然后与水或容器（20）中的酸性溶液（22）反应，以在溶液中形成溶解的熔融或半熔融盐，其中可以送到萃取器（26）等，并提供有价值或污染成分（30）的浓缩流。

4. 发明授权

US5774815A Dry halide method for separating the components of spent nuclear fuels 失效
标题翻译：用于分离乏核燃料组分的干卤法
公开(公告)号：US5774815A
公开(公告)日：1998-06-30
申请号：US696187
申请日：1996-08-13
申请人： Jerry Dale Christian , Thomas Russell Thomas , Glen F. Kessinger
发明人： Jerry Dale Christian , Thomas Russell Thomas , Glen F. Kessinger
IPC分类号： C22B60/02 , G21C19/48 , G21F9/30 , G21F9/00
CPC分类号： C22B60/0213 , G21C19/48 , G21F9/30 , G21F9/305 , Y02W30/884
摘要： The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.
摘要翻译：本发明是通过从非放射性和易裂变铀产物中分离裂变和超铀产物来处理多种乏燃料类型的非水单一方法。本发明有四个主要操作：在优选大于1200℃的温度下将废燃料暴露于氯气中以形成挥发性金属氯化物; 在约400℃下在熔盐洗涤器中除去裂变产物氯化物，超铀产物氯化物和任何氯化镍和氯化铬; 剩余挥发性氯化物在164℃至2℃的温度范围内部分冷凝。并通过真空蒸馏再生和回收转移的废熔融盐。剩余的裂变产物，超铀产物和镍铬和铬酸盐被转化为氟化物或氧化物用于玻璃化。该方法提供了单一，紧凑的过程的显着优点，适用于大多数不同的核燃料，最大限度地减少二次废物，将高能废物中的裂变铀分离以解决潜在的关键问题，将非放射性废物与高级废物隔离体积减少，并产生常见的废玻璃或玻璃陶瓷。

5. 发明授权

US5582706A Electroseparation of actinide and rare earth metals 失效
标题翻译：锕系元素和稀土金属的电分离
公开(公告)号：US5582706A
公开(公告)日：1996-12-10
申请号：US458527
申请日：1995-06-02
申请人： LeRoy F. Grantham , Jerold Guon , David L. Grimmett , Lowell R. McCoy , Jeffery J. Roy
发明人： LeRoy F. Grantham , Jerold Guon , David L. Grimmett , Lowell R. McCoy , Jeffery J. Roy
IPC分类号： C22B60/02 , C22B60/04 , C25C3/34 , G21C19/48
CPC分类号： C22B60/0213 , C22B60/0295 , C22B60/04 , C25C3/34 , G21C19/48 , Y02W30/884
摘要： A pyrochemical process is utilized to recover 99% of the remaining transuranium materials from plutonium-uranium extraction waste. One step of the overall pyrochemical process involves the electrochemical separation of the waste components. A solid anode and a solid cathode are used in this electrochemical separation step to electrorefine in single or multiple steps. The solid anode and solid cathode are selected from certain preferred anodic and cathodic materials. Where multiple electrorefining steps are used to obtain more complete electroseparation, this is achieved by employing in the multiple electrorefining steps both a solid anode, suitably graphite, and a molten metal anode containing a mixture of the actinide and rare earth elements. This results in greater separation than can be realized through electroseparation by use of either anode alone.
摘要翻译：利用热化学工艺从铀 - 铀萃取废物中回收99％的剩余的铀钛材料。整个焦化过程的一个步骤涉及废物组分的电化学分离。在该电化学分离步骤中使用固体阳极和固体阴极，以单步或多步进行电致精制。固体阳极和固体阴极选自某些优选的阳极和阴极材料。当使用多个电解精制步骤来获得更完整的电分离时，这通过在固体阳极，合适的石墨和含有锕系元素和稀土元素的混合物的熔融金属阳极的多个电解精炼步骤中使用来实现。这导致比通过仅使用任一阳极的电分离可以实现的更大的分离。

6. 发明授权

US5454914A Method of removal of heavy metal from molten salt in IFR fuel pyroprocessing 失效
标题翻译：在IFR燃料热处理中从熔盐中除去重金属的方法
公开(公告)号：US5454914A
公开(公告)日：1995-10-03
申请号：US172313
申请日：1993-12-23
申请人： Eddie C. Gay
发明人： Eddie C. Gay
IPC分类号： C22B59/00 , C22B60/02 , C25C3/34 , G21C19/48
CPC分类号： C22B59/00 , C22B60/0213 , C25C3/34 , G21C19/48 , Y02W30/884
摘要： An electrochemical method of separating heavy metal values from a radioactive molten salt including Li halide at temperatures of about 500.degree. C. The method comprises positioning a solid Li--Cd alloy anode in the molten salt containing the heavy metal values, positioning a Cd-containing cathode or a solid cathode positioned above a catch crucible in the molten salt to recover the heavy metal values, establishing a voltage drop between the anode and the cathode to deposit material at the cathode to reduce the concentration of heavy metals in the salt, and controlling the deposition rate at the cathode by controlling the current between the anode and cathode.
摘要翻译：一种在约500℃的温度下从包括Li卤化物的放射性熔融盐分离重金属值的电化学方法。该方法包括将固体Li-Cd合金阳极定位在含有重金属的熔盐中，定位含Cd 阴极或固体阴极，位于熔融盐中的捕获坩埚上方，以回收重金属值，在阳极和阴极之间建立电压降，以在阴极处沉积材料以减少盐中重金属的浓度，并控制通过控制阳极和阴极之间的电流在阴极处的沉积速率。

7. 发明授权

US5322545A Method of producing uranium metal 失效
标题翻译：生产铀金属的方法
公开(公告)号：US5322545A
公开(公告)日：1994-06-21
申请号：US888818
申请日：1992-05-27
申请人： Paul Gilchrist
发明人： Paul Gilchrist
IPC分类号： C22B60/02
CPC分类号： C22B60/0213 , C22B60/0286 , Y02P10/214
摘要： Uranium chloride is reacted with either magnesium, sodium or calcium in the presence of a molten salt comprising light metal chlorides including lithium chloride. The temperature is maintained below the melting point of uranium. The magnesium may be in the form of magnesium-cadmium alloy, the temperature being maintained below the temperature at which magnesium and cadmium vaporize. The components of the molten salt may be first fused together so as to form the molten salt eutectic. Subsequently after separation of the uranium, products of the reaction may be recovered and recycled.
摘要翻译：在包括氯化锂的轻金属氯化物的熔融盐的存在下，氯化铯与镁，钠或钙反应。温度保持在铀的熔点以下。镁可以是镁镉合金的形式，其温度保持在镁和镉蒸发的温度以下。熔融盐的组分可以首先熔合在一起以形成熔融盐共晶。随后在分离铀后，可以回收和回收反应产物。

8. 发明授权

US5290337A Pyrochemical processes for producing Pu, Th and U metals with recyclable byproduct salts 失效
标题翻译：用可再生副产物盐生产Pu，Th和U金属的热化学工艺
公开(公告)号：US5290337A
公开(公告)日：1994-03-01
申请号：US941959
申请日：1992-09-08
申请人： Ram A. Sharma
发明人： Ram A. Sharma
IPC分类号： C22B60/02
CPC分类号： C22B60/0278 , C22B60/0213
摘要： In the pyrochemical reduction of uranium dioxide or other actinide metal oxides by reaction with magnesium, magnesium oxide byproduct is produced. The use of a salt flux comprising magnesium chloride and a rare earth element trichloride such as neodymium chloride is disclosed. The neodymium chloride reacts with magnesium oxide to form magnesium chloride and neodymium oxychloride. The resulting magnesium chloride-neodymium oxychloride salt mixture can readily be subjected to electrolysis to regenerate magnesium and neodymium chloride for reuse in the pyrochemical reduction process. Other uses of the magnesium chloride-neodymium chloride salt flux are also proposed.
摘要翻译：在通过与镁反应的二氧化铀或其它锕系金属氧化物的焦化还原中，产生氧化镁副产物。公开了使用包含氯化镁和稀土元素三氯化物如氯化钕的盐通量。氯化钕与氧化镁反应形成氯化镁和氯氧化钕。所得到的氯化镁 - 氯氧化钕盐混合物可以容易地进行电解以再生镁和氯化钕，以在焦化还原过程中重新使用。还提出氯化镁 - 氯化钕盐通量的其他用途。

9. 发明授权

US5164050A Method of obtaining uranium from oxide using a chloride process 失效
标题翻译：使用氯化法从氧化物中获得铀的方法
公开(公告)号：US5164050A
公开(公告)日：1992-11-17
申请号：US547186
申请日：1990-07-03
申请人： Yves Bertaud , Jean Boutin , Pierre Brun , Roger Durand , Antoine Floreancig , Airy-Pierre Lamaze , Roland Tricot
发明人： Yves Bertaud , Jean Boutin , Pierre Brun , Roger Durand , Antoine Floreancig , Airy-Pierre Lamaze , Roland Tricot
IPC分类号： C01G43/00 , C22B60/02 , C25C3/34
CPC分类号： C25C3/34 , C22B60/0213 , C22B60/0286
摘要： A method of obtaining uranium metal from an oxidized uranium compound, characterized in that the oxidized compound is treated with chlorine and carbon at a first stage, to obtain a chloride which is reduced by electrolysis or metallothermy using a reducing metal at a second stage.
摘要翻译：一种从氧化铀化合物中获得铀金属的方法，其特征在于，在第一阶段用氯和碳处理氧化的化合物，以获得在第二阶段通过电解或金属热还原的氯化物。

10. 发明授权

US5147616A Magnesium transport extraction of transuranium elements from LWR fuel 失效
标题翻译：镁离子从LWR燃料中提取超铀元素
公开(公告)号：US5147616A
公开(公告)日：1992-09-15
申请号：US770387
申请日：1991-10-03
申请人： John P. Ackerman , James E. Battles , Terry R. Johnson , William E. Miller , R. Dean Pierce
发明人： John P. Ackerman , James E. Battles , Terry R. Johnson , William E. Miller , R. Dean Pierce
IPC分类号： C22B60/02 , G21C19/48
CPC分类号： C22B60/0213 , G21C19/48 , Y02P10/212 , Y02W30/884
摘要： A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.
摘要翻译：将铀的锕系元素值与含有稀土和贵金属裂变产物的废核氧化物燃料中存在的铀值分离的过程。在CaCl 2和含有不少于约84重量％铀的U-Fe合金的存在下，在约800℃至约850℃的温度范围内，氧化物燃料用Ca金属还原以产生溶解在U-Fe合金中的另外的铀金属提高铀浓度并且具有铀锕系金属和稀土裂变产物金属以及其中溶解的贵金属裂变产物。将碳酸钙和碱金属的裂变产物和碱土金属和碘溶解在其中的CaCl 2分离并用碳电极进行电解处理，以将碳电极转化为CO和CO 2，从而将CaO还原为Ca金属。 Ca金属和CaCl2被再循环以减少额外的氧化物燃料。具有锕系金属和稀土裂变产物金属的U-Fe合金和溶解在其中的贵金属裂变产物与吸收锕系元素和稀土裂变产物金属的金属Mg接触。 U-Fe合金保留贵金属裂变产物，并在Mg蒸馏和再循环时储存，分离出铀锕系和稀土裂变产物。

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