The United States of America As Represented by the United States

Other

U.S. Department of Energy

The United States of America As Represented by the Department of Energy

The United States of America As Represented by the United States Energy

Terry R Johnson

John P Ackerman

William E Miller

James E Battles

R Dean Pierce

James L Willit

Mark A Williamson

Stanley G Wiedmeyer

Javier Figueroa

Zygmunt Tomczuk

Terry Alan Cruse

Michele A Lewis

Donald F Fischer

According to one aspect of the invention, a system to separate salt from uranium. The system has a vessel, a heater, a pump, and a condenser. The vessel is adapted to receive a uranium that has a salt concentration. The heater heats the uranium for a period of time, causing the salt to turn into a salt vapor and the uranium to melt. The melted uranium releases the salt vapor. The pump circulates an inert gas that carries the salt vapor away from the melted uranium. The condenser is adapted to receive the salt vapor.

System for salt removal from uranium metal

According to one aspect of the invention, a method to create a ceramic waste form from used nuclear fuel. An active metal salt waste, a rare earth metal waste, and raw materials are received. The active metal salt waste is combined with the rare earth metal waste, forming a waste salt. The waste salt is then heated to approximately 500&#xb0; C. The raw materials are also heated to approximately 500&#xb0; C. The waste salt and raw materials are then blended to form a homogenous waste mixture. The homogenous waste mixture is heated to a first predetermined temperature for a predetermined amount of time, creating a ceramic waste form. The ceramic waste form is cooled to a second predetermined temperature.

Ceramic waste form production from used nuclear fuel

An underlayment sheet (') is applied in overlapping courses to a roof deck () for establishing a weather-resistant membrane for use as a final exposed surface or below a roof covering (). The underlayment sheet (&#x2032;) has a bottom surface (&#x2032;) to which a pressure sensitive adhesive section (&#x2032;) is applied generally continuously along its length. A non-adhesive section (&#x2032;) also extends the length of the bottom surface (&#x2032;) adjacent its upper long edge (&#x2032;) for providing a contact surface with the roof deck () which is devoid of any adhesive material. The non-adhesive section (&#x2032;) is attached to the roof deck () using tin tags (&#x2032;) or other mechanical fastening components. When subsequent underlayment sheets (&#x2032;) are applied in overlapping courses, the adhesive sections (&#x2032;) bond over the preceding underlayment sheet (&#x2032;) to establish a water tight seal. The top surface () of the underlayment sheet () may include a supplemental bonding strip () which adheres directly to the adhesive section () of an overlying underlayment sheet () to establish a tenacious seal. The top surface (&#x2032;) may also be impregnated with a course granular material () to serve as a final, exposed surface for the roof in lieu of shingles or the like. If a future re-roofing operation is required, the subject underlayment system can be removed from the roof deck () without damaging or delaminating the roof deck () material.

Roof underlayment

The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.

Process to remove rare earth from IFR electrolyte

The invention is a method for the encapsulation of soluble radioactive waste chloride salts containing radionuclides such as strontium, cesium and hazardous wastes such as barium so that they may be permanently stored without future threat to the environment. The process consists of contacting the salts containing the radionuclides and hazardous wastes with certain zeolites which have been found to ion exchange with the radionuclides and to occlude the chloride salts so that the resulting product is leach resistant.

Method for immobilizing mixed waste chloride salts containing

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 Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of 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 Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including Mg Cl.sub.2 to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy.

Salt transport extraction of transuranium elements from lwr fuel

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.

Magnesium transport extraction of transuranium elements from LWR fuel

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

Uranium chloride extraction of transuranium elements from LWR fuel

A compact, portable magnetization/demagnetization device comprises a tubular housing internally of which at one end is mounted an electrical coil which surrounds a cavity accessible to screwdrivers and the like. At the opposite end of the housing there is mounted a pair of AC prongs. A manually operated switch on the housing permits the coil to be energized when the prongs are energized to establish a magnetic field in the cavity for the purpose of magnetizing and demagnetizing tools and other objects.

Magnetization/demagnetization device

A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

Process to separate transuranic elements from nuclear waste

Alkali metals such as those employed in liquid metal coolant systems can be safely reacted to form hydroxides by first dissolving the alkali metal in relatively inert metals such as lead or bismuth. The alloy thus formed is contacted with a molten salt including the alkali metal hydroxide and possibly the alkali metal carbonate in the presence of oxygen. This oxidizes the alkali metal to an oxide which is soluble within the molten salt. The salt is separated and contacted with steam or steam-CO.sub.2 mixture to convert the alkali metal oxide to the hydroxide. These reactions can be conducted with minimal hydrogen evolution and with the heat of reaction distributed between the several reaction steps.

Growing community of inventors

Wheaton, IL, United States of America

Terry R Johnson