U.S. Department of Energy

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

Steven Herrmann

Shelly Xiaowei Li

Robert Dominick Mariani

Ronald Scott Herbst

Brian Robert Westphal

Jan-fong Jue

Brenda Serrano-Rodriguez

Guoping Cao

Kevin Norbash

According to one aspect of the invention, a potentiometric sensor having a cathode and an anode. The cathode is configured to provide a summary voltage representative of at least two voltage points. The anode is configured to provide a first voltage. The cathode is in communication with the anode by a first electrolyte forming an open circuit having an open circuit potential. Within the first electrolyte is a concentration of a target ion. The open circuit potential mathematically corresponds to the concentration of the target ion.

Potentiometric sensor

According to one aspect of the invention, a method for separating and recovering uranium from a nuclear fuel element. The method includes immersing a nuclear fuel element containing nuclear fuel and cladding in a molten metal. The nuclear fuel includes uranium. The cladding is selectively dissolved from the nuclear fuel element when immersed in the molten metal. The nuclear fuel is separated from the cladding. The method then includes loading the nuclear fuel into a permeable basket that is electrically configured as an anode of an electrolytic cell. There are also a molten salt electrolyte and a cathode in the electrolytic cell. Then, the method includes applying an electric charge across the electrolytic cell. The molten salt electrolyte selectively transfers uranium from the anode to the cathode.

Method of separating and recovering uranium from aluminum-clad metallic nuclear fuel

A method and device for the real-time, in-situ monitoring of Plutonium content in U&#x2014;Pu Alloys comprising providing a crucible. The crucible has an interior non-reactive to a metallic U&#x2014;Pu alloy within said interior of said crucible. The U&#x2014;Pu alloy comprises metallic uranium and plutonium. The U&#x2014;Pu alloy is heated to a liquid in an inert or reducing atmosphere. The heated U&#x2014;Pu alloy is then cooled to a solid in an inert or reducing atmosphere. As the U&#x2014;Pu alloy is cooled, the temperature of the U&#x2014;Pu alloy is monitored. A solidification temperature signature is determined from the monitored temperature of the U&#x2014;Pu alloy during the step of cooling. The amount of Uranium and the amount of Plutonium in the U&#x2014;Pu alloy is then determined from the determined solidification temperature signature.

Real-time monitoring of plutonium content in uranium-plutonium alloys

An apparatus for the real-time, in-situ monitoring of actinide-ion concentrations. A working electrolyte is positioned within the interior of a container. The working electrolyte is separated from a reference electrolyte by a separator. A working electrode is at least partially in contact with the working electrolyte. A reference electrode is at least partially in contact with the reference electrolyte. A voltmeter is electrically connected to the working electrode and the reference electrode. The working electrolyte comprises an actinide-ion of interest. The separator is ionically conductive to the actinide-ion of interest. The separator comprises an actinide, Zr, and Nb. Preferably, the actinide of the separator is Am or Np, more preferably Pu. In one embodiment, the actinide of the separator is the actinide of interest. In another embodiment, the separator further comprises P and O.

Actinide-ion sensor

Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

A basket, for use in the reduction of UO to uranium metal and in the electrorefining of uranium metal, having a continuous annulus between inner and outer perforated cylindrical walls, with a screen adjacent to each wall. A substantially solid bottom and top plate enclose the continuous annulus defining a fuel bed. A plurality of scrapers are mounted adjacent to the outer wall extending longitudinally thereof, and there is a mechanism enabling the basket to be transported remotely.

Growing community of inventors

Idaho Falls, ID, United States of America

Steven Herrmann