Steven D. Herrmann

Electrolytic Reduction of Spent Nuclear Oxide Fuel as Part of an Integral Process to Separate and Recover Actinides from Fission Products

Abstract Bench‐scale tests were performed to study an electrolytic reduction process that converts metal oxides in spent nuclear fuel to metal. Crushed spent oxide fuel was loaded into a permeable stainless steel basket and submerged in a molten salt electrolyte of LiCl–1 wt% Li2O at 650°C. An electrical current was passed through the fuel basket and a submerged platinum wire, effecting the reduction of metal oxides in the fuel and the formation of oxygen gas on the platinum wire surface. Salt and fuel samples were analyzed, and the extent of fission product separation and metal oxide reduction was determined.

SEPARATION AND RECOVERY OF URANIUM METAL FROM SPENT LIGHT WATER REACTOR FUEL VIA ELECTROLYTIC REDUCTION AND ELECTROREFINING

Abstract A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs performed in succession with a single salt loading of molten LiCl-1 wt% Li2O at 650°C. Analysis of salt samples following the series of electrolytic reduction runs identified the partitioning of select fission products from the spent fuel to the molten salt electrolyte. The extent of metal oxide conversion in the posttest fuel was also quantified, including a 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500°C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

Actinide recovery experiments with bench-scale liquid cadmium cathode in real fission product-laden molten salt

Abstract This article summarizes the observations and analytical results from a series of bench-scale liquid cadmium cathode experiments that recovered transuranic elements together with uranium from a molten electrolyte laden with real fission products. Variable parameters such as the ratio of Pu3+/U3+ in the electrolyte, liquid cadmium cathode voltage, and feed materials were tested in the liquid cadmium cathode experiments. Actinide recovery efficiency and Pu/U ratio in the liquid cadmium cathode product under variable conditions are reported in this paper. Separation factors for actinides and rare earth elements in the molten LiCl-KCl/cadmium system are also presented.

Modeling the Pyrochemical Reduction of Spent UO2 Fuel in a Pilot-Scale Reactor

Abstract A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be dependent only upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.

Diffusion Model for Electrolytic Reduction of Uranium Oxides in a Molten LiCl-Li2O Salt

Abstract In this study, a diffusion-based kinetic model essential for design and operational analysis of spent nuclear fuel reduction has been developed. The model considers the cathode side of the system to be rate limiting and deals with diffusion of lithium metal through the basket loaded with uranium oxide (UO2 or U3O8). Faraday’s law was implemented into the model to observe the electrochemical effect on the model. Solutions with different conditions are developed, and detailed results are presented. These solutions were compared against experimental bench scale data. At high operating current conditions (I > 0.8 A), the model fits the data well. The fitting resulted in estimated effective lithium diffusion coefficients for high and low void fraction UO2 crushed fuels of 8.5 × 10−4 cm2/s and 2.2 × 10−4 cm2/s, respectively. The effective diffusion coefficient for U3O8 is estimated to be 8.6 × 10−4 cm2/s. In some experiments, a porous magnesium oxide basket was used for containing the U3O8. It was estimated that the lithium diffusion coefficient through this magnesia basket is 3.3 × 10−5 cm2/s.

Separation and Recovery of Uranium and Group Actinide Products From Irradiated Fast Reactor MOX Fuel via Electrolytic Reduction and Electrorefining

A series of bench-scale tests was conducted with irradiated fast reactor MOX fuel to separate and recover refined uranium and group actinide products via electrolytic reduction and electrorefining. The fuel was declad, crushed, immersed in a pool of molten LiCl −1 wt% Li2O at 650°C, and electrolyzed to convert the mixed oxide fuel to metal. The reduced fuel was then electrorefined in LiCl-KCl-UCl3 at 500°C, yielding a refined uranium metal product. Additional electrorefining experiments were performed in which actinides (that is, uranium, neptunium, plutonium, and americium) were recovered as a group metal product.

Electrochemical analysis of actinides and rare earth constituents in liquid cadmium cathode product from spent fuel electrorefining

Abstract The results of a recently reported series of bench-scale actinide recovery experiments with liquid cadmium cathodes (LCCs) are subjected to a more detailed analysis in this paper. It is suggested that separation efficiency (SE), not separation factor (SF), should be used to assess the effectiveness of an LCC to separate actinides from rare earth (RE) elements. The common definition of SF for any pair of actinide and RE elements in the molten salt/liquid Cd system is the ratio of their distribution coefficients, which are measured under equilibrium conditions. The definition of SE is broader than that of SF. For any pair of actinide and RE elements in the molten salt/liquid Cd system, SE is the ratio of their distribution coefficients, such as SEPu-U = DPu/DU, where DPu and DU are measured at either equilibrium or nonequilibrium conditions. The relationship of SE with SF is linear and can be expressed as SEPu-U = DPu/DU and DPu = SFPu-UDU + b. When DPu and DU are measured under equilibrium conditions, SE is equal to SF. The physical or chemical meaning of the intercept b is not clear. From a mathematical point of view, the absolute values of b reveal the differences between the measured DPu/DU or SE and SF. The negative values of b indicate that the SE measurement results are smaller than the associated SF. The values of b may be used to evaluate the SE of LCC on electrochemically recovered actinides from fission product elements. An electrochemical model was developed to investigate the mechanism of RE contamination of the actinides collected by the LCC. It was confirmed that REs were electrochemically transported into the Cd phase. A more negative LCC voltage has a stronger impact on the quantities of REs transported into the Cd than those of the actinides.