Shelly X. Li

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.

Engineering-Scale Liquid Cadmium Cathode Experiments

Abstract Recovery of uranium and transuranic (TRU) actinides from spent nuclear fuel by an electrorefining process was investigated as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative. Experiments were performed in a shielded hot cell at the Materials and Fuels Complex at Idaho National Laboratory. The goal of these experiments was to collect, by an electrochemical process, kilogram quantities of uranium and plutonium into what is called a liquid cadmium cathode (LCC). For each experiment, a steel basket loaded with chopped spent nuclear fuel from the Experimental Breeder Reactor II acted as the anode in the electrorefiner. The cathode was a beryllium oxide crucible containing ~26 kg of cadmium metal (the LCC). In the three experiments performed to date, between 1 and 2 kg of heavy metal was collected in the LCC after passing an integrated current between 1.80 and 2.16 MC (500 and 600 A h) from the anode to the cathode. Sample analysis of the processed LCC ingots measured detectable amounts of TRUs and rare earth elements.

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.

Development of Computational Models for the Mark-IV Electrorefiner—Effect of Uranium, Plutonium, and Zirconium Dissolution at the Fuel Basket-Salt Interface

Abstract The electrochemical processing of spent metallic nuclear fuel has been demonstrated by and is currently in operation at the Idaho National Laboratory (INL). At the heart of this process is the Mark-IV electrorefiner (ER). This process involves the anodic dissolution of spent nuclear fuel into a molten salt electrolyte along with a simultaneous deposition of pure uranium on a solid cathode. This allows the fission products to be separated from the fuel and processed into an engineered waste form. A one-dimensional model of the Mark-IV ER has begun to be developed. The computations thus far have modeled the dissolution of the spent nuclear fuel at the anode taking into account uranium (U3+), plutonium (Pu3+), and zirconium (Zr4+). Uranium and plutonium are the two most important elements in the system, whereas zirconium is the most active of the noble metals. The model shows that plutonium is quickly exhausted from the anode, followed by dissolution of primarily uranium, along with small amounts of zirconium. The total anode potential as calculated by the model has been compared to experimental data sets provided by INL. The anode potential has been shown to match the experimental values quite well with root-mean-square (rms) values of 2.27 and 3.83% for two different data sets, where rms values closer to zero denote better fit.

A Computational Model of the Mark-IV Electrorefiner: Phase I―Fuel Basket/Salt Interface

Spent driver fuel from the Experimental Breeder Reactor-II (EBR-II) is currently being treated in the Mk-IV electrorefiner (ER) in the Fuel Conditioning Facility (FCF) at Idaho National Laboratory. The modeling approach to be presented here has been developed to help understand the effect of different parameters on the dynamics of this system. The first phase of this new modeling approach focuses on the fuel basket/salt interface involving the transport of various species found in the driver fuels (e.g. uranium and zirconium). This approach minimizes the guessed parameters to only one, the exchange current density (i0). U3+ and Zr4+ were the only species used for the current study. The result reveals that most of the total cell current is used for the oxidation of uranium, with little being used by zirconium. The dimensionless approach shows that the total potential is a strong function of i0 and a weak function of wt% of uranium in the salt system for initiation processes.

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.

COMPUTATIONAL MODEL OF THE MARK-IV ELECTROREFINER: TWO-DIMENSIONAL POTENTIAL AND CURRENT DISTRIBUTIONS

Abstract A computational model of the Mark-IV electrorefiner is currently being developed as a joint project between Idaho National Laboratory, Korea Atomic Energy Research Institute, Seoul National University, and the University of Idaho. As part of this model, the two-dimensional potential and current distributions within the molten salt electrolyte are calculated for U3+, Zr4+, and Pu3+ along with the total distributions, using the partial differential equation solver of the commercial Matlab software. The electrical conductivity of the electrolyte solution is shown to depend primarily on the composition of the electrolyte and to average 205 mho/m with a standard deviation of 2.5 × 10–5% throughout the electrorefining process. These distributions show that the highest potential gradients (thus, the highest current) exist directly between the two anodes and cathode. The total, uranium, and plutonium potential gradients are shown to increase throughout the process, with a slight decrease in that of zirconium. The distributions also show small potential gradients and very little current flow in the region far from the operating electrodes.

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.

INTEGRATED EFFICIENCY TEST FOR PYROCHEMICAL FUEL CYCLES

Abstract An engineering-scale pyroprocessing integrated efficiency test was conducted with sodium-bonded, spent Experimental Breeder Reactor II drive fuel elements. The major pieces of equipment used to conduct the test were the element chopper, Mk-IV electrorefiner, cathode processor, and casting furnace. Four batches of the spent fuel (containing 50.4-kg heavy metal) were processed under a set of fixed operating parameters. The primary goal of the test was to demonstrate the actinide dissolution and recovery efficiencies typical of the fixed operating parameters that have been developed for this equipment based on over a decade’s worth of processing experience. The total mass balance for the test was 101.28% (slightly more output than input). The uranium mass balance for the test was 100.13%. The test results indicate that 99.3 wt% of uranium in the feed was electrochemically dissolved and 98.4 wt% of the uranium was collected as metal ingots. The complexity of zirconium behavior during electrorefining was confirmed by the test results. More than 85 wt% of the zirconium was electrochemically dissolved during the later stages of the electrorefining process. However, only 33.7 wt% of the zirconium was collected as metal in the ingots. The balance of the zirconium is believed to reside in the cadmium pool. The test also identified that the dross streams from the cathode processor and casting furnace account for ˜2.4 wt% of the uranium relative to the feed.