Yoshiharu Sakamura

Electrochemical Reduction of UO2 in Molten CaCl2 or LiCl

To develop an electrochemical reduction technique for the reprocessing of nuclear fuels, the reduction behavior of UO 2 at the cathode and the anode reactions were investigated in both CaCl 2 and LiCl salt baths. In the CaCl 2 at about 800°C, UO 2 was reduced into metal over the potential range <0.6 V vs the Ca 2+ /Ca. The reduced uranium metal cohered due to the high temperature and a dense metal skin covered the surface of the UO 2 disk sample. It prevented the transportation of oxygen from the inside to the bulk salt and the reduction often stopped with UO 2 remaining inside. A significant underpotential deposition of calcium metal was observed. In the LiCl at 650°C, UO 2 was reduced into metal over the potential range <0.15 V vs the Li + /Li. The UO 2 disk sample was satisfactorily reduced because the LiCI melt could permeate into the sample. The current efficiency of UO 2 reduction in the LiCl was much better than in the CaCl 2 . The anodic currents for oxygen and carbon oxide gas evolutions were verified in cyclic voltammograms of the platinum and glassy carbon electrodes. At the platinum surface, Pt 3 O 4 and Li 2 PtO 3 yielded in the CaCl 2 and LiCl, respectively.

Measurement of standard potentials of actinides (U,Np,Pu,Am) in LiCl–KCl eutectic salt and separation of actinides from rare earths by electrorefining

Abstract Pyrochemical separation of actinides from rare earths in LiCl–KCl eutectic–liquid metal systems has been studied. The electromotive forces of galvanic cells of the form, Ag|Ag(I), LiCl–KCl‖actinide(III), LiCl–KCl|actinide, were measured and standard potentials were determined for uranium, neptunium and plutonium to be −1.283 V, −1.484 V and −1.593 V (at 450°C vs. Ag/AgCl (1wt%–AgCl)), respectively. A typical cyclic voltammogram of americium chloride has two cathodic peaks, which suggests reduction Am(III)→Am(II) occurs followed by reduction of Am(II) to americium metal. Standard potential of Am(II)/Am(0) was estimated to be −1.642 V. Electrorefining experiments to separate actinides (U, Np, Pu and Am) from rare earths (Y, La, Ce, Nd and Gd) in LiCl–KCl eutectic salt were carried out. It was shown that the actinide metals were recovered on the cathodes and that americium was the most difficult to separate from rare earths. The actinide separation will be achieved by means of the combination of electrorefining with multistage extraction.

Distribution behavior of uranium, neptunium, rare-earth elements ( Y, La, Ce, Nd, Sm, Eu, Gd) and alkaline-earth metals (Sr,Ba) between molten LiClKCI eutectic salt and liquid cadmium or bismuth

Abstract Distribution coefficients of uranium neptunium, eight rare-earth elements (Y, La, Ce, Pr, Nd, Sm, Eu and Gd) and two alkaline-earth metals (Sr and Ba) between molten LiCl-KCI eutectic salt and either liquid cadmium or bismuth were measured at 773 K. Separation factors of trivalent rare-earth elements to uranium or neptunium in the LiCl-KCl/Bi system were by one or two orders of magnitude larger than those in the LiCl-KCl/Cd system. On the contrary, the separation factors of alkaline-earth metals and divalent rare-earth elements to trivalent rare-earth elements were by one or two orders of magnitude smaller in the LiCl-KCl/Bi system.

Electrolytic Reduction and Electrorefining of Uranium to Develop Pyrochemical Reprocessing of Oxide Fuels

Abstract Two series of pyrochemical reprocessing tests for oxide fuels, consisting of pretreatment, electrolytic reduction, and electrorefining processes, were conducted using ~100 g of UO2. In the pretreatment process, UO2 pellets of the starting material were oxidized into U3O8 powder, which simulated fuel decladding by voloxidation. Then, UO2 sinter with a porosity of 30 to 38% was fabricated from the U3O8 powder. Two cathode baskets charged with ~100 g of the UO2 sinter were prepared, and two electrolytic reduction tests were carried out in a LiCl-Li2O electrolyte at 650°C. The results suggested that the reduction to uranium metal could be completed within 10 h with the current efficiency >62%. It was verified that the porous UO2 sinter was of great advantage to the electrolytic reduction process. In the subsequent electrorefining process, the reduction products were charged in two anode baskets, and electrolysis was carried out in a LiCl-KCl-UCl3 electrolyte at 500°C. Within 8 h, most of the uranium metal was anodically dissolved into the electrolyte with the current efficiency >88%. Dendritic uranium metal was collected on a stainless steel cathode. Consequently, it was demonstrated that a refined uranium metal could be produced from UO2 pellets with a high degree of efficiency.

Thermodynamic quantities of actinides and rare earth elements in liquid bismuth and cadmium

Abstract The activity coefficients of uranium, lanthanum, cerium, praseodymium, gadolinium and yttrium in liquid bismuth were measured electrochemically as 10−4.65, 10−13.44, 10−13.00, 10−12.81, 10−11.04 and 10−10.17 respectively at 773 K. The activity coefficients of neodymium and neptunium were estimated as 10−12.85 and 10−7.43 respectively from the standard potentials in LiClKCl eutectic molten salt and the distribution coefficients in the LiClKCl/bismuth system. The partial molar enthalpy of mixing and the partial molar excess entropy of solutes in both liquid bismuth and cadmium were evaluated from the change in the activity coefficients with temperature.

Zirconium Behavior in Molten LiCl-KCl Eutectic

Some oxidation states (0, +1, +2, and +4) of zirconium exist in a LiCl-KCl eutectic system over the temperature range 450-550°C, and the behavior is complicated. In cyclic voltammograms at 500°C, a cathodic peak was observed at about -1.2 V vs. Ag/AgCl reference electrode, which might be due to the reduction of Zr(IV) to ZrCl and zirconium metal. Two anodic peaks might correspond to the oxidation of ZrCl and zirconium metal, respectively. The electrolysis at a cathode potential of about -1.1 V yielded a nodular deposit identified as ZrCl, which appeared to be a metastable compound in this system. When the potential was sufficiently negative (i.e., < -1.35 V), zirconium metal was obtained. The deposited zirconium metal was fine black powder, and adhesion to the cathode wire was poor. In the presence of cadmium metal at the cathode, an intermetallic compound that might be Cd 3 Zr was obtained. The collection efficiency of zirconium is improved using cadmium because the adhesion of the intermetallic compound was much better. Zirconium metal reacted with Zr(IV) to give Zr(II) whose solution was light brown, and Zr(II) was easily disproportionated into Zr(IV) and zirconium metal. The anodic dissolution test indicated that the zirconium metal primarily dissolved into the electrolyte salt as Zr(IV). The Zr(II)/Zr(IV) ratio seemed to be very low and to increase with increasing temperature.

Distribution behavior of plutonium and americium in LiCl–KCl eutectic/liquid cadmium systems

Abstract The thermodynamics of plutonium and americium in LiCl–KCl eutectic/liquid cadmium systems was studied with interest in the oxidation state of americium in the salt phase. The standard potential of plutonium vs. the Ag/AgCl (1 wt% AgCl) electrode, E 0 Pu/Pu(III) , in the LiCl–KCl eutectic was measured in the temperature range of 400–500°C and given by the equation with a standard deviation, σ =0.0009 V: E 0 Pu/Pu(III) (V)=−2.204+0.000845 T (K). The Ag/AgCl electrode had been carefully calibrated using the Li–Al electrode. The potential of the cadmium containing plutonium and americium, E Cd , was measured at 500°C as a function of the distribution coefficient ( D : mole fraction in salt divided by mole fraction in cadmium), and represented by the following equations. over the range of E Cd >−1.45 V: E Cd =−1.360 (±0.004)+0.0511 log D Am =−1.348 (±0.002)+0.0511 log D Pu . It is indicated that americium as well as plutonium is present in the trivalent oxidation state in the salt under this condition. Based on the potential data, the activity coefficient of plutonium in liquid cadmium and the separation factor of americium relative to plutonium were determined to be (1.74±0.28)×10 −4 and 1.77±0.46, respectively. Under the reducing conditions (i.e. E Cd E Cd and log D Am , indicates that divalent americium is possibly present in the salt phase.

Application of Electrochemical Reduction to Produce Metal Fuel Material from Actinide Oxides

Abstract The electrochemical reduction process has been recently developed for converting oxide nuclear fuels to metals. In order to characterize the reduction mechanism and to investigate appropriate conditions for improving the reduction rate, several reduction tests were conducted in a LiCl-Li2O electrolyte at 650°C using various types of cathode baskets containing 10 to 100 g of UO2. The reduction progressed from the outside to the center of the cathode basket, and the reduction rate might be determined by the transportation of oxygen ion to the bulk salt. It was verified that feeding in small UO2 particles and reducing the thickness of the UO2 layer in the cathode basket improved the reduction rate. The completion of UO2 reduction was indicated by the open circuit potential of the cathode basket exhibiting lithium deposition potential for a long time. A salt distillation test was conducted using the reduction product comprising a mixture of porous uranium metal particles and the electrolyte. The reduction product loaded in an yttria crucible was heated to 1400°C in an argon stream. The residue in the crucible consisted of a uranium metal ingot and a small amount of dross. The adhering LiCl seemed to be completely removed. Consequently, it was demonstrated in the electrochemical reduction followed by the salt distillation that a uranium metal ingot could be produced from the UO2 feed with a high degree of efficiency.

Thermodynamic assessment of systems of actinide or rare earth with Cd

A pyrometallurgical process is being developed for recycling nuclear reactor fuels. Thermodynamic information on multicomponent systems of actinides and rare earths (REs) with liquid Cd is very useful in the design of a process in which a liquid Cd electrode is used for the selective recovery of Pu and U, and a reductive extraction process using a molten salt/liquid Cd system for the recovery of minor actinides, such as Np, Pu, etc. A key issue in the design of these processes is a variation in solubility or activity of actinides or REs in multielement systems. In the present study, phase diagrams of U-Cd, Pu-Cd, Np-Cd, Y-Cd, La-Cd, Ce-Cd, Pr-Nd, Nd-Cd, and Gd-Cd were optimized by the CALPHAD method. For these systems, thermodynamic data, such as the activity of solutes in liquid Cd and the Gibbs energies of formation of the intermetallic compounds as well as the phase diagram data were available for the optimization. For optimization, the calculated primary results were entered into a database. Then, some ternary systems were preliminarily assessed through the use of the optimized data for the binary systems. Two extreme conditions were assumed: one condition was complete miscibility between the compounds that have the same mole ratio between solutes and Cd; the other condition was no solid solubility between the compounds. The results indicated the tendencies toward solubility and activity of actinides and REs in multielement systems.

Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology

A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO2) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO2 is separately collected for ~60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which has the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO2 spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to ~10 wt% of the total spent fuel owing to the prior UO2 recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process.