Tadafumi Koyama

Actinides recovery from molten salt/liquid metal system by electrochemical methods

Abstract Electrochemical methods were examined for the recovery of actinides from the electrorefiner which is used in pyrometallurgical reprocessing of spent metal fuel for fast reactors. Uranium was successfully collected at the solid steel cathode from both liquid cadmium and molten salt solvents. In electrotransport from liquid cadmium, the behavior of uranium and rare earths was as expected by a computer simulation code based on the diffusion layer model at the interface between the electrolyte and the electrodes. In electroreduction from the molten salt electrolyte, a considerable amount of uranium was reduced at the CdLi anode by direct chemical reduction with lithium, especially at a lower anodic current density. The decrease in collection efficiency of uranium due to the direct chemical reduction would be avoided by maintaining the anode potential higher than the deposition potential of uranium.

Development of an environmentally benign reprocessing technology—pyrometallurgical reprocessing technology

Abstract Present status of research and development on pyrometallurgical reprocessing technology in Central Research Institute of Electric Power Industry is described with emphasis on electrorefining and waste salt immobilization. As for the electrorefining, three different electrodes — anode basket, solid cathode, and liquid-cadmium cathode — have been being investigated; prismatic anode basket with faster rotation was found to accelerate metal-fuel dissolution. Morphology and collection efficiency of electrodeposited uranium on solid cathode were found to vary with ratio of cathode to anode surface area. Liquid-cadmium cathode with a paddle-shape stirrer was developed and determined maximum uranium concentration into the cathode without dendrite formation. As for waste salt immobilization, sodalite is proposed as waste form and synthesized by dry reaction without gas generation. Measured leachability of the synthesized sodalite is as low as those of vitrified waste form.

Electrodeposition of uranium in stirred liquid cadmium cathode

Abstract The electrodeposition of U in a liquid Cd cathode was known to be hampered by the formation of dendritic U on the Cd surface. Electrotransports of uranium to the stirred liquid Cd cathode were carried out at 773 K for different cathode current densities and different Reynolds number of stirring. The maximum amount of U taken in the liquid Cd cathode without forming dendrites was found to increase with an increasing Reynolds number of stirring and decrease with increasing cathode current density.

Pyrometallurgical Reprocessing of Fast Reactor Metallic Fuel—Development of A New Electrorefiner with A Ceramic Partition

A new electrorefiner with a ceramic partition has been developed for pyrometallurgical reprocessing of metallic fuel. In this electrorefiner, dissolution of spent fuel and deposition take place simultaneously, resulting in an increase of the processing rate. The feasibility of this electrorefiner was confirmed by a polarization profile and a current efficiency of an electrotransportation of uranium from a pure uranium anode to an iron cathode through a liquid cadmium pool. Separation of active fission products from actinide was confirmed by a transportation of simulating fission products elements with and without imposing electropotential. The maximum cathode current density onto a liquid cadmium pool without formation of a dendrite was measured against the concentration, and it was found to decrease with increasing concentration of uranium in cadmium. The estimated time required to process 50 kg of heavy metal by the new electrorefiner was less than that of the original electrorefiner.

Study of Molten Salt Electrorefining of U-Pu-Zr Alloy Fuel.

Electrorefining of unirradiated metal alloy fuel, U-20Pu-10Zr (% by weight), was carried out in an Ar atmosphere cell to obtain a basic knowledge of spent fuel treatment. Before electrorefining, the Pu ion concentration in LiCl-KCl electrolyte was adjusted to 4 wt.% using pure Pu metal and CdCl2. Cylindrical alloy fuel was anodically dissolved in the salt electrolyte. Electrotransport to either liquid Cd cathode or solid cathode was carried out several times in galvanic mode. The deposit on the solid cathode mostly consisted of metal U and entrained salt and took place at high current efficiency. The recovery of Pu and Am into a liquid cadmium cathode was also demonstrated with high current efficiency. The anodic dissolution of the fuel alloy was found to progress from the outside whilst leaving a dense salt layer on the alloy surface. The peculiar fluctuation of anode potential during U-Pu-Zr dissolution was explained by the competitive oxidation of U-Pu and Zr.

Development of Fast Reactor Metal Fuels Containing Minor Actinides

Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phénix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU.

Integrated Experiments of Electrometallurgical Pyroprocessing Using Plutonium Oxide

Electrometallurgical pyroprocessing is a promising technology to realize actinide fuel cycle. Integrated experiments to demonstrate electrometallurgical pyroprocessing of PuO2 in continuous operation were carried out. In each test, 10–20 g of PuO2 was reacted with Li reductant to form metal product. The reduction products were charged in an anode basket of the electrorefiner with LiCl-KCl-UCl3 electrolyte. Using the anode, deposition of uranium on the solid cathode was carried out when PuCl3/UCl3 concentration ratio was low. After the Pu/U ratio in the salt electrolyte was increased enough, Pu and U were recovered simultaneously on a liquid cadmium cathode. By heating up the deposits for distillation of the salt and the cadmium, U metal or Pu-U alloyed metal was obtained as residues in the crucible. It was the first result to demonstrate the recovery of metal actinides in the continuous operation of pyroprocessing of oxide fuels.

Equilibrium Distribution of Actinides Including Cm Between Molten LiCl-KCl Eutectic and Liquid Cadmium

Equilibrium distribution of actinides both in molten LiCl-KCl eutectic and liquid cadmium were measured from the concentration data obtained in electrorefining tests and reductive extraction tests. Separation factors for U, Np, Am, Cm against Pu were derived in the practical temperature range of 700 K to 783 K. The derived separation factors are consistent with the reported values measured at 773 K and 723 K. The temperature dependence for Cm is different compared to the other actinides (U, Np and Am). This behavior remains unclear and additional experimental measurements of distribution coefficient of Cm are required before ruling on the real behavior.

Recovery of Transuranium Elements from Real High-Level Liquid Waste by Pyropartitioning Process

A pyropartitioning process is under development to recover minor actinide elements from high-level liquid waste (HLLW) generated by Purex reprocessing. This pyropartitioning process consists of a denitration step that converts various elements in the HLLW into oxides, a chlorination step that converts the oxides into chlorides, and a reductive-extraction step that separates the actinide elements from fission products (FPs) in the chlorination product. The feasibility of each step was confirmed using simulating FPs and unirradiated transuranium elements (TRUs). In the present study, approximately 520 g of real HLLW was prepared to demonstrate the feasibility of the pyropartitioning process. Almost 100% of each TRU originally contained in the HLLW was recovered in the liquid cadmium phase in the reductive-extraction step, which showed that the expected chemical reactions were completed and the mass loss of TRUs was negligible in the denitration, chlorination, and reductive-extraction steps. The separation behaviors of actinide elements, including americium and curium, from FPs in the reductive-extraction step were quite similar to those observed in previous experiments using unirradiated materials. Hence, the pyropartitioning process was successfully verified.

Early construction and operation of highly contaminated water treatment system in Fukushima Daiichi Nuclear Power Station (I) – Ion exchange properties of KURION herschelite in simulating contaminated water

To support the design and operation of the decontamination system using KURION media for the treatment of highly contaminated water accumulated in Fukushima Daiichi Nuclear Power Station, Central Research Institute of Electric Power Industry has urgently carried out many kinds of research and development programs to support the operation of the decontamination system using columns filled with three kinds of KURION media (H, AGH and SMZ). Since the contaminated water at Fukushima Daiichi Nuclear Power Station contained seawater and oil, the effects of sea salt and dissolved oil on Cs adsorption behavior were examined closely by batch type. The concentration of sea salt in the solutions was varied between 0.0 and 3.4 wt%. The Cs adsorption capacity of KURION herschelite in seawater decreased to nearly 1/10th of that in pure water, but it was still concluded that herschelite has sufficient adsorption capacity to remove Cs from the contaminated water. The effect of dissolved oil could be ignored because of its low solubility in seawater. Langmuir-type adsorption isotherm equations, which can be applied for estimating Cs adsorption in sea salt containing water, were developed.