Takatoshi Hijikata

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.

Equilibrium distribution of rare earth elements between molten KCl-LiCl eutectic salt and liquid cadmium

Abstract Distribution experiments for several rare earth elements (La, Ce, Pr, Nd and Y) between molten KCl-LiCl eutectic salt and liquid Cd were carried out at 450, 500 and 600°C. The material balance of rare earth elements after reaching the equilibrium and their distribution and chemical states in a Cd sample frozen after the experiment were examined. The results suggested the formation of solid intermetallic compounds at the lower concentrations of rare earth metals dissolved in liquid Cd than those solubilities measured in the binary alloy system. The distribution coefficients of rare earth elements between two phases (mole fraction in the Cd phase divided by mole fraction in the salt phase) were determined at each temperature. These distribution coefficients were explained satisfactorily by using the activity coefficients of chlorides and metals in salt and Cd. Both the activity coefficients of metal and chloride caused a much smaller distribution coefficient of Y relative to those of other elements.

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.

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.

Development of High-Temperature Transport Technologies for Liquid Cadmium in Pyrometallurgical Reprocessing

Pyrometallurgical reprocessing is one of the most promising technologies for the advanced fuel cycle with favorable economic potential and intrinsic proliferation-resistance. The feasibility of pyrometallurgical reprocessing has been studied through many laboratory-scale experiments. Hence the development of the engineering technology necessary for pyrometallurgical reprocessing is a key issue for its industrialization. The development of high-temperature transport technologies for molten salt and liquid cadmium is crucial for pyrometallurgical processing; however, there have been a few transport studies on high-temperature fluids. In this study, a metal transport test rig was installed in an argon glove box with the aim of developing technologies for transporting liquid cadmium at approximately 773 K. The transport of liquid Cd using gravity was controlled by adjusting the valve. The liquid Cd was transported by a suction pump against a 0.93 m head and the transport amount of Cd was well controlled with the Cd amount and the position of the suction tube. The transportation of liquid cadmium at approximately 700 K could be controlled at a rate of 0.5–2.5 dm3/min against a 1.6 m head using a centrifugal pump.

Early construction and operation of the highly contaminated water treatment system in Fukushima Daiichi Nuclear Power Station (II) – dynamic characteristics of KURION media for Cs removal in simulated contaminated water

The kinetic characteristics of the column were necessary property to be understood before actual operation. Hence, a functional small-scale zeolite column system was installed for conducting the experiments to understand decontamination behaviors. Each column has a 2 cm inner diameter and a 12 cm height, and 12 g of zeolite-type media was packed into the column. The column experiments were carried out with Kurion-zeolite, herschelite, at different feed rates of simulated water with different concentrations of Cs and sea salt. As expected from equilibrium ion-exchange isotherms obtained for KURION-herschelite, the adsorption of Cs is hampered by the existence of sea salt ratio. The difference in breakthrough behaviors can be ascribed to the difference in sea salt ratio. Above 1000 bed volumes, the adsorption rate of Cs was the same at a solution velocity of between 14 and 81 cm/min. Under the condition of a 3.4 wt% sea salt ratio, the performance of the media supplied by KURION was in the order surfactant modified zeolite < silver-impregnated engineered herschelite = herschelite (H). This result was suggested to evaluate the performance of KURION media on the actual columns.

Conversion of Simulated High-level Liquid Waste to Chloride for the Pretreatment of Pyrometallurgical Partitioning Process

A pyrometallurgical partitioning process is being developed for recovering minor actinides from high-level liquid waste resulting from PUREX reprocessing. Since the high-level liquid waste consists of concentrated raffinate, concentrated alkaline waste and insoluble residues, the various elements in the waste must be converted to chlorides before they can be sent on to the pyrometallurgical partitioning process. The conversion to chlorides is done by a combination of denitration and chlorination. The mass balance of these processes was measured in the present study using simulated high-level liquid waste. The results indicate that almost all of the alkali elements and Re, substituting for Tc, and significant amounts of Se, Cr, and Mo were separated by denitration, and that Cr, Fe, Zr, Mo, and Te were separated by chlorination. The remaining noble metals, Ni, U, and alkaline-earth and rare-earth elements were efficiently converted to chlorides, which were then supplied to the reductive extraction test using a molten salt/liquid-Cd system to demonstrate that the obtained chlorides are appropriate for processing by pyrometallurgical partitioning. In further reduction, noble metals and Ni were reductively extracted into the liquid-Cd phase, and the rare-earth elements and U into the liquid-Cd phase by adding Li reductant. These elements were completely separated from the alkaline-earth elements remaining in the chloride phase.

Early construction and operation of the highly contaminated water treatment system in Fukushima Daiichi Nuclear Power Station (III) – a unique simulation code to evaluate time-dependent Cs adsorption/desorption behavior in column system

A simulation code was developed to evaluate the performance of the cesium adsorption instrument operating in Fukushima Daiichi Nuclear Power Station. Since contaminated water contains seawater whose salinity is not constant, a new model was introduced to the conventional zeolite column simulation code to deal with the variable salinity of the seawater. Another feature of the cesium adsorption instrument is that it consists of several columns arranged in both series and parallel. The spent columns are replaced in a unique manner using a merry-go-round system. The code is designed by taking those factors into account. Consequently, it enables the evaluation of the performance characteristics of the cesium adsorption instrument, such as the time history of the decontamination factor, the cesium adsorption amount in each column, and the axial distribution of the adsorbed cesium in the spent columns. The simulation is conducted for different operation patterns and its results are given to Tokyo Electric Power Company (TEPCO) to support the optimization of the operation schedule. The code is also used to investigate the cause of some events that actually occurred in the operation of the cesium adsorption instrument.

Measurement of Molten Chloride Salt Flow and Demonstration of Simulated Fission Product Removal Using a Zeolite Column Apparatus for Spent Salt Treatment in Pyroprocessing

Abstract A zeolite column system is under development to realize both a high decontamination factor and high throughput for the treatment of the spent salt generated in the pyroprocessing of the metal fuel cycle. To study the feasibility of the zeolite column system from an engineering aspect, an engineering-scale zeolite column apparatus was installed. Measurements of the superficial velocities of molten salt passing through the columns filled with granular form type-A zeolite at various driving pressures showed that the conventional relationship of the velocity and pressure loss in the components of the apparatus can be useful for the molten salt system. Then, a demonstration test to simulate the decontamination of a fission product, which was represented by cesium, was performed using a zeolite that had been pretreated in aqueous solutions to remove the sodium. Although the absorbed amount of cesium was not as high as previously reported, the concentration of cesium in the effluent salt exhibited a breakthrough curve. Therefore, some of the cesium in the salt was absorbed into the zeolite, and accordingly, the feasibility of the zeolite column system was demonstrated.