Koichi Uozumi

Electrode reaction of the Np3+/Np couple at liquid Cd and Bi electrodes in LiCl–KCl eutectic melts

The electrode reactions of the Np3+/Np couple at liquid Cd and Bi electrodes were investigated by cyclic voltammetry at 723, 773 and 823 K in the LiCl–KCl eutectic melt. It was found that the diffusion of Np3+ in the salt phase was the rate-determining step in the cathodic reaction when the concentration of NpCl3 was less than about 1 wt % and the liquid Cd or Bi phase was not saturated with Np. The redox potentials of the Np3+/Np couple at the liquid Cd electrode at 723, 773 and 823 K were observed to be more positive than those at the Mo electrode by 0.158, 0.140 and 0.126 V, respectively. The potential shift results from a lowering of the activity of Np in the Cd phase according to the formation of the NpCd11 alloy at 723 K and NpCd6 at 773 and 823 K. The redox potentials of the Np3+/Np couple at the liquid Bi electrode at 723, 773 and 823 K were more positive than those at the Mo electrode by 0.427, 0.419 and 0.410 V, respectively, which is attributable to a lowering of the activity of Np in the Bi phase due to the formation of NpBi2.

Development of an Innovative Electrorefiner for High Uranium Recovery Rate from Metal Fast Reactor Fuels

To increase the uranium recovery rate of molten salt electrorefining step in pyrometallurgical reprocessing of metallic fast reactor fuels, tests were carried out using electrorefiners equipped with mechanisms for scraping cathode deposits. After the modifications in the design of the anode basket and scraper mechanism, no stalling of the anode and scraper rotation due to interference by cathode deposits occurred. Under the condition that codissolution of zirconium and uranium was allowed in order to obtain maximum throughput, a current of 400–450A was maintained until 82% of the initially loaded uranium was recovered. The uranium recovery rate for the same duration reached 789 g U/h (32.9 g U/h_L per electrode volume). On the assumption that an electrorefiner operates for 20 h/d and 200 d/y in an actual pyrometallugical reprocessing facility, this result corresponds to a uranium recovery rate of 3.16 t U/y using only one electrode assembly of about 30 cm diameter, which should be a sufficiently high performance for practical use. From these results, the engineering feasibility of uranium recovery using an electrorefiner with cathode deposit scraper mechanism has been demonstrated.

Pyrometallurgical Partitioning of Uranium and Transuranic Elements from Rare Earth Elements by Electrorefining and Reductive Extraction

High-level liquid waste generated from PUREX reprocessing contains a small amount of transuranic elements, such as Np, Pu, Am, and Cm, with long-lived radioactivities. A pyrometallurgical partitioning process is being developed to recover transuranic elements from such waste. Small amounts of U contained in the high-level liquid waste are also recovered in the process. A key issue for developing the process is effective separation of U and the transuranic elements from the rare-earth elements, because the two elemental groups are chemically analogous. A series of process tests were carried out in the present study to demonstrate that a combination of electrorefining and reductive extraction is useful for separating U and transuranic elements from the rare-earth elements. The results indicate that 99% of U and each transuranic element is recovered by the combination process as a product, and that the quantity of rare-earth elements contained in the product is smaller than the transuranic elements by weight. The overall mass balance of U and transuranic elements in the system ranged within the experimental errors assigned to sampling and analysis.

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.

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.

Recovery of U by electrolysis of UN in LiCl-KCl eutectic melts

The electrochemical behavior of UN in the LiCl-KCl eutectic melt containing UC13 at the temperature between 723 and 823 K was investigated by cyclic voltammetry. The redox reaction of UN seemed to be irreversible, since the generated N2gas was escaped from the salt phase. By the controlled-potential electrolysis of UN at -0.55 V vs. Ag/AgCl reference electrode in the UCl3-LiCl-KCl eutectic melt, UN was dissolved into the salt as U3+ at the anode and U metal was recovered at the cathode.

Thermodynamic Evaluation of Liquid Cd Cathode Containing U and Pu

In our previous study, a mixture of U and Pu was recovered in liquid Cd cathode from molten salt under various conditions of the U:Pu ratio. Two important things were observed. The first was that three kinds of precipitated phase had been detected in the saturated liquid Cd cathode, such as a U metal and two kinds of U-Pu-Cd compound. The compositions of the compounds were roughly (U+Pu):Cd = 1:11 and (U+Pu):Cd = 1:6. The second was that the U metal had selectively precipitated in the saturated liquid Cd cathode under the condition that the U:Pu ratio is higher than about 0.8 in the liquid Cd phase. In the present study, phase diagrams were evaluated by the CALPHAD method on the liquid Cd cathode containing U and Pu. The U-Pu-Cd compounds were modeled as MCd11-type and MCd6-type, respectively, based on the reported binary phase diagrams of U-Cd and Pu-Cd. The calculated result reasonably agreed with the experimental observations. The variations in the U and Pu activities were estimated along with the U:Pu ratio, which is related to the precipitation of various phases in the liquid Cd cathode.

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.

Parameter Surveys on Glass-Bonded Sodalite Synthesis Conditions from Spent Salt Generated in Pyroprocess

Abstract A parameter-based survey of the synthesis conditions by a so-called pressureless consolidation method to fabricate glass-bonded sodalite waste form for stabilizing fission products generated in pyrometallurgical reprocessing of spent metal fuel was performed. The maximum temperature, the heating duration at the maximum temperature, the glass fraction in the initial material, and the weight load used for pressing the material were chosen as the variable parameters. Accordingly, modified conditions to reduce the maximum temperature and increase the weight load were selected for reducing the volatilized-salt ratio during the heating and the free-salt ratio in the product. By fabricating a simulated waste under the modified conditions, the effect of changing the conditions was confirmed. Leaching tests in pure water using the consolidated products fabricated under both reference and modified conditions showed that the stability of the products was not significantly deteriorated by modifying the heating conditions.