Mineo Fukushima

Gamma-ray irradiation test of electric components of rescue mobile robot Quince

On March 11, 2011, a massive earthquake and tsunami hit eastern Japan, particularly the Tohoku area. Since then, the Fukushima Daiichi Nuclear Power Station has been facing a crisis. To respond to this situation, we began a project to redesign our mobile robots for disaster response missions. A key issue to be addressed was to check the radiation hardness of the electric components of our robot. Initially, no information was available in this regard. Therefore, we conducted gamma-ray irradiation tests for the electric components using cobalt-60. In this paper, we introduce the procedure for the irradiation test and report the results of the test.

Recovery of U-Pu Alloy from MOX Using a Pyroprocess Series

Abstract A series test of the pyroprocess was carried out to recover U-Pu alloy from mixed oxide (MOX) pellets. In the Li-reduction step, the reduction behavior of MOX was similar to that of UO2. In the electrorefining step, the separation factor between U and Pu was 1.9 for the combination of the reduced MOX anode and the liquid cadmium cathode, which agrees well with the value obtained in previous studies. Approximately 99% of the HM (U and Pu) initially present in the anode or molten salt was detected in the electrodes or molten salt after the electrolysis. Considering the analytical error of inductively coupled plasma-atomic emission spectroscopy, this mass balance is reasonable. The amount of U remaining in the anode was slightly larger than that of Pu, due to the reoxidation. The U-Pu alloy ingot was successfully formed by distillation of Cd.

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.

ELECTROCHEMICAL REDUCTION OF MOX PELLETS IN MOLTEN LITHIUM CHLORIDE BASED ON A PRACTICAL OPERATING CONDITION

Abstract Previous studies for electrochemical reduction using uranium oxide have shown that reduction was completed within several tens of hours when particles or powders of oxide were used for the cathode material. In the case of mixed oxide (MOX) fuel prepared for fast reactors, there are two significant differences with respect to uranium oxide fuel for light water reactors. The MOX fuel contains ~30% Pu and a small amount of Am. The density of the uranium oxide pellet and MOX pellet is ~98% and ~85% with respect to theoretical values, respectively. These differences decrease the electroconductivity of oxide and the reaction rate. Also, the behavior of transuranic elements has not been certified. In the present study, electrochemical reduction of MOX pellets was performed by setting the pellets directly on the cathode in a molten lithium chloride bath. Reduction was completed after ~15 h, even when using MOX pellets. This value compares closely to the previous values for uranium oxide particles or powders. Current efficiency was varied at ~60%, which is slightly higher than in the previous study. The lower density of MOX allows better diffusion of the molten salt into the pellet and contributes to efficient electrolysis. Concerning actinide behavior during electrolysis, the uranium and plutonium concentrations in the molten salt bath were lower than their detection limits. Although a small amount of americium was dissolved in the molten salt bath and gradually accumulated, the amount was <1% with respect to the initial amount. The oxygen concentration in the molten salt decreased gradually during electrolysis. These variations in the salt hardly affected the current efficiency and the actinide recovery ratio. These observations indicate that the electrochemical reduction of MOX pellets is applicable to industrial processes.

CHEMICAL FORM OF ACTINIDE ELEMENTS CONTAINED IN ANODE RESIDUE GENERATED IN ELECTROLYSIS AND THE CONVERSION TO CHLORIDES USING ZrCl4

Abstract Various residues containing uranium and transuranic are considered to be generated in pyroprocessing, and provided that the actinide elements are recovered from the residues, this can contribute to increasing the recovery ratio in the entire process. In this study the chemical form of the anode residues generated in our previous electrolysis test was investigated. The anode residue consisted of PuOCl, PuO2, and UO when electrolysis was performed using reduced oxide fuels, which are thought to be formed by the reaction between the anode residue and U-chloride contained in the molten salt. By adding ZrCl4 the actinide contained in the residue was converted to chloride. The chlorination reaction took ~10 h to complete.

Sequential Electrolysis of U-Pu Alloy Containing a Small Amount of Am to Recover U- and U-Pu-Am Products

Sequential electrolysis testing was performed in a LiCl-KCl eutectic molten salt bath, in which variation in current efficiency, composition of product or molten salt, etc. were measured successively. Porous or dense U-Pu alloys containing a small amount of Am were dissolved one after another from the anode and, simultaneously, U dendrites or U-Pu-Am alloys were recovered on the iron rod or in the liquid cadmium cathode, respectively. The current efficiencies were mostly maintained to be 100% except for the anodic dissolution of the porous alloy and the cathodic recovery of U-Pu-Am in the liquid cadmium. The Pu and Am concentrations in the U dendrites varied in the ranges of 120–350 and 4–15 ppm, respectively. The separation factors for U/Pu and Am/Pu in the liquid cadmium varied in ranges of 1.69–1.74 and 0.66–0.78, respectively. The sum of U, Pu, and Am concentrations in the molten salt bath was maintained in the entire sequence, although the U concentration decreased or increased when using the iron rod or liquid cadmium cathode, respectively. The variation in the Pu concentration compensated for the U concentration variation. Material balance of Pu was mostly maintained at 100% with respect to the initial amount at each step of the sequence. These results indicate that the sequence used in the present study can be recycled.

Low Current Efficiency in MOX Deposition Tests

A series of MOX deposition tests has been performed since 2001 at RIAR to clarify its complex phenomena and to improve its poor current efficiency. In the 2001 tests, the cathode current efficiency was between 60 and 100% but the Pu fraction in the MOX was between 5 and 20%. In 2002 tests, the fraction was raised to more than 30% by modifying the test conditions but the current efficiency fell to between 20 and 60%. A new method was proposed to simulate the parasitic current due to the electrode reactions of UO2 2+/UO2+, Pu4+/Pu3+ and Fe3+/Fe2+ at the cathode. It was found that the parasitic current due to the UO2 2+/UO2+ reaction significantly lowers the current efficiency especially when the cathode potential is kept near the equilibrium value during the electrolysis to increase the Pu fraction in the MOX deposit.

Electro-deposition behavior of minor actinides with liquid cadmium cathodes

Transuranic elements have been simultaneously electro-deposited on a liquid cadmium cathode and decontaminated from fission products dissolved in molten 3LiCl-2KCl in the electrorefining of a pyrochemical reprocessing. Some lab-scale electrolysis experiments were carried out using uranium, plutonium and minor actinide elements in order to evaluate the performance of liquid cadmium cathodes. The results of several experiment confirmed that neptunium, americium and curium could be recovered together with Pu by the liquid cadmium cathodes through electrolysis operation. The separation factors of minor actinide vs. Pu were estimated to be about 0.5 to 2.5 and those of rare earth element vs. total actinides 18 to 32.