Masaki Kurata

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.

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 interaction between granular B4C and 304L-type stainless steel materials used in BWRs in Japan

Chemical reactions between stainless steel and boron carbide were investigated using the materials applied for control rods in BWRs in Japan, specifically 304L-type stainless steel and granular boron carbide. The reaction region consisted of 2–4 layers, in which the significant composition variation of each element was detected, especially for B and C. Assuming that the reaction layer growth obeys the parabolic law, the effective rate constant between 304L-type stainless steel and granular boron carbide was evaluated to be approximately one order of magnitude smaller than the previously reported values for boron carbide pellets or powers. This difference might originate from the loose contact between the stainless steel and the granular boron carbide in the present study. Regarding liquefaction progress, the stainless steel components were selectively dissolved in the melt; consequently, the unreacted boron carbide tended to remain.

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.

Pyrochemical treatment of spent nitride fuels for MA transmutation

Nitride fuels have several advantages including high thermal conductivity and high metal density (like metallic fuels) and high melting point and isotropic crystal structure (like oxide fuels). Since the late 1990s, the partitioning and transmutation of minor actinides (MA) has been studied to decrease the long-term radio-toxicity of high-level waste and to mitigate the burden of final disposal. Japan Atomic Energy Agency (JAEA) has proposed a dedicated transmutation cycle using an accelerator-driven system (ADS) with nitride fuels containing MA. The nitride fuel cycle we have developed includes a pyrochemical process. Our focus is on the electrolysis of nitride fuels and their refabrication from the recovered actinides; other processes are similar to the technology for metal fuel treatment and have been studied elsewhere. Here, we summarize our activity on the development of the pyrochemical treatment of spent nitride fuels.

Chlorination of UO2 and (U,Zr)O2 solid solution using MoCl5

Conversion of UO2 and (U0.5Zr0.5)O2 solid solution into chlorides by MoCl5 was performed in order to confirm the applicability of the chlorination by MoCl5 to a pretreatment of fuel debris by pyrochemical methods. Chlorination of (U0.5Zr0.5)O2 powders and dense pieces was successfully achieved at 573 and 773 K, respectively, based on the following chemical reaction: 2(U0.5Zr0.5)O2 + 4MoCl5 = UCl4 + ZrCl4 + 4MoOCl3. Rough separation of MoCl5, ZrCl4 and MoOCl3 from UCl4 was achieved by volatilization under temperature gradient. From these results, fundamental feasibility of the chlorination method using MoCl5 as a pretreatment of fuel debris was shown.

Fundamental experiments on phase stabilities of Fe–B–C ternary systems

To understanding the control blade degradation mechanism of a boiling-water reactor (BWR), a thermodynamic database for the fuel assembly materials is a useful tool. Although the iron, boron, and carbon ternary system is a dominant phase diagram, phase relation data are not sufficient for the region in which boron and carbon compositions are richer than the eutectic composition. The phase relations of three samples such as Fe0.68B0.06C0.26 (at%), Fe0.68B0.16C0.16 (at%), and Fe0.76B0.06C0.18 (at%) were analyzed by X-ray diffraction, scanning electron microscopy, and energy–dispersive X-ray spectrometry. The results indicate that the Fe3(B,C) phase exists only in the intermediate region at 1273 K and that the solidus temperature widely maintains at approximately 1400 K for all three samples; these results differ from the calculated data using the previous thermodynamic database. The difference might originate from the overestimation of the interaction parameter between boron and carbon in Fe3(B,C). Proper titling was performed using the present data.