Tadahiro Washiya

Extraction and Stripping Tests of Engineering-Scale Centrifugal Contactor Cascade System for Spent Nuclear Fuel Reprocessing

The Japan Atomic Energy Agency has been developing centrifugal contactors for solvent extraction to apply to next-generation reprocessing plants. The centrifugal contactor has some attractive advantages such as more compact design and shorter liquid residence times than conventional contactors. Many kinetic studies using a miniature centrifugal contactor have been carried out worldwide. However, there are few engineering-scale studies in which stage efficiencies, transition behavior of concentration profiles, and robustness under maloperation conditions have been comprehensively discussed for a contactor cascade system. In this study, we carried out extraction and stripping tests of an engineering-scale centrifugal contactor cascade system based on a flowsheet of 10 kg-HM/h using uranyl nitrate solution. As a result, the stage efficiencies on uranium extraction and stripping were quite high, nearly 100% for extraction and 97–98% for stripping. The uranium concentration profiles became stable within 10 minutes for both extraction and stripping sections. No overflow or entrainment was observed under normal operation during the extraction and stripping tests. During the stripping test, it was estimated that an increase in temperature of the feed stripping solution from 308 to 333K or a decrease in the flow rate ratio of the organic to aqueous phase from 1.0 to 0.8 corresponded to the distribution capacity of the two contactors. The maloperation test, in which a motor at a stage of the contactor cascade system was intentionally stopped, showed that the system could maintain stable operation with no emergency shutdown following the installation of at least two additional stages.

Influence of nitric acid and plutonium concentrations in dissolver solution of mixed oxide fuel on decontamination factors for uranyl nitrate hexahydrate crystal

Abstract In order to examine the decontamination behavior of the Pu and fission products (FPs) that contained with the uranyl nitrate hexahydrate (UNH) crystals in the U crystallization process, experiments were carried out using mixed oxide (MOX) fuel dissolver solution. The experiments confirmed that Eu was adequately decontaminated by washing the UNH crystals with a HNO3 solution. However, Ba crystallized as Ba(NO3)2 and the washing was ineffective for the decontamination of Ba. High HNO3 and Pu concentrations in the mother liquor, the decontamination factor (DF) of Cs was low because Cs precipitated with Pu as Cs2Pu(NO3)6. The Pu clearly showed a reasonable DF because the amount of Pu in the dissolver solution was higher than that of Cs. Almost all the amount of Pu remains in the mother liquor except the one in the Cs2Pu(NO3)6 precipitate.

Dissolution behavior of (U,Zr)O2-based simulated fuel debris in nitric acid

To explore the possibility of dissolving fuel debris into nitric acid as a potential pre-treatment for waste treatment in which the U and Pu are removed from the inventory, dissolution tests of U1−xZrxO2 and (U,Pu)1−xZrxO2 were carried out in 6 M HNO3 at 353 K. At the end of the dissolution test (after 4 h), the ratio of dissolved uranium decreased with an increase in the Zr contents, x. While the dissolution of U-rich samples was congruent, a preferential leaching of U was observed with Zr-rich samples. Taking into account these different dissolution phenomena, the dissolution rate analysis was carried out using surface-area model to calculate the instantaneous dissolution rate (IDR). The IDR decreased from 10−5 down to 10−10 mol cm−2 min−1 as x increased from 0 to 0.95. From these findings, dissolution with HNO3 is expected to be only applicable in U-rich part of fuel debris (x < 0.3) if the dissolution in 6 M HNO3 at 353 K is assumed. Application of complexing acids, such as mixture of HNO3 and HF, should be considered to increase the dissolution rate of the Zr-rich part.

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm−3) will be treated. This dissolution process involves short stroke shearing of fuels (∼10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 102–103 times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio.

Removal of Liquid and Solid Impurities from Uranyl Nitrate Hexahydrate Crystalline Particles in Crystal Purification Process

The purification behavior of uranyl nitrate hexahydrate (UNH) was investigated to evaluate the decontamination performance of liquid and solid impurities using a dissolver solution of mixed oxide (MOX) fuel in batch experiments. The UNH crystal recovered from the MOX fuel dissolver solution containing simulated fission products (FPs) was purified by a sweating and melt filtration process. Although the decontamination factors (DFs) of Pu, Cs, and Ba did not change in the sweating process, that of Eu increased with increases in temperature and time. These results indicate that liquid impurities such as Eu were effectively removed by the sweating method, but solid impurities such as Pu, Cs, and Ba were minimally affected in the batch experiments. On the other hand, the DF of Ba increased with 0.45 and 5.0 μ filters in the melt filtration process. Since Pu and Cs formed as Cs2Pu(NO3)6 in the course of U crystallization and was accompanied with the UNH crystal, these behaviors were similar to each other. Although the DFs of Pu and Cs did not change with the 5.0 μ filter, it increased approximately twofold with the 0.45 μ filter. The particle size of Cs2Pu(NO3)6 is relatively small and might pass through the 5.0 μ filter in the melt filtration process. The liquid impurities as Eu remained in the molten UNH crystal with some filters.

Development of Crystallizer for Advanced Aqueous Reprocessing Process

Crystallization is one of the remarkable technologies for future fuel reprocessing process that has safety and economical advantages. Japan Atomic Energy Agency (JAEA) (former Japan Nuclear Cycle Development Institute), Mitsubishi Material Corporation and Saitama University have been developing the crystallization process. In previous study, we carried out experimental studies with uranium, MOX and spent fuel conditions, and flowsheet analysis was considered. [1, 2, 3] In association with these studies, an innovative continuous crystallizer and its system was developed to ensure high process performance. From the design study, an annular type continuous crystallizer was selected as the most promising design, and performance was confirmed by small-scale test and engineering scale demonstration at uranium crystallization conditions. In this paper, the design study and the demonstration test results are described.Copyright

Applicability of Single Mode Fiber Laser for Wrapper Tube Cutting

Japan Atomic Energy Agency (JAEA) has been developing a fuel disassembly system with reliability for FBR fuel reprocessing. Laser technology has a high cutting performance and stable operation and was apply to disassembly system in our previous studies. However, it was hard to produce the impeccable disassembly system, because it was occurred the problems, such as pin damage and dross adhesion between a wrapper tube and fuel pins.After that, the advance of the laser cutting technology has recently attracted a great deal of attention from industry. In particular, single mode fiber laser (SMFL), which has a small beam size and high beam quality, has been reported as a new oscillator. Then, it was presumed that SMFL might provide a possible to prevent the original matters in the disassembly.The main purpose of this study is to reevaluate an applicability of laser for the wrapper tube cutting by the basic cutting tests. Concretely, the authors researched whether every cutting condition such as SMFL etc, has the effects on the original matters or not. This experimental results show that the kerf width of SMFL is still narrower than that of multi mode fiber laser (MMFL). It is an important phenomenon to decrease the amount of dross. Therefore, the authors confirmed that SMFL is suitable for prevention of the original matters and the new feasibility method of wrapper tube cutting.© 2012 ASME

Continuous-Operation Test at Engineering Scale Uranium Crystallizer System

Uranium crystallization based on solubility difference is one of the remarkable technologies which can provide simple process to separate uranium in nitric acid solution since the process is mainly controlled by temperature and concentration of solute ions. Japan Atomic Energy Agency (JAEA) and Mitsubishi Materials Corporation (MMC) are developing the crystallization process for elemental technology of FBR fuel reprocessing.[1–3] The uranium (U) crystallization process is a key technology for New Extraction System for TRU Recovery (NEXT) process that was evaluated as the most promising process for future FBR reprocessing.[4–6] We had developed an innovative crystallizer and carried out several fundamental investigations. On the basis of the results, we fabricated an engineering-scale crystallizer and have carried out continuous operation test to investigate the stability of the equipment at steady and non-steady state conditions by using depleted uranium. As for simulating typical failure events in the crystallizer, crystal accumulation and crystal blockage were occurred intentionally, and monitoring method and resume procedure were tried and selected in this work. As the test results, no significant phenomenon was observed in the steady state test. And in the non-steady state test, process fluctuation could be detected by monitoring of screw torque and liquid level in the crystallizer, and all failure events are proven to be recovered by appropriate resumed procedures.© 2009 ASME

Development of Short Stroke Shearing Technology for FBR Fuel Pins

The short stroke shearing tests with simulated fuel pin bundle were carried out in engineering scale. The shearing device was designed to handle the simulated Monju (FBR prototype reactor) type fuel pin bundle. Monju type and Commercial reactor type simulated fuel pins were used for the test. The length of sheared pin and the opening ratio of sheared section were measured under several shearing settings such as the pressure to hold pin bundle, the shearing speed and the filling-ratio of pins in the pin magazine. Both types of fuel pin were able to be sheared accurately at the length of about 10mm, and the opening ratio of sheared section was not significantly reduced. As the results, fundamental data of the short stroke shearing characteristics were obtained and that shearing method was confirmed to be promising with the reliable shearing device.Copyright

Design and Fabrication of the FBR Fuel Disassembly System

Japan Atomic Energy Agency (JAEA) has been developing a reliable disassembly system for FBR fuel reprocessing as a part of Fast Reactor Cycle Technology Development (FaCT). As FBR fuel pins are installed in a hexagonal shaped wrapper tube made of stainless steel, the fuel pins should be separated from the wrapper tube prior to the shearing process. JAEA has been developing the laser beam cutting method and the mechanical cutting method as the disassembly system. Although Fiber laser system showed a good cutting performance, it couldn’t completely avoid fuel pin damage and adhesion during the cutting operation. So we focused on the mechanical method to minimize such troubles. Two types of mechanical cutting modes have to be developed to realize the disassembly procedure, namely, the slit-cut for the wrapper tube and the crop-cut for the end plug region of the fuel pin bundle. To ensure disassembly technology of commercial reactor fuel assemblies, we designed and fabricated the testing machine of disassembly system having the cutting modes in engineering scale. We confirmed basic functions of this machine and improved its performance. We will soon demonstrate engineering operation by a series of disassembling and pin bundle handling procedure; separating fuel pins from wrapper tube, transferring them to the fuel magazine for shearing. Scattering of cutting dust cause machine troubles and transition of it to the dissolution process together with pins causes unknown problems. To resolve the problems, collection device of cutting dust will be tested and the cutting condition to make the disassembly easy to cut will be improved.Copyright