Tsutomu Koizumi

Separation of Actinide Elements by Solvent Extraction Using Centrifugal Contactors in the NEXT Process

Using the advanced aqueous reprocessing system named NEXT process, actinides recovery was attempted by both a simplified solvent extraction process using TBP as an extractant for U, Pu and Np co-recovery and the SETFICS process for Am and Cm recovery from the raffinate. In U, Pu and Np co-recovery experiments a single cycle flow sheet was used under high nitric acid concentration in the feed solution or scrubbing solution. High nitric acid concentration in the feed solution aided Np oxidation not only in the feed solution, but also at the extraction section. This oxidation reaction accomplished Np extraction by TBP with U and Pu. Most of Np could be recovered into the product solution. In the SETFICS process, a TRUEX solvent of 0.2 mol/dm3 CMPO and 1.4 mol/dm3 TBP in n-dodecane was employed instead of 0.2 mol/dm3 CMPO and 1.0 mol/dm3 TBP in n-dodecane in order to increase the loading of metals. Instead of sodium nitrate, hydroxylamine nitrate was applied to this experimental flow sheet in accordance with a “salt-free” concept. The counter current experiment succeeded with the Am and Cm product. On the high-loading flow sheet, compared with the previous flow sheet, the flow of the aqueous effluents and spent solvent were expected to decrease by about one half. Two solvent extraction experiments for actinides recovery demonstrated the utility of the flow sheet of these processes in the NEXT process.

Uranium Crystallization Test with Dissolver Solution of Irradiated Fuel

The crystallization process has been developed as a part of the advanced aqueous process, NEXT (New Extraction System for TRU recovery) for fast reactor (FR) cycle. In this process, a large part of U is separated from dissolver solution by crystallization as UO2(NO3)2.6H2O. The U crystallization test was carried out with real dissolver solution of irradiated FR fuel to investigate the influence of cooling rate on the crystal size and the behavior of fission product (FP) compared with that of Pu(IV). In regard to the influence of the cooling rate, it was confirmed that the crystal size was smaller as the cooling rate is faster. Although it was expectable that the decontamination performance was improved by diminishing the specific surface of the crystals, it was suggested that a large crystal produced by crystallization was not always high purity. Concerning the behavior of FPs, Eu behaved similarly to Pu(IV). Cs accompanied with U into the crystals under the condition in this test.

Enhancement of Decontamination Performance of Impurities for Uranyl Nitrate Hexahydrate Crystalline Particles by Crystal Purification Operation

Abstract A crystal purification process consisting of sweating and melt filtration was developed to improve decontamination factors (DFs) of fission product impurities from uranyl nitrate hexahydrate (UNH) crystal recovered from a dissolver solution of irradiated fast reactor core fuel. Batch experiments on the sweating and melt filtration processes were carried out at 56 to 80°C. Although the DFs of solid impurities such as Cs and Ba remain the same in the sweating process, those of liquid impurities such as Zr, Nb, Ru, Ce, and Eu were 2.32, 2.40, 2.50, 2.45, and 2.60 at 60°C. On the other hand, the DF of Pu for the UNH crystal slightly increased to 1.25 at 60°C. Because Pu incorporated the UNH crystal in both the solid impurities such as Cs2Pu(NO3)6 and in the liquid impurities, Pu in the liquid fraction was removed by the sweating operation. Decontamination of liquid impurities was effective with sweating time and with a rise in sweating temperature. In the melt filtration process, 0.45- to 5.0-μm–diam filters were used for the separation of the molten UNH crystal. The DF of Ba was approximately ten times as high as the crude crystal with 0.45- to 5.0–μm-diam filters. The particle size of Pu and Cs formed as Cs2Pu(NO3)6 was quite small. As a proof of this, although the decontamination of Pu and Cs was not effective with a 5.0-μm-diam filter, their DFs rose 2.7 times using a 0.45-μm-diam filter.

Evaluation of precipitation behavior of zirconium molybdate hydrate

In the dissolution step of spent nuclear fuel, there is a world-concern problem that zirconium molybdate hydrate precipitates as a byproduct, and accumulates in some reprocessing equipments. In order to prevent this accumulation, we have developed a new method based on the controlled reaction crystallization of zirconium molybdate hydrate (ZMH) in the reprocessing solution, followed by solid liquid separation. In order to measure the particle size of ZMH, batch crystallization experiments were conducted by varying nitric acid concentration and operating temperature. In result, almost all particle sizes scatter around 1 μm on average, despite the higher concentration of nitric aid and operating temperature, and then small particles grow up as an aggregate sticking to the crystallizer. Moreover, polymorph and color changing were observed by varying the concentration of nitric acid and reaction time. These results suggest that crystal color and adhesiveness are closely related to the particle size of ZMH. And the control of nitric acid concentration and small particle growth would be the useful technique to prevent the ZMH sticking.

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.

Behavior of Actinide Elements and Fission Products in Recovery of Uranyl Nitrate Hexahydrate Crystal by Cooling Crystallization Method

Abstract To elucidate various kinds of actinide element and fission product behavior, U crystallization experiments were carried out with a uranyl nitrate solution and with a solution in which irradiated fast reactor core fuel was dissolved. Insoluble residue simulating that found in actual reactor operation was not incorporated into the uranyl nitrate hexahydrate (UNH) crystal in the course of the U crystallization. However, the decontamination factors (DFs) were below 10 even when the UNH crystal was washed because the mother liquor containing the simulated insoluble residue occupied the interspaces of the agglutinated UNH crystal. In the U crystallization process, the DF of Pu was >40 when the UNH crystal was washed. But, Np was not removed from the UNH crystal because Np was oxidized to Np(VI) in the feed solution and thus was co-crystallized with U(VI). Cesium exhibited different behavior depending on whether Pu was present. Although a high DF of Cs was obtained in the case of uranyl nitrate solution without Pu, Cs was hardly separated at all from the UNH crystal formed from the dissolver solution of irradiated fast reactor core fuel. It is likely that crystals of a mixed salt of Pu and Cs, Cs2Pu(NO3)6, precipitated from the dissolver solution. Since Ba precipitated as Ba(NO3)2 during the crystallization process, its DF was low after the UNH crystal was washed. On the other hand, Am, Cm, Rb, Sr, Zr, Nb, Ru, Sb, and rare earth elements remained in the mother liquor at the time of U crystallization. Therefore, portions of these elements in the mother liquor that was attached to the surface of the UNH crystal were washed away with HNO3 solution.

Precipitation Behavior of Dicesium Tetravalent Plutonium Hexanitrate in Cooling Crystallization of Uranyl Nitrate Hexahydrate

Abstract There is concern that a binary salt of Pu(IV) and Cs forms deposits on the uranyl nitrate hexahydrate (UNH) crystal formed in the dissolver solution for U crystallization containing Cs. Precipitation behavior of dicesium tetravalent plutonium hexanitrate, Cs2Pu(NO3)6, in the U crystallization process is studied. In this work, the solubility of Cs2Pu(NO3)6 was measured in a HNO3 solution, and influence of Pu valence and Cs concentration in the dissolver solution on decontamination factors (DFs) of Pu and Cs in the crystal was examined in the U crystallization process. The solubility of Cs2Pu(NO3)6 increased with a decrease in the concentration of HNO3 in the mother liquor and a rise in temperature of the mother liquor. In the U crystallization process, although the DF of Cs was low where there was Pu(IV) since the two were difficult to separate in the feed solution, Cs was removed thoroughly where there was Pu(VI) in the feed solution. The Cs concentration in the feed solution affected the DFs of Pu and Cs after the UNH crystal was washed. The DFs of Pu and Cs had a tendency to decrease with increase of Cs concentration in the feed solution, because large amounts of Cs+ contributed to the formation of Cs2Pu(NO3)6.

Physicochemical properties of dicesium tetravalent plutonium hexanitrate in uranium crystallization process

Tetravalent Pu reacts with Cs ions to form the crystalline precipitate of Cs2Pu(NO3)6under certain chemical conditions during the U crystallization process. The Cs2Pu(NO3)6precipitate reduces the decontamination factor (DF) of Cs to U in the crystal after being washed. The solubility and thermal properties of Cs2Pu(NO3)6 were studied with the aim of providing a characterization estimate. The solubility of Cs2Pu(NO3)6 increased with decreases in HNO3 concentration. Loss in weight of the compound caused by thermal degradation of Cs2Pu(NO3)6 to Cs2PuO2(NO3)4 was observed at 245 °C in thermal analysis. A uranyl nitrate hexahydrate (UNH) crystal was obtained by cooling irradiated fast reactor core fuel dissolver solution. The DF of Cs decreased with increasing the HNO3 concentration of the mother liquor because more Cs2Pu(NO3)6 precipitates with high concentration of HNO3.

Development of New Decladding System in the Reprocessing Process for FBR Fuel

Seiya YAMADA, Tadahiro WASHIYA, Masayuki TAKEUCHI, Tsutomu KOIZUMI and Shinichi AOSE As a part of the feasibility study on commercialized fast reactor cycle systems, Japan Nuclear Cycle Development Institute (JNC) has been developing the fuel decladding technology for the dry reprocessing process (oxide electrowinning process) and aqueous reprocessing process. Particularly, in the oxide electrowinning process, the spent fuel should be reduced to powder for quick dissolution in the molten salt at electrolyzer. Therefore, JNC proposes new decladding system with innovative mechanical decradding devices. The decladding system consists of fuel crushing stage, hull separation stage and hull rinsing stage. In the fuel crushing stage, disassembled spent fuel pins are crushed and powdered by mechanical decladding device, then the following stage, the hull and the fuel powder are separated by magnetic separator. Only the fuel powder is fed to the electrolyzer. On the other side, the separated hull is melted by induction heating method, and the small amount of oxide included in the hull fragments is recovered at the hull rinsing stage. The recovered oxide fuel is fed back to the electrolyzer. In this paper, the basic performance of the element equipment that composes this new decladding system will be descried.