Masaumi Nakahara

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.

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.

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.

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.

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.

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.

Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

A solvent extraction flowsheet for Pu partitioning, based on the acid split method without reductant, originally proposed by the Oak Ridge National Laboratory (ORNL), was tested for sodium-cooled fast reactor fuel reprocessing. To enhance resistance to nuclear proliferation, a flowsheet for co-processing was developed that controls Pu content in the products while avoiding Pu polymerization and formation of a third phase during extraction. In this method, Pu is partitioned using the difference in distribution coefficients of U and Pu. It is effective for selective Pu stripping from U at low temperatures and HNO3 concentrations. The flowsheet with a supply of 0.15 mol/dm3 HNO3 solution at 21°C for Pu partitioning was tested experimentally using miniature centrifugal contactors and a highly radioactive solution. Neither a Pu(IV) polymer nor a third phase was observed during the experiment. The Pu content in the U/Pu product increased to 2.28 times that in the feed solution. The leakage ratio of Pu to the U product was slightly less in the U stripping section. Some fission products (FPs) were effectively decontaminated; e.g., decontamination factors (DFs) of Cs in U/Pu and U products were 4.51×105 and 2.42×105, respectively.

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.