Kazunori Nomura

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.

Separation of Trivalent Minor Actinides from Fission Products Using Single R-BTP Column Extraction Chromatography

As part of the Fast Reactor Cycle Technology Development (FaCT) Project, research and development has been underway on a system for reprocessing spent fuel from fast breeder reactors. In this system, the method of extraction chromatography is used to recover minor trivalent actinides (MA(III) = Am(III) + Cm(III)) from acidic raffinate in the solvent extraction process that recovers U, Np, and Pu. In general, extractants for solvent extraction can be used as impregnated adsorbents for the extraction chromatography system. The principle of separation used in extraction chromatography is similar to that of solvent extraction. Because of the similarity in chemical properties between MA(III) and Lanthanides(III), MA(III) recovery processes using solvent extraction consist of two steps, namely, MA(III) Ln(III) recovery and MA(III)/Ln(III) separation. Of these, MA(III)/ Ln(III) separation is one of the most challenging issues. Since nitrogen and sulfur donors bind more readily to MA(III) than to Ln(III), a large number of N-donor extractants have been developed in many research projects. Kolarik et al. have reported that a new N-donor ligand, 2,6-bis(5,6-dialkyl-1,2,4-triazine-3-yl)pyridine (R-BTP), shows high selectivity for MA(III) over Ln(III). To develop a partitioning process using the extraction chromatography technique, Wei et al. reported an excellent adsorbent for extraction chromatography. The support particle for the adsorbent consisted of porous silica supports coated with styrenedivinylbenzene polymer (SiO2-P). In many partitioning methods using extraction chromatography, MA(III) separation from the raffinate is achieved using a two-column unit system corresponding to MA(III) Ln(III) recovery and MA(III)/Ln(III) separation, which is similar to the solvent extraction process. In this study, the authors attempted MA(III) direct separation from simulated raffinate using a novel single-column unit system. The partitioning plant for the fast reactor cycle using the singlecolumn unit system will be a simple and compact structure compared with the two-column unit system. The column for the single-column system was packed with the adsorbent, which was impregnated with R-BTP into SiO2-P, for the separation of MA(III) from the raffinate. The behaviors of Am(III) and Cm(III) in separating from the raffinate were examined through a column experiment using a single column packed with the R-BTP/SiO2-P adsorbent.

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.

Experimental Study on U-Pu Cocrystallization Reprocessing Process

A new reprocessing system with a 2-stage crystallization process has been developed. In the first stage of the system, U and Pu are recovered from dissolver solution by U-Pu cocrystallization. Laboratory-scale experiments were carried out with U and Pu mixed and irradiated fuel dissolver solutions to obtain fundamental data on the U-Pu cocrystallization process. Pu(VI) was cocrystallized with U, but crystallization yields of Pu were lower than those of U. FPs were separated from U and Pu by cocrystallization, and decontamination factors of Cs and Eu to U in crystals were over 100.

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.

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.

Am, Cm recovery from genuine HLLW by extraction chromatography

The extraction chromatography experiments for Am(III) and Cm(III) recovery from genuine HLLW were carried out in order to demonstrate modified 2 flow-sheets using CMPO/SiO2-P and HDEHP/SiO2-P adsorbents for obtaining DTPA-free MA(III) product solution. The first flow-sheet achieved about 90% MA(III) recovery with more than 103 of decontamination factor for 155Eu. However, further modification is necessary for separation of light lanthanides. Purification of MA(III) from both of heavy and light lanthanides was successfully done by the 2nd flow-sheet although recovery yields of MA(III) was almost the same with the current flow-sheet i.e. 70%. The recovery yield is expected to be improved by some optimizations in operation conditions such as column length, flow rate of the mobile phase and etc.