Takahiro Chikazawa

Selective Extraction of Americium(III) over Macroscopic Concentration of Lanthanides(III) by Synergistic System of TPEN and D2EHPA in 1‐Octanol

Abstract The synergistic extraction with N,N,N′,N′‐tetrakis(2‐methylpyridyl)‐ethylenediamine (TPEN) and di(2‐ethylhexyl)phosphoric acid (D2EHPA) in 1‐octanol demonstrates good selectivity for Am(III) over Ln(III) in solutions containing macroscopic concentrations of the latter. The maximum apparent separation factor, which is defined as the ratio of the distribution ratio of Am(III) to that of Eu(III), is ca. 80 at the molar ratio of D2EHPA to TPEN, 2.0. This ratio corresponds to the result of the spectrophotometric titration, which indicates that TPEN and D2EHPA associate in 1‐octanol and two D2EHPA molecules coordinate to Eu(III)–TPEN complex, [Eu(TPEN)]3+. In the present study, association of D2EHPA and TPEN is one of the most important factors for the synergistic extraction and the complexation of TPEN and D2EHPA with Eu(III).

Batch Crystallization of Uranyl Nitrate

Batch crystallization of uranyl nitrate is carried out in order to obtain fundamental data required for the development of reprocessing involving crystallization. Particular attention is paid to the development of a method for predicting the concentrations of uranium and nitric acid in the mother liquor and the amount of uranyl nitrate crystals produced. Initial concentrations of uranyl nitrate and nitric acid are 500–600 g/l and 4–6 mol/l, respectively, corresponding to the condition of a dissolver solution of spent fuel. Steady-state mass balance equations including the correlation equation for the equilibrium solubility of uranium nitrate are applied to the prediction. The calculated concentrations of uranium and nitric acid are in close agreement with the experimental ones. The recovery of uranium is accurately predicted by the calculated concentrations, with an error of less than 10%.

Flowsheet Study of U-Pu Co-Crystallization Reprocessing System

A U-Pu co-crystallization reprocessing system is proposed for light water reactor fuels and its flowsheet study is carried out. This reprocessing system is based on experimental evidence indicating that Pu(VI) in a nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate is not present in the solution. The system consists of five steps: dissolution of spent fuel, Pu oxidation, U-Pu co-crystallization, re-dissolution of the crystals, and U re-crystallization. The proposed system does not require the use of organic solvents, resulting in a relative increase in safety and cost-effectiveness. The system requires recycling of the mother liquor from the U-Pu co-crystallization step to recover almost the entire amount of U and Pu at this step. The appropriate recycling ratio is determined, such that satisfactory decontamination is achieved. A consistent ratio of Pu to U in the mother liquor from the U re-crystallization is maintained by regulating the temperature, suggesting that the quality of the liquor, which can be a source of mixed oxide fuels, can be controlled despite differences in the composition of the spent fuel. The size of a plant utilizing the proposed system is estimated to be about 30% less than that of the PUREX system.

A Study on Precipitation Behavior of Plutonium and Other Transuranium Elements with N-Cyclohexyl-2-pyrrolidone for Development of a Simple Reprocessing Process

Precipitation behavior of Pu and other transuranium elements with a combustible organic reagent Ncyclohexyl-2-pyrrolidone (NCP) has been examined to develop a simple reprocessing process of spent nuclear fuel based only on precipitation method. Experiments on the precipitation behavior of Pu in HNO3 solution containing only Pu showed that both Pu(VI) and Pu(IV) were precipitated with NCP, but they required more NCP than in the U(VI) precipitation. Selective U(VI) precipitation from HNO3 solution containing U(VI) and Pu(IV) (U(VI)-Pu(IV) solution) was achieved by stirring the solution for sufficient time after addition of NCP with ratio of [NCP]/[U] = 1.4. Addition of an enough amount of NCP to U(VI)-Pu(VI) or U(VI)-Pu(IV) solutions gave a quantitative precipitation of both U and Pu. From a viewpoint of the physical property of the precipitate, Pu(VI) was found to be favorable for the U-Pu precipitation. Hence, we have studied the oxidation of Pu(IV) to Pu(VI). As a result, heating of the solution was found to be a promising method for the present process. Furthermore, it was clarified that neither Am(III) nor Np(V) was precipitated in the selective U precipitation and the simultaneous U-Pu precipitation. These results demonstrate the feasibility of the reprocessing by precipitation with NCP.

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.

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.

Purification of uranium products in crystallization system for nuclear fuel reprocessing

Uranium crystallization system has been developed to establish an advanced aqueous reprocessing for fast breeder reactor (FBR) fuel cycle. In the crystallization system, most part of uranium in dissolved solution of spent FBR-MOX fuels is separated as uranyl nitrate hexahydrate (UNH) crystals by a cooling operation. The targets of U yield and decontamination factor (DF) on the crystallization system are decided from FBR cycle performance and plutonium enrichment management. The DF is lowered by involving liquid and solid impurities on and in the UNH crystals during crystallization. In order to achieve the DF performance (more than 100), we discuss the purification technology of UNH crystals using a Kureha Crystal Purifier (KCP). Results show that more than 90% of uranium in the feed crystals could be recovered as the purified crystals in all test conditions, and the DFs of solid and liquid impurities on the purified UNH crystals are more than 100 under longer residence time of crystals in the column of KCP device. The purification mechanism is mainly due to the repetition of sweating and recrystallization in the column under controlled temperature.

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

Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System

A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined by flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced.Copyright