**1. Introduction**

Batch crystallization is widely used for the separation and high purification of organic and inorganic materials in the fine chemical, food, pharmaceutical and biochemical industries. In the atomic power industry, application of crystallization to U purification of the Plutonium Uranium Reduction Extraction (PUREX) first cycle product was attempted in Kernforschungszentrum Karlsruhe (KfK), Germany (Ebert et al., 1989). The feed solution had 240−480 g/dm3 U concentration and 0.1 g/dm3 fission products (FPs) concentration in 5−6 mol/dm3 HNO3 solution. Reducing conditions were achieved with 2.4 g/dm3 of U(IV) which was added to change the Pu valence to Pu(IV) which was required for good separation of Pu from U. In a six-stage cascade crystallizer, the feed solution was cooled down in steps from 30 to −30°C in the course of about 30 min. More than 90% of U was recovered in form of uranyl nitrate hexahydrate (UNH) crystals with an average diameter of 0.2 mm, while a much greater proportion of the transuranium (TRU) elements and FPs remained in the mother liquor. The decontamination factors (DFs) of several of the FPs were determined for one crystal step plus several crystal washing operations. The measured DFs of Pu and Cs were 102 and 103, respectively.

An advanced aqueous reprocessing for a fast neutron reactor fuel cycle named "New Extraction System for TRU Recovery (NEXT)" has been proposed as one fast neutron reactor fuel reprocessing method (Koyama et al., 2009) and is being developed in Japan Atomic Energy Agency (JAEA). On the advanced aqueous reprocessing for fast neutron reactor fuel cycle, it is supposed to recover not only U and Pu but also minor actinides (MAs; Np, Am and Cm) for the efficient utilization of resources. It will be also effective in decreasing the environmental impact because of their long half-life and high radiotoxicity. These elements are loaded in a fast neutron reactor and are burned as core fuel. Figure 1 shows schematic diagram of NEXT process for fast neutron fuel reprocessing. The NEXT consists of highly efficient dissolution of fuel with HNO3 solution (Katsurai et al., 2009), U crystallization for partial U recovery (Shibata et al., 2009), simplified solvent extraction for U, Pu and Np co-recovery using tri-*n*-butyl phosphate (TBP) as an extractant (Sano et al., 2009), and extraction chromatography for mutual separation of actinide elements and lanthanide elements from a raffinate (Koma et al., 2009). The powdered fuel was dissolved

Separation of Uranyl Nitrate Hexahydrate Crystal

**2. Principal of uranium crystallization** 

to left as the HNO3 concentration increases.

Pu is crystallized on the UNH crystal.

paper.

process.

from Dissolver Solution of Irradiated Fast Neutron Reactor Fuel 385

solution. In this study, the feed solution was changed in HNO3 concentration and the influence on the UNH crystal ratio was examined in the cooling batch crystallization. Two experiments, crystal ratio and the co-existing element behavior, were carried out with a dissolver solution derived from irradiated fast neutron reactor "JOYO" core fuel in a hot cell of the Chemical Processing Facility (CPF), JAEA. Additionally, current status of crystallization apparatus and crystal purification method for the NEXT is described in this

In a HNO3 solution, U ions are crystallized as UNH by the following reaction.

( ) <sup>2</sup>

Figure 2 shows the solubility curves of U in HNO3 solution (Hart & Morris, 1958). The results represent the mean of two temperatures observed for the first formation and final disappearance of crystal on, respectively, slowly cooling and warming solutions with vigorous agitation. The U ions concentration decreases with decreasing temperature in the solution before reaching the eutectic point, where H2O and HNO3 start to crystallize. Thus, U crystallization process should be performed in the right region of the minimum point in this figure. A high HNO3 solution is desirable for achieving a low U concentration in the solution, therefore yielding more UNH crystals because the eutectic point shifts from right

In Pu(NO3)4-HNO3-H2O system, the crystallization behavior of plutonyl nitrate hexahydrate (PuNH) was examined. Figure 3 shows the solubility curves of Pu in HNO3 solution (Yano et al., 2004). The Pu solution was prepared by dissolving PuO2 powder with 4 mol/dm3 HNO3 solution containing 0.05 mol/dm3 AgNO3 electrochemically. In the experiments, the Pu valence was adjusted as following methods. The valence of Pu was changed to Pu(IV) with a few drops of 100% H2O2. On the other hand, the Pu solution was oxidized to Pu(VI) by Ag2+ ion and Ag in the solution was separated by ion exchange. The Pu solution was cooled quickly to −20°C and then cooled at −1 °C/min to −55°C. In the Pu(IV) solution appeared to be a green quasi-liquid (crystals in liquid). In all runs, PuNH was not crystallized in the experimental conditions but crystals of H2O and HNO3·3H2O were observed. In the NEXT, PuNH would not precipitate solely in the U crystallization

The influence of Pu valence in the feed solution was examined in the U crystallization process (Yano et al., 2004). When Pu(IV) existed in the feed solution, the yellow crystal was observed. On the other hand, the appearance of the crystal was orange in the feed solution adjusted so that Pu valence was Pu(VI), this color likely resulting from the mixture of the yellow crystal of UNH and the red crystal of PuNH. Plutonium(VI) in the feed solution was co-crystallized with U(VI) in the course of U crystallization. The crystal yields of Pu were smaller than those of U (Ohyama et al., 2005). The fact that the crystal ratio of Pu is smaller than that of U suggests a mechanism of U-Pu co-crystallization in which U begins to crystallize when the saturation point of U is reached by cooling the feed solution, and then

2 3 2 23 2 <sup>2</sup> UO 2NO 6H O UO NO 6H O + − ++↔ ⋅ (1)

by the highly efficient dissolution process and the dissolver solution was adjusted to high heavy metal concentration. Then, U is recovered as UNH crystals from dissolver solution derived from fast neutron reactor fuel. Since the amount throughput will be reduced in the simplified solvent extraction process, the adoption of the crystallization process is expected to reduce the radioactive waste, equipment, and hot cell volume. In addition, U/Pu ratio in the dissolver solution is adjusted in the crystallization process to be a suitable Pu content for core fuel fabrication. In the NEXT, Np is changed to Np(VI) in the high HNO3 concentration feed solution and is co-extracted with U and Pu in the simplified solvent extraction system. The FPs in the raffinate obtained from the simplified solvent extraction process is removed using *N,N,N',N'*-tetraoctyl-3-oxapentane-1,5 diamide (TODGA) absorbent in the extraction chromatography I. The actinide elements such as Am and Cm is recovered from the solution containing actinide and lanthanide elements by chromatography with 2,6-bis-(5,6-dialkyl-1,2,4-triazine-3-yl)pyridine (R-BTP) absorbent in the extraction chromatography II.

Fig. 1. Schematic diagram of the NEXT process

A dissolver solution of irradiated fast neutron reactor mixed oxide (MOX) fuel in JAEA contains a number of TRU elements and FPs than in KfK. Since U is used as blanket fuel and TRU elements are supposed to recover by other chemical process, it is need to remove TRU elements and FPs from UNH crystals in the U crystallization process. It would be also bring about reduction in the cost for the recovered U storage and the blanket fuel fabrication due to decreased radiation shielding. Therefore, the behavior of TRU elements and FPs in the U crystallization process must be confirmed experimentally.

Since U is recovered as UNH crystal for a blanket fuel fabrication in the U crystallization process, the crystal ratio of U should be evaluated with a dissolver solution of irradiated fast neutron reactor. The crystal ratio of UNH affects HNO3 concentration in the feed solution. In this study, the feed solution was changed in HNO3 concentration and the influence on the UNH crystal ratio was examined in the cooling batch crystallization. Two experiments, crystal ratio and the co-existing element behavior, were carried out with a dissolver solution derived from irradiated fast neutron reactor "JOYO" core fuel in a hot cell of the Chemical Processing Facility (CPF), JAEA. Additionally, current status of crystallization apparatus and crystal purification method for the NEXT is described in this paper.
