Masanobu Nogami

Electrochemical Properties of Uranyl Ion in Ionic Liquids as Media for Pyrochemical Reprocessing

We have proposed a new reprocessing process by using ionic liquids (ILs) instead of molten salts of alkali chlorides in pyrochemical process. In the proposed process, spent nuclear fuels are dissolved in ILs by using Cl2 as an oxidant, and UO2 2+ and PuO2 2+ ions in ILs are recovered as UO2 and PuO2 by electrochemical reduction. In order to examine applicability of ILs as media for reprocessing, we have studied electrochemical behavior of UO2 2+ in 1-butyl-3-methylimidazolium chloride (BMICl), 1-butyl-3-methylimidazolium tetrafluoroborate (BMIBF4), and 1-butyl-3-methylimidazolium nonafluorobutanesulfonate (BMINfO). Electrochemical properties of uranyl chloride dissolved into ILs were examined by cyclic voltammetry. In BMICl, an almost reversible redox couple was observed, and the formal potential and the diffusion coefficient were evaluated as _0:758V vs. Ag/AgCl and 4:8 × 10−8 cm2s−1, respectively. On the other hand, the electrochemical reactions of UO2 2+ in BMIBF4 and BMINfO were irreversible. In BMINfO, some reduction peaks and one sharp oxidation peak were observed in the range of −0:6∼–0:2V and around 0.85V vs. Ag/AgCl, respectively. The reduction and oxidation peaks were assigned to multi step reduction of UO2 2+ to U(IV) via U(V) and/or direct reduction of UO2 2+ to U(IV), and the oxidative dissolution of the resulting U(IV) compounds, respectively. The electrochemical reduction of UO2 2+ in BMINfO at −1:0V vs. Ag/AgCl produced the deposits on a carbon electrode as a cathode. Analyses of the deposits with the scanning electron microscope and the energy dispersive X-ray spectrometer indicated that the deposits are compounds containing uranium, oxygen, and chlorine. As a result, it is expected that the UO2 2+ in IL can be recovered electrolytically as uranium compounds such as UO2 and uranium oxychlorides.

RADIATION RESISTANCE OF PYRIDINE TYPE ANION EXCHANGE RESINS FOR SPENT FUEL TREATMENT

The resistance against radiation of the tertiary pyridine resins synthesized for the treatment of spent nuclear fuels and high level radioactive waste was evaluated. After irradiation at 10 MGy, only approximately 10% or less of the exchange groups were lost in HCl solutions regardless of their concentrations, while 30∼40% were lost in HNO3. The pyridine resin has shown remarkable resistance against radiation particularly in HCl solution. It has been revealed that the decomposition of pyridine type resins results from the scission of the principal chains. An irradiation study was conducted also on the quaternary ammonium resins. Quatemization ratio was found to be reduced in HNO3 solutions at 10 MGy irradiation.

Ion-Exchange Selectivity of Tertiary Pyridine-Type Anion-Exchange Resin for Treatment of Spent Nuclear Fuels

The tertiary pyridine-type anion-exchange resin has been synthesized for the treatment of spent nuclear fuels and high-level radioactive waste. This resin, with a uniform diameter of 60 {micro}m, is mechanically strong enough and shows no swelling or shrinking regardless of its chemical forms. Systematic analysis was made of the adsorption selectivities of the resin in HCl solutions for a number of cations that exist in spent fuels, such as uranium and fission product elements. The results indicate that the resin is suitable to be used for the treatment of spent fuels and high-level radioactive waste.

Complex Formation of U(VI) with Benzamidoxime in Non-aqueous Solvents

The structural and kinetic studies of U(VI) complex with benzamidoxime (Hba) as ligand in CD3COCD3 have been studied by means of 1H and 13 C NMR. The Hba molecule was found to coordinate to UO2+ 2 in the form of anionic benzamidoximate (ba), and the number of ba coordinated to UO2+ 2 was determined to be 3 by analyzing the chemical shift of 13C NMR signal for Hba in the presence of UO2+ 2. The exchange rate constants (kex) of ba in [UO2(ba)3]- were determined by the NMR line-broadening method. The kinetic parameters were obtained as follows: kex (25°C)=1.3× 103 s-1, ΔH≠ =35.8±3.5kJ·mol-1, and ΔS≠=-65±13.7 J·K-1·mol-1. The UV-visible absorption spectra of solutions containing UO2+ 2 and Hba were also measured. The molar extinction coefficient of the complex was found to be extremely large compared with those of UO2(L)2+ 5 (L=unidentate oxygen donor ligands) complexes. This is due to the strong electron withdrawing of UO2+ 2 from Hba and suggests that an interaction between UO2+ 2 and Hba is very strong. Such a high affinity of monomeric amidoxime to UO2+ 2 reasonably explains the high adsorptivity of amidoxime resin to U(VI) species, and is considered to result in the high recovery of U(VI) species from sea water using amidoxime resin.

Anodic Dissolution of UO2 Pellet Containing Simulated Fission Products in Ammonium Carbonate Solution

We have proposed a reprocessing process based on the dissolution of spent fuels in aqueous carbonate solution. In order to develop such a reprocessing process, anodic dissolution experiments were carried out by using a simulated spent fuel pellet in (NH4)2CO3 solution. The dissolution rate constant of the simulated fuel pellet at 323 K was estimated as 3.1 × 10-6 mol·cm−2·min−1, which is comparable to that in 5 mol·dm−3 HNO3 solution containing 0.01 mol·dm−3 HNO2 at 323 K. The dissolution rate and the current efficiency for the dissolution of simulated fuel pellet in (NH4)2CO3 solution were found to decrease with the elapse of time. Furthermore, in order to examine dissolution behavior, the surface conditions of the pellet before and after the dissolution experiments were analyzed with the scanning electron microscopy and the energy dispersive X-ray spectroscopy. It was found that neat UO2 matrix exists on the dissolved surface and a decrease in the dissolution rate and the current efficiency is caused by the increase in such surface area on the pellet. During the dissolution experiments, precipitates of the simulated fission products were observed on the pellet and in (NH4)2CO3 solution used as the electrolytic solution. Analyses of the electrolytic solution revealed that most of the simulated fission products, i.e. alkaline earth and rare earth elements, are precipitated in high ratios. From these results, it is expected that the anodic dissolution of spent fuels and the separation of fission products by precipitation can be performed simultaneously.

Development of Advanced Reprocessing System Based on Precipitation Method Using Pyrrolidone Derivatives as Precipitants: —Precipitation Behavior of U(VI), Pu(IV), and Pu(VI) by Pyrrolidone Derivatives with Low Hydrophobicity—

An advanced reprocessing system for spent FBR fuels based on two precipitation processes has been proposed. In the first process, only U(VI) species is precipitated using a pyrrolidone derivative (NRP) with lower hydrophobicity and donicity, which should yield a lower precipitation ability. In the second process, residual U(VI) and Pu(IV, VI) are precipitated simultaneously using an NRP with higher hydrophobicity and donicity, which should yield a higher precipitation ability. In order to select the precipitants for the first precipitation process, we have examined the precipitation behavior of U(VI), Pu(IV), and Pu(VI) species in HNO3 using N-n-propyl-2-pyrrolidone (NProP), N-n-butyl-2-pyrrolidone (NBP), and N-isobutyl-2-pyrrolidone (NiBP) with lower hydrophobicity and donicity than N-cyclohexyl-2-pyrrolidone (NCP) previously proposed as the precipitant. It was found that NRPs could precipitate U(VI) nearly stoichiometrically and that the decontamination factors for simulated fission products were higher than those in NCP systems. Furthermore, as seen in NCP, it was found that in the U(VI)-Pu(IV) mixtures, a small amount of Pu(IV) was temporarily coprecipitated with U(VI) by NRPs in spite of their lower precipitation ability and then the coprecipitated Pu(IV) component was redissolved with continuous stirring. From these results, NRPs can be proposed as candidate precipitants for the first precipitation process. In particular, NBP is considered to be the most promising precipitant, because of the relatively high solubility of the NProP precipitant, the increases in viscosity of NiBP slurry with stirring, and the relatively fast sedimentation rate of NBP precipitates.

Effect of pore diameter of TMMA chelating resin beads on the adsorption properties of U(VI) and Ce(III) from different media

In order to investigate the relationship between the pore diameter and the rate of uptake of U(VI) ions and Ce(III) ions, N ,N ,N ?,N ?-tetramethylmalonamide (TMMA) chelating resins with different pore diameters were synthesized, using different pore producing solvents. The batch technique was used to study the kinetics of the adsorption U(VI) and Ce(III) ions from 3 M HNO3, acid, and 3 M NaNO3� /0.05 M HNO3, salt, media. The process was found to be controlled by the particle diffusion model. The calculated values for the diffusion coefficient were found to be directly proportional to the pore diameter, i.e. the resin with the largest pore diameter showed the fastest rate of uptake. # 2002 Elsevier Science B.V. All rights reserved.

Solubility of Uranyl Nitrate Precipitates with N-Alkyl-2-pyrrolidone Derivatives (Alkyl = n-Propyl, n-Butyl, iso-Butyl, and Cyclohexyl)

The solubility of UO2(NO3)2(NRP)2 (NRP = N-alkyl-2-pyrrolidone) in aqueous solutions with HNO3 (0–5.0 M) and the corresponding NRP (0–0.50M) has been studied. As a result, the solubility of each speciesof UO2(NO3)2(NRP)2 generally decreases with increasing concentrations of HNO3 and the corresponding NRP (C HNO3 and C NRP, respectively) in the supernatant. The solubility of UO2(NO3)2(NRP)2 also depends on the type of NRP; a higher hydrophobicity of NRP generally leads to a lower solubility of UO2(NO3)2(NRP)2. The logarithms of effective solubility products (K eff) of UO2(NO3)2(NProP)2, UO2(NO3)2(NBP)2, UO2(NO3)2(NiBP)2, and UO2(NO3)2(NCP)2 at different CHNO3 values and 293K were evaluated. For instance, at CHNO3 = 3:0 M, logK NProP eff = −1:07 ± 0:03, log K NBP eff = −2:23 ± 0:02, log K NiBP eff = −2:59 ± 0:03, and log K NCP eff = −3:80 ± 0:05. The solubility of UO2(NO3)2(NRP)2 is determined by the balance among the common-ligand effect, ionic strength, and variation of log K eff with C HNO3.

A study on selective precipitation ability of cyclic urea to U(VI) for developing reprocessing system based on precipitation method

We have proposed a new reprocessing system based on precipitation method. In order to find out precipitants with high selectivity to U(VI) and to investigate factors controlling precipitation ability to U(VI) and U(IV), properties of 3,4,5,6-tetrahydro-1,3-dimethyl-2(1H)-pyrimidinone (DMPU) as a precipitant have been examined by using U(VI), U(IV) as a simulant of Pu(IV), and simulated fission products (FPs). We have evaluated precipitation ratios (P.R.) for U(VI) and U(IV), solubility of U(VI) precipitates to 3.0 mol dm−3 (M) HNO3 solution, melting points (MPs) of U(VI) precipitates, log P (distribution ratio of a substance in 1-octanol/water biphasic system, a measure of hydrophobicity) of precipitants, and decontamination factors (DFs) of FPs. The properties of DMPU were compared with those in systems using N-n-butyl-2-pyrrolidone (NBP), N-cyclohexyl-2-pyrrolidone (NCP), and other pyrrolidone derivatives as the precipitant. The P.R. values of DMPU to U(VI) and U(IV) in 3.0 M HNO3 solutions were around 99% at [DMPU]/[U(VI)] = 2.0 and 0% at [DMPU]/[U(IV)] = 5.0, respectively. In DMPU system, the DF values of the most of simulated FPs [Rb(I), Cs(I), Sr(II), Ba(II), Ru(III), Rh(III), La(III), Ce(III), Pr(III), Nd(III), and Sm(III)] used in the present study were found to be more than 100. Even in U(VI)–U(IV) coexisting system, the selectivity of DMPU to U(VI) was higher than those of NBP and NCP. This selective precipitation ability of DMPU to U(VI) was evaluated by the solubility of U(VI) precipitates on the basis of their MPs and the log P values of precipitants. As a result, it was found that the precipitants having low hydrophobicity and forming the U(VI) precipitates with high MPs have highly selective precipitation ability to U(VI).

Extraction of Uranium(VI) from Nitric Acid Solution Using Pyrrolidone Derivatives

The extraction behavior of uranyl species from HNO3 aqueous solutions using CH2 Cl2 containing pyrrolidone derivatives as extractants has been studied to find an alternative to tributyl phosphate as an extractant. In this study, we used N-cyclohexyl-2-pyrrolidone (NCP), N-octyl-2-pyrrolidone (NOP), and N-dodecyl-2-pyrrolidone (NDP) as the extractants. As a result, it was found that NCP, NOP, and NDP (abbreviated as NRP) extract U(VI) as UO2(NO3)2(NRP)2, that a part of NCP in CH2 Cl2 phase is moved to the aqueous phase, and that the extractability (corresponding distribution ratio) of U(VI) increases with increasing [HNO3]int from 7% (0.08) ([HNO3]int = 0.1 M, [NRP]int = 1.0 M) to more than 90% (11) ([HNO3]int = 3.0 M, [NRP_int = 1.0 M). From these results, it is expected that NOP and NDP extract U(VI) effectively from HNO3 solution of more than 3M and that the extracted U(VI) species are stripped by using HNO3 aqueous solution of low concentration.