Dissolution Behavior of Uranium Oxides with Supercritical CO2 Using HNO3-TBP Complex as a Reactant

Dissolution behavior of U₃O₈ and UO₂ using supercritical CO₂ medium containing HNO₃-TBP complex as a reactant was studied. The dissolution rate of the oxides increased with increasing the HNO₃/TBP ratio of the HNO₃-TBP complex and the concentration of the HNO₃-TBP complex in the supercritical CO₂ phase. A remarkable increase of the dissolution rate was observed in the dissolution of U₃O₈ when the HNO₃/TBP ratio of the reactant was higher than ca. 1, which indicates that the 2:1 complex, (HNO₃)₂TBP, plays a role in facilitating the dissolution of the oxides. Half-dissolution time (t_½ ) as an indication of the dissolution kinetic was determined from the relationship between the amount of uranium dissolved and the dissolution time (dissolution curve). A logarithmic value of a reciprocal of the t_½ was proportional to the logarithmic concentration of HNO₃, C_HNO3, in the supercritical CO₂. The slopes of the (l/t_½ ) vs. ln C_HNO3 plots for U₃O₈ and UO₂ were different from each other, indicating that the reaction mechanisms or the rate-determining steps for the dissolution of U₃O₈ and UO₂ are different. A principle of the dissolution of uranium oxides with the supercritical CO₂ medium is applicable to a method for the removal of uranium from solid matrices.     

Structural and Electrochemical Studies on Uranium(VI) Nitrato Complex with n-Octyl(phenyl)-N,N-Diisobutylcarbamoylmethylphosphine Oxide in Non-aqueous Solvents

The structure of uranyl nitrato complex with CMPO [n-Octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide] in solid state and in non-aqueous solvents without containing free CR/IPO has been studied by using IR spectrophotometer, ¹³C- and ³¹P-NMR. The carbonyl(vcO) and phosphoryl(vpO) stretching bands of coordinated CMPO were observed at lower wavenumber than the corresponding bands of free CMPO in both the states. The ¹³C and ³¹P peaks assigned to the carbonyl carbon and phosphoryl phosphine of coordinated CMPO was detected in the lower field than that of free CMPO. From these results, it was concluded that the uranyl nitrato complex with CMPO in both the states has the structure with two nitrate and one CMPO coordinated as bidentate in the equatorial plane of uranyl ion, i.e., UO₂(NO₃)₂·CMPO. Furthermore, the electrochemical studies of UO₂(NO₃)₂·CMPO complex in CH₃CN have been carried out using cyclic and normal pulse voltammetric methods. It was found that the UO₂(NO₃)₂·CMPO complex is reduced to U(V) complex at around ?1.22V vs. Fc/Fc⁺ (ferrocene/ferrocenium) and that the resulting reductant is oxidized to U(VI) at around +0.04V vs. Fc/Fc⁺.     

Extraction of Thorium and Uranium from Chloride Solutions by Tri-n-Butyl Phosphate and Tri-n-Octyl Phosphine Oxide

The extraction of thorium and uranium chlorides by TBP and TOPO was studied. The composition of complexes extracted from the chloride solutions of low acid concentration was established by partition study to be UO₂Cl₂ (TOPO)₂, UO₂Cl₄ (TOPO)₂, UO₂Cl₄ (TOPO)₂ and UCl₄ (TBP)₂. Composition of the thorium complex in the TBP phase free from hydrochloric acid was revealed by infrared study to be ThCl₄ (TBP)4. The extraction behavior of thorium chloride by TBP was different from that of U (N) and Pu(N) chloride, and the composition of the complex was presumed to be HThCl₅(TBP)₄ in the extraction from concentrated chloride solution containing hydrochloric acid.     

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 C_HNO3 values and 293K were evaluated. For instance, at C_HNO3 = 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.     

Precise determination of precipitation yield of uranyl nitrate with N-alkylated 2-pyrrolidone derivatives

Yields of precipitated UO₂(NO₃)₂(NRP)₂ (NRP = N-alkylated 2-pyrrolidone) were precisely determined by considering reduction of the solution volume through the precipitation, which can be estimated from difference in acid concentrations of the liquid phases before and after the precipitation. The studied NRPs were N-n-butyl (NBP) and N-n-propyl (NProP) derivatives. In both systems, the precipitation yields precisely determined were always higher than those simply calculated from the ratio of uranium concentrations before and after the precipitation. However, the differences between them are in the range of 0.6%–2.6%. If such a difference is practically negligible, the volume reduction through the precipitation does not have to be taken into account for simplicity of the analytical manipulation.     

Sorption of U(VI) on the 4-Mercaptopyridine Self-Assembled Monolayer

The sorption of U(VI) on the 4-mercaptopyridine self-assembled monolayer (4-PyS-SAM) on Au(111) was studied by cyclic voltammetry. Cyclic voltammograms (CVs) of the 4-PyS-SAM working electrode were obtained by contact with 1mM UO₂(NO₃)₂ solution, 1mM UO₂(NO₃)₂ and 50mM acetic acid solution, or 1mM UO₂(NO₃)₂ and 50mM oxalic acid solution for 6 h at pH 4. The reduction current of U(VI) to U(V) was detected in the CV. The CV of the U(VI) associated 4-PyS-SAM after transport to U(VI)-free 0.1M NaClO₄ solution showed that the reduction current was detected in the cases of UO₂(NO₃)₂ and U(VI)-acetate, but not in the case of U(VI)-oxalate solution, indicating that U(VI) was adsorbed on the 4-PyS-SAM from the UO₂(NO₃)₂ and U(VI)-acetate solutions, but not from U(VI)-oxalate solution. These results suggest that stability of U(VI)-4-PyS-SAM is not so high that U(VI)-4-PyS-SAM cannot be formed in the presence of 50mM oxalate.     

“The Third Phase” of Extraction Processes in Fuel Reprocessing, (I)

Formation conditions, compositions and structures of precipitates or “the third phase” were systematically examined in the systems of Zr and radiation depleted products of TBP, such as HDBP, H₂MBP, H₃PO₄, by means of elemental analysis, X-ray diffraction, infrared spectra and ¹H-NMR. It was confirmed that one of the most important origins for the third phase is a complexation between Zr ion and the depleted products of TBP. Followings were also elucidated:

(1) When the depleted products co-exist with each other, the cooperative effects on the precipitate formation appear in low acid solutions.

(2) Precipitate formation depends on the mole ratio of HDBP/Zr. The amount of precipitate reaches the maximum at the mole ratio of approximately 2 and decreases with increasing concentration of HDBP and finally disappears at approximately 10.

(3) Precipitate formed at the mole ratio of approximately 2 has the chemical formula, Zr (NO₃)₂ (HDBP)₂ (OH)₂.

(4) Precipitates of the Zr-H₂MBP system begin to appear at the concentration of H₂MBP in one order of magnitude smaller than that of HDBP in Zr-HDBP system.

(5) Precipitates of Zr-H₂MBP system have no NO₃ ion and a basic structure of Zr-(HMBP)₂(OH)₂ with an interlayer distance of 16 Å.     

Treatment of UO2 pellets used for preparing fuel elements of HTR-10 followed by supercritical fluid extraction

International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO₂ pellets used for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) followed by supercritical fluid extraction was investigated. When UO₂ pellets were dissolved and extracted with tri-n-butyl phosphate (TBP)–HNO₃ complex in supercritical CO₂ (SC-CO₂), the extraction efficiency was less than 7% under experimental conditions. After UO₂ pellets were ground into UO₂ fine powders, the extraction efficiency of the UO₂ fine powders with TBP–HNO₃ complex in SC-CO₂ could reach 92%. After UO₂ pellets broke spontaneously into U₃O₈ powders under O₂ flow and 600 °C, the extraction efficiency of the U₃O₈ powder with TBP–HNO₃ complex in SC-CO₂ could reach more than 98%.     

Neutron Diffraction Studies of NaUO3

Neutron diffraction studies have been made on a powdered NaUO₃ sample. It has been shown that an octahedron composed of six oxygen atoms around pentavalent uranium in this compound is distorted and four of six oxygen atoms are much closer to an uranium atom (U-O distance of 2.15Å) than the other two (U-O distance of 2.24Å). A unit of the UO₂ ⁺ ion does not exist in the crystal, though the pentavalent uranium has been expected to be in the form of UO₂ ⁺ as in the case of UO₂ ²⁺ of a hexavalent uranium Ion. A neutron diffraction pattern at temperature of liquid helium has shown no appreciable difference from that at temperature of liquid nitrogen.     

Thermal Conductivities of Granular UO2 Compacts with/without Uranium Particles

The thermal conductivities of granular UO₂ compacts with and without uranium particles were measured to evaluate the thermal performance of vibro-packed granular MOX fuels containing metallic fine particle oxygen getters. The thermal conductivity of the compact with 10 wt% of uranium particles was higher than that of the compact without uranium particles. After heating beyond 1,408 K, the melting point of the uranium particles, the thermal conductivity increased further. The evaluation model for analyzing such phenomena was developed. The model predicted that once the UO₂ compact with uranium particles was exposed to a temperature beyond 1,408 K, the uranium particles should melt and provide interconnecting areas between the UO₂ granules and uranium particles, and between other uranium particles. The resulting increase of the thermal conductivity was reasonably expressed by the effect of necks in the compact on the heat conduction.     

A study of the generation of 232U in UO2 and MOX fuels

To clarify the generation pathway of ²³²U, an important nuclide for dose evaluation at various stages in the reuse of uranium, concentrations of ²³²U generated through various pathways were evaluated for UO₂ and mixed oxide (MOX) fuels. Burnup calculation was conducted with ORIGEN2.2 code adopting ORLIBJ40 library, a set of cross-section libraries based on JENDL-4.0. It was found that differences in ²³²U concentrations in UO₂ and MOX fuels mainly arise from differences in the initial compositions of ²³⁴U, ²³⁵U, and ²³⁶U. It was also found that the contribution of plutonium and americium isotopes in MOX fuels is small compared with that of uranium isotopes. The results clarified that the capture cross sections of ²³⁰Th, ²³¹Pa, ²³⁵U, and ²³⁶U, as well as the (n,2n) cross sections of ²³⁷Np and ²³⁸U, have a large effect on the generation of ²³²U. Additional investigation showed that ²³²U concentration is strongly affected not only by time after irradiation but also by time before irradiation.     

Effects of UO2/Zircaloy-4 Mole Ratios on Reaction of UO2 Fuel with Molten Zircaloy-4

Dissolution of UO₂ crucibles by molten Zircaloy-4 (Zry) was investigated in the temperature range of 2,223-2,373 K and for specimens having UO₂/Zry mole ratios between 7 and 18.2. The uranium concentration in the Zry melt rapidly increased during initial reaction time and approached saturated values, depending on reaction temperature and UO₂/Zry mole ratio. Kinetics of uranium concentration increase in the melt was analyzed based on a natural convection mass transfer model that takes into account the change of contact surface area/melt volume ratio with reaction time. The saturated uranium concentration in the Zry melt was inversely proportional to the U0₂/Zry mole ratio. An empirical correlation of saturated uranium concentration in the Zry melt was obtained as a function of UO₂/Zry mole ratios and reaction temperature. This study of the empirical correlation was intended to estimate maximum UO₂ fuel dissolution by molten Zry cladding during severe fuel damage accidents for three different reactor type fuels.     

Extraction properties of diisoamyl methylphosphinite

The properties of one of the new, efficient extractants for uranium, diisoamyl methylphosphinate (DAMP), are described. It was shown that uranyl nitrate is extracted from nitric acid solutions by this extractant with considerably higher distribution coefficients than tributyl phosphate (TBP). The extraction of uranyl nitrate and HNO³ with DAMP solutions in hydrogenated kerosene was studied. It was shown that uranyl nitrate is extracted in the form of the complex [UO₂(NO₃)₂ (DAMP)₂], whose stability constant equals 2540±200, and HNO₃ is extracted in the form of the compound HNO₃ DAMP, whose stability constant is 0.30±0.03.     

Oxidation of Uranium Dioxide

The oxidation of UO₂ was studied by thermogravimetry and X-ray diffraction. It was clarified that the thermal history covering the first stage of the oxidation from UO₂ to U₃O UO₇ significantly influenced the rate of the oxidation of the second stage from U₃O₇ to U₃ O₈.

The entire oxidation reaction proceeded in what to all appearances, was a single stage when the specimen temperature was raised rapidly, whereas at slower rates of heating up, two distinct stages of oxidation were observed, separated by an intermediate induction period. These findings suggest the existence of a close connection between the rate of formation of the U₃O₇ phase and the rate of the subsequent oxidation of this phase: A slow formation of U₃O₇ would tend to prolong the induction period preceding the second stage of the oxidation. A similar effect was observed also with annealing of the intermediate U₃O₇ at 200°C: The increase of annealing time prolonged the induction stage.

The rate of the second stage oxidation was fairly well expressed by Johnson & Mehl's equation, log (1/(1-y/)=(1/2.303)kⁿtⁿ . The time exponent n in this equation varied in the range of 1.0~2.5, and the rate constant k of 1.15×10^?4~2.04 ×10^?1 min^?1, depending on the experimental conditions.     

Search for cold fusion using Pd-D2O cells and Ti-D mixtures

We have searched for cold fusion produced in an electrolytic cell with Pd cathode and Pt anode. The electrolyte was 0.1 molar LiOD in 99.8% D₂O. A 2-mm rod of polycrystalline Pd and a 4-mm rod of single crystal Pd were used. No radiation was detected above background by a BF₃ neutron and Ge -X detector. The D₂ loading of the Pd was 0.8 D per Pd atom reaching saturation after 4 hours. We also attempted to duplicate the work of Scaramuzzi and co-workers on the Ti-D₂ system. Both powder and pieces of Ti were used. The material was cycled several times between 1100 K and 77 K. No neutron, - or x-ray emission above background was observed. The results of a barrier penetration calculation for H-like atoms are presented. The high fusion rates reported for PdD _x . are much larger than those expected from theoretical calculations on these systems.This work supported by USDOE under Contract nos. W-7405-ENG-82 and DE-FG02-87ER40371.     

Improvement of UO2 Pellet Properties by Controlling the Powder Morphology of Recycled U3O8 Powder

Powder morphology evolution of recycled U₃O₈ according to the thermal treatments has been studied. The defective UO₂ pellets are oxidized to U₃O₈ powders at a conventional temperature of 350 or 450°C in air. Those powders are pressed into green pellets and then sintered at 1,500 and 1,730°C in H₂ gas flow. Final reoxidized U₃O₈ powers are obtained by reoxidizing those sintered pellets at 450°C in air. This paper shows that the reoxidized U₃O₈ powder morphology and the BET surface areas are greatly dependent on the density of sintered UO₂ pellets before reoxidation. Reoxidized U₃O₈ powders are added to virgin UO₂ powders to fabricate UO₂ pellets and the effect of such addition on the UO₂ pellet properties is investigated. The reoxidized U₃O₈ powders having a certain range of BET surface area significantly promote the grain growth of UO₂ pellets.     

Critical Experiments on Light-Water Moderated PuO2-UO2 Lattices

A series of critical experiments using mixed-oxide (PuO₂-UO₂) plutonium fuels was carried out at the Tank type Critical Assembly (TCA) in the Japan Atomic Energy Research Institute in cooperation with the Power Reactor and Nuclear Fuel Development Corporation. Isotopie compositions of the Pu were 68, 22, 7 and 2w/0 for ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu and ²⁴²Pu, respectively, and the U was natural uranium.

Critical mass, power distribution and neutron activation distribution of Au wire were measured for five kinds of lattices in which the ratios of atomic density of H to Pu were 295, 402, 494, 704 and 922, respectively. The codes LASER, UGMG42-THERMOS and GTB-2 were used for obtaining three neutron group constants and the PDQ-5 code was used in two-dimensional diffusion calculations for the lattices.

The maximum differences of the calculated multiplication factors from the measured values distributed from –1.32 to 1.72%δk/k. Calculated thermal neutron density, epithermal neutron flux and power distributions showed good agreement with measured data.     

Alkali Metal Ion and Lithium Isotope Selectivity of HZr2(PO4)3

HZr₂(PO₄)₃ has been synthesized by the heat treatment of NH₄Zr₂(PO₄)₃ and its properties as an ion exchanger have been examined with the main focus on its alkali metal ion and lithium isotope selectivity. The distribution coefficients for alkali metal ions revealed that HZr₂(PO₄)₃ was lithium ion-specific and showed little affinity toward potassium, rubidium or cesium ion. The lithium and sodium ion uptakes from aqueous solutions were monotonously increasing functions of pH. Isotopically, HZr₂(PO₄)₃ was ⁶Li-specific. Contrary to ion uptake, the lithium isotope effect was a monotonously decreasing function of pH; a larger separation factor was observed at a lower pH. This result was consistent with the existence of two different ion exchange sites formed in lithium ion-inserted HZr₂(PO₄)₃.     

Pyrohydrolysis Reactions of UF4 and UO2F2

This study was made for clarification of the pyrohydrolysis behaviors of UF₄ and UO₂F₂. The progress of the reaction was measured by titrating the amount of produced hydrogen fluoride. When nitrogen was used as carrier gas of water, the pyrohydrolysis of UF₄ proceeded at the temperatures of 350–400°C, and UO_2+x (x?0.3) was formed; the value of x decreased with the decrease of oxygen dissolved in the water. In the case of oxygen carrier gas, the pyrohydrolysis of UF₄ formed U₃O₈-UO₂F₂ mixture as the reaction product in the above temperature region and the reaction was markedly retarded in the course because of lower rate in the pyrohydrolysis of UO₂F₂. The pyrohydrolysis of UO₂F₂ proceeded at a little [higher temperature of 450–500°C in both cases of nitrogen and oxygen carrier gas and α-UO₃ was formed.     

Evaluation of UF6 Vapor Release in Postulated Accident

The rupture of UF₆ gas line connected to hot UF₆ cylinder, being one of various accidents in UF₆ vapor leak-out, is considered as a postulated accident for uranium enrichment plants. For this type of rupture, we will estimate the amount of UF₆ vapor release based on a simplified calculation model and then make an evaluation of UF₆ vapor release through a ventilation system of feed vaporization facility. Assuming an instantaneous steady state for the change of UF₆ states, an unsteady state thermodynamics process is solved. Numerical examples show that about 52% of the initial UF₆ quantity are vaporized at 80°C (the temperature of the liquid UF₆ in the cylinder). Furthermore, by using the amount of released UF₆ vapor and the collection capacity of HEPA filter for IiF gas, the amount of gaseous UO₂F₂, HF which may be dissipated to the environment are conservatively estimated.