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.     

Effect of the preparation conditions of zirconium phosphate on the characteristics of Sr immobilization

The proton-type crystalline zirconium phosphate, HZr₂(PO₄)₃, was obtained by a thermal decomposition of NH₄Zr₂(PO₄)₃ at different temperatures from 400 to 800 °C, where NH₄Zr₂(PO₄)₃ was obtained in advance by a hydrothermal synthesis using a mixed solution of ZrOCl₂, H₃PO₄ and H₂C₂O₄ with different processing times from 5 to 72 h. Sr ion was immobilized to HZr₂(PO₄)₃ by treating the mixture of HZr₂(PO₄)₃ and Sr(NO₃)₂ aqueous solution in an autoclave at 250 °C. Immobilizing and leaching performance of St in HZr₂(PO₄)₃ were discussed.     

Immobilization of Cs and Sr to HZr2(PO4)3 using an autoclave

The proton-type crystalline zirconium phosphate, HZr₂(PO₄)₃, was prepared by a thermal decomposition of NH₄Zr₂(PO₄)₃ at about 450 °C, where NH₄Zr₂(PO₄)₃ was obtained in advance by a hydrothermal synthesis using a mixed solution of ZrOCl₂, H₃PO₄ and H₂C₂O₄. Cs or Sr ion was immobilized to HZr₂(PO₄)₃ by mixing HZr₂(PO₄)₃ with an aqueous solution of CsNO₃ or Sr(NO₃)₂ under the molar ratio CsNO₃/HZr₂(PO₄)₃ = 1.0 or Sr(NO₃)₂/HZr₂(PO₄)₃ = 0.5. The mixtures were treated thermally in an autoclave at different temperatures from 200 to 275 °C and Arrhenius equation was applied to the Cs and Sr immobilization process to HZr₂(PO₄)₃. The activation energy for the immobilization process of Cs or Sr was estimated as 179 kJ mol^?1 and 186 kJ mol^?1, respectively.     

Kinetics and Equilibrium of Reaction between Octakis (dimethyl sulfoxide) uranium(IV) and Nitrite Ion in Dimethyl Sulfoxide

The kinetics and equilibrium of the reaction between octakis (dimethyl sulfoxide) uranium (IV), U(dmso)₈ ⁴⁺, and nitrite ion N0₂ ^? were studied in dmso solutions by using stopped-flow, diode-array and UV-visible spectrophotometers. Three different reactions were observed under the condition of an excess NaNO₂, and U(IV) was finally oxidized to U(VI) by N0₂ ^?. The first reaction is the formation of 1: 2 adduct complex [U (dmso₈)(NO₂)₂ ²⁺), the rate of which was so fast that the rate constant k_l obs was not determined accurately even by the stopped-flow method. If the concentration of N0₂ ^? was as low as that of U(IV), 1:1 adduct complex was formed. However, 1:1 complex was inert for further redox process, i.e. U(IV) was not oxidized by N0₂ ^? under this condition. The second reaction represents the substitution of coordinated dmso with N0₂ ^? forming mixed ligand complex [U(dmso) _8-x(N0₂ ^?)_x+2(^{2-x +}) The rate constant k ₂0bs for the second reaction was measured by the conventional method using a diode-array spectrophotometer and k ₂0bs increased linearly with the concentration of free N0₂ ^?. The third reaction coincides with the intra-molecular oxygen transfer from coordinated N0₂ ^? to U(IV) and this process is accompanied by the oxidation of U(IV) to U(VI). The oxygen transfer mechanism was confirmed by ¹⁷O labeled experiments using ¹⁷O-NMR. It was found from acid dependence experiments that only nitrite ion oxidized U(IV) and that nitrous acid was inert as an oxidant for U(IV).     

Dissolution behavior of (U,Zr)O2-based simulated fuel debris in nitric acid

To explore the possibility of dissolving fuel debris into nitric acid as a potential pre-treatment for waste treatment in which the U and Pu are removed from the inventory, dissolution tests of U_1?xZr_xO₂ and (U,Pu)_1?xZr_xO₂ were carried out in 6 M HNO₃ at 353 K. At the end of the dissolution test (after 4 h), the ratio of dissolved uranium decreased with an increase in the Zr contents, x. While the dissolution of U-rich samples was congruent, a preferential leaching of U was observed with Zr-rich samples. Taking into account these different dissolution phenomena, the dissolution rate analysis was carried out using surface-area model to calculate the instantaneous dissolution rate (IDR). The IDR decreased from 10^?5 down to 10^?10 mol cm^?2 min^?1 as x increased from 0 to 0.95. From these findings, dissolution with HNO₃ is expected to be only applicable in U-rich part of fuel debris (x < 0.3) if the dissolution in 6 M HNO₃ at 353 K is assumed. Application of complexing acids, such as mixture of HNO₃ and HF, should be considered to increase the dissolution rate of the Zr-rich part.     

Light ion irradiation effects on stuffed Lu2(Ti2−xLux)O7−x/2 (x = 0, 0.4 and 0.67) structures

We have recently synthesized “stuffed” (i.e., excess Lu) Lu₂(Ti_2−xLu_x)O_7−x/2 (x = 0, 0.4 and 0.67) compounds using conventional ceramic processing. X-ray diffraction measurements indicate that stuffing more Lu³⁺ cations into the oxide structure leads eventually to an order-to-disorder (O-D) transition, from an ordered pyrochlore to a disordered fluorite crystal structure. At the maximum deviation in stoichiometry (x = 0.67), the Lu³⁺ and Ti⁴⁺ ions become completely randomized on the cation sublattices, and the oxygen “vacancies” are randomized on the anion sublattice. Samples were irradiated with 400 keV Ne²⁺ ions to fluences ranging from 1 × 10¹⁵ to 1 × 10¹⁶ ions/cm² at cryogenic temperatures (∼77 K). Ion irradiation effects in these samples were examined by using grazing incident X-ray diffraction. The results show that the ion irradiation tolerance increases with disordering extent in the non-stoichiometric Lu₂(Ti_2−xLu_x)O_7−x/2.     

Sorption and Colloidal Behavior of Np(IV) in a Bentonite-Carbonate Solution System

Abstract

Sorption of Np(IV) on bentonite in the presence of carbonate was studied by a batch method. The sorption was considered mainly due to Np(OH)₂(CO₃)₂ ^2- and Np(OH)₄(CO₃)₂ ^4- The distribution coefficient (K_d of Np(OH)₂(CO₃)₂ ^2- and Np(OH)₄ (CO₃)₂ ^4- was determined to be 10^4.01±0.24 ml/g and 10^3.93±0.37 ml/g respectively. The increasing K_d at high pH was attributed to the sorption of NpO₂(OH)_-2 with the K_d of 10^5.82±0.39 ml/g

The formation of colloidal species was observed under a reducing condition. By assuming Np(OH)₄ as the colloidal species, the formation constant was evaluated to be β_c≤10^53.85 Under an oxidizing condition, the formation of the colloidal species was less observed possibly due to the transformation of Np(OH)₄ to NpO₂OH.     

Al2O3对独居石玻璃陶瓷固化体的影响

本文研究了Al₂O₃掺量对独居石玻璃陶瓷固化体结构和化学稳定性的影响。用傅里叶变换红外光谱（FTIR）和X射线衍射（XRD）方法表征样品结构,用溶解速率法和全谱直读等离子体发射光谱（ICP-OES）分别测定样品在浸出液中浸泡后的失重速率及各元素的浸出浓度,以研究固化体的化学稳定性。研究结果表明：当Al₂O₃掺量为4%（摩尔分数）时,在980 ℃下保温3 h得到的独居石玻璃陶瓷固化体具有较高的化学稳定性,浸泡14 d时其质量浸出率最低,约为8.1 ng/(cm²•min),其中Ce、La元素在浸出液中均未检出;固化体的主晶相为独居石,结构中含有大量稳定的正磷酸基团［PO₄］^3-和少量的焦磷酸基团［P₂O₇］^4-,不存在偏磷酸基团［PO₃］^-。     

Precipitation of Thorium or Uranium(VI) Complex Ion with Cobalt(III) or Chromium(III) Complex Cation, (I)

The precipitation of uranium(VI) peroxo complex ion with both tris (ethylenediamine) cobalt (III) and tris(trimethylenediamine) cobalt(III) salts has been studied.

The precipitates obtained immediately are, respectively, [Co(en)₃]₄ [(UO₂)₂ (O₂)₄]₃nH₂O and [Co(tn) ₃]₄ [(UO₂)₂ (O₂)₃]₃nH₂O, which change to [CO(en) ₃]₄ [(UO₂)₂ (O₂)₂ (OH) ₄]₃nH₂O and [CO(tn) ₃]₄ [(UO₃)₂ (O₂)₂ (OH) ₄]₃nH₂O with time.

Carbonate ion affects the precipitation reactions by forming stable outer-sphere complex ion with cobalt (III) complex cation.     

Hydrous Titanium Oxide Ion Exchanger, (I)

Abstract

Synthesis of hydrous titanium oxide ion exchanger was tried from three systems: (a) TiCl₄-NaOH-H₂O, (b) ATS (ammonium titanyl sulphate monohydrate)-NaOH-H₂O and (c) TiCl₄-NH₄OH-H₂O. The first method gave the best results under the conditions covered in the present work.

It was found that repeated washing and aging of the hydrous titanium oxide precipitate were indispensable in order to obtain reproducible results, and that this operation further obviated the need of precisely adjusting the conditions for mixing the reagents.

Both the yield of exchanger and the ion exchange capacity of the resulting product increase with concentration of the reagents. A cation exchange capacity of about 3 meq·Na⁺/g could be obtained, which is considerably higher than any corresponding value reported in literature. The value decreased with increasing drying temperature. Irrespectively of the conditions of synthesis, the chemical composition was TiO₂·(2.0–2.2) H₂O, and the impurities contained in the product were found to be less than 0.1%. The exchanger produced is in granular form suitable for use in column operation. It is fairly stable in alkaline solution, and also in mineral acid solutions of <0.1 n concentration.     

An investigation of the reaction of UO2(NO3)2 with Na2HPO4 in aqueous solution

The system UO₂(NO₃)₂- Na₂HPO₄-H₂O was investigated by determining the solubility, pH, electrical conductivity and apparent volume of the precipitates. It was found that the reaction in this system takes place in three stages, with successive formation of (UO2)a(PO4), a mixed sodium uranyl acid phosphate UO₂H_xNa_1–xPO₄ and, finally, NaUO₂PO₄·nH₂0.Coneiusions are drawn regarding the possibility of titrimetric determination of the uranyl ions and the free acidity in its salts, also the conditions for obtaining the most closely compacted umnyi phosphate precipitate.Translated from Atomnaya Énergiya, Vol. 14, No. 4, pp. 395–399, April, 1963.     

Ion Exchange Separation for Decontamination of Centrifuge Enrichment Plant

Ion exchange separation of uranyl ion (UO²⁺ ₂) from metal cations has been carried out by the columnar operation using ion exchange resins in 0.1 mol/dm³ sulfuric acid medium. Uranyl ion was adsorbed in an anion exchange resin and the rest metal cations, Fe(III), Al(III), Cr(III), Ni(II) and Cu(II) ions, were adsorbed in a cation exchange resin in this system. Desorption of uranyl ion and metal cations adsorbed in the resins were tested by 2 mol/dm³ sulfuric acid solution. Desorbed elements were confirmed to be precipitated by appropriate alkaline solutions. On the basis of the results obtained, a concept was made on a decontamination system for uranium-contaminated waste solution from centrifuge enrichment plant.     

Precipitation of Thorium or Uranium(VI) Complex Ion with Cobalt(III) or Chromium(III) Complex Cation, (II)

The precipitations of thorium and uranium(VI) sulfito complex ions with hexammine cobalt(III) chloride as the precipitant have been studied.

The orange-colored uranium(VI) precipitate obtained is [Co(NH₃)₆]₄[UO₂(SO₃)₃]₃22H₂O, which is in the form of square bipyramid, about 4 μm across in a cubic symmetry of the diamond type with a=10.40Å It decomposes to an oxide mixture of Co₃O₄ and U₃O₈ above 850°C in the air through a sulfate mixture of CoSo₄ and UO₂SO₄.

Composition of the thorium precipitate varies with the precipitation conditions. Therefore, it is considered that the thorium precipitate contains thorium hydroxide and basic thorium sulfite.     

Synthesis and characterization of low-temperature precursors of thorium-uranium (IV) phosphate-diphosphate solid solutions

Several compositions of new precursor of thorium-uranium (IV) phosphate-diphosphate solid solutions (Th_4−xU_x(PO₄)₄P₂O₇, called β-TUPD) were synthesized in closed PTFE containers either in autoclave (160 °C) or on sand bath (90-160 °C). All the samples appeared to be single phase. From XRD data and TEM observations, the diffraction lines matched well with that of pure thorium phosphate-hydrogenphosphate hydrate (TPHPH), Th₂(PO₄)₂(HPO₄) · H₂O, which confirmed the preparation of a complete solid solution between pure thorium and uranium (IV) compounds. TGA/DTA experiments showed that samples of thorium-uranium (IV) phosphate-hydrogenphosphate hydrate (TUPHPH) prepared at 150-160 °C were monohydrated leading to the proposed formula Th_2−x/2U_x/2(PO₄)₂(HPO₄) · H₂O. The variation of the XRD diagrams versus the heating temperature showed that TUPHPH remained crystallized and single phase from room temperature to 200 °C. After heating between 200 °C and 800 °C, the presence of diphosphate groups in the solid was evidenced. In this range of temperature, the solid was transformed into the low-temperature monoclinic form of thorium-uranium (IV) phosphate-diphosphate (α-TUPD). This latter compound finally turned into well-crystallized, homogeneous and single-phase β-TUPD (orthorhombic form) above 930-950 °C for x values lower than 2.80. For higher x values, a mixture of β-TUPD, α-Th_1−zU_zP₂O₇ and U_2−wTh_wO(PO₄)₂ was obtained. By this new chemical route of preparation of β-TUPD solid solutions, the homogeneity of the samples is significantly improved, especially considering the distribution of thorium and uranium.     

ESR Study of the Structure of γ-Irradiated Sodium-Aluminum- and Sodium-Boro-Phosphate Glasses

The ESR spectra of the anion radicals PO₃ ^2– and PO₄ ^2–, formed as a result of irradiating the terminal (Q ⁽¹⁾) and central (Q ⁽²⁾) phosphorus-oxygen groups of chains of the phosphate glass, are obtained for -irradiated sodium-aluminum- and sodium-boron-phosphate glasses with Na/P = 1 and variable ratios Al/P and B/P. The concentration of linear (2Q ⁽¹⁾ + Q ⁽²⁾) and ring structural groupings (3Q ⁽²⁾) in the irradiated sodium-boron-phosphate glass is determined to be independent of the ratio B/P, while in the sodium-aluminum-phosphate glass the concentration of ring structural groups increases as the ratio Al/P. The latter phenomenon could serve as an explanation of the higher chemical stability of sodium-aluminum-phosphate glass as compared with sodium-boron-phosphate glass.     

The Uranium-Carbon-Nitrogen System

In the previous work⁽¹⁾ dealing with the equilibrium phase relations in the ternary system U-C-N, a qualitative phase diagram was constructed from thermodynamical calculations and experiments. In the thermodynamical calculations, it was assumed that the free energy of formation of the solid solution between UC and UN, i.e., UC_1-x N _x , is the sum of (1-x)ΔG ⁰(UC), xΔG ⁰ and a term for the entropy of mixing RT{x logx+(1-x)log(1-x)}, neglecting the heat of mixing.

The purpose of the present work is to consider the same U-C-N system statistico-thermodynamically, and to evaluate the heat of mixing from the results of several experiments. A value of about ?1.3 kcal/mole was obtained as the heat of mixing corresponding to {2φ(C-N)-φ(C-C)-φ(N-N)}N, where φ(C-N), φ(C-C) and φ(N-N) are the bond energies of C-N, C-C and N-N respectively.     

Thermodynamic study of UO2 +x by solid state emf technique

A complete set of thermodynamic parameters of UO_{2 +x} — the relative partial molar thermodynamic quantities of oxygen: g(O₂), h(O₂) and s(O₂) as a function of nonstoichiometry x and temperature T — have been determined with sufficient accuracy by the precise emf measurements of the solid state galvanic cell of the type Ni.NiO/Stabilized ZrO₂/UO_{2 + x}, at 0.0030 ⩽ x ⩽ 0.23 between 500 and 1100°C. Nonstoichiometry x was controlled and determined by the coulometric titration of oxide ions at 1000°C by using NiO in the Ni/NiO reference mixture as a source of oxygen.UO_{2 +x} samples of two different preparation procedures give almost identical results and show that g(O₂) versus T plots at various compositions x are not a linear function of temperature, but curve downwards with temperature in the composition and temperature ranges studied. This tendency becomes more pronounced with decreasing nonstoichiometry x and indicates that both h(O₂) and s(O₂) of UO_{2 +x} increase with temperature, their temperature dependence becoming stronger with decreasing nonstoichiometry x.The statistical analysis on about forty emf versus T plots at various compositions x confirms that g(O₂), h(O₂) and s(O₂) of UO_{2 +x} at 0.0030 ⩽ x ⩽ 0.23 and 500 ⩽ T ⩽ 1100°C are accurately expressed by the following equations utilizing the polynomial forms of log x for the temperature independent heat capacity-, entropy- and enthalpy-parameters: c_p(O₂), s₀ and h₀ g(O₂) = h(O₂)−Ts(O₂), s(O₂) =s₀ +c_p(O₂) In T, h(O₂) = h₀ +c_p(O₂)T, where temperature T is in Kelvin, and c_p(O₂), s₀ and h₀ are given by c(O₂) = −43.4642−120.129 log x−60.9395(log x)²− 19.4064(log x)³ (J/mol·K). s₀= −93.807 + 281.272 log x + 119.575(log x)² + 72.651(log x)³ (J/mol·K). h₀ = 373.411 + 277.706 log x + 23.4894(log x)² + 654.207(log x)⁻¹ + 198.212(log x)⁻² (kj/mol).The present results are extensively discussed in comparison with the available literature data on these quantities and also in connection with the point defect model of UO_{2 +x} recently proposed by the present authors.     

Mass Spectrometric Study of Ions Formed from Cesium Metaborate Vapor under Electron Impact

An ionization behavior of cesium metaborate vapor under electron impact has been studied by a mass spectrometric method. Formations of Cs²⁺, CsB⁺, CsO⁺, Cs⁺ ₂, Cs₂O⁺, B⁺ and BO^? ₂ ions have been identified in addition to the well known ions of Cs⁺, CsBO⁺, CsBO⁺ ₂ and Cs₂BO⁺ ₂. Ionization processes and vapor precursors for these ions have been given from ionization efficiency curves, appearance energies, temperature dependence of ion intensities and energetics of the ionization processes as follows: the process for the formations of Cs⁺ with AE(Cs⁺)=3.9± 1.0 eV and BO^? ₂ ions is the ion pair formation from CsBO₂(g), that for CsBO⁺ ₂ ion is the simple ionization of CsBO₂(g), that for Cs⁺ with AE (Cs⁺) =9.1±0.5 eV, Cs²⁺, CsBO⁺, CsB⁺, CsO⁺ and B⁺ ions is the dissociative ionization from CsBO₂(g) and that for Cs⁺ ₂, Cs₂O⁺ and Cs₂BO⁺ ₂ ions is the dissociative ionization from Cs₂(BO₂)₂(g). The knowledge of the ionization behavior of cesium metaborate vapor under electron impact is very useful in the mass spectrometric study of vaporization behaviors of CsBO₂(s) and simulated radioactive waste borosilicate glasses.     

Ageing of a phosphate ceramic used to immobilize chloride contaminated actinide waste

A process for the immobilization of intermediate level waste containing a significant quantity of chloride using Ca₃(PO₄)₂ as the host material has been developed. Waste ions are incorporated into two phosphate-based phases, chlorapatite [Ca₅(PO₄)₃Cl] and spodiosite [Ca₂(PO₄)Cl]. Non-active trials performed using Sm as the actinide surrogate demonstrated the durability of these phases in aqueous solution. Trials of the process, in which actinide-doped materials were used, were performed at PNNL which confirmed the wasteform resistant to aqueous leaching. Initial leach trials conducted on ²³⁹Pu/²⁴¹Am loaded ceramic at 313 K/28 days gave normalized mass losses of 1.2 × 10⁻⁵ g m⁻² and 2.7 × 10⁻³ g m⁻² for Pu and Cl, respectively. In order to assess the response of the phases to radiation-induced damage, accelerated ageing trials were performed on samples in which the ²³⁹Pu was replaced with ²³⁸Pu. No changes to the crystalline structure of the waste were detected in the XRD spectra after the samples had experienced an α radiation fluence of 4 × 10¹⁸ g⁻¹. Leach trials showed that there was an increase in the P and Ca release rates but no change in the Pu release rate.     

Solvation of Lithium Ion in Organic Electrolyte Solutions and Its Isotopie Reduced Partition Function Ratios Studied by ab initio Molecular Orbital Method

To explore local structures around lithium ions and to estimate lithium isotopic reduced partition function ratios (RPFRs) ofsolvated lithium ions in ethylene carbonate (EC), methylethyl carbonate (MEC) and EC/MEC mixed solvent systems, ab initio molecular orbital calculations at the HF/6-31G(d) level of theory were carried out. Both EC and MEC were coordinated to lithium ions using their carboxyl oxygens and the Li-O bond distance increase with increasing solvation number up to 4 in the primary solvation sphere both in EC and MEC systems. Binding energy calculations suggested that EC was preferentially coordinated to the lithium ion in the EC/MEC mixed solvent system. RPFRs of solvated lithium ions were convex functions of the solvation number between 1 and 4 and took the maxima at 3 both in EC and MEC systems. The RPFR value in EC/MEC mixed solvent system was estimated to be 1.07818 at 25°C.