Recovery of neutron-irradiation-induced defects of Al2O3, Y2O3, and yttrium-aluminum garnet

SiC fiber-reinforced SiC matrix composites (SiC_f/SiC) are considered as one of the candidates for blanket materials in future fusion reactors and as an advanced fuel cladding material for next-generation fission reactors. Generally, the densification of SiC needs sintering additives and oxides such as Al₂O₃, Y₂O₃, and yttrium-aluminum garnet (YAG, Y₃Al₅O₁₂), which are frequently added to SiC. However, the effects of neutron irradiation on sintering additives are still unclear. In this study, we performed the neutron irradiation of Al₂O₃, Y₂O₃, and YAG at fluences up to 2.0–2.5 × 10²⁴ n/m² (E > 0.1 MeV) at 60–90 °C. The isochronal recovery of the macroscopic volume of Al₂O₃ against annealing temperature showed smooth and continuous shrinkage at a temperature of up to 1200 °C, and the volume slightly increased above that temperature. In contrast, the volume of Y₂O₃ showed quick shrinkage at the low temperature range, and slower and smooth recovery was observed up to ～1100 °C. In the case of YAG, the recovery of volume occurred in a step-wise manner at 600–750 °C, and continuous shrinkage occurred at temperatures lower and higher than that temperature range. The activation energies for the macroscopic volume recoveries of three oxides were obtained from the Arrhenius plots of the rate coefficients. Two-stage recovery was observed for Al₂O₃, whereas more complicated recovery processes were suggested for Y₂O₃ and YAG.     

Adsorption Characteristics of Trivalent Rare Earths and Chemical Stability of a Silica-Based Macroporous TODGA Adsorbent in HNO3 Solution

In order to separate trivalent minor actinides (Am and Cm) from high-level liquid waste generated in a nuclear fuel reprocessing process, a silica-based macroporous TODGA (N,N,N′,N′-tetraoctyl-3-oxapentane-1,5-diamide) adsorbent (TODGA/SiO₂-P) was prepared. In this study, the adsorption characteristics of some trivalent rare earths (Y, Nd, and Eu), whose separation behaviors were similar to those of the minor actinides from HNO₃ solution with the TODGA/SiO₂-P adsorbent, and the chemical stability of the adsorbent against the HNO₃ solution were evaluated experimentally. It was found that the adsorbent exhibited a quite strong adsorption especially for Y(III). The standard enthalpy change for the Y(III) adsorption was determined to be ?2.5kJ·mol^?1 using the van't Hoff equation, which indicates that the adsorption was an exothermic reaction. The results of chemical stability experiments showed that the adsorbent had relatively excellent properties in long-time contact with the HNO₃ solution.     

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.     

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%.     

Optimization for removal of ruthenium from nitric acid solution by volatilizing with electrochemical oxidation

Ruthenium is a major fission product element among the platinum group elements (PGEs) in high-level liquid waste (HLLW). Ru tetra-oxide, RuO₄, has high vapor pressure, which is high enough to be run off from its solution even at room temperature. Electrochemical oxidation method, to oxidize nitrosyl ruthenium to the tetra-oxide and then to remove ruthenium from liquid phase to gas phase, was studied to separate Ru from the HLLW. The advantage of this method requires neither additional reagents nor adjustment its valency before the oxidation and disadvantage is necessity of long time for oxidation. In order to improve oxidation rate, we carried out the experiments to clarify the effects following fundamental conditions to the electrochemical oxidation, which are (a) electrolyte temperature, (b) presence of promoter elements, (c) evaporation or reflux of condensed phase, and (d) using or not using of diaphragm at counter electrode. We found the fast oxidation conditions as follows: (1) higher temperature; 95°C, (2) Ce coexistence; 3000 ppm; and (3) usage of a diaphragm for counter electrode. However, evaporation or reflux conditions did not directly affect the electrochemical oxidation efficiency.     

Recovery of palladium by silica/polymer-based 2,6-bis(5,6,7,8-tetrahydro-5,8,9,9-tetramethyl-5,8-methano-1,2,4-benzotriazin-3-yl)pyridine adsorbents from high level liquid waste

The adsorption properties of a porous silica-based adsorbent 2,6-bis(5,6,7,8-tetrahydro-5,8,9,9-tetramethyl-5,8-methano-1,2,4-benzotriazin-3-yl)pyridine (Me₂-CA-BTP/SiO₂-P) towards Pd(II) were investigated. The adsorption ability of Me₂-CA-BTP/SiO₂-P towards Pd(II) increased dramatically with the increase of HNO₃ concentration and showed a saturated uptake of Pd(II) after the HNO₃ concentration reached 3 mol/dm³. It was found that nitrate ion would participate in the adsorption reaction. Palladium adsorption kinetic and isotherm were in accordance with the pseudo-second-order rate law and Langmuir isotherm adsorption model, respectively. Thiourea could effectively elute the adsorbed palladium from Me₂-CA-BTP/SiO₂-P. Adsorption studies with high level liquid waste elements showed a high selectivity towards Pd(II) over other fission product elements in ≥3 mol/dm³ HNO₃ solution. The stability of Me₂-CA-BTP/SiO₂-P against HNO₃ was satisfied for the reason of the adsorption performance keeping for Pd(II) after a long-time contact with HNO₃ solution. The adsorbent performed good γ-irradiation stability in dry state and low concentrations of HNO₃. The presence of HNO₃ enhanced the radiolysis and the adsorbent still reserved a presentable adsorption amount after γ-irradiation in 3 mol/dm³ HNO₃.     

Radiation Treatment of Exhaust Gases, (XII) NO Removal in Moist NO-SO2-O2-N2 Mixtures Containing NH3

The NO removal by electron beam irradiation was studied in the moist NO-SO₂-O₂-N₂ mixtures containing NH₃. The NO removal was promoted markedly by addition of NH₃ and at the same time, SO₂ was removed. The formations of NO₂, N₂O, NH₄NO₃ and (NH₄)₂SO₄ were observed in the mixtures containing NH₃. The NO removal increased with H₂O and SO₂ concentrations and was hardly affected by the presence of 3–19.5% O₂. The degree of the NO_x(NO+NO₂) removal became larger with decreasing temperature in the range of 80–150°°C. The NO removal was independent of dose rate in the range of 3.1×10⁴–2.4×10⁶ rad/s. The promotion of the NO removal by addition of NH₃ is attributable to the effective decomposition of NO by NH₂ radical formed by the reaction of NH₃ with OH radical. The NO₂ yield decreased by addition of NH₃ and the N₂O yield increased.     

Crystalline phases in a devitrified simulated high-level waste glass containing the elements of the platinum group

A borosilicate glass containing 20 wt% simulated high-level waste oxides was subjected to heat treatment at 700°C for 1000 h. Seven crystalline phases were newly formed by the treatment in addition to two phases, (Ru, Rh)O₂ and (Pd, Rh, Te), which had already existed in the as-prepared simulated high-level waste glass. Among the new seven phases, five phases were certainly identified to be (RE)BSiO₅, CeO₂, SiO₂, (RE)PO₄ and (Sr, Ba, RE)MoO₄. Of two unidentified phases, one was rich in silicon, chromium and rare earth elements (RE), and the other was rich in nickel and chromium.The crystalline phases of the elements of the platinum group facilitated the occurrence of the other phases and suppressed crystal growth.     

Feasibility studies on the selective separation of fission palladium(II) by isoHex-BTP/SiO2-P adsorbent from HLLW

Aiming at the selective recovery of fission palladium(II) from high-level liquid waste (HLLW), the silica/polymer (SiO₂-P)-based isoHex-BTP adsorbent (isoHex-BTP/SiO₂-P) was synthesized by impregnating complexing agent isoHex-BTP into the multiporous SiO₂-P inert support. The feasibility of separation of Pd(II) from HLLW by isoHex-BTP/SiO₂-P was evaluated by batch experiment method. The results showed that isoHex-BTP/SiO₂-P exhibited much higher adsorption selectivity for Pd(II) than the other fission products, even Am(III) and Pu(IV) presented in HLLW. The ideal nitric acid concentration for the adsorption of Pd(II) by the adsorbent was shown to be ≥2 mol dm^–3. The adsorption of Pd(II) fits well to the pseudo-second-order kinetic model and Langmuir isotherm model. Quantitative Pd(II) desorption was achieved by using 0.5 mol dm^?3 SC(NH₂)₂ - 0.1 mol dm^?3 HNO₃ solution.     

Gd2Zr2-xCexO7(0.0≤x≤2.0)的制备与表征

为获得钆锆烧绿石基固化体固化Pu后的物理性能、物相变化及微观形貌,本研究用Ce⁴⁺模拟Pu⁴⁺,以Gd₂O₃、ZrO₂和CeO₂粉体为原料,采用高温固相法,制备不同固溶度（0～100%,按摩尔计）的系列钆锆烧绿石固化体,并对密度、硬度、物相和微观形貌等进行表征。结果表明：实验所得系列固化体的密度和硬度均随着x的增大而增大,硬度H_V与x满足关系式H_V＝661.272 73+223.936 36x（R²＝0.946 38）。在x＝0.0时,所得固化体为单一的烧绿石结构;在x＝0.2时,固化体从烧绿石结构转变为萤石型结构;在0.2≤x≤2.0范围内,固化体均为单一萤石型结构。固化体微观形貌不规则,呈板块状。     

EPR Study of Uranium (V) Species in Photo-and Electrolytic Reduction Processes of UO2 NO3)2-2TBP

Abstract

Electron Paramagnetic Resonance (EPR) and optical spectra of uranium(V) species were observed in both processes of photo- and electrolytic reduction of UO₂(NO₃)₂-2tributylphosphate (TBP) in 80%TBP-n-dodecane solution. The formation of U(V) species was detected by an optical spectrum (λ_max: 770, 970 and 1,420 nm). EPR signal with the value of ff-factor –2.3 and a linewidth of approximately 1,100 Gs was observed during the electrolytic reduction. On the other hand, during the photoreduction the signal with the value of fil-factor –1.94 was observed and there was found a superhyperfine structure with the intensity ratio of 1:2:1, that is caused by the superhyperfine coupling with nuclear spin, I=1/2, of the strongly coordinated ³¹P to the central uranium through oxygen atom. The superhyperfine coupling constant was estimated to be 27 Gs. Moreover, the signal with the value of g-factor –2.00 due to an organic radical was observed. The residue after the thermo-gravimetric analysis of UO₂(NO₃)₂-2TBP was identified as α-UP₂O₇ by the powder X-ray diffraction analysis, indicating the strong coordination of TBP to the central uranium atom.     

Thermal conductivity of Na2O–B2O3–SiO2 melts

Borosilicate glasses are candidate materials for the immobilization of high-level radioactive waste. The values of thermal conductivity of different borosilicate melts are thus indispensable information when optimizing the temperature distribution in a glass melting furnace. In this study, the thermal effusivity of Na₂O–B₂O₃–SiO₂ melts was measured using a front heating–front detection laser-flash method. The thermal conductivity, which can be obtained by combining the measured thermal effusivity with the specific heat capacity and density, was calculated using the least-squares method; the values for the Na₂O–B₂O₃–SiO₂ melts either slightly decreased linearly with increasing temperature or remained almost constant over the investigated temperature range. The values of thermal conductivity of the Na₂O–B₂O₃–SiO₂ melts were higher than those of B₂O₃–SiO₂ melts and lower than those of CaO–B₂O₃–SiO₂ melts. Furthermore, the thermal conductivity of the Na₂O–B₂O₃–SiO₂ melts was compared with those of the B₂O₃–SiO₂ and CaO–B₂O₃–SiO₂ samples.     

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.     

Thermochemical and Experimental Considerations of NOx Composition and Iodine Species in the Dissolution of Spent PWR-Fuel Specimens

Abstract

The NO_x composition and iodine species in the dissolution of spent fuels are discussed on the basis of thermochemical calculations and experimental results. The influence of N0_x sparging on the expulsion of iodine is also discussed. The dissolution of a spent PWR-fuel specimen (–3g) in 30 ml of 3.5_MHNO₃ at 100°C is calculated to yield a concentration of 7×10^?2atm of N0₂, which is 80% of the total NO_X in the dissolver. This N0₂ fraction is much higher than experimental values of 15% or less that were reported for dissolver off-gas cooled near to room temperature. The high N0₂ fraction suppresses the formation of iodate (IO₃ ^?) in the dissolution. The calculations predict that IO₃ ^?) is not formed in 3.5 M HNO₃ at 100°C at an NO₂ pressure ≥3×10^?2 atm (3kPa). Attempts to expel iodine from the fuel solution indicated that the main iodine species in the solution was colloidal iodine and not iodate (I0₃ ^?) which earlier workers postulated. The obtained experimental results are consistent with the thermochemical predictions. For the decomposition of the colloidal iodine in the expulsion process, NO₂ sparging has a negative effect. This is because an increase in NO₂ pressure promotes the formation of colloidal Agl.     

Investigation of high-temperature chemical interaction of calcium silicate insulation and cesium hydroxide

ABSTRACT

The interaction of cesium hydroxide and a calcium silicate insulation material was experimentally investigated at high-temperature conditions to evaluate the possibility of unprecedented cesium retention under severe accident of boiling water reactor. The temperature where the interactions occurred and chemical species of cesium after the interaction were examined in this study. A thermogravimetry equipped with differential thermal analysis was used to analyze thermal events in the samples of mixed calcium silicate and cesium hydroxide under oxidizing and reducing atmospheric conditions with a maximum temperature of 1100°C. Before being mixed with cesium hydroxide, a part of calcium silicate was pretreated at high temperature to evaluate the effect of possible structural changes of this material due to a preceding thermal history and also for the sake of thermodynamic evaluation. It was found that for the original material, as xonotlite (Ca₆Si₆O₁₇(OH)₂) crystal, the endothermic reaction with cesium hydroxide occurred over the temperature range 575–730°C; meanwhile, for heat-treated material, which varied the crystal phase of original material to wollastonite (CaSiO₃), the interaction occurred over the temperature range 700–1100°C. The X-ray diffraction analyses have indicated that both types of calcium silicates regardless of the atmospheric conditions, cesium aluminum silicate, CsAlSiO₄, was formed with aluminum in the samples as an impurity or adduct. The insolubility of this formed cesium suggested the potency of cesium localization in the primary containment vessels on other structural materials that possess similar elements to that of calcium silicate insulation; hence, an effective decommissioning process could be developed.     

Glass matrices for immobilizing nuclear waste containing molybdenum and phosphorus

Vitrification has been selected in France as the process for immobilizing high-level waste arising from spent fuel reprocessing. Some high-level solutions generated by reprocessing legacy fuel contain high molybdenum concentrations. Molybdenum is known to be sparingly soluble in conventional borosilicate glass, and work is in progress to find suitable glass formulations for such waste. The results of a basic study to identify borosilicate glasses composition zones of potential interest are discussed. A vast composition range was investigated by defining a fine mesh. The limits considered to delimit the range of the study were intentionally extended to identify formulations such as SiO₂-B₂O₃-Al₂O₃-Na₂O-P₂O₅ that are of interest for vitrifying molybdenum-rich waste. Observation of more than 50 tested mixtures revealed two composition zones of potential interest. One forms a homogeneous glass after melting at 1300 °C and rapid cooling; the other vitreous material comprises unconnected microbeads uniformly dispersed in a borosilicate glass.     

Study on effects of swift heavy ion irradiation on the crystal structure in CeO2 doped with Gd2O3

To simulate the effects of Gd₂O₃-doping and high-energy fission products in UO₂, Gd₂O₃-doped CeO₂ pellets were irradiated with 200-MeV Xe¹⁴⁺ ions. Doping and irradiation effects were analyzed using X-ray diffraction (XRD) and extended X-ray absorption fine structure (EXAFS). The lattice constant of CeO₂ decreases and the local structure is disordered with increased doping levels. However, the irradiation induces an expansion of the lattice and a disordering of atomic arrangement near the Gd atoms. The effects of the irradiation become more pronounced with increasing Gd₂O₃-dopant levels. Our results are compared with those of a study involving Er₂O₃-doped CeO₂.     

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.     

Relative Reactivities of Radioiodine Released from U3O8 in Oxygen and Helium Atmospheres toward Propane

Chemical form of radioiodine released from U₃O₈ will be UI_x in helium and I₂ in oxygen atmospheres. Relative reactivities of these iodines toward organic compounds have been studied at a temperature range of 250~500°C, by choosing propane as a reactant. The UI_x has a higher reactivity than I₂, particularly at a temperature range of 300~400°C. Above that temperature, both iodine species show a similar reactivity. In the reactions, CH₃I, C₂H₅I, i, and n-C₃H₇I are formed: relative yields depend on both temperature and atmosphere. Two reaction processes participate in the formation of organic iodides. One is an iodine-initiation reaction and is relatively important at a lower temperature. The other is a radical-initiation reaction and become predominant as the temperature rises.     

In-Line Uranium Monitoring in Reprocessing Waste Solution by Time-Resolved Laser-Induced Fluorometry

In-line monitoring by fluorometry of uranium concentration in reprocessing waste solution has been realized. The reduction of U0₂ ²⁺ fluorescence by coexisting ions and solution temperature can be corrected by measuring fluorescence lifetime, absorbance of excitation beam wavelength and absorbance of fluorescence wavelength. The method applicability was examined by using a sample solution simulating the waste solution of the codecontamination process. When the coexisting ion concentrations were increased in the sample solution which included 50 mg//of uranium, the corrected value of the uranium concentration was constant in the range of 0~1.5 times the coexisting ion concentrations. When the temperature was changed in the range of 30~45°C, the corrected value was also constant. The precision of the correcting method was ±15%'. These results verified that the correcting method could be applied to in-line monitoring of uranium concentration even though species and amounts of the coexisting ions and solution temperature were changed in the waste solution.