Dissolution kinetics of Indian PHWR natural UO2 fuel pellets in nitric acid – Effect of initial acidity and temperature

Understanding the dissolution behaviour of plutonium rich MOX fuel is one of the major challenges in fast reactor spent fuel reprocessing. As an initial step, kinetics for the dissolution of Indian PHWR natural UO₂ sintered pellets in nitric acid was studied experimentally. In this paper we have reported the effect of initial nitric acid concentration and temperature on dissolution rate of sintered UO₂ pellets. The shrinking core model was used to correlate the experimental results. The apparent activation energy estimated from the temperature effect of the chemical reaction was found to be about 90.5 kJ/mol. The estimated apparent activation energy indicates that the dissolution rate of UO₂ pellets was controlled by chemical reaction under the experimental conditions.     

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm^?3) will be treated. This dissolution process involves short stroke shearing of fuels (～10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 10²–10³ times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio.     

基于地球化学模拟分析杂质矿物对酸法地浸采铀的影响

为探明酸法地浸采铀过程中杂质矿物对铀浸出的影响,以分批浸出试验为基础,采用反应路径模拟探讨杂质矿物对铀浸出机制的影响,利用反应溶质运移模型探讨杂质矿物对铀浸出化学场时空特征的影响。模拟结果表明：方解石、黄铁矿、赤铁矿会与铀矿竞争酸,竞争由强至弱依次为方解石、赤铁矿、黄铁矿,其中黄铁矿在酸浸条件下溶解较弱,但生成的低价硫和亚铁离子能降低溶浸液的Eh值,导致铀浸出减少,赤铁矿在酸浸条件下因耗酸而降低溶浸液的酸度,但又促进黄铁矿的溶解,进而影响铀的浸出;在时间上,杂质矿物会使铀的溶解迁移存在不同程度的滞后,铀的溶解-沉淀旋回周期延长,整个模拟矿层沥青铀矿完全溶解时间更长,铀浸出速率降低;在空间上,杂质矿物会使模拟矿层中铀矿溶解范围减小,铀矿溶解-沉淀旋回过程中沉淀量增加,U（Ⅵ）迁移浸出所需时间延长,浸出铀的迁移累积峰值变化不同。     

Composition and structure of fission product precipitates in irradiated oxide fuels: Correlation with phase studies in the Mo-Ru-Rh-Pd and BaO-UO2-ZrO2-MoO2 Systems

Composition and crystal structure of fission product precipitates in irradiated oxide fuels were studied by X-ray microanalysis and X-ray diffraction using instruments shielded for α-contamination and β-γ-radiation. Pin cross-sections, fuel micro-samples from 300 μm hollow drillings and residues from the dissolution of irradiated material in HNO₃ were investigated. The metallic phases found are hcp ?-Ru(Mo,Tc,Rh,Pd) solid solutions with broad variations in concentration of the components, bcc β-Mo(Tc,Ru) and fcc α-Pd(Ru,Rh). The dominating ceramic precipitate is composed of ($\text{Ba}{}_{\mathrm{1-x-y}}\text{Sr}{}_{\mathrm{x}}\text{Cs}{}_{\mathrm{y}}$)(U,Pu,RE,Zr,Mo)O₃ which crystallizes in the cubic perovskite type. The Mo fraction of these phases is related to the local oxygen potential of the fuels. The in-pile observed results agree well with phase studies in the quaternary Mo-Ru-Rh-Pd system where complete solid solubility exists between hcp Ru and the hexagonally stabilized Mo-Rh and Mo-Pd phases. Agreement is further attained with phase studies in the pseudoquaternary BaO-UO₂-ZrO₂-MoO₂ system which is characterized by a cubic perovskite phase Ba(U,Zr,Mo)O₃.     

Sequential Ion-Exchange Separation of Heavy Elements and Selected Fission Products for Burnup Measurement

A sequential ion-exchange separation method was developed for use in burnup measurements of nuclear fuels. Group separation by anion-exchange resin column with hydrochloric acid solutions containing small amounts of nitric acid and hydrochloric acid was followed by various cation and anion- exchange processes. The heavy elements, such as U, Np and Pu, and some fission products selected as burnup monitors, such as Cs, Mo and Nd, could be sequentially and quantitatively separated from a sample taken from spent fuel. The recovery of these elements through the separation processes were examined. The sampling ratio of an aliquot in reference to the whole fuel specimen was determined by adding as sampling monitor a known amount of Cu to the sample during dissolution. The validity of the ion-exchange separation technique for routine analysis for burnup measurements is also discussed.     

Continuous separation of molybdenum and zirconium from simulated high-level liquid waste with a Taylor–Couette contactor

Reprocessing of spent nuclear fuels generates high-level liquid waste (HLLW) which undergoes vitrification into borosilicate glass before final geological disposal. To ensure the quality of the glass, control of the concentration of chemical species such as molybdenum (Mo), which has an adverse impact on the vitrification process, is critical. Also, zirconium (Zr) can cause crud in washing process and Zr-93 is a long-lived fission product needed to be separated. In this study, a liquid–liquid countercurrent centrifugal contactor with Taylor–Couette flow (TC contactor) was applied to practical multi-species cases. Continuous separation of Mo and Zr from a simulated HLLW with bis(2-ethylhexyl) phosphoric acid (HDEHP) as extractant has been performed. Among a variety of metals in simulated HLLW, Mo, Zr, Y, and Fe are extractable, Mo and Zr were separated from HLLW by equilibrium, and Fe/Y separation was achieved by the effect of non-equilibrium state in TC contactor. Addition of tributyl phosphate could improve extraction of Mo. This study has expanded the scope of the TC contactor to multi-species separation processes.     

Performance of FCCI barrier foils for U-Zr-X metallic fuel

Diffusion couple tests of U-Zr or U-Zr-Ce alloys vs. ferritic martensitic steels such as HT9 or T91 were carried out in order to evaluate the performance of the diffusion barrier candidates. Elemental metal foils of Zr, Nb, Ti, Mo, Ta, V and Cr were very effective in inhibiting interdiffusion between these fuels and steels. Eutectic melting between the fuels and steels was not observed in any of the diffusion couples using these diffusion barrier foils at annealing temperatures up to 800 °C. Among the metallic foils evaluated in this study, V and Cr exhibited the most promising performances as a diffusion barrier material for eliminating the fuel cladding chemical interaction problem. However, Zr, Nb and Ti showed an active interaction with the fuel mainly due to the large U solubility.     

二(2-乙基己基)磷酸酯从硝酸体系中萃取Mo(Ⅵ)

用低浓缩铀靶代替高浓缩铀靶辐照进行~(99)Mo的生产是一个必然的趋势,但采用低浓缩铀靶辐照后裂变体系的组成可能发生改变,从而影响~(99)Mo的分离提取过程。为此,本工作以低浓缩铀辐照后溶解的模拟溶液为研究对象,在U(Ⅵ)大量存在的情况下,考察了二(2-乙基己基)磷酸酯(P_(204))从硝酸体系中萃取Mo(Ⅵ)的行为,重点研究了不同Mo(Ⅵ)浓度下萃取时间、萃取剂浓度、硝酸浓度、温度、其他主要元素(Cs(Ⅰ),Zr(Ⅳ),Y(Ⅲ),Nd(Ⅲ),Al(Ⅲ))等因素对萃取的影响。实验结果表明,不同Mo(Ⅵ)浓度下,P_(204)-磺化煤油对硝酸体系中Mo(Ⅵ)的萃取行为相似;在相比为1时,φ=10%P_(204)-磺化煤油对Mo(Ⅵ)即有较好的萃取效果;硝酸浓度不大于2mol/L时分配比随着硝酸浓度的增加而减少,但硝酸浓度进一步增大时对萃取无显著影响;萃取反应的ΔH和ΔG均为负值,表明该萃取是一个常温下能自发进行的放热反应;溶液中U(Ⅵ)和本工作考察的其它主要元素存在及其浓度的改变不会显著影响P204对Mo(Ⅵ)的萃取行为,且采用P_(204)可将Mo(Ⅵ)与Y(Ⅲ)、Nd(Ⅲ)、Al(Ⅲ)选择性地分离。     

Electrolysis of Burnup-Simulated Uranium Nitride Fuels in LiCl-KCl Eutectic Melts

The electrochemical behavior of burnup-simulated uranium nitride fuels containing representative solid fission product elements, UN+Mo (Mo = 2.84 wt%), UN+Pd (Pd = 4.6 wt%) and (U, Nd)N (NdN = 8.0 wt%), was investigated in the molten LiCl-KCl eutectic salt with 0.54 wt% UCl₃ in order to clarify the effects of fission products on the dissolution of actinide nitrides and the behavior of FPs in the electrorefining of spent nitride fuel. The rest potentials of burnup-simulated UN pellets were similar to that of pure UN. The electrochemical dissolution of UN began at about _0:75V vs Ag/AgCl reference electrode in all samples as well as that of pure UN. After the electrolyses at the constant anodic potential of ?0:65––0:60V vs Ag/AgCl, most of UN was dissolved into LiCl-KCl as UCl₃ at the anode, and U was recovered in the liquid Cd cathode in all samples. Furthermore, Nd was dissolved at the anode and accumulated into LiCl-KCl as NdCl₃, while Mo and Pd were not dissolved but remained at the anode.     

Separation of Precipitate-forming Elements during High Level Waste Treatment Using Acidic Zirconium Salt of Dibutyl Phosphoric Acid

Iron extraction and the limits of iron dibutyl phosphate precipitation were investigated for use in TPE/RE recovery and partitioning with the use of the acidic Zr salt of dibutyl phosphoric acid (ZS HDBP, Zr:HDBP = 1:9), dissolved in 30% TBP with Isopar-L. The presence of Mo and the solvent loading with RE affected the Fe extraction in the opposite ways. Slow kinetics and process irreversibility were found for the Fe extraction with the ZS HDBP solution. To increase acceptable Fe concentration in a HLW, reduction of Fe(III) by ascorbic acid (AA) was studied to be carried out continuously in the feed flow, just before entering the head contactor of the partitioning extraction cycle. The Mo extraction is of the same order as that of TPE and RE; so its selective stripping prior to the TPE/RE separation is required. This can be done using the DTPA or H₂O₂ solution in diluted nitric acid. The effectiveness of the process was verified by laboratory scale trials on the centrifugal contactor rig using simulated HLW. The Zr-to-HDBP ratio of 1:6 was found to be useful to decrease the Fe and Mo extractability for their better backwashing with complexants.     

Basic study of photochemistry for application to advanced nuclear fuel cycle technology

In this study, we have accomplished for the first time the photochemical valence adjustment of Pu and Np for the separation and coextraction of these elements in a nitric acid solution using UV light irradiation. Also, the separation and coextraction of Pu and Np were substantiated in principle by the solvent extraction using 30% TBP/n-dodecone after or during the photochemical valence adjustment. By only one photochemical separation operation, about 86% of Pu and about 99% of Np were distributed into the organic phase and the aqueous phase, respectively, and then by only one photochemical coextraction operation, about 86% of Pu was distributed together with about 99% of Np into the same organic phase. Based on these experimental data, we determined that the photochemical oxidation reaction was due to the photoexcted nitric acid species, ′NO₃⁻.

To confirm the strong oxidative ability of this species, the photochemical dissolution tests of UO₂ powder in a nitric acid solution by UV light irradiation were carried out. The irradiation rate and the concentration of nitric acid solution significantly effects the photochemical dissolution reaction, we have also accomplished for the first time the photochemical dissolution of UO₂ at room temperature.     

Preparation, sintering and leaching of optimized uranium thorium dioxides

Mixed actinide dioxides are currently studied as potential fuels for several concepts associated to the fourth generation of nuclear reactors. These solids are generally obtained through dry chemistry processes from powder mixtures but could present some heterogeneity in the distribution of the cations in the solid. In this context, wet chemistry methods were set up for the preparation of U_1−xTh_xO₂ solid solutions as model compounds for advanced dioxide fuels. Two chemical routes of preparation, involving the precipitation of crystallized precursor, were investigated: on the one hand, a mixture of acidic solutions containing cations and oxalic acid was introduced in an open vessel, leading to a poorly-crystallized precipitate. On the other hand, the starting mixture was placed in an acid digestion bomb then set in an oven in order to reach hydrothermal conditions. By this way, small single-crystals were obtained then characterized by several techniques including XRD and SEM. The great differences in terms of morphology and crystallization state of the samples were correlated to an important variation of the specific surface area of the oxides prepared after heating, then the microstructure of the sintered pellets prepared at high temperature. Preliminary leaching tests were finally undertaken in dynamic conditions (i.e. with high renewal of the leachate) in order to evaluate the influence of the sample morphology on the chemical durability of the final cohesive materials.     

Acidic dissolution behavior of U containing ZrO2–MgO ceramics

This study explores the possibility of dissolving zirconia–magnesia inert matrix fuel containing uranium oxide as a fissile material and plutonium homolog and erbium oxide as a burnable poison with nitric and sulfuric acid as a potential first step in a reprocessing scheme. The progress of the dissolution is followed by monitoring the amount of material in solution by inductively coupled plasma-atomic emission spectroscopy, assessing the speciation of the material by time resolved laser fluorescence spectroscopy, and determining and quantifying the crystalline phases present in the remaining residue by X-ray diffraction. This study has shown a linear incongruent dissolution of the cubic zirconia phase in concentrated nitric acid under certain chemical compositions, while the magnesium oxide phase is completely soluble. In sulfuric acid uranium, erbium, and magnesium are soluble to different extents while zirconium forms a colloidal suspension that conglomerates and settles out of solution. The feasibility of the dissolution of zirconia–magnesia inert matrix fuel with nitric and sulfuric acid for reprocessing is discussed.     

Dissolution of ThO2Based Oxides in Nitric Acid Solutions at Elevated Temperatures

Highly-dense spherical particles of thorium-based oxides, ThO₂ and (Th, U)O₂, prepared by the sol-gel method were subjected to dissolution with nitric acid containing 0–0.05 mol/l NaF at high temperatures above 120°C. The dissolution rate depended upon temperature, fluoride concentration and UO₂ content. High-temperature in the range of 120–200°C enhanced the dissolution of the ThO₂-based fuels. At low temperatures and/or low U0₂ concentrations, insoluble tetrafluoride precipitates were formed on the particle surfaces and they resulted in the decrease of the dissolution rates. In the present study, the apparent activation energies for the high-temperature dissolution were obtained.     

Dissolution of spent nuclear fuel in carbonate–peroxide solution

This study shows that spent UO₂ fuel can be completely dissolved in a room temperature carbonate–peroxide solution apparently without attacking the metallic Mo–Tc–Ru–Rh–Pd fission product phase. In parallel tests, identical samples of spent nuclear fuel were dissolved in nitric acid and in an ammonium carbonate, hydrogen peroxide solution. The resulting solutions were analyzed for strontium-90, technetium-99, cesium-137, europium-154, plutonium, and americium-241. The results were identical for all analytes except technetium, where the carbonate–peroxide dissolution had only about 25% of the technetium that the nitric acid dissolution had.     

Axial Dispersion Coefficient and Mass Transfer Rate in Pulsed Column

This work has investigated an improvement for the usual batch denitration by formic acid. It studied several destruction characteristics of nitric acid and formic acid in a continuous denitration process newly suggest in this work which consisted of a continuous denitration by formic acid and a residual acid-electrolytic destruction system. Also, the precipitation behaviors of a few metal ions such as Mo, Zr, Nd, and Fe during the denitration were investigated. The continuous denitration by formic acid reached a steady state in 30 min and showed a dependency of the final acidity on the residence time of the feeding solution in the reaction. In a Ti-IrO₂ electrolytic cell, the destructive rates of formic acid and nitric acid were 1.37×10^-2M-cm²/h-mA and 9.33×10^-3 Mcm²/hmA, respectively. The nitric acid was reduced at the Ti cathode and the formic acid was oxidized at the IrO₂ anode. The suggested denitration process combining a chemical system and an electrolytic system could treat continuously a feeding nitric acid of 2.0 M below about 0.1 M. The precipitation of the metal ions occurred almost totally in the denitration column.     

Modified TRUEX Process for the Treatment of High-Level Liquid Waste

The TRUEX process has been examined to recover Am and Cm from the high-level liquid waste of Purex reprocessing plant. Continuous counter current extraction and back-extraction experiments were carried out by a mixersettler using simulated waste solution for three process flowsheets, i.e. a process flowsheet developed by Argonne National Laboratory and other two process flowsheets which were modified in the scrub stage. The result indicates that the process flowsheet of Argonne National Laboratory cannot be applied for the high-level liquid waste containing high concentrations of lanthanide and actinide elements because of the formation of insoluble salts of these elements with oxalic acid, which is added to restrict the extraction of fission products such as Mo and Zr. A modified process flowsheet, which had only one scrub stage with high concentration of nitric acid, was found to be the best of three process flowsheets examined, where Nd as a simulated element of Am and Cm was sufficiently recovered and any precipitation of oxalate salt was not observed.     

Synthesis,characterization and solubility of mixed zirconium-cerium molybdate precipitates

The reaction between fission products Mo(VI) and Zr(IV) in hot nitric acid solution during the spent fuel dissolution process leads to the formation of hydrated zirconium molybdate precipitate. This solid can include other metallic cations at the 4th oxidation state (+ IV) inside the crystal such as Pu(IV).

Precipitations in the inactive surrogate system of mixed zirconium-cerium molybdates Mo-Zr(IV)-Ce(IV) were performed all over the cerium molar percentage scale from 0 to 100%, Ce(IV) being used as a surrogate of Pu(IV). The crystal structures were identified from powder X-ray diffraction patterns and complementary investigations by SEM and elemental analyses were carried out. The solid composition pointed out two phases that are ZrMo₂O₇(OH)₂(H₂O)₂ and Ce₃Mo₆O₂₄(H₂O)₄. Each of these compounds was proved being a solid solution within which the Zr^IV and Ce^IV atoms can substitute each other.

Moreover, the influence of the cerium molar fraction on the precipitate solubility was investigated and showed a strong evolution of solubility not only with the nature of the precipitate and the Ce content, but also with the nitric acid concentration.     

Effect of Si and Zr on the interdiffusion of U-Mo alloy and Al

We investigated the effect of Zr additions to U-Mo and Si additions to Al on interdiffusion between U-Mo and Al by employing diffusion couple tests. We examined the phase stability of the γ-heat-treated alloys by high-temperature annealing tests. Using X-ray diffraction, we observed that the γ-phase U-7Mo-Zr alloys with more than 2 wt% Zr decomposed faster than the U-7Mo alloys. The diffusion couples showed that a Zr addition to U-7Mo and the addition of Si in Al reduced the interaction layer growth rates. However, Zr additions to U-Mo are most effective in reducing the overall interdiffusion rates when combined with Si additions to Al. The decomposition of the metastable U-Mo γ-phase during the diffusion test appears to have a significant effect on the overall interdiffusion rates.     

Fabrication of zirconia-based targets for transmutation

The fabrication of homogeneous (Am,Y,Zr)O_2−x pellets and heterogeneous pellets, containing (Am,Y,Zr)O_2−x spheres dispersed in an inert matrix, by dust-free processes has been investigated. Due to the high activity of americium, the preparatory fabrication tests and process development are being carried out using cerium analogue element. The sol gel route is used to produce highly porous Y_0.15Zr_0.85O_2−x spheres, which are then infiltrated with a cerium nitrate solution to give (Ce,Y,Zr)O_2−x. The goal (28 wt% Ce) can be achieved. Homogeneous targets with densities up to 94 %TD have been obtained.