玻璃固化过程中Mo的化学行为研究进展

高放废液中的Mo在玻璃中的溶解度较低,在高放废液的玻璃固化中易形成黄相,黄相的出现对玻璃固化过程和固化体性能均不利,限制了玻璃固化体中废物的包容量。通过改变玻璃配方或开发研究新的配方提高固化体中Mo含量,可以消除黄相。本文综述了近年来国内外针对玻璃固化过程中Mo的化学行为研究所取得的研究进展。     

Angular Distribution of Gadolinium Vapor Produced by Electron Beam Heating

Static corrosion tests were performed for the glass phase of a simulated waste form of non-combustible radioactive low-level waste to study a basic aqueous corrosion behavior. The waste form, which was fabricated from simulated waste sample by use of in-can type induction-heated melting, consists of two separated phases; a glass phase and a metal phase. Tests were performed for the glass phase from two types of the waste form with different chemical composition at 35°C and S/V ratio of 2,600 m^?1. The glass phase with both types showed an incongruent dissolution similar to conventional high-level radioactive waste (HLW) glasses, i.e., the normalized elemental mass loss (NL_i) for soluble elements such as B and Na continued to increase after the saturation of insoluble elements such as Si, A1 and Ca. The NL_i for B increased in proportion to the square root of time except for early stage, which suggests that the rate of the long-term dissolution or alteration may be controlled by a diffusion process. Potential secondary phases forming as the results of incongruent dissolution were estimated to be kaolinite and calcite by comparison of the measured solution data with the thermodynamically calculated phase stability relationships. These results suggest that the glass phase has a potential chemical durability not so different from conventional HLW glasses.     

Temperature Effect on Plutonium Leach Rate of Nuclear Waste Glass

The two kinds of nuclear waste glass with similar composition, a ²³⁸Pu-doped and nonradioactive waste glass, were leached under the ISO-test conditions at temperature between 23 and 90°C. An activation energy of 22±10kJ/mole was obtained from the initial leach rates of Pu, which was much lower than the 78±9kJ/mole obtained from those of Si, Na, Sr and Cs, It is suggested that in the initial stages of leaching. Pu is not released from the waste glass with the same mechanisms as the releases of Si, Na, Sr and Cs, but the dissolution of hydrous plutonium dioxide PuO₂·_xH₂O formed on the glass surface becomes predominant. In the long duration tests (<32d), the release of Pu appears to be affected by the solubility of PuO₂·_xH₂O remaining in the leached surface layers.     

Vitrification of high-level liquid waste: glass chemistry, process chemistry and process technology

The management of high-active liquid waste includes the solidification and the subsequent isolation in a deep geological formation. Vitrification is internationally accepted as the technology of choice for its immobilization. It is the best-demonstrated available technology. The advantages of the glass waste form are its tolerance of chemical variability, chemical durability and processability. About one dozen of vitrification plants are worldwide being operated or in the planning stage. One of the latter is the Karlsruhe vitrification plant (VEK) at Karlsruhe, which makes use of the liquid-fed ceramic melter process. This process is described with respect to the process technique, especially the melter technique and the process chemistry. Also basic aspects of the glass chemistry are discussed.     

Vitrification of high-level radioactive waste considering the behavior of platinum group metals

To demonstrate a method using liquid metal for the removal of PGMs during the vitrification process of high-level radioactive waste, removal of Pd was performed using Cu from molten glass containing fission products such as Nd, Sr, Zr, Mo, Te and Ni. Almost all the Pd, 93%, was extracted into liquid Cu and removed as a separable Cu–Pd metal button from the glass. Tellurium and Ni were also extracted 42 and 5.6%, respectively. Nearly 100% of the other elements, especially the heat generating elements such as Sr and Cs, for which Na was used as a substitute of Cs remained in the glass.     

Corrosion Behavior of a Powdered Simulated Nuclear Waste Glass in the Presence of Bentonite

Static corrosion tests of a powdered simulated waste glass were performed in deionized water with and without bentonite at 90°C for periods of up to 130 days. The glass irradiated with thermal neutron for activating Cs in the glass was used as a glass specimen in order to determine the sorption of Cs on bentonite. In the corrosion tests without bentonite, it was observed that normalized elemental mass loss: (NL) values for soluble elements (B, Li, Na and Mo) were larger than those for Si by a factor of three and continued to increase after saturation of Si. In the corrosion tests in the presence of bentonite, it was observed that the glass corrosion was enhanced, and a large amount of Cs was sorbed on bentonite.

The experimental results were analyzed by use of some corrosion models. The analysis showed that diffusion of the soluble elements is expected to be a dominant process for the glass corrosion, as well as the dissolution/precipitation reactions. In addition, it is expected that the glass corrosion in the presence of bentonite is largely affected by both ion-exchange equilibrium of the aqueous phase with Na-montmorillonite and precipitation of sepiolite from dissolution of dolomite.     

Long-term product consistency test of simulated 90-19/Nd HLW glass

Chemical durability of 90-19/Nd glass, a simulated high-level waste (HLW) glass in contact with the groundwater was investigated with a long-term product consistency test (PCT). Generally, it is difficult to observe the long term property of HLW glass due to the slow corrosion rate in a mild condition. In order to overcome this problem, increased contacting surface (S/V = 6000 m⁻¹) and elevated temperature (150 °C) were employed to accelerate the glass corrosion evolution. The micro-morphological characteristics of the glass surface and the secondary minerals formed after the glass alteration were analyzed by SEM-EDS and XRD, and concentrations of elements in the leaching solution were determined by ICP-AES. In our experiments, two types of minerals, which have great impact on glass dissolution, were found to form on 90-19/Nd HLW glass surface when it was subjected to a long-term leaching in the groundwater. One is Mg-Fe-rich phyllosilicates with honeycomb structure; the other is aluminosilicates (zeolites). Mg and Fe in the leaching solution participated in the formation of phyllosilicates. The main components of phyllosilicates in alteration products of 90-19/Nd HLW glass are nontronite (Na_0.3Fe₂Si₄O₁₀(OH)₂·4H₂O) and montmorillonite (Ca_0.2(Al,Mg)₂Si₄O₁₀(OH)₂·4H₂O), and those of aluminosilicates are mordenite ((Na₂,K₂,Ca)Al₂Si₁₀O₂₄·7H₂O)) and clinoptilolite ((Na,K,Ca)₅Al₆Si₃₀O₇₂·18H₂O). Minerals like Ca(Mg)SO₄ and CaCO₃ with low solubility limits are prone to form precipitant on the glass surface. Appearance of the phyllosilicates and aluminosilicates result in the dissolution rate of 90-19/Nd HLW glass resumed, which is increased by several times over the stable rate. As further dissolution of the glass, both B and Na in the glass were found to leach out in borax form.     

Liquid Metal Extraction for Removal of Molybdenum from Molten Glass Containing Simulated Nuclear Waste Elements

Laboratory-scale experiments for removing Mo and MoO3 from molten borosilicate glass were performed using liquid Cu as an extractant. Removal of Mo from the simulated HLW glass containing oxides of Nd, Fe, Zr, Mo, Sn, Ni, Sr, Cd, Ru, and Se was also performed, and the fractions of these elements transferred into Cu were examined. Mixtures of Cu anda ternary SiO₂-B₂O₃-Na₂O glass containing metallic Mo or MoO₃ were heated in an alumina crucible at 1,673K in an Ar environment. The amounts of Mo and MoO₃ added to 10 g of the ternary glass were fixed at 0.1 and 0.15 g, respectively. As for the glass containing metallic Mo, more than 90% of Mo was extracted into liquid Cu. Spherical Cu metal buttons containing Mo formed on the bottom of the crucible when Cu was added at more than 10 times that of Mo on a mass basis. Removal of Mo from the glass containing MoO3 was also achieved by the addition of Si as a reducing agent for the reduction from MoO₃ to Mo. The fraction of Mo extracted into liquid Cu depended on the molar ratio of Si to Cu added to the glass. The fraction increased up to 84% with an increase in the molar ratio of Si/Cu. However, the excess addition of Si may enhance the chemical interaction between the metal phase and the glass phase, and some of the metal phase containing Mo remained in the glass phase without forming a metal button. The optimum molar ratio of Si/Cu that produces the highest removal fraction was found to be approximately 0.5. Almost the same removal fraction of 88% was obtained from the simulated HLW glass under the condition of Si/Cu = 0.5. Nearly 100% of Ru was extracted into Cu with Mo, while Sr, Zr, and Nd were hardly extracted and remained in the glass.     

Candidate waste forms for immobilisation of waste chloride salt from pyroprocessing of spent nuclear fuel

Sodalite/glass bodies prepared by hot isostatic pressing (HIPing) at ～850 °C/100 MPa are candidates for immobilising fission product-bearing waste KCl–LiCl pyroprocessing salts. To study the capacity of sodalite to structurally incorporate such pyroprocessing salts, K, Li, Cs, Sr, Ba and La were individually targeted for substitution in a Na site in sodalite (Na vacancies targeted as charge compensators for alkaline and rare earths) and studied by X-ray diffraction and scanning electron microscopy after sintering in the range of 800–1000 °C. K and Li appeared to enter the sodalite, but Cs, Sr and Ba formed aluminosilicate phases and La formed an oxyapatite phase. However these non-sodalite phases have reasonable resistance to water leaching.Pure chlorapatite gives superior leach resistance to sodalite, and alkalis, alkaline and rare earth ions are generally known to enter chlorapatite, but attempts to incorporate simulated waste salt formulations into HIPed chlorapatite-based preparations or to substitute Cs alone into the structure of Ca-based chlorapatite were not successful on the basis of scanning electron microscopy. The materials exhibited severe water leachability, mainly in regard to Cs release. Attempts to substitute Cs into Ba- and Sr-based chlorapatites also did not look encouraging. Consequently the use of apatite alone to retain fission product-bearing waste pyroprocessing salts from electrolytic nuclear fuel reprocessing is problematical, but chlorapatite glass–ceramics may be feasible, albeit with reduced waste loadings. Spodiosite, Ca₂(PO₄)Cl, does not appear to be suitable for incorporation of Cl-bearing waste containing fission products.     

Heats of formation of (U,Mo)Al3 and U(Al,Si)3

The heats of formation of (U,Mo)Al₃ intermetallic compounds were obtained by measuring the reaction heats of U-Mo/Al dispersion samples by differential scanning calorimetry. Based on literature data for the reaction heats of U₃Si/Al and U₃Si₂/Al dispersion samples, the heats of formation of U(Al,Si)₃ as a function of the Si content were calculated. The heat of formation of (U,Mo)Al₃ becomes less negative as the Mo content increases. Conversely, the heat of formation of U(Al,Si)₃ becomes more negative with increasing Si content.     

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.     

Leach Rates and Thermal Properties of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate (LIP) glasses loaded with a simulated high-level nuclear waste were studied on their leach rates and thermal properties.

The obtained results showed that the phosphate glass matrix consisting of lead monoxide, phosphorus pentoxide and ferric oxide of 56:35:9w/₀ is able to vitrify the waste, pretreated with formic acid to remove Zr, to about 15 w/₀ at 950°C. The leach rate of the vitrified waste glass was in the order of 10^?7 g/cm².d at 110°C, which is low compared with that of the borosilicate glass waste form. Increasing the phosphorus pentoxide content of the matrix to higher than 35% enabled it to produce the glass form with the waste near 20 w/₀ at 950°C, but this increase rendered the glass waste form more soluble than the former. Thermal properties such as thermal expansion coefficient, critical cooling rate for vitrification and temperatures of glass transition, softening and maximum rate of crystallization were measured and discussed.

Removing Na ions from wastes improves considerably both the leach rate and the thermal stability of the LIP glass waste form.     

Initial dissolution rate of a Japanese simulated high-level waste glass P0798 as a function of pH and temperature measured by using micro-channel flow-through test method

Aqueous dissolution tests were performed for a Japanese type of simulated high-level waste (HLW) glass P0798 by using a newly developed test method of micro-channel flow-through (MCFT) method, and the initial dissolution rate of glass matrix, r ₀, was measured as a function of solution pH (3–11) and temperature (25–90°C) precisely and consistently for systematic evaluation of the dissolution kinetics. The MCFT method using a micro-channel reactor with a coupon shaped glass specimen has the following features to provide precise and consistent data on the glass dissolution rate: (1) any controlled constant solution condition can be provided over the test duration; (2) the glass surface area actually reacting with solution can be determined accurately; and (3) direct and totally quantitative analyses of the reacted glass surface can be performed for confirming consistency of the test results. The present test results indicated that the r ₀ shows a “V-shaped” pH dependence with a minimum at around pH 6 at 25°C, but it changes to a “U-shaped” one with a flat bottom at neutral pH at elevated temperatures of up to 90°C. The present results also indicated that the r ₀ increases with temperature according to an Arrhenius law at any pH, and the apparent activation energy evaluated from Arrhenius relation increases with pH from 54 kJ/mol at pH 3 to 76 kJ/mol at pH 10, which suggests that the dissolution mechanism changes depending on pH.     

Experimental investigation of reaction behavior between carbon dioxide and liquid sodium

Reaction behavior of carbon dioxide (CO₂) with a liquid sodium pool was experimentally investigated to understand the consequences of boundary tube failure in a sodium-CO₂ heat exchanger. In this study, two kinds of experiments were carried out to investigate the reaction behavior.In one experiment, about 1-5 g of liquid sodium pool were poured into flowing CO₂ to obtain the information mainly about the thermo-chemical conditions to initiate the reaction and the chemical constituents of reaction products. During the experiment, visual observation was made using video-camera and the temperature change of the sodium pool and near the surface was measured by thermocouples. The experimental parameters were the sodium pool diameter, the initial temperature of sodium and CO₂, the CO₂ flow direction against pool surface, and the initial moisture concentration in CO₂. The solid products of sodium-CO₂ reaction were sampled and analyzed by X-ray diffraction (XRD), energy dispersion X-ray analysis (EDX), total organic carbon analysis (TOC), and chemical analysis. The reaction gas products were also sampled and analyzed by gas chromatography.In the other experiment, CO₂ was injected into about 200 g of liquid sodium pool to simulate the boundary failure in the sodium-CO₂ heat exchanger. The CO₂ was fed through a helical coil-type tube dipped into the pool to adjust the temperature to the sodium pool temperature, and injected upward into the pool from a pool bottom using a nozzle attached at the end-side of the tube. The experimental parameters were the initial temperature of sodium, the diameter of the nozzle, the flow rate and the injection time of CO₂. The temperature change of sodium pool and the cover gas was measured by thermocouples during the experiment, and the reaction products were sampled and analyzed by the same manner as in the former experiments after the experiment.From these experiments, it became clear that the exothermic reaction occurred above a threshold temperature, and useful and indispensable information such as the resulting temperature and pressure rise and the behavior of solid reaction products in the pool was obtained to evaluate the consequences of boundary tube failure incident in a sodium-CO₂ heat exchanger.     

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.     

Characterization of monazite glass-ceramics as wasteform for simulated α-HLLW

Two monazite glass-ceramic wasteforms were sintered by mixing the lanthanum metaphosphate glass powder with the oxide powder of the components in simulated α-HLWs. The co-existence of components Al and Mo in an iron phosphate melt separated the melt into two immiscible glass melts, namely aluminum iron phosphate glass (Gb) and molybdenum iron phosphate glass (Gg). 24 wt% of ZrO₂, together with P₂O₅ and proper amounts of Fe and Mo formed a zirconium pyrophosphate glass (Gg1), which was immiscible with the phase Gg. The iron ions in the wasteforms were all in Fe³⁺, 1/3 of which was in 4-fold coordination. The O/P and O/(P + 1/3Fe³⁺) ratios for the glass phases were Gg1 3.70, Gb 3.89-3.98, Gg 4.23-4.25, and Gg1 3.58, Gb 3.47-3.42, Gg 3.74-3.69, respectively. The dissolution rates of two wasteforms were 0.3008 and 0.2598 g/m²d, respectively.     

Vitrification of nuclear waste with low heat generation

Analyses of thermal processes in the glass melter and storage container are given for vitrification of defense waste.     

Analysis of Leaching Behavior of Simulated LILW Glasses by Using the MCC-1 Test Method

Glasses developed for the treatment of low- and intermediate-level radioactive waste (LILW) from nuclear power plants were evaluated by using the Material Characterization Center-1 (MCC-1) leaching method. Tests were conducted at temperatures of 40, 70, and 90°C for three weeks in pH buffer solutions spanning the range from pH 4 to pH 11. Normalized mass losses and forward dissolution rates of major glass elements (B, Na, Al, Si, Co, Cs) were analyzed under each leaching condition. From these data, the forward rate equations depending on pH and temperature were defined using a nonlinear regression method. This equation provided an overall diagram of the leach rate with these parameters (i.e., pH and temperature). The forward dissolution rates of the glasses were found to have a V-shaped pH dependence. The glasses in the pH ranges were found to have a forward dissolution rate below 10 g/m²·d, when the temperatures were between 40 and 90°C and the leachant condition was pH 4–11. Except for the DG2 glass, the minimum forward dissolution rate (0.01–1 g/m²·d) was obtained at approximately pH 7–8. Compared with previously reported results, the developed glasses showed relatively high forward dissolution rates at the neutral region, while showing similar or lower rates compared with other glasses and ceramic waste forms at both extremes of pH.     

Leach Rates of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate glasses loaded with simulated high-level nuclear wastes at temperatures between 900 and 1,100°C were studied on their soaking behavior in distilled water by means of leachate solution analysis.

The obtained results showed that the leach rates of the glass waste forms were at least 10 to 100 times lower than that of the currently investigated borosilicate glass, even though the selective release of Na ion from the forms was observed. Zirconium of the waste led the glass to partial crystallization at 900°C, but was able to be incorporated in the glass at near 1,100°C.

The liquid chromatographic analysis of poly-phosphate ions in the leachate solution revealed that the low leachability of the glass forms was brought about by a certain degree of depoly-merization of long poly-phosphate chains of lead metaphosphate caused by the addition of ferric oxide.     

Sequential Cation-Exchange Separation of Molybdenum-99 and Tellurium-132 from Neutron Irradiated U3O8

The method of sequential cation-exchange separation of fission products proposed by Natsume et al. was applied to the preparation of fission-produced ⁹⁹Mo and ¹³²Te, with particular attention paid to increasing the recovery of ⁹⁹Mo and ¹³²Te, and to reducing contamination with ⁹⁵Zr-⁹⁵Nb and ¹⁰³Ru. The cation-exchange behavior of these nuclides was found to be influenced by the particle size of the target U₃O₈ powder, the method of dissolution, the standing time allowed between dissolution and separation, and the quantity of uranyl ion treated in one batch. In order to enhance the distribution of ¹³²Te in the Te fraction, and to reduce the contamination of the Mo and Te fractions with ⁹⁵Zr- ⁹⁵Nb and ¹⁰³Ru, the ion-exchange separation should be applied immediately after dissolution of the U₃O_s in nitric acid and upon treatment of the solution with concentrated HCl. Relatively coarse particles of U₃O_s were found more suitable for the present purpose of preparing ¹³²Te. Batches of U₃0₈ smaller than about 0.5 g proved to result in better separation of ⁹⁹Mo and ¹³²Te, for a column bed volume of 25 ml.