Inhibition of Nuclear Waste Glass Leaching by Chemisorption

The durability of a simulated nuclear waste containment glass was investigated by static leaching tests in six types of leaching containers. Glass corrosion in lead containers was reduced by more than two orders of magnitude from that measured in Teflon, whereas aluminum containers resulted in nearly a one order of magnitude reduction. Relatively minor effects on glass leaching were measured in copper, tin, and titanium containers. The results have strong implications concerning the choice of materials in possible waste package designs.     

Crystallization of Nuclear Waste Disposal Glass

The rate of bulk crystallization and shear viscosity η of a zinc borosilicate simulated nuclear waste disposal glass (SNWDG) were measured above the glass transition temperature T_g . Using recent theories and experimental data for the temperature dependence of η at < T_g , the crystallization rate below T_g was calculated; well below T_g the SNWDG (and, more generally, any high-silica glass) will not crystallize appreciably over any important time period for radioactive decay of fission products.     

Volatility of Simulated High-Level Nuclear Waste Glass by Thermogravimetric Analysis

The volatilities of simulated, high-level nuclear waste glasses have been measured using thermogravimetric analysis (TGA). These volatilities were measured in the region of the glass transition temperature ( T _g) of the waste glasses, which is between 450° and 500°C. These data were obtained because the Waste Acceptance Preliminary Specifications require that no foreign materials be released into the canistered waste form upon heating of the canister to this glass transition temperature. In fact, all of the waste glass samples studied actually exhibited a net weight gain upon heating. This weight gain was shown to be due to oxygen uptake through oxidation of FeO. Powdered glass samples did show a small weight loss which was smaller in magnitude than the weight gain and was associated with water desorption. No true volatility was detected to the level of sensitivity (0.01 wt%) of the TGA instrument. This converts to a sensitivity of 330 μ g/m² of glass surface and a corresponding minimum value of 41 mg of volatiles for each Defense Waste Processing Facility conistered waste form. TGA experiments carried out at higher temperatures (800°C) revealed that organic concentrations in the waste glasses are less than 0.01 wt%. Thus, these results demonstrate that the Defense Waste Processing Facility will be able to comply with the Waste Acceptance Preliminary Specifications on the exclusion of foreign materials from the canistered waste forms, after exposure to T _g.     

Redox Characterization of Simulated Nuclear Waste Glass

Ferrous/ferric equilibria were determined in an alkali-iron borosilicate glass of the type that is presently being considered for nuclear waste disposal. Theoretical redox behavior is found at conditions more reducing than 1 kPa. Deviation from this behavior at more oxidizing conditions results from mutual interactions between the iron redox couple and other multivalentions in the glass. The results suggest that the ferrous/ferric ratio can be used as a quantitative indicator of the glass oxidation state in these complex glasses. Implications regarding redox control of the vitrification process are discussed.     

Kinetics of Surface-Layer Formation on Simulated Nuclear Waste Glass

Waste glass immersed in water at high temperatures and pressures rapidly develops a surface layer. The growth of this layer was followed semi quantitatively using an optical interference technique. Growth rates were measured as a function of temperature and pressure in distilled water and 1% NaCl solution. Pressures of 25 to 85 × 10³ Pa had no measurable effect at 200°C on glass immersed in water. At a pressure of 25 × 10⁵ Pa the growth rate increased with temperature in both water and salt solution. The rates were described in terms of activation energies (83.4 kj/mol for water and 53.0 for salt solution) with approximately equal rates in both solutions at 240°C. The rates of growth were linear in all cases. The surface layer began to exhibit cracks when its thickness exceeded =2.5 μm and it eventually spalled.     

Sodium Diffusion and Leaching of Simulated Nuclear Waste Glass

Oxides such as CaO, ZnO, Fe₂0O₃, and Zr0O₂ were substituted for Si0O₂ into a 14 Na₂0-10 B₂0O₃-76 Si0O₂ (wt%) glass in proportion to their weight percent in PNL-7668 simulated nuclear waste glass. No correlation was found between the Na diffusion coefficient (D_Na) in the bulk glass and sodium leaching in water for any of the iron-containing borosilicate glasses. Effective diffusion coefficients calculated from the sodium leach rates in deionized water were 500 to 2000 times larger than D_Na. The experimental leaching data agreed with a diffusion-convection model where the convective term was a measured "phase-boundary" velocity moving into the glass.     

Effects of α-Radiolysis on Leaching of a Nuclear Waste Glass

The effects of α-radiolysis on leaching were studied in compositionally identical Pu-doped waste glass samples which were prepared at different α-activities (dose rates) by varying the ²³⁹Pu/²³⁸Pu isotopic ratio in the Pu dopant. The results of 14-, 28-, and 56-d static leach tests at 40°C, in both deionized water and brine solution, indicate that α-radiolysis at the samplelsolution interface increased Pu and Si release by a factor of ≅3.     

Gamma Radiation-Induced Changes in Trombay Nuclear Waste Glass Containing Iron

Alkali-based barium borosilicate glasses having similar composition to the Trombay nuclear research reactor waste base glass containing iron were prepared and irradiated by gamma rays. The radiation-induced defect centers were investigated using electron paramagnetic resonance (EPR) technique. The results showed the formation of silicon hole centers and electron trap centers apart from boron-based oxy hole centers in the glass due to the irradiation. The EPR Hamiltonian parameters for these radicals were evaluated by simulation technique using Bruker SIMFONIA computer program. Glasses with varying iron content were irradiated with the same dose to evaluate the effect of iron content on the defect center concentration. A spin counting technique was employed to evaluate the number of defect centers produced in each glass. It was observed that the defect center concentration reduced as a function of increasing iron content. This was attributed to the charge trapping process of ferric ions in the amorphous system.     

The Interaction Between Nuclear Waste Glass and Ordinary Portland Cement

The interaction between simulated reference waste glasses SON68 and SM539, and cement has been studied in suspensions of ordinary Portland cement and synthetic cement water with pH 13.5 at 30°C. The cement appears to trigger glass dissolution by consumption of glass matrix components. This leads to fast glass dissolution at a constant rate with formation of a porous gel layer on the glass. A glass dissolution model has been proposed considering that the silicon coming from the glass reacts with portlandite to form C-S-H phases. The transformation of C-S-H into C-A-S-H phases is a second parallel driving force especially for the Al-rich SM539 glass. After consumption of the portlandite, the glass alteration rate is expected to decrease.     

Investigation of Foaming in Liquid-Fed Melting of Simulated Nuclear Waste Glass

Melt foaming in alkali-iron borosilicate glasses considered for nuclear waste disposal was investigated to identify the causative reactions and stabilizing mechanisms. Foaming is attributed to the release of oxygen and water vapor from the melt. The oxygen is suggested to originate from oxidation-reduction reactions of species other than iron. Foaming problems are largely solved under reducing conditions, which result in less stable foams. The results suggest that a greater proportion of water vapor is released into the foam under more reducing conditions. The water destabilizes the foam by viscosity or surface tension effects and causes the bubbles to rupture on the melt surface.     

Dissolution of a Nuclear Waste Ceramic: An Experimental and Modeling Study

The degree of alteration of a multiphase ceramic waste form (Synroc-D) exposed to flowing water at 75°C is reported. After 117 days, a 25- to 30-μm deep leach zone developed from which all silicate phases have congruently dissolved. The associated spinel and titanate phases appear unaltered, whereas an Al(OH)₃ precipitate ∼ 5-μm thick develops at the sample surface. The dissolution of silicates and deposition of the precipitate are simulated using a modification of the crackling-core model. The model predictions closely match the observed behavior of the material. Implications of these results for long-term stability of the waste form are discussed.     

利用回收平板废玻璃制备泡沫玻璃的研究

泡沫玻璃是一种新型的环境友好型多孔材料,其性能与气孔结构息息相关.本文利用回收的平板玻璃,通过引入石英砂,添加碳酸钙发泡剂,制备了泡沫玻璃,同时对气孔调控进行了研究.研究结果表明,随着碳酸钙发泡剂质量分数的增加,泡沫玻璃的孔径增大,开气孔数量增加,总气孔率增大,泡沫玻璃的保温性能良好.但当发泡剂添加量进一步增加时,泡沫玻璃孔壁变薄,相邻气孔易融合塌陷,致使抗压强度降低.通过引入石英砂提高发泡时熔体的空间骨架强度,可以调节泡沫玻璃中开气孔、闭气孔比例,提升抗压强度,满足不同领域对泡沫玻璃的性能要求,进而拓宽泡沫玻璃应用范围.     

Sodium Diffusion in the Nuclear Waste Glass GP 98/12

The diffusion of²² Na was measured in the nuclear waste borosilicate glass GP 98/12 in the temperature range of 350° to 450°C. The boundary condition of a thin initial tracer layer and serial sectioning were employed to measure the diffusion profiles. The activation enthalpy was found to be 1.08 eV. A parallel experiment with the base glass VG 98/12 not containing added (chemically simulated, inactive) waste showed that Na diffusion does not vary significantly when adding the waste.     

废玻璃烧结法制备多孔玻璃

本文采用回收的废玻璃在高于玻璃化转变温度下短时间烧结制备出具有一定孔隙度和强度的过滤用多孔玻璃.废玻璃破碎成细颗粒后过筛,并在32 MPa的压力下模压成型,压坯在640℃到670℃的温度区间内烧结.实验结果表明,试样在670℃烧结30 min可以获得合适的抗弯强度(12.84 MPa)和开孔率(29.4％),此时的多孔玻璃具有明显的烧结颈,平均孔径为10 μm左右,平均氮气通量和水通量分别为13600 m3 m-2h-1bar-1和693.4 m3 m-2h-1 bar-1.     

Stabilizing Glass Bonded Waste Forms Containing Fission Products Separated from Spent Nuclear Fuel

Abstract

A model has been developed to represent the stresses developed when a molten, glass-bonded brittle cylinder (used to store nuclear material) is cooled from high temperature to working temperature. Large diameter solid cylinders are formed by heating glass or glass-bonded mixtures (mixed with nuclear waste) to high temperature (915°C). These cylinders must be cooled as the final step in preparing them for storage. Fast cooling time is desirable for production; however, if cooling is too fast, the cylinder can crack into many pieces. To demonstrate the capability of the model, cooling rate cracking data were obtained on small diameter (7.8 cm diameter) glass-only cylinders. The model and experimental data were combined to determine the critical cooling rate which separates the non-cracking stable glass region from the cracked, non-stable glass regime. Although the data have been obtained so far only on small glass-only cylinders, the data and model were used to extrapolate the critical-cooling rates for large diameter ceramic waste form (CWF) cylinders. The extrapolation estimates long term cooling requirements. While a 52-cm diameter cylinder (EBR-II-waste size) can be cooled to 100°C in 70 hours without cracking, a 181.5-cm diameter cylinder (LWR waste size) requires 35 days to cool to 100°C. These cooling times are long enough that verification of these estimates are required so additional experiments are planned on both glass only and CWF material.     

玻璃熔窑余热发电技术开发和设计应用

玻璃熔窑余热发电可以高效利用玻璃生产中的烟气余热,减少玻璃厂对环境的热污染以及粉尘污染,同时将电能回用于生产,给企业带来巨大的经济效益。     

Effect of Alumina Source on the Rate of Melting Demonstrated with Nuclear Waste Glass Batch

The selection of raw materials affects the rate of batch-to-glass conversion. In all-electric melters, foam under the batch blanket limits the heat flux from the molten glass, thus slowing the rate of melting. Our study, in which we compare the melting behaviors of three batches formulated to vitrify high-alumina high-level waste, shows that a slowly dissolving refractory component can cause excessive foaming. Faster melting batches with gibbsite [Al(OH)₃] or boehmite [AlO(OH)] as an alumina source produced substantially less foaming than a batch with corundum (Al₂O₃). Although gibbsite and boehmite dissolved below 500°C, corundum was still present in the batch up to 900°C; hence, the glass-forming melt lacked alumina in the batch with corundum. The low viscosity of that batch caused the open pores to close prematurely at 660°C, trapping gases and expanding to foam. This would explain the literature-reported slow melting rate of a batch with corundum, as compared to batches with gibbsite and boehmite.     

废旧建筑玻璃制备泡沫玻璃的研究

利用废旧建筑玻璃为原料,以碳酸钙为发泡剂、硼酸为助溶剂,采用模压成型、高温发泡工艺成功制备出了泡沫玻璃,研究了发泡剂含量与发泡温度对泡沫玻璃的体积膨胀率、体积密度、气孔率、抗折强度等性能的影响.结果 表明:在800 ～840℃的发泡温度范围内,随发泡剂含量的增加,泡沫玻璃的体积膨胀率与气孔率先增大后降低,而体积密度和抗折强度先降低后增大,在发泡剂含量为3％,发泡温度在820℃时,泡沫玻璃综合性能良好,其体积膨胀率达476％,体积密度为0.21 g/cm3,气孔率达87.26％,抗折强度为(1.38 ±+0.15) MPa,孔径介于1.7 ～2.2 mm之间.     

利用废显像管研制泡沫玻璃

介绍了以度显像管玻璃为主要原料制备泡沫玻璃，通过对泡沫玻璃气泡结构的观察和性能测试，分析了发泡剂的作用机理，探讨发泡剂和硼砂对气泡结构和性能的影响，得出了较佳泡沫玻璃配方。     

废弃玻璃引发混凝土碱-硅反应风险

废弃玻璃在当今城镇固体废弃物中占有很大比例,大量的堆积引起了一系列的环境和资源浪费问题,而将玻璃用作混凝土原材料可解决废弃玻璃的回收问题。部分发达国家针对玻璃混凝土已开展过大量相关研究,并取得可用于实践的研究成果。随着中国经济的快速发展,开发玻璃混凝土技术是必由之路。通过概括当前废弃玻璃混凝土的主要研究成果,分析废弃玻璃应用于混凝土生产引发的碱-硅反应(ASR)风险,并论述ASR膨胀的影响因素及控制措施。