钕在玻璃-陶瓷各相中的分布研究

乏燃料循环过程中，不可避免地产生含钚废物。钚在玻璃中的溶解度非常低，需研制新型固化基材——玻璃一陶瓷来固化含钚废物。     

硼硅酸盐玻璃中硫存在形态的XANES方法研究

硼硅酸盐玻璃固化高放废物是目前国际公认的较好方法，采用这种方法所产生的玻璃固化体具有良好的化学耐久性，熔制温度在可接受范围。但是很多放射性废物中，包括我国的高放废液，含有一定量的硫，由于硫在硼硅酸盐熔体中溶解度较低，使得硫常常成为废物包容量的限制因素。因此提高玻璃中硫包容量的方法对于含高浓硫酸盐的高放废液玻璃固化来说是非常重要的。     

高放废液玻璃固化玻璃配方及固化玻璃性能研究

介绍了由三元体系到多元体系的硼硅酸盐基础玻璃,并用基础玻璃包熔废物氧化物。在包熔无硫酸盐的废物氧化物时,在多种配方中优选出704~#配方,包熔量达(20～30)％Wt。在研究含SO_4~(2-)的高放废液时,用3种硼硅酸盐基础玻璃包熔。分析结果表明,固化玻璃中SO_3含量为(0.3～0.6)％wt,加入的SO_4~(2-),除部分进入固化玻璃和挥发到尾气中以外,其余则以黄相形式出现在玻璃表面。黄相将浓集放射性元素,严重影响产品固化玻璃质量。本试验详细研究了黄相的组成和生成因素,并详细介绍了3种固化玻璃的综合性能。试验证明,704~#固化玻璃性能最佳,被部定为罐式玻璃固化工程规模试验材料。     

高放废液玻璃固化玻璃配方及固化玻璃性能研究

介绍了由三元体系到多元体系的硼硅酸盐基础玻璃,并用基础玻璃包熔废物氧化物。在包熔无硫酸盐的废物氧化物时,在多种配方中优选出704~#配方,包熔量达(20～30)％wt。在研究含SO_4~(2-)的高放废液时,用3种硼硅酸盐基础玻璃包熔。分析结果表明,固化玻璃中SO_3含量为(0.3～0.6)％wt,加入的SO_4~(2-),除部分进入固化玻璃和挥发到尾气中以外,其余则以黄相形式出现在玻璃表面。黄相将浓集放射性元素,严重影响产品固化玻璃质量。本试验详 细研究了黄相的组成和生成因素,并详细介绍了3种固化玻璃的综合性能。试验证明,704~#固化玻璃性能最佳,被部定为罐式玻璃固化工程规模试验材料。     

高硅玻璃陶瓷对模拟高放废液中锕系核素的固化性能测试

为了提高玻璃固化体性能,将玻璃质的Si O2在1 200℃条件下烧结制备成高硅玻璃陶瓷固化体。测试结果表明:高硅玻璃陶瓷固化体的密度、包容量及浸出率等性能均优于硼硅酸盐玻璃固化体。     

高放核废料磷酸盐玻璃陶瓷固化研究进展

磷酸盐玻璃陶瓷是固化“难溶”核废料的优异基材,具有高的废料包容量和优异的稳定性,因而,磷酸盐玻璃陶瓷固化是高放核废料固化处理的重要研究方向之一。本文简要综述了高放核废料磷酸盐玻璃陶瓷固化体的类型、固化机理、固化体设计、稳定性及其制备,并对其研究做了展望。其今后研究方向主要包括：（1） 磷酸盐玻璃固化体的中长期化学稳定性、蚀变规律和抗腐蚀机制的研究;关注其物理性能、热稳定性和辐照稳定性;（2） 磷酸盐玻璃陶瓷固化体的简洁制备工艺技术及其工艺原理,及其对设备和电极的侵蚀和寿命的影响。     

模拟高放废物玻璃固化体在处置条件下的浸出行为研究（Ⅰ）

用模拟高放废物硼硅酸盐玻璃固化体和介质（包括膨润土、凝灰岩、沸石、氧化铁粉、去离子水和模拟地下水）构成模拟处置条件下的９个浸泡体系，研究了在有介质存在条件下，玻璃固化体浸泡后的失重，玻璃体的元素浸出和浸出液的ｐＨ值变化；研究了温度和ｐＨ对浸出的影响，求出了玻璃、水反应的表观活化能为７３．０ＫＪ／ｍｏｌ。对高放废物处置库的回填材料的选择提供了优选方案。     

SiO2和B2O3含量对模拟高放废物玻璃结构的影响

硼硅酸盐玻璃具有特别高的化学稳定性、较好的热稳定性、较大的放射性废物包容量等优点,被广泛作为固化高放废液的基础玻璃料。废物玻璃固化体的结构与其组成存在一定的内在依存关系,它将对废物玻璃的性质产生影响。以组分含量作为变量,所引起的废物玻璃固化体的某些结构特征的变化,是探索影响高放废物玻璃固化体性能的内在线索。为了获取废物玻璃组分含量对其结构特征参数（非桥氧键数）的影响规律,以对废物玻璃固化体的性能进行估计,进而指导设计高效处理高放废液的玻璃固化体配方,以某模拟高放废物玻璃固化体的配方为基础,通过改变其中SiO₂和B₂O₃含量制备一系列的废物玻璃样品,并进一步采用拉曼光谱法对废物玻璃的硅酸盐网络结构单元（Qn,Q和n分别代表四面体单元和每个四面体结构单元所具有的桥氧键数）进行分析,讨论SiO₂和B₂O₃含量对此废物玻璃固化体网络结构的影响。结果表明：随着SiO₂含量（SiO₂的质量与剩余氧化物质量的比值）由0.721增加到1.037,废物玻璃固化体中桥氧键的比例随之而增多,同时玻璃网络结构的聚合度（N,非桥氧/桥氧的比例）减小,但是,随着B₂O₃的含量（B₂O₃的质量与剩余氧化物质量的比值）由0.117增加到0.143,再到0.176,桥氧键的比例先减小后增大。在本实验的范围内,废物玻璃固化体的密度变化不明显,无析晶现象。     

高放废物玻璃固化过程中玻璃组成对硫酸盐溶解度的影响

硼硅酸盐玻璃固化的高放废物固化体能进行长期安全存储是已为国际所公认。然而，对于含有较高浓度硫酸根的高放废液，熔制过程中会产生分离黄色第二相（简称黄相），这是一种易溶于水的结晶物质。分析表明，玻璃固化体黄相含有碱金属和碱土金属的硫酸盐、铬酸盐和钼酸盐，并有一定量的铯、锶等裂片元素。玻璃固化体在深地质处置后，一旦受到地下水侵蚀，这些核素易浸泡出来，进入生物圈，因此，它严重危害玻璃固化体包容和隔离核素的作用，这是必须克服和避免的。     

ZnO和CaO对模拟高放废液硅酸盐玻璃固化体性能的影响研究

针对有些高放废液含有较多Fe、Cr、Ni过渡金属元素,在玻璃固化工艺过程中易于形成晶体,导致熔融玻璃体的黏度增加、化学稳定性变差以及工艺过程中易出现出料口堵塞等问题,研究了废物包容量为15%和20%、添加ZnO(5.6%)和CaO(1.75%)的配方对形成的4种玻璃固化体的物理性能(密度、硬度、断裂韧性)、化学性能(产品一致性测试和蒸汽腐蚀测试)和结构(X射线衍射析晶分析、拉曼光谱分析)的影响。研究分析显示,提高废物包容量至20%以及添加ZnO和CaO均可促进硼硅酸盐玻璃固化体网络结构的稳定性和化学稳定性,并增强玻璃体的密度,提高硬度;但玻璃固化体的高温黏度升高,断裂韧性下降。     

Ultraphosphate glasses for immobilization of liquid plutonium-containing wastes

Ultraphosphate glass is proposed for immobilization, taking account of the composition of the plutonium-containing wastes at the Mining-Chemical conglomerate. Some parameters of the process of immobilizing in ultraphosphate glass model wastes with plutonium content up to 150 mg/liter are investigated. A preliminary study is made of the degree of inclusion of plutonium dioxide, the chemical stability of the ultraphosphate glass obtained, and the distribution of plutonium in it. It is shown that almost 2 times more plutonium dioxide is included in ultraphosphate glass than in phosphate glass. With respect to the main physicochemical parameters, specifically, component leachability, ultraphosphate glass can be regarded as promising for immobilizing liquid plutonium-containing wastes. __________ Translated from Atomnaya énergiya, Vol. 100, No. 5, pp. 368–372, May, 2006.     

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.     

Properties of phosphate glass waste forms containing fluorides from a molten salt reactor

Radioactive fluoride wastes are generated during the operation of molten salt reactors(MSRs) and reprocessing of their spent fuel.Immobilization of these wastes in borosilicate glass is not feasible because of the very low solubility of fluorides in this host.Alternative candidates are thus an active topic of research including phosphatebased glasses,crystalline ceramics,and hybrid glass-ceramic systems.In this study,mixed fluorides were employed as simulated MSRs waste and incorporated into sodium aluminophosphate glass to obtain phosphate-based waste form.These waste forms were characterized by X-ray diffraction,Raman spectroscopy,and scanning electron microscopy.Leaching tests were performed in deionized water using the product consistency test A method.This study demonstrates that up to 20 mol%of simulated radioactive waste can be introduced into the NaA1 P glass matrix,and the chemical durability is much better than that of borosilicate.The addition of Fe_2O_3 in the NaAlP glass matrix results in increases of the chemical durability at the expense of fluoride loading(to 6.4 mol%).Phosphate glass vitrification of radioactive waste containing fluorides is a potential method to treat and dispose of MSR wastes.     

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.     

Release of Neptunium from Neptunium-Doped Borosilicate Waste Glass

The release of neptunium from a neptunium-doped borosilicate waste glass was studied at 90°C in deionized water and silicate water. The standard MCC-1 static leach method was applied to the tests for durations up to 91 days with the SA/V ratio of 10 m^?1.

The normalized elemental mass loss obtained for neptunium was about 5 g/m² for both the deionized and the silicate water leachates. This value is similar to those for currently typical borosilicate waste glasses. That is, the studied glass is comparable with the typical glasses in terms of the ability to immobilize neptunium.

The time dependence of the release of neptunium from the glass was different from those of soluble glass components such as sodium, boron and cesium, but similar to that of strontium. A part of neptunium, like strontium, probably remained in the surface layer formed on the leached glass. The neptunium species in the surface layer was predicted to be NpO₂.xH₂O(am) based on available solubility data.     

Plutonium solubility and self-irradiation effects in borosilicate glass

This article presents the limitations for the immobilization of plutonium in borosilicate glasses. A first one is related to the solubility of this element in glass. The effects of the temperature and redox conditions during glass processing were studied. Glass specimens containing plutonium and plutonium surrogates are fabricated. The results show that trivalent elements (La, Gd, Nd, etc.) exhibit greater solubility than tetravalent elements (Pu, Th, Hf). Fabricating the plutonium-doped glass samples under reductive conditions reduced the Pu to trivalent oxidation state and increased its solubility to 4 wt% PuO₂ without reaching the solubility limit. A structural approach based on the results of EXAFS and NMR spectroscopy suggests that the structural role of the trivalent and tetravalent elements corresponds to that of intermediate network modifiers and intermediate network formers, respectively.

The second factor is the effect of actinide decay on the long-term behavior of the glass. Borosilicate glass samples were doped with different curium contents (0.05, 0.5, 1.5 and 4.1 wt% of CmO₂). The macroscopic properties (density, microhardness and initial dissolution rate) of the glasses were characterized up to 4 × 10¹⁸  g⁻¹. No significant effect on the initial alteration rate was detected. The glass swelled slightly, saturating at about 0.5% after receiving a dose of about 2 × 10¹⁸  g⁻¹.

Further studies are ongoing to confirm the satisfactory long-term behavior of the borosilicate glass matrix at higher doses, and to determine the solubility limit of plutonium in reducing conditions.     

Vitrification and testing of a Hanford high-level waste sample. Part 1: Glass fabrication, and chemical and radiochemical analysis

The Hanford radioactive tank waste will be separated into low-activity waste and high-level waste that will both be vitrified into borosilicate glasses. To demonstrate the feasibility of vitrification and the durability of the high-level waste glass, a high-level waste sample from Tank AZ-101 was processed to glass, analyzed with respect to chemical composition, radionuclide content, waste loading, and the presence of crystalline phases and then tested for leachability. The glass was analyzed with inductively coupled plasma-atomic emission spectroscopy, inductively coupled plasma-mass spectrometry, γ-energy spectrometry, α-spectrometry, and liquid scintillation counting. The WISE Uranium Project calculator was used to calculate the main sources of radioactivity to the year 3115. The observed crystallinity and the results of leachability testing of the glass will be reported in Part 2 of this paper.     

钙含量对钙钛锆石-钡硼硅酸盐玻璃陶瓷结构的影响

采用熔融-热处理工艺制备SiO₂-Na₂O-B₂O₃-BaO-CaO-TiO₂-ZrO₂体系玻璃陶瓷,利用差热分析法（DTA）、傅氏转换红外线光谱分析仪（FTIR）、X射线衍射仪（XRD）、扫描电子显微镜（SEM）等技术手段研究了晶核剂（CaO、TiO₂和ZrO₂）总含量为45%时,不同钙含量（CaO∶TiO₂∶ZrO₂=x∶2∶1（摩尔比）,x=0.5～6）对玻璃陶瓷晶相和显微结构的影响,并采用粉末静态浸泡法（PCT）测试了C₂（即x=2）玻璃陶瓷样品的抗浸出性能。结果表明：玻璃网络结构主要由［SiO₄］、［BO₃］和［BO₄］构成,随着Ca含量的增加,更多的B以［BO₄］加入玻璃网络中,玻璃转变温度T_g逐渐升高,放热峰温度逐渐降低,但峰强逐渐增高;x＜2时,样品中除CaZrTi₂O₇晶相外还有其他晶相（如TiO₂和ZrO₂）出现;当x=2和4时,样品中只有单一的CaZrTi₂O₇晶相;x=6时,有星状的CaZrTi₂O₇和柱状的CaTiO₃晶相生成。PCT实验结果表明：B、Na、Ca元素的归一化浸出率随浸泡时间的增加而降低,并在28 d后保持不变,分别为8.4、7.8、2.2 mg/(m²•d),与硼硅酸盐玻璃固化体浸出率处于同一数量级。