氧化硼对铁磷酸盐玻璃陶瓷固化体的影响

研究了不同B₂O₃掺量对铁磷酸盐玻璃陶瓷高放废物固化体结构和性能的影响。应用溶出速率法（DR）对固化体进行了化学稳定性测试,使用傅里叶变换红外光谱（FTIR）和X射线衍射（XRD）方法研究了样品的结构。研究结果表明：玻璃陶瓷固化体的主晶相为独居石;B₂O₃的引入对玻璃陶瓷固化体的化学稳定性影响较大,以10%（摩尔分数）的B₂O₃代替Fe₂O₃制得的固化体化学稳定性最佳,其28d的质量浸出率约为7.81×10^-9g•cm^-2•min^-1;试样中存在大量正磷酸基团［PO₄］^3-和少量焦磷酸基团［P₂O₇］^4-,无偏磷酸基团［PO₃］^-存在,固化体中的B主要以［BO₄］四面体基团形式存在。     

含Pu废物的玻璃和玻璃陶瓷固化基材研究进展

对于239Pu含量较高且很难回收利用的含Pu废物,在安全处置前须进行妥善的固化处理。玻璃和玻璃陶瓷因在制备方面具有较陶瓷简单的工艺、低廉的成本和高效的产出被认为是目前处理含Pu废物综合优势明显的固化基材,因而得到了广泛和深入的研究。本文对碱硼硅酸盐玻璃、镧硼硅酸盐玻璃、铁磷酸盐玻璃以及含钙钛锆石、烧绿石或独居石结晶相的玻璃陶瓷等在含Pu废物固化方面的研究进展进行了综述,包括其组分、Pu包容量和化学稳定性,并进行了对比分析,认为在对玻璃固化基材继续研究与应用的基础上,玻璃陶瓷有望成为固化绝大多数含Pu废物的较佳选择。     

高放废液玻璃固化贵金属沉积研究进展

动力堆乏燃料后处理产生的高放废液因富含贵金属,在玻璃固化时可能发生贵金属沉积,从而造成出料口堵塞并导致熔炉底部电极的损坏。本文广泛调研了国内外高放废液玻璃固化中贵金属的沉积问题,对相关解决方案进行了总结,主要包括熔炉结构的改进与贵金属分离回收。高燃耗的乏燃料贵金属含量高,须在研究贵金属沉积行为的基础上,综合考虑以上两种解决方案。相关结果可为国内高放废液玻璃固化的运行提供一定的参考。     

高放废物玻璃固化与人造岩石固化技术的研究进展

高放废物的处理和处置是世界各核能国家面临的重大挑战。高放废物处理和处置的技术路线是先将其固化,再将其深埋。高放废物的固化有玻璃固化及人造岩石固化两种。玻璃固化已发展成一项成熟的技术,人造岩石固化尚在研发中。由于人造岩石固化工艺更简单,固化体抗浸出性能更优,稳定性更好,已引起世界各核能国家的关注,有可能取代玻璃固化而成为新一代固化技术。本文对玻璃固化和人造岩石固化的类型、机理和优缺点进行了系统的分析,对人造岩石固化的发展方向提出了建议。     

Al2O3对独居石玻璃陶瓷固化体的影响

本文研究了Al₂O₃掺量对独居石玻璃陶瓷固化体结构和化学稳定性的影响。用傅里叶变换红外光谱（FTIR）和X射线衍射（XRD）方法表征样品结构,用溶解速率法和全谱直读等离子体发射光谱（ICP-OES）分别测定样品在浸出液中浸泡后的失重速率及各元素的浸出浓度,以研究固化体的化学稳定性。研究结果表明：当Al₂O₃掺量为4%（摩尔分数）时,在980 ℃下保温3 h得到的独居石玻璃陶瓷固化体具有较高的化学稳定性,浸泡14 d时其质量浸出率最低,约为8.1 ng/(cm²•min),其中Ce、La元素在浸出液中均未检出;固化体的主晶相为独居石,结构中含有大量稳定的正磷酸基团［PO₄］^3-和少量的焦磷酸基团［P₂O₇］^4-,不存在偏磷酸基团［PO₃］^-。     

模拟含锶废物铁磷酸盐玻璃固化体的化学稳定性

针对我国高放废液全分离流程中产出的锶废物组成特点,设计了用铁磷酸盐玻璃固化锶废物的配方。用红外光谱（IR）研究了玻璃固化体的结构,用Product Consistency Test(PCT）试验方法研究了玻璃固化体的化学稳定性。研究表明,在所选的配方组成范围内,所熔制的玻璃固化体均有较好的化学稳定性。当配料中模拟含锶废物的含量为24～28%(wt)、FeO3的含量大于24%(wt)、O/P（氧磷摩尔比）为3.5～3.6时,玻璃固化体的化学稳定性最好。     

高放废液和锕系核素玻璃-陶瓷固化技术研究简况

高放废液在20世纪50年代出现后，人们首先想到的是将核素固定在晶体内。美国的阿贡实验室将高放废液用流化床煅烧成粉末，英国将放射性的铯交换到黏土上，加拿大则是将高放废物在1350℃熔融，制成霞石，法国则是在1300℃下制备云母，目的是将铯、锶固定在云母的晶体内，     

日本高放废液玻璃固化技术

本文介绍了日本东海后处理厂和六所村后处理厂的玻璃固化设施和运行情况,以及针对当前玻璃固化中出现的熔炉底部出料口堵塞、乏燃料储存池泄露等问题,日本在固化设施、运行方式、原料组分等方面采取的措施。在未来的研究计划中,日本将侧重于低放废物玻璃固化研究的标准制定,冷坩埚以及等离子体固化等先进技术的研发,同时加强配方研究及先进工艺流程的开发,从而实现玻璃固化的减容性和经济性,其经验和教训对我国玻璃固化的发展有重要的借鉴意义。     

美国高放废液玻璃固化概述

从美国高放废液的储量及特点、设施运行经验及教训等方面进行阐述,并从中得到启示,为我国高放废液玻璃固化的发展提供建议。     

玻璃-陶瓷固化工艺研究

随着核能的发展，乏燃料日益增多；核燃料燃耗逐渐增加，锕系核素的量会逐渐增大。锕系核素在玻璃中的溶解度很低，将锕系核素从高放废液中分离出来再进行人造岩石固化技术难度大，有必要开发高放废液和长寿命核素的处理新技术。本研究的目的是利用玻璃固化工艺制备玻璃一陶瓷，用结晶相固定锕系核素，玻璃相固定裂片核素。当高温玻璃熔融体被浇注到玻璃产品罐时，由于玻璃量大，降温速度很慢，这为晶核的形成、晶胞的生长提供了条件。     

Research on Solidification of High-Level Liquid Waste and Actinides by Glass-Ceramics

In the 1950s, the high-level liquid waste （HLLW） was intended to be immobilized in the crystals of synthetic minerals. HLLW was calcinated in the fluid bed into dry powder in the Argonne laboratory of USA and the cesium was absorbed onto the clays by ion exchanges in Brittan. The Canadian melted the mixture of HLLW and other additives at 1 350 ℃ to form nepheline. The French synthized mica at 1 300 ℃ to immobilize the cesium and strontium in the crystals and the rare earth elements in the layers of mica.     

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.     

A Thermoelectric-Conversion Power Supply System Using a Strontium Heat Source of High-Level Radioactive Nuclear Waste

A thermoelectric-conversion power supply system with radioactive strontium in high-level radioactive waste has been proposed. A combination of Alkali Metal Thermo-Electric Conversion (AMTEC) and a strontium fluoride heat source can provide a compact and long-lived power supply system. A heat source design with strontium fluoride pin bundles with Hastelloy cladding and intermediate copper has been proposed. This design has taken heat transportation into consideration, and, in this regard, the feasibility has been confirmed by a three-dimensional thermal analysis using Star-CD code. This power supply system with an electric output of 1MW can be arranged in a space of 50m² and approximately 1.1m height and can be operated for 15 years without refueling. This compact and long-lived power supply is suitable for powering sources for remote places and middle-sized ships. From the viewpoint of geological disposal of high-level waste, the proposed power supply system provides a financial base for strontiumcesium partitioning. That is, a combination of minor-actinide recycling and strontium-cesium partitioning can eliminate a large part of decay heat in high-level waste and thus can save much space for geological disposal.     

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.     

Formation of Precipitate in High-Level Liquid Waste from Nuclear Fuel Reprocessing

Stability of high-level liquid waste (HLW) from nuclear fuel reprocessing was studied by using a simulated HLW. Fundamental works disclosed that precipitates formed during aging at ambient temperature or refluxing the simulated HLW in 2 mol/lHNO₃ solution consist mainly of Mo, Zr and Te contributing significantly to the formation of precipitate. When the simulated HLW was denitrated with formic acid or deacidified with NaOH, fractions of precipitated Mo, Zr and Te increased with pH and amounted to over 85% at pH 0.5, where the fraction of precipitated La was below 0.1%. For further treatment of HLW such as partitioning, denitration of HLW to pH 0.5 might be useful for removing Mo, Zr and Te from the solution without significant contamination with rare earths, Am and Cm.     

铁元素对铝热剂SHS核废物固化体性能的影响

研究了铁元素对用自蔓延高温合成（SHS）技术固化处理的核废物固化体的组成结构及化学稳定性的影响。借助XRD、SEM和TG-DSC等测试手段对固化体的矿相组合、微观结构和热稳定性进行了测试分析,采用PCT浸泡法对固化体的抗浸出性能进行了研究。结果表明：不同配比的固化体矿相结构基本相同,矿相的主要组成为Al₂O₃,Fe元素对固化体的矿相结构无明显影响。固化体的微观结构非常致密,结晶良好,但固化产物的气孔数量及孔径随Fe元素含量的增大而增大。在小于1400 ℃时,固化体无明显失重,具有良好的热稳定性。固化体中示踪核素Ce的28 d浸出速率为10^-5～10^-6 g/(m²•d),固化体中Fe的含量在1%~10%范围内波动时,其对Ce、Ca、Si、Al、Fe的浸出速率无明显影响。     

核设施运行及退役中的废物最小化技术

核设施的运行及退役不可避免会产生放射性废物,废物管理的代价以及对公众、工作人员和对环境的危害取决于废物的数量及废物中所含的放射性核素,在核燃料循环过程中进行废物最小化管理是降低这些影响的一项必须的活动。在有些国家,废物最小化已作为一项国策。本文介绍了放射性废物最小化的环境效益及核设施运行和退役过程中废物最小化的方法,重点介绍了已研发的部分有效的废物最小化技术。通过总结美国等发达国家的放射性废物最小化的经验,提出了如何在我国实现放射性废物最小化的建议。     

核设施退役废石墨的处理与处置

石墨有成为核反应堆的慢化剂和反射层的较好的综合性能,早期发展核反应堆的国际原子能机构成员国拥有大量的石墨慢化反应堆,现在要安排退役,退役废石墨的处理和处置,成为人们共同关注的问题。由于废石墨存量大,放射性活度大,它的处理与处置有若干疑难问题有待解决。这些问题的解决关系到环境保护。我国有类似的疑难问题,为此应积极跟踪并开展必要的研究开发工作。     

桶装核废物层析γ扫描技术研究

层析γ扫描技术是桶装中低放核废物无损分析技术的主要分析方法,能够准确测量桶装容器内中、高密度非均匀核废物中的核素及其含量。本文综述了国内外桶装核废物层析γ扫描技术研究现状,简述国内现有研究基础和研究进展,并对层析γ扫描技术的发展趋势作了简要分析。提出了层析γ扫描技术研究发展方向和我国桶装核废物层析γ扫描技术面临的问题,并对所面临问题提出了相应的对策。