高放废物固化技术研究

随着核能源的广泛应用,特别是民用核能的发展,放射性废物的安全处置已经成为当今各国普遍关注的问题。目前各国公认的较为安全的方式是将放射性废物固化后进行深地质处置。对于高放射性废物而言,主要的固化方式有玻璃固化和人造岩石固化。本文从化学稳定性、机械稳定性、热稳定性、辐射稳定性等几个方面介绍了两种固化体的性能,为高放废物固化的进一步研究提供参考。     

微波技术在放射性废物处理中的应用研究

放射性废物的安全有效处理是世界各国关注的重要课题,同时也是人类能否继续开发和利用核能的关键。简要概述了微波技术的加热机理及特点,介绍了微波在低中放废物处理的应用,重点阐述了微波技术应用于高放废物的玻璃固化和人造岩石固化领域的研究进展,最后就目前微波技术处理高放废物的研究提出建议和展望。     

放射性废物固化处理的研究及应用现状

放射性废物安全有效的处置是世界各国关注的重要课题,也是核工业健康、可持续发展的重要保证.对放射性废物进行固化处理后埋入地下已经成为放射性废物处置的发展趋势.对水泥固化、沥青固化、塑料固化、玻璃固化、人造岩石固化等5种固化处理方法的固化机理、研究现状、应用情况、适用领域及优缺点进行了较系统的分析探讨.水泥固化、沥青固化、塑料固化适用于中低放废物的固化处理,玻璃固化和人造岩石固化适用于高放废物的固化处理.     

富钙钛锆石型人造岩石固化体的制备现状

富钙钛锆石型人造岩石将放射性核素固定在其晶格中作为晶体的一部分固定起来,大大提高了放射性废物处置的长期安全性,是固化高放射性废物理想的固化介质.对制备得到的性能稳定的富钙钛锆石型人造岩石固化体的研究现状进行了较为全面的综述和评价.     

矿物固化含Sr、Cs放射性废物研究进展

放射性元素锶和铯具有半衰期长、生物危害性大等特点,是目前放射性废物安全处理处置的重点之一,利用人造岩石固化含锶、铯放射性废物已成为当前的研究热点。简述了锶、铯的特点及危害,着重介绍了近年来几种典型矿物固化基材(碱硬锰矿、磷灰石、钙钛矿、铯榴石)的研究进展,探讨了现阶段矿物固化锶、铯的研究进展和发展方向,并对未来关于矿物固化锶、铯的材料领域发展趋势进行了展望。     

矿物混合材在放射性废物水泥固化中的作用

水泥固化是核电站放射性废物处理的常用技术,在放射性废物的水泥固化过程中,矿物混合材在固化操作性、固化体机械性能及核素浸出性等方面表现出优异性能。综述了活性和非活性矿物混合材的特性,及其在放射性废物水泥固化中的应用和对固化体性能的影响。活性矿物混合材的形态效应、火山灰效应和微集料效应以及非活性矿物混合材的填充作用可改善水泥固化体性能,降低水泥水化热升温,提高固化体抗压强度、抗裂性和耐久性,降低核素浸出率。混合材种类和掺加比例极大地影响着固化效果,通过配方实验保证混合材的适当掺量,可以保证水泥固化体性能的提升,保证固化体在处置期间的长期稳定性。     

烧绿石固化高放废物的研究进展

自然界中稳定存在的烧绿石由于在高放废物固化研究中呈现出的良好特性,近年来得到了大量的研究.在前人研究成果的基础上,对烧绿石的晶体结构与固化机理、锕系核素和稀土类似物的包容情况以及固化体的合成方法和相关性能等进行了系统的研究总结.已有的研究表明,烧绿石人造岩石固化体具有较高的密度和包容量,良好的抗浸出和抗辐照性能,是固化高放废物和进行最终地质处置的理想固化体,具有良好的应用前景.     

标准导读

GB／T7023--2011（低、中水平放射性废物固化体标准浸出试验方法》。本标准规定了在实验室条件下,低、中水平放射性废物固化体（以下简称为废物固化体）浸出性能检测的试验方法。本标准适用于比较和评价废物固化体在实验室控制条件下的抗浸出性能。定价：18．00元     

放射性废物的SHS固化处理研究及应用

采用自蔓延高温合成技术(SHS)固化处理放射性废物是固化方法新的研究方向。分析探讨了SHS固化处理方法的固化机理、研究现状及固化特点。SHS固化具有工艺简单、能量利用效率高、处理过程快速、成本低廉等优点,可针对不同类型的放射性废物选择合适的反应体系,进行产物设计,可直接应用到废物处置点或实现废物就地处置。介绍了近期笔者采用铝热剂自蔓延高温合成固化处理爆炸过程产生的有毒物质和受锕系核素污染的砂土及时两种形态的核废物的模拟固化实验研究。     

合成榍石的研究现状

榍石是一种稳定的矿物,是人造岩石固化高放射性废弃物较理想的基材之一。本文中从榍石的结构、性质及合成方法等方面进行全面的综述,同时对合成榍石的最新进展进行评述并展望了榍石在高放射性废弃物固化处理中的应用前景,为相关的研究工作提供参考依据。     

Advances in cement solidification technology for waste radioactive ion exchange resins: a review

Treatment and disposal of waste radioactive ion exchange resins is one of the most urgent problems for nuclear industries in China. Cement solidification technology has many advantages, such as requiring simple equipment, easy scaling-up, low working temperature, no trouble of gas cleaning and low cost. It is a suitable technology for treatment of waste radioactive resins, and has been widely used. The new developments and theoretical basis of cement solidification of radioactive resins were introduced in this paper. The cement solidification technology suitable for China and the questions needed to solve were also discussed.     

Study of Extraction of Palladium from Nitric Acid Solutions with Nitrogen-Containing Compounds,as Applied to Recovery of Fission Palladium from Spent Nuclear Fuel of Nuclear Power Plants: 3. Optimization of Extraction Process for Palladium Recovery and Refining

An extraction process for recovery of fission Pd is proposed. The process parameters are optimized with simulated high-level liquid radioactive waste from nuclear power plants using tri-n-octylamine (TOA) and tricaprylmethylammonium nitrate (Aliquat 336 nitrate) as extractants. The degree of Pd decontamination from fission and corrosion products is estimated and the major interfering components are determined.     

自蔓延高温合成技术在高放废物处理领域的应用进展

高放废物难以安全高效处理一直是制约核工业发展的关键因素之一。目前,自蔓延高温合成技术(SHS)作为一种高效、简单、低耗能的高放废物固化体合成手段,成为当下高放废物处理研究的热点领域之一。简述了SHS技术的原理及特点,着重介绍了近年来SHS技术在高放废物固化领域的应用,探讨了现阶段SHS技术的研究进展和发展方向,并对其未来发展趋势进行了展望。     

等离子体技术处理放射性废物的研究进展

等离子体技术处理放射性废物具有高效、环保、节能等特点,符合放射性废物处理的技术理念,具有广阔的应用前景。简述了等离子体技术处理放射性废物的原理及特点,详细介绍了等离子体技术在矿山开采放射性废水、核工业放射性废物和医疗放射性废物方面的应用,并着重分析了该技术的优势,最后就目前等离子体技术处理放射性废物过程中的问题进行了阐述。     

An additional performance of HTRS: the waste radiotoxicity minimisation

The management of radioactive waste is a key issue for the present and future use of nuclear energy. In this frame, high temperature reactors (HTRs) have, among others, the capability to burn actinides. After a short introduction on HTRs, the performances of two MC-based burnup codes (Monte Carlo continuous energy burnup and MONTEBURNS) in assessing the ability of these reactors to burn actinides are compared. These codes are necessary for performing ultra-high burnup calculations on HTRs. The best one, in this specific case, results to be MONTEBURNS. It was analysed using HTRs loaded with the following: (1) 1st generation Pu, 600 equivalent full power days; (2) 2nd generation Pu, 645 equivalent full power days; and (iii) 33% 1st generation Pu and 67% Th, 705 equivalent full power days. Finally, it is possible to conclude that HTRs can reduce time when the waste is considered dangerous. Even if the amount of reduction does not solve the whole problem, it represents an important step in the management of radioactive waste.     

Modelling of the shielding capabilities of the existing solid radioactive waste storages at Ignalina NPP

There is only one nuclear power plant in Lithuania--Ignalina NPP (INPP). The INPP operates two similar units with design electrical power of 1500 MW. The units were commissioned in 1983 and 1987 respectively. From the beginning of the INPP operation all generated solid radioactive waste was collected and stored at the Soviet type solid radwaste facility located at INPP site. The INPP solid radwaste storage facility consists of four buildings, namely building No. 155, No. 155/1, No. 157 and No. 157/1. The buildings of the INPP solid radwaste storage facility are reinforced concrete structures above ground. State Nuclear Safety Inspectorate (VATESI) has specified that particular safety analysis must be performed for existing radioactive waste storage facilities of the INPP. As part of the safety analysis, shielding capabilities of the walls and roofs of these buildings were analysed. This paper presents radiation shielding analysis of the buildings No. 157 and No. 157/1 that are still in operation. The buildings No. 155 and No. 155/1 are already filled up with the waste and no additional waste loading is expected.     

Concept of autonomous processing of liquid radioactive waste from ship nuclear power installations

The concept of autonomous processing of liquid radioactive waste from ship nuclear power installations is substantiated, and its implementation in the form of a module membrane sorption installation consisting of micro- and ultrafiltration, reverse-osmosis, and ion-exchange modules is suggested. Data on treatment of liquid radioactive waste of complex physicochemical composition using this installation are reported.