铁磷摩尔比对铁磷酸盐玻璃陶瓷固化体的影响

研究了不同铁磷摩尔比对铁磷酸盐玻璃陶瓷固化体结构和化学稳定性的影响。用溶解速率法(DR)研究了固化体的化学稳定性,用傅立叶变换红外光谱(FT-IR)和X射线衍射(XRD)方法表征了样品的结构,用全谱直读等离子体发射光谱(ICP-OES)测定浸出液中各元素的含量。研究结果表明,当铁磷摩尔比为0.67时,在980℃下保温3h得到的铁磷酸盐玻璃陶瓷固化体具有较高的化学稳定性,浸泡42d的质量浸出率变化幅度不大且浸出率较低,约为7.05×10-9g/(cm2.min),其中Ce、La元素均未检出,其余浸出元素来自玻璃相;固化体的主晶相为独居石,结构中主要含有大量的正磷酸基团[PO4]3-和少量的焦磷酸基团[P2O7]4-,不存在偏磷酸基团[PO3]-。     

熔盐法低温合成掺钕ZrSiO4陶瓷的物相演变和化学稳定性

相较于传统固相烧结方法,熔盐在较低的温度下提供了快速的传质和成核过程,可合成用于固化高放废物(HLW)的陶瓷固化体。本工作采用熔盐法(MSS)在不同烧结温度(1100、1200、1300、1400、1500℃)和不同烧结时间(3、6、9、12、15 h)下制备了掺Nd的锆石(ZrSiO₄)陶瓷(Zr_1-xNdxSiO_4-x/2 (0≤x≤0.1)),并采用静态浸出试验(PCT)研究掺Nd的ZrSiO₄陶瓷在模拟地质处置环境下的化学稳定性。在熔盐与氧化物最佳摩尔比为10:1、烧结温度为1200℃、烧结时间为6 h的较温和条件下,利用熔盐法成功合成了Zr1-xNdxSiO4-x/2,可将Nd在ZrSiO₄中的固溶摩尔分数提高到8%,结果显示MSS法能够降低陶瓷合成温度,缩短合成时间,提高固溶量。ZrSiO₄陶瓷对三价锕系核素的固化机理为晶格固化。浸出实验结果显示, Nd的归一化浸出率(LRNd)低至～10^-5 g·m^-2     

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     

温度对含模拟α-高放核废液的磷酸镁水泥固化体性能的影响

利用磷酸镁水泥(Magnesium potassium phosphate cement,MPC)对模拟α-高放核废液(HLW)进行固化,研究温度对固化体力学性能、物相组成、微观形貌及核素Cs~+浸出率的影响。BET、XRD、SEM及AAS等测试结果表明,室温下MgO、KH2PO4与高放核废液反应形成致密结构;随着温度的升高,固化体脱水,400℃时孔道结构增多,平均孔径增大,抗压强度降低,Cs~+浸出率增加;温度继续升高,磷酸镁水泥烧结陶瓷化,平均孔径逐渐减小,抗压强度增大;900℃时固化体表现出良好的陶瓷结构特征,晶粒完全熔融,晶粒间没有明显界线,Cs~+的28d浸出率为7.21×10~(-6) g/(cm~2·d)。不同温度下高放核废液的磷酸镁水泥固化体核素Cs~+的浸出率均能达到玻璃固化体的性能要求,表明磷酸镁水泥用于固化高放核废液具有明显优势。     

回归自然:人造岩石固化核素的思考与进展

核工业生产、核能开发、核武器研制等不可避免会产生放射性废物,高放废物是现存放射性废物中最难处理的废物之一。随着我国"积极发展核电"战略的实施,放射性废物的安全有效处理处置成为关系到我国核能可持续发展的关键问题。人造岩石固化体(SYNROC)弥补了玻璃固化体低化学耐久性和亚稳态性能的缺点。本文在综述人造岩石固化的概念、候选矿物固化体分类的基础上,重点介绍了SYNROC固化体快速合成方法、固核机理和长期稳定性评价等方面的最新研究进展。"道阻且长,行则将至"。最后,指出了SYNROC固化存在的不足,并针对今后应重点关注的研究方向与发展趋势提出了建议。     

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。     

核电厂典型中低放射性废物等离子体熔融处理试验研究

选取国内压水堆核电厂两种典型的中低放射性废物开展等离子体熔融中试试验,在1250℃温度条件下熔融100 min,熔制成完全玻璃态的固化体,两种玻璃固化体的XRD衍射谱均呈现为典型的非晶态谱;并选取了非放射性的Co_2O_3、SrCO_3、CsCl作为放射性核素示踪剂,模拟放射性核素~(137)Cs、~(90)Sr、~(58)Co、~(60)Co在核电站放射性废物等离子熔融处理过程中的包容情况;经检测,玻璃固化体物理性能、抗浸出性能以及机械性能满足高放玻璃固化体要求,且机械性能优于水泥固化体标准;最后对后续试验进行了展望,并提出了需要进一步解决的问题。     

自蔓延高温合成掺钐烧绿石陶瓷固化体及其化学稳定性

为了探索Gd₂Ti1.3Zr0.7O₇烧绿石快速固化高放废物中锕系核素的方法,采用Sm模拟三价锕系核素,以氧化铜为氧化剂、钛粉为还原剂,氧化钆、氧化钐和二氧化锆为原料,利用自蔓延高温结合快速加压合成技术制备掺钐烧绿石陶瓷固化体,分别采用X射线衍射仪(XRD)、场发射扫描电镜(FESEM)、能谱仪(EDS)、电感耦合等离子体发射光谱-质谱(ICP-MS)和维氏硬度计研究样品的物相、微观组织形貌、元素分布、化学稳定性和力学性能。结果表明:自蔓延高温快速加压合成法在极短的时间(5 min)内合成完全固溶的Gd2-xSmxTi1.3Zr0.7O₇(0≤x≤0.2)烧绿石陶瓷固化体,烧绿石陶瓷固化体(x=0.2)的相对密度达92.8%,具有优异的化学稳定性;浸出时间为42 d时,Gd和Sm的归一化浸出率分别为2.23×10-5 g·m-2·d-1、1.57×10-5 g·m-2·d-1。     

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.     

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.     

Accelerated chemical aging of crystalline nuclear waste forms

Nuclear waste disposal is a significant technological issue, and the solution of this problem (or lack thereof) will ultimately determine whether nuclear energy is deemed environmentally friendly, despite significantly lower carbon emissions than fossil fuel energy sources. A critical component of any waste disposal strategy is the selection of the waste form that is tasked with preventing radionuclides from entering the environment. The design of robust nuclear waste forms requires consideration of several criteria, including: radiation tolerance, geological interaction and chemical durability; all of these criteria ensure that the radionuclides do not escape from the waste form. Over the past 30 years, there have been numerous and thorough studies of these criteria on candidate waste forms, including radiation damage and leaching. However, most of these efforts have focused on the performance of the candidate waste form at t = 0, with far less attention paid to the phase stability, and subsequent durability, of candidate waste forms during the course of daughter product formation; that is, the chemical aging of the material. Systematic understanding of phase evolution as a function of chemistry is important for predictions of waste form performance as well as informing waste form design. In this paper, we highlight the research challenges associated with understanding waste form stability when attempting to systematically study the effects of dynamic composition variation due to in situ radionuclide daughter production formation.     

A carbon composite based on pyrolyzed diphthalocyanines for immobilization of high-level waste from nuclear industry

A new procedure for immobilization of high-level waste from spent nuclear fuel (SNF) of nuclear power plants (NPP) is suggested. The HLW immobilization process consists in the radionuclide transfer into the form of acetates, followed by synthesis of their diphthalocyanines and their subsequent pyrolysis in an inert medium. Real raffinate solutions from reprocessing of SNF of Novovoronezh NPP were used as HLW. The efficiency of the immobilization of radionuclides (mainly REE and minor actinides) was found to be higher than 99%. The thermal stability, chemical durability (leaching with aqueous solutions with pH from 2 to 9), and radiation resistance of the samples were evaluated using α-, β-, and γ-ray spectrometry. The evolution of the structure of the pyrolyzed diphthalocyanines in the temperature range 800–1600°C was studied by small-angle neutron scattering, X-ray diffraction analysis, and atomic force microscopy. The results obtained have confirmed the hypothesis that the pyrolysis of diphthalocyanines yields a branched network of carbon atom nanoclustes of size from 5 to ~150 nm, ensuring efficient retention of radionuclides in the host material. The host material is comparable in the thermal stability, i.e., in the degree of radionuclide retention at 1000–1200°С, with borosilicate or phosphate glasses, which are the most widely used host materials for radioactive waste immobilization, and considerably surpasses them in the resistance to leaching and radiation.     

Characterization and remediation of soils contaminated with uranium

Environmental contamination caused by radionuclides, in particular by uranium and its decay products is a serious problem worldwide. The development of nuclear science and technology has led to increasing nuclear waste containing uranium being released and disposed in the environment. The objective of this paper is to develop a better understanding of the techniques for the remediation of soils polluted with radionuclides (uranium in particular), considering: the chemical forms of uranium, including depleted uranium (DU) in soil and other environmental media, their characteristics and concentrations, and some of the effects on environmental and human health; research issues concerning the remediation process, the benefits and results; a better understanding of the range of uses and situations for which each is most appropriate. The paper addresses the main features of the following techniques for uranium remediation: natural attenuation, physical methods, chemical processes (chemical extraction methods from contaminated soils assisted by various suitable chelators (sodium bicarbonate, citric acid, two-stage acid leaching procedure), extraction using supercritical fluids such as solvents, permeable reactive barriers), biological processes (biomineralization and microbial reduction, phytoremediation, biosorption), and electrokinetic methods. In addition, factors affecting uranium removal from soils are furthermore reviewed including soil characteristics, pH and reagent concentration, retention time.     

Modeling the long-term leaching behavior of (137)Cs, (60)Co, and (152,154)Eu radionuclides from cement-clay matrices

Leaching characteristics of some radionuclides that are commonly encountered in radioactive waste streams from immobilized waste matrices in different cement-clay grouts have been assessed to investigate the influence of the clay additives on the leaching behavior of the solid waste matrices. The International Atomic Energy's Agency (IAEA) standard leach method has been employed to study the leach pattern of (137)Cs, (60)Co, and (152,154)Eu radionuclides immobilized in ordinary Portland cement (OPC), OPC-bentonite, and OPC-red clay grouts. The examination of the leaching data revealed that adding clays to OPC reduces the leach pattern as OPC-bentonite相似文献   

Magnesium potassium phosphate matrices for immobilization of high-level liquid wastes

Physicochemical properties of magnesium potassium phosphate (MPP) matrices for conservation of highly saline high-level liquid wastes (HLW) after their solidification at room temperature were studied comprehensively. The matrices showed high chemical stability to leaching of radionuclides and other components at various temperatures. The leaching indices of ²³⁹Pu, ²³⁷Np, ²⁴¹Am, Sr, and Cs are in the range 12–14, and those of Tc, I, and Se, in the range 10–11. The mechanical strength (>20 MPa) and radiation resistance of the matrices and the chemical yield of radiolytically generated hydrogen (0.004 molecules/100 eV) were determined. The phase composition of the matrices and the character of radionuclide distribution in their bulk were determined. An enlarged installation was developed and fabricated at the Mayak Production Association, and solidification of simulated liquid radioactive wastes from this plant was performed in volumes of up to 2001.     

The immobilization of high level radioactive wastes using ceramics and glasses

An overview is given of the immobilization of high level radioactive waste (HLW) and surplus materials from a variety of commercial and defence sources employing glass and ceramic hosts. A number of specific host materials are reviewed, including borosilicate and phosphate glasses, glass-ceramics and crystalline ceramics. Topics covered include wasteform processing and manufacture, in addition to wasteform stability, durability and mechanical behaviour. Although, at the present time, borosilicate glass is the generally accepted first generation wasteform for the immobilization of HLW, the emergence of new sources of radioactive materials requiring immobilization has renewed interest in many of the alternative candidates. These include, in particular, titanate, zirconate and phosphate based ceramics, together with iron phosphate based glasses and basaltic glass-ceramics. The relative merits and limitations of each host material are compared and discussed, with particular reference to processing considerations and to current and likely future requirements. This revised version was published online in November 2006 with corrections to the Cover Date.     

高放废料固化基材的研究现状

简述了高放废料的产生、基本特点,几种高放废料固化基材(玻璃、陶瓷、增压养护水泥、碱矿渣水泥和水合陶瓷)各自的特性和研究现状,介绍了固化基材技术要求和经济条件等方面的一些情况.     

Metallurgical recovery of metals from electronic waste: a review

Waste electric and electronic equipment, or electronic waste, has been taken into consideration not only by the government but also by the public due to their hazardous material contents. In the detailed literature survey, value distributions for different electronic waste samples were calculated. It is showed that the major economic driver for recycling of electronic waste is from the recovery of precious metals. The state of the art in recovery of precious metals from electronic waste by pyrometallurgical processing, hydrometallurgical processing, and biometallurgical processing are highlighted in the paper. Pyrometallurgical processing has been a traditional technology for recovery of precious metals from waste electronic equipment. However, state-of-the-art smelters are highly depended on investments. Recent research on recovery of energy from PC waste gives an example for using plastics in this waste stream. It indicates that thermal processing provides a feasible approach for recovery of energy from electronic waste if a comprehensive emission control system is installed. In the last decade, attentions have been removed from pyrometallurgical process to hydrometallurgical process for recovery of metals from electronic waste. In the paper, hydrometallurgical processing techniques including cyanide leaching, halide leaching, thiourea leaching, and thiosulfate leaching of precious metals are detailed. In order to develop an environmentally friendly technique for recovery of precious metals from electronic scrap, a critical comparison of main leaching methods is analyzed for both economic feasibility and environmental impact. It is believed that biotechnology has been one of the most promising technologies in metallurgical processing. Bioleaching has been used for recovery of precious metals and copper from ores for many years. However, limited research was carried out on the bioleaching of metals from electronic waste. In the review, initial researches on the topic are presented. In addition, mechanisms and models of biosorption of precious metal ions from solutions are discussed.     

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。