中国高放废物深地质处置的缓冲材料选择及其基本性能

人类的许多生产、生活活动均可能产生不同活度的放射性废物,其中高放废物的安全处置倍受全球科学家和广大公众所重视。目前深地质处置被国际上公认为处置高放废物的最有效可行的方法。借鉴国外成熟的技术和经验,我国采用多重工程屏障系统(包括废物固化体、废物容器及其外包装和缓冲/回填材料)和适宜的围岩地质体共同作用,来确保高放废物与生物圈的安全隔离。膨润土由于具有极低的渗透性和优良的核素吸附等性能而被国际上选作缓冲材料的基础材料。经过全国膨润土矿床筛选,我国高放废物深地质处置库缓冲材料的研究以产自高庙子膨润土矿床深部的钠基膨润土作为基本组成材料。本文介绍了高庙子膨润土矿床的地质特征以及高庙子钠基膨润土的基本特征。该膨润土与国外同类型材料相比具有蒙脱石含量高(75%左右)、杂质矿物相对较少的特点,该材料的系统和深入研究对于开发我国缓冲回填材料技术、确保高放废物的安全有效处置有重要意义。     

中国高放废物处置库缓冲材料选择与基本性能

人类的许多生产、生活活动均可能产生不同活度的放射性废物。其中高放废物由于具有放射性水平高,发热量大,并含有对生物极有害的α放射性的长寿命核素等特点,其安全处置倍受全球科学家和广大公众所重视。目前深地质处置被国际上公认为处置高放废物的最有效可行的方法。借鉴国外成熟的技术和经验,我国采用多重工程屏障系统（包括废物固化体、废物容器及其外包装和缓冲/回填材料）和适宜的地质围岩地质体共同作用来确保高放废物与生物圈的安全隔离。膨润土由于具有极低的渗透性和优良的核素吸附等性能而被国际上选作缓冲材料的基础材料。经过全国膨润土矿床筛选,高庙子膨润土矿床被选作我国缓冲材料供应基地,我国高放废物深地质处置库缓冲材料的研究以产自该矿床的深部钠基膨润土作为基本组成材料。本文介绍了高庙子膨润土矿床的地质特征以及高庙子钠基膨润土的基本特征。该膨润土与国外同类型材料相比具有蒙脱石含量高（75%左右）,杂质矿物相对较少的特点,这对系统和深入研究该材料以开发我国缓冲回填材料技术,确保高放废物的安全有效处置具有重要意义。     

类似物研究和矿物学问题

本文叙述了核废物地质处置研究领域中涉及天然和人为类似物研究中的一些矿物学问题。在自然界的天然玻璃、膨润土、晶质铀矿等分别作为高放废物玻璃固化体、高放废物处置库的缓冲／回填材料和乏燃料的天然类似物，考古遗址中的玻璃和青铜器文物作为人为类似物。通过这些天然非晶质结晶物质和人造制品的稳定性研究来预测未来10000－100000a间处置库中废物和缓冲／回填物质的变化，放射性核素迁移规律，为高放废物处置库的设计和建造提供重要科学依据，提高公众对高放废物安全处置的信心。     

内蒙古高庙子膨润土作为高放废物处置库回填材料的可行性

本文简述高放废物处置工程要求和缓冲／回填材料在高放废物深地质处置库中的作用及要求，介绍了内蒙古兴和县高庙子膨润土的性能，认为内蒙古兴和县高庙子膨润土矿床可作为高放废物处置库回填材料的供给基地。     

大气条件下高庙子钠基膨润土化学缓冲特性研究

深地质处置目前被国际上公认为是处置高放废物的最有效可行的方法。我国采用多重工程屏障系统和适宜的地质体共同作用来确保与生物圈的安全隔离。缓冲材料是高放废物重要的工程屏障材料之一,我国选用高庙子钠基膨润土作为缓冲材料的基础材料。膨润土作为缓冲材料的一个重要性能表现为缓冲孔隙水的化学变化。介绍了GMZ-1钠基膨润土大气条件下与蒸馏水的反应试验,并对试验结果进行了讨论。批式试验反应溶液中钠离子来源于钠基膨润土层间阳离子和矿物溶解,镁离子来源于钠基膨润土层间阳离子,钾离子和钙离子来源于矿物溶解,相关研究认识对于高放废物处置库近场核素迁移研究和评价工程屏障的长期稳定性具有重要意义。     

中国高放废物地质处置:现状和展望

本文阐述了中国的高放废物处置政策,回顾了中国高放废物地质处置在选址和场址评价、缓冲回填材料等方面的进展,提出了中国高放废物地质处置的初步规划,指出我国计划在2015年确定地下实验室场址.还展望了国际合作前景.     

中国高放废物地质处置缓冲材料大型试验台架和热-水-力-化学耦合性能研究

缓冲材料作为高放废物深地质处置库中一道重要的人工屏障,与高放废物容器和处置库围岩直接接触,在高放废物衰变热、辐射作用和地下水等影响下产生复杂的热-水-力-化学耦合作用,为了验证缓冲材料是否能长期有效地发挥其屏障材料的作用,核工业北京地质研究院利用高庙子钠基膨润土组装并运行了模拟中国高放废物地质处置室 尺寸的大型缓冲材料膨润土试验台架（China-Mock-Up）。建立了缓冲材料试验台架的安装和试验方法,依据实测数据和理论分析,揭示了热-水-力-化学耦合作用条件下膨润土中的相对湿度是在加热器的热效应和外部供水的湿效应共同作用下发生变化的,压实膨润土中应力的变化主要是由于膨润土遇水膨胀和加热器的热效应引起的,试验验证了模拟高放废物地质处置室内加热器（废物罐）运行初期的位移过程,为缓冲材料和高放废物地质处置库的设计提供了重要的工程参数和理论依据。     

新疆地区选建高放射性核废物处置库的可能性

高放废物选址是一项万年大计工程,需要多个预选场地的综合比较,决策废物处置库的具体选建场地。论述了高放废物处置库选址过程中需要考虑的主要问题,然后针对新疆的自然环境,水资源、花岗岩体和缓冲/回填材料的分布,以及区域地壳稳定性等因素,讨论了在新疆选建高放废物地质处置库的可能性,并提出了选址方法。     

新疆地区选建高放射性核废物处置库的可能性

高放废物选址是一项万年大计工程,需要多个预选场地的综合比较,决策废物处置库的具体选建场地.论述了高放废物处置库选址过程中需要考虑的主要问题,然后针对新疆的自然环境,水资源、花岗岩体和缓冲/回填材料的分布,以及区域地壳稳定性等因素,讨论了在新疆选建高放废物地质处置库的可能性,并提出了选址方法.     

高放废物深地质处置：回顾与展望

文章对我国高放废物地质处置研究的历史进行了回顾,并对未来发展进行了展望。我国的高放废物深地质处置研究开发从1985年开始,迄今为止可初步分为3个阶段：①起步和跟踪研究阶段（1985～1998）;②逐步发展阶段（1999～2005）;③政府规划指导阶段（2006至今）。20多年来,我国在国家法律法规、战略规划、选址、工程屏障、核素迁移研究等方面取得了显著进展。我国已经提出在2020年前建成地下实验室、21世纪中叶建成高放废物处置库的目标。研究开发和处置库工程建设分成3个阶段：试验室研究开发和处置库选址阶段（2006～2020）;地下现场试验阶段（2021-2040）和处置库建设阶段（2041～本世纪中叶）。经过全国筛选对比,已初步选定甘肃北山地区为重点预选区,系统的场址评价工作正在进行。已确定采用膨润土作为处置库的回填材料,并初步确定内蒙古高庙子膨润土为我国高放废物处置库的首选缓冲回填材料。工程设计、核素迁移研究和安全评价也取得了一定进展。1999年起与国际原子能机构开展了3期高放废物地质处置技术合作项目,对提高我国的技术水平起到了积极作用。20多年的研究开发工作为我国在21世纪完成高放废物地质处置任务奠定了基础。     

Estimation of loading density of underground well repositories for solid high-level radioactive wastes

The convective transfer of radionuclides by subsurface water from a geological repository of solidified high-level radioactive wastes (HLW) is considered. The repository is a cluster of wells of large diameter with HLW disposed of in the lower portions of the wells. The safe distance between wells as a function of rock properties and parameters of well loading with wastes has been estimated from mathematical modeling. A maximum permissible concentration of radionuclides in subsurface water near the ground surface above the repository is regarded as a necessary condition of safety. The estimates obtained show that well repositories allow for a higher density of solid HLW disposal than shaft storage facilities. Advantages and disadvantages of both types of storage facilities are considered in order to estimate the prospects for their use for underground disposal of solid HLW.     

Selection and Basic Properties of the Buffer Material for High-Level Radioactive Waste Repository in China

Radioactive wastes arising from a wide range of human activities are in many different physical and chemical forms, contaminated with varying radioactivity. Their common features are the potential hazard associated with their radioactivity and the need to manage them in such a way as to protect the human environment. The geological disposal is regarded as the most reasonable and effective way to safely disposing high-level radioactive wastes in the world. The conceptual model of geological disposal in China is based on a multi-barrier system that combines an isolating geological environment with an engineered barrier system. The buffer is one of the main engineered barriers for HLW repository. It is expected to maintain its low water permeability, self-sealing property, radio nuclides adsorption and retardation properties, thermal conductivity, chemical buffering property, canister supporting property, and stress buffering property over a long period of time. Bentonite is selected as the main content of buffer material that can satisfy the above requirements. The Gaomiaozi deposit is selected as the candidate supplier for China＇s buffer material of high level radioactive waste repository. This paper presents the geological features of the GMZ deposit and basic properties of the GMZ Na-bentonite. It is a super-large deposit with a high content of montmorillonite （about 75 %）, and GMZ-1, which is Na-bentonite produced from GMZ deposit is selected as the reference material for China＇s buffer material study.     

Natural and artificial minerals as matrices for immobilization of actinides

The possibility of insulation of long-lived actinides for the entire period of their potential hazard, i.e., over a virtually infinite time, is the crucial problem in safe disposal of high-level radioactive wastes (HLW). The reality of such a possibility is corroborated by the study of natural radioactive minerals that firmly retain U, Th, and REE, which are close in geochemical properties to transuranic actinides, for millions of years despite the effects of groundwater. The natural analogues of actinide HLW matrices are minerals where U, Th, and REE are contained as major elements or isomorphic admixtures. The study of these minerals is helpful for synthesis of durable artificial forms of wastes that ensure reliable insulation of HLW up to the complete decay of actinides independently of such engineering barriers of underground repositories as containers and bentonite buffers. The main requirements on confinement matrices include a high isomorphic capacity with respect to actinides and other HLW components, chemical and radiation stability, and technological feasibility of their industrial production. The natural and artificial minerals—uraninite, monazite, zirconolite, pyrochlore, britholite, garnet, and murataite—characterized in this paper may serve as a basis for efficient matrices for immobilization of actinide wastes.     

花岗岩型地质处置库情景试构建与分析

高水平放射性废物(高放废物)地质处置库关闭后长期演变情景的构建和定量分析是安全评价的关键。研究以拟建于北山预选区的花岗岩型处置库为研究对象,对情景开发及其分析进行了初步探索。采用"自下而上"的情景开发方法,构建了处置库关闭后预期演变情景和3类典型的非预期演变情景,对各类情景条件下的核素释放率进行了计算和分析。结果表明,现有条件下,处置库具备较好的安全性;为提升安全评价的可靠度,需要进一步加强情景不确定性的分析。     

Confinement matrices for low- and intermediate-level radioactive waste

Mining of uranium for nuclear fuel production inevitably leads to the exhaustion of natural uranium resources and an increase in market price of uranium. As an alternative, it is possible to provide nuclear power plants with reprocessed spent nuclear fuel (SNF), which retains 90% of its energy resource. The main obstacle to this solution is related to the formation in the course of the reprocessing of SNF of a large volume of liquid waste, and the necessity to concentrate, solidify, and dispose of this waste. Radioactive waste is classified into three categories: low-, intermediate-, and high-level (LLW, ILW, and HLW); 95, 4.4, and 0.6% of the total waste are LLW, ILW, and HLW, respectively. Despite its small relative volume, the radioactivity of HLW is approximately equal to the combined radioactivity of LLW + ILW (LILW). The main hazard of HLW is related to its extremely high radioactivity, the occurrence of long-living radionuclides, heat release, and the necessity to confine HLW for an effectively unlimited time period. The problems of handling LILW are caused by the enormous volume of such waste. The available technology for LILW confinement is considered, and conclusion is drawn that its concentration, vitrification, and disposal in shallow-seated repositories is a necessary condition of large-scale reprocessing of SNF derived from VVER-1000 reactors. The significantly reduced volume of the vitrified LILW and its very low dissolution rate at low temperatures makes borosilicate glass an ideal confinement matrix for immobilization of LILW. At the same time, the high corrosion rate of the glass matrix at elevated temperatures casts doubt on its efficient use for immobilization of heat-releasing HLW. The higher cost of LILW vitrification compared to cementation and bitumen impregnation is compensated for by reduced expenditure for construction of additional engineering barriers, as well as by substantial decrease in LLW and ILW volume, localization of shallow-seated repositories in various geological media, and the use of inexpensive borosilicate glass.     

高放废物地质处置研究中的矿物学问题

在高放废物地质处置研究中的一些矿物学问题应引起矿物学家的注意 。高放废物地 质处置库的缓冲/回填材料是选择以钠质蒙脱石为主要成分的钠基膨润土,还是以钙质蒙脱 石为主的钙基膨润土?怎样选择对99Tc、129I有较好吸附能力的硒汞矿、脆 硫锑铅矿和辉锑矿的替代物来作缓冲/回填材料中的添加剂,以能阻滞99Tc和129 I的迁移?沸石对某些放射性核素的吸附特征也需进一步研究。     

Unsaturated permeability of Gaomiaozi bentonite under partially free-swelling conditions

Acta Geotechnica - In the deep geological disposal of high-level radioactive wastes (HLW), Na-bentonite produced from Gaomiaozi (GMZ) deposit has been selected as a possible buffer and backfill...     

核废物处置试验场环境地质研究综述

介绍了国外核废物处理试验场有关环境地质研究现状与进展。据核废物的放射性不同，目前或未来处置的方式不一样：中－低放射性废物（ＩＬＷ－ＬＬＷ）一般采用浅层处置，多置于粘土层或沉积岩层中；高放射性废物（ＨＬＷ）一般通过竖井或平巷处置于深部的花岗岩或岩盐中。根据母岩的类别不同，本文分花岗岩、岩盐、粘土及其它四部分，对不同的试验场的研究计划、内容、方法、进展等进行了评述。