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。     

铀矿床作为高放废物处置库天然类似物的国内外研究进展

铀矿床具有目前国际上开发的高放废物深地质处置概念的相似特征，对其开展天然类比研究，是认识核素在地质环境中迁移行为的一种有效方法。本文阐述了高放废物深地质处置库天然类似物的一般概念及天然类比研究拟解决的科学问题。重点介绍了30年来国内外在天然类比研究中取得的主要成果、认识以及国内外研究进展。     

加强环境地质学科研究 开发环保产业

８０年代以来随着环境保护作为我国的一项基本国策的确立，核工业北京地质研究院把环境评价和一半保护确定为５大技术开发项目之一。基于多年从事放射性铀矿床的成矿理论和找矿研究中积累的技术和经验，引入了国际上新近发展的新技术，开发了环境评价、高效废物及中低放废物处置库的选址、勘查和评价、核设施退役中核废物的填直处置等方面的研究工作。在承担项目的过程中促进科研成果的转化。经过不断学习、摸索和提高，我们初步打开     

今天的核废物地质处置研究

凡有核设施的国家都会产生核废物.由于核废物具有一定的毒性,所以国际上对核废物的地质分量研究极为关注,这也是第28届国际地质大会讨论的热点之一.本文就是根据这次大会的有关论文编写的,并按下列6部分作了简要的介绍:核废物地质处置研究概况,处置库场地预选,预选场地的水文地质研究,核素迁移,场地的特性评价和安全评价.文中侧重介绍了核废物地质分量研究中的一些新思想和新方法.     

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。     

日本的核电和核废物处置及其研究现状

日本的核电和核废物处置及其研究现状陈璋如（核工业北京地质研究院１０００２９）笔者作为日本科技厅１９９３年财政年度科学家交流计划的研究人员在东京东北部的ＰＮＣ东海事业所工作半年，以下仅就日本核电和核废物处置状况及研究动态予以简介。１日本的核电日本由于天...     

玻璃陨石的成因争论及可能的彗星撞击模型

玻璃陨石的成因争论及可能的彗星撞击模型李春来（中国科学院地球化学研究所，贵阳５５０００２）关键词玻璃陨石成因彗星撞击事件玻璃陨石是一种棕黑色或浅绿色的天然玻璃，一般为厘米级大小的块状，表面多具空气动力学熔蚀刻痕。长期的研究证明，玻璃陨石在化学和结构特...     

6124铀矿床构造应力场研究

６１２４铀矿床构造应力场研究梁良，刘成东，李建红（华东地质学院江西抚州３４４０００）６１２４铀矿床位于相山矿田的西北部，含矿主岩为上侏罗统的碎斑熔岩。根据区内共轭节理的系统测量和应力场的数学模拟，提出了矿区的ＮＥ向基底断裂为过时针向的压剪性断裂；ＮＷ...     

加拿大陈田博士介绍SYVAC_3模式

放废物处置库性能评价是处置库选址乃至废物安全处置中的重要研究内容。针对这一研究内容,核工业北京地质研究院于1998年5月11日邀请了加拿大多年从事高放废物处置的专家——陈田博士来院交流。他详细介绍了加拿大高放废物处置的研究进展,特别是关于处置库系统性能评价模式方面。目前,加拿大所采用的评价模式是SYVAC。,该模式应用概率系统变量分析方法对关闭后的处置库系统特征、事件和过程进行评价。该模式由3个子模式构成,即：处置库子模式,岩石圈子模式和生物圈子模式。处置库子模式模拟废物罐的腐蚀、核素从废物体中的释放以及…     

加拿大Atikokan地区花岗深成岩体内断裂带中的钍石：在核废物处理…

对整个Ｅｙｅ－Ｄａｓｈｗａ岩体内断裂带中的钍石进行了研究，钍石周围反应边结构未受到变形，这表明它是动力作用后生长的。ＴｈＯ２和ＳｉＯ２是岩体中钍石的主要氧化物组合，热力学分析表明花岗岩体内钍石在与周围地下水接触中是稳定相。钍石的这种稳定性（约２．３Ｇａ）和化学惰性可作为核废物处置某些领域中天然类似物的实例。     

世界高放废物地质处置库选址研究概况及国内进展

高放废物是核能事业发展的必然产物。它的安全处置是核能事业持续发展的前提 ,已受到世界各国的高度重视。文中阐述了高放废物深地质处置的一般概念。同时重点介绍了世界上一些国家处置库选址研究的主要内容和研究进展 ,例如 ,美国把处置库建造过程分为场地推荐、场地的特征评价、处置库场地的选择和批准、领取场地执照和处置库建造设计的审批、处置库的建造 5个阶段 ;德国的选址研究工作包括地电和地热研究 ,重力、地震、地球化学、水文地质、同位素地球化学及微生物研究等 ;瑞典在花岗岩中建成了地下实验室 ,并制定了实验室的总体研究目标等等。另外也介绍了中国在甘肃省北山进行的高放废物地质处置库选址工作的情况 ,研究表明北山地区为一地壳稳定区 ,也是地下水贫水区且地下水流速缓慢 ,有利于处置库的建造 ,进一步的地面地质、水文地质勘察工作及钻探工程工作正在进行中。伴随着这些工作的完成 ,中国将大大缩短在高放废物地质处置研究方面与发达国家的距离。     

高放废物地质处置黏土岩处置库围岩研究现状

世界上很多国家都对处置库的可能围岩进行了详细研究。通过对比,认为花岗岩、黏土岩、岩盐比较适合作为处置库围岩,而黏土岩由于具有自封闭性、渗透率低等其他岩石类型不可比拟的优点,因而将黏土岩作为高放废物地质处置库围岩越来越受到各国的关注。文章同时介绍了瑞士、法国、比利时等国家在黏土岩中所进行的大量研究,均认为在黏土岩中处置高放废物和乏燃料是安全的。文章还对黏土岩处置库概念设计、黏土岩处置库围岩地下实验室研究,以及我国开展黏土岩处置库研究的意义等进行了综述。     

The Antei uranium deposit: A natural analogue of an SNF repository and an underground geodynamic laboratory in granite

The estimation of the long-term stability of crystalline rock massifs with respect to natural and technogenic loads in the course of long-term storage of spent nuclear fuel (SNF) is a special area of surveys at underground research laboratories (URLs). In parallel with these surveys, data on uranium deposits—natural analogues of repositories of SNF consisting of 95% UO₂—are used for obtaining insight into the dynamics of radionuclide migration and validating barrier properties of host rocks. Examples of URLs located in granitic massifs of Sweden (Äspö), Canada (Whiteshell), Switzerland (Grimsel), Japan (Mizunami), and Finland (ONKALO), as well as the El Berrocal (Spain), Palmottu (Finland), Sanerliu (China), and Kamaishi (Japan) deposits, are considered in the paper. The objects listed above are distinct in tectonic settings, geology, control of ore mineralization, redox conditions of uranium migration, and character and intensity of filtration and transportation, which predetermine the direction and specific features of research conducted therein. A variant in which a URL and a natural analogue are combined in one object is especially promising for validation of safe long-term isolation of SNF. The Antei vein-stockwork uranium deposit in the southeastern Transbaikal region, localized in Paleozoic granite at a depth of 400–1000 m and opened by mine workings at six levels, is such an object. Its geological features, stress-strain state, and infrastructure of mine workings offer an opportunity to study the entire spectrum of processes proceeding in near-and far-field of an SNF repository. The structural geology, mineralogy and petrography, and petrophysical and tectonophysical features of the deposit at its three lower levels are considered. The sequence of metasomatic alteration of rocks and the dynamics of formation of ore-bearing faults that crosscut prototectonic elements, as well as relationships of physicomechanical properties of rocks as a function of the intensity of their metasomatic alteration and the distance from master fault planes, have been established. A 3D geological model of the deposit in combination with estimated parameters of the present-day stress field and physicomechanical properties of particular rock blocks serves as the basis for prediction of the geomechanical behavior of the massif. The practical implications of the results obtained for assessment of the long-term safety of SNF repositories in granites are discussed.     

Geoscientific R&D for high level radioactive waste disposal in Sweden — current status and future plans

SKB (Svensk Kärnbränslehantering AB) is responsible for all handling, transport and storage of the nuclear wastes outside the Swedish nuclear power stations. According to Swedish law, SKB is responsible for an R&D-programme needed to take care of the radwastes. The programme comprises, among others, a general supportive geo-scientific R&D and the Äspö Hard Rock Laboratory (HRL) for more in-situ specific tasks.

Sweden is geologically located in the Fennoscandian shield which is dominated by gneisses and granitoids of Precambrian age. The Swedish reference repository concept thus considers an excavated vault at ca. 500 m depth in crystalline rocks. In this concept (KBS-3), copper canisters with high level waste will be emplaced in deposition holes from a system of tunnels. Blocks of highly compacted swelling bentonite clay are placed in the holes leaving ample space for the canisters. At the final closure of the repository, the galleries are backfilled with a mixture of sand and bentonite. This repository design aims to make the disposal system as redundant as possible. Although the KBS-3 concept is the reference concept, alternative concepts and/or repository lay-outs are also studied. The main alternative, currently under development at SKB, is disposal in boreholes with depths of 4–5 km. The geoscientific research will to a great extent be guided by the demands posed by the performance and safety assessments, as well as the constuctability issues. Some main functions of the geological barrier are fundamental for the long-term safety of a repository. These are: bedrock mechanical stability, a chemically stable environment as well as a slow and stable groundwater flux. The main time-table for the final disposal of long-lived radioactive waste in Sweden foresees the final selection of the disposal system and site during the beginning of next decade.     

地质系统热-水-力耦合作用的随机建模初步研究

热-水-力(THM)耦合作用是岩石力学与环境地质中的重要基础理论问题,核废料地质处置库周围的缓冲材料和围岩中的热-水-力耦合现象将影响其力学稳定性、热传导性和渗透性,进而影响放射性核素在裂隙岩体中的迁移规律。核废料或放射性废料的地下深埋处置是国际上正在研究的永久性隔离的有效方法之一。因此,对核废料地质处置法安全性评估的一个重要内容就是对裂隙岩体中力学稳定性与构造应力、地下水渗流及热载荷等的耦合作用之数值模拟和评估。这已成为当前刻不容缓的重要的环境影响评价课题。笔者研究了温度场-渗流场-应力场中热传导系数和渗透率以及岩体力学参数的空间变异性,用实验方法研究三场耦合效应及裂隙岩体的场性能等效处理,试图建立热-水-力耦合作用的随机性数学模型及可视化数值模拟方法,为核废料地质处置安全性评估提供直观的新方法。     

The COUPLEX Test Cases: Nuclear Waste Disposal Simulation

The models appearing in the COUPLEX benchmark are a set of simplified albeit realistic test cases aimed at simulating the transport of radionuclides around a nuclear waste repository. Three different models were used. The first test case is related to simulations based on a simplified 2D far-field model close to those used for safety assessments in nuclear waste management. It leads to a classical convection diffusion type problem, but with highly variable parameters in space, highly concentrated sources in space and time, very different time scales and accurate results expected even after millions of years. The second test case is a simplification of a typical 3D near-field computation, taking into account the glass dissolution of vitrified waste, and the congruent release of several radionuclides (including daughter products), with their migration through the geological barrier. The aim of the third test case is to use the results of the near-field computation (COUPLEX 2) to drive the behavior of the nuclide source term in the far-field computation (COUPLEX 1). The modeling of this last case was purposely left rather open, unlike the previous two, leaving the choice to participants of the way the coupling should be made.     

Natural analogues for containment-providing barriers for a HLW repository in salt

In 2005, a German research project was started to develop a novel approach to prove safety for a HLW repository in a salt formation, to refine the safety concept, to identify open scientific issues and to define necessary R&D work. This project aimed at identifying the key information for a HLW repository in salt. One important question is how this information may be best fulfilled by natural analogue studies. This question is answered by starting a review of the required key information needs of the safety case (post-closure phase) in order to assess whether or not these requirements can be supported by natural analogues information. In order to structure the review and to address the key elements of the safety concepts, three types of natural analogues are distinguished: (i) natural analogues for the integrity of the geological barrier, (ii) natural analogues for the integrity of the geotechnical barriers and (iii) natural analogues for release scenarios. For the safety case in salt type (i) and (ii) are of highest importance and are treated in this paper. The assessment documented in this paper on the one hand indicates the high potential benefit of natural analogues for a safety case in salt and on the other hand helps to focus the available human and financial resources for the safety case on the most safety-relevant aspects.     

High-level waste disposal,ethics and thermodynamics

Moral philosophy applied to nuclear waste disposal can be linked to paradigmatic science. Simple thermodynamic principles tell us something about rightness or wrongness of our action. Ethical judgement can be orientated towards the chemical compatibility between waste container and geological repository. A container-repository system as close as possible to thermodynamic equilibrium is ethically acceptable. It aims at unlimited stability, similar to the stability of natural metal deposits within the Earth’s crust. The practicability of the guideline can be demonstrated.     

Glasses for immobilization of low- and intermediate-level radioactive waste

Reprocessing of spent nuclear fuel (SNF) for recovery of fissionable elements is a precondition of long-term development of nuclear energetics. Solution of this problem is hindered by the production of a great amount of liquid waste; 99% of its volume is low- and intermediate-level radioactive waste (LILW). The volume of high-level radioactive waste (HLW), which is characterized by high heat release, does not exceed a fraction of a percent. Solubility of glasses at an elevated temperature makes them unfit for immobilization of HLW, the insulation of which is ensured only by mineral-like matrices. At the same time, glasses are a perfect matrix for LILW, which are distinguished by low heat release. The solubility of borosilicate glass at a low temperature is so low that even a glass with relatively low resistance enables them to retain safety of under-ground LILW depositories without additional engineering barriers. The optimal technology of liquid confinement is their concentration and immobilization in borosilicate glasses, which are disposed in shallow-seated geological repositories. The vitrification of 1 m³ liquid LILW with a salt concentration of ～300 kg/m³ leaves behind only 0.2 m³ waste, that is, 4–6 times less than by bitumen impregnation and 10 times less than by cementation. Environmental and economic advantages of LILW vitrification result from (1) low solubility of the vitrified LILW in natural water; (2) significant reduction of LILW volume; (3) possibility to dispose the vitrified waste without additional engineering barriers under shallow conditions and in diverse geological media; (4) the strength of glass makes its transportation and storage possible; and finally (5) reliable longterm safety of repositories. When the composition of the glass matrix for LILW is being chosen, attention should be paid to the factors that ensure high technological and economic efficiency of vitrification. The study of vitrified LILW from the Kursk nuclear power plant with high-power channel reactors (HPCR; equivalent Russian acronym, RBMK) and the Kalinin nuclear power plant with pressurized water reactors (PWR; equivalent Russian acronym VVER) after their 14-yr storage in the shallow-seated repository at the MosNPO Radon testing ground has confirmed the safety of repositories ensured by confinement properties of borosilicate matrix. The most efficient vitrification technology is based on cold crucible induction melting. If the content of a chemical element in waste exceeds its solubility in glass, a crystalline phase is formed in the course of vitrification, so that the glass ceramics become a matrix for such waste. Vitrified waste with high Fe; Na and Al; Na, Fe, and Al; Na and B is characterized. The composition of frit and its proportion to waste depends on waste composition. This procedure requires careful laboratory testing.     

高放废物处置库选址中低渗透介质地质研究的几个问题

低渗透介质是阻碍有害物质在地下迁移良好的天然屏障, 因此成为高放废物处置库围岩类型的首选。本文通过对高放废物处置库选址中地质研究的回顾, 阐述了低渗透介质地质研究的特点, 对地质参数测定、取样、水流模拟、地球化学模拟进行了重点介绍。