Minimal alteration of montmorillonite following long-term interaction with natural alkaline groundwater: Implications for geological disposal of radioactive waste

Bentonite is one of the more safety-critical components of the engineered barrier system in the disposal concepts developed for many types of radioactive waste. Bentonite is utilised because of its favourable properties which include plasticity, swelling capacity, colloid filtration, low hydraulic conductivity, high retardation of key radionuclides and stability in geological environments of relevance to waste disposal. However, bentonite is unstable under the highly alkaline conditions induced by Ordinary Portland Cement (OPC: initial porewater pH > 13) and this has driven interest in using low alkali cements (initial porewater pH9-11) as an alternative to OPC. To build a robust safety case for a repository for radioactive wastes, it is important to have supporting natural analogue data to confirm understanding of the likely long-term performance of bentonite in these lower alkali conditions. In Cyprus, the presence of natural bentonite in association with natural alkaline groundwater permits the zones of potential bentonite/alkaline water reaction to be studied as an analogy of the potential reaction between low alkali cement leachates and the bentonite buffer in the repository. Here, the results indicate that a cation diffusion front has moved some metres into the bentonite whereas the bentonite reaction front is restricted to a few millimetres into the clay. This reaction front shows minimal reaction of the bentonite (volumetrically, less than 1% of the bentonite), with production of a palygorskite secondary phase following reaction of the primary smectites over time periods of 10⁵–10⁶ years.     

极低放射性废物地质处置中的几个问题

极低放废物是国际上近年来新提出的一种放射性废物,目前尚未形成公认的定义和处置方案。针对地质填埋中几个关键科学问题进行了初步探讨,涉及到的内容主要包括区域地壳的稳定性、基础地质条件、水文地质条件、地球化学条件、地质灾害、安全评价等方面,最后提出了今后几个重要的研究方向。     

极低放射性废物地质处置中的几个问题

极低放废物是国际上近年来新提出的一种放射性废物,目前尚未形成公认的定义和处置方案。针对地质填埋中几个关键科学问题进行了初步探讨,涉及到的内容主要包括区域地壳的稳定性、基础地质条件、水文地质条件、地球化学条件、地质灾害、安全评价等方面,最后提出了今后几个重要的研究方向。     

System reliability analysis for near-surface radioactive waste disposal facilities

The low- and intermediate-level radioactive wastes are generally disposed in near-surface disposal facility (NSDF). The NSDF is composed of engineered barriers. The probabilistic safety assessment model has been developed to analyse the performance of NSDF. The endpoints of the assessment are the concentration of the radionuclide in the groundwater and the corresponding dose rate of the radionuclide. The barrier system can have multiple failure modes but practically the possible failure modes are failure of top cover, failure of waste container, degradation of waste form and failure of bottom cover, which are usually considered as independent failure events. Through a sensitivity analysis, the most critical parameters affecting the design reliability for failure criteria are identified as the groundwater velocity and distribution coefficient. The study shows that for the NSDF considered, there is a high degree of dependence between the failure modes, and demonstrates the probability of the simultaneous occurrence of failures. Thus, the need to consider the system reliability in the NSDF is highlighted. The study also advocates the use of optimisation techniques to evaluate the probability of failure, which provides a better estimate of the probability of failure, as validated from the results obtained from Monte Carlo simulations.     

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.     

包气带黄土环境研究与放射性废物处置

黄土中的粘粒含量较高、比表面积较大、吸附性较强、阳离子交换容量也较高,这些因素对迟滞放射性核素的迁移有利。而且黄土的非饱和渗透系数K(θ)远远低于它的饱水渗透系数K_(?)。使得放射性核素在黄土中的迁移很慢。因此,包气带黄土作为中、低放射性废物处置库的环境屏障是可行的。     

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。     

高放废物地质处置系统核素迁移模型研究

高放废物地质处置系统核素迁移研究是高放废物安全处理和处置的重要内容.从整个系统的角度出发,把高放废物地质处置系统核素迁移过程分为三种模式:工程屏障中的释放-扩散模型;地质屏障中的对流-弥散模型和在生物圈中的分区传递模型.在分区传递模型中利用转移系数描述核素在各分区之间的迁移.通过对每种模型核素迁移机理的分析得出概化模型,并给出了相应的数学描述.     

Principles of geomechanical safety assessment for radioactive waste disposal in salt structures

Today, a large amount of knowledge is available concerning various sites of potential high active waste (HAW) repositories in salt media. Domal Zechstein salt formations have been examined at several sites in Germany. Extensive R&D work was initiated in the former Asse Salt Mine in order to explore the suitability of salt for waste isolation by laboratory tests, theoretical studies and in-situ tests with test results forming a technological base for future repository development.

Resulting from the inhomogeneity of salt structures the demanded safety of a permanent repository for radioactive wastes can be demonstrated only by a specific site analysis in which the entire system, “the geological situation, the repository, and the form and amount of the wastes” and their interrelationships are taken into consideration.

The site analysis has three essential tasks: (1) Assessment of the thermomechanical load capacity of the host rock, so that deposition strategies can be determined for the site; (2) Determination of the safe dimensions of the mine (e.g. stability of the caverns and safety of the operations); and (3) Evaluation of the barriers and the long-term safety analysis for the authorization procedure.

The geotechnical stability analysis is a critical part of the safety assessment. Engineering–geological study of the site, laboratory and in situ-experiments, geomechanical modelling, and numerical static calculations comprise such an analysis.

Within a scenario analysis — according to the multi-barrier principle, the geological setting is checked to be able to contribute significantly to the waste isolation over long periods. The assessment of the integrity of the geological barrier can only be performed by making calculations with geomechanical and hydrogeological models. The proper idealization of the host rock in a computational model is the basis of a realistic calculation of stress distribution and excavation damage effects. The determination of water permeability along discontinuities is necessary in order to evaluate the barrier efficiency of each host rock.

In this paper some important processes for the performance assessment are described, namely creep and fracturing, permeability and infiltration, and halokinesis and subrosion.

For the future, the role and contributions of geoscientific and rock mechanics work within the safety assessment issues (e.g. geomechanical safety indicators) must be identified in greater detail, e.g. considerations of geomechanical natural analogy for calibration of constitutive laws.     

Subseabed disposal of radioactive waste: effects of consolidational fluid flow

Subseabed disposal of radioactive waste applies a multiple-barrier concept with the sediment being the most important barrier for preventing a release of nuclides into the biosphere. While many investigations have been carried out to analyze the risk potential in this type of disposal, the effects of sediment consolidation and associated fluid flow have not fully been taken into consideration. Here, possible effects of consolidational fluid flow in the penetrator disposal option and possible consequences to the transport of nuclides in the sediment are analyzed. Results of numerical experiments demonstrate that consolidation contributes to the transport of radioactive nuclides released from containers buried in the sediment and to the release of nuclides at the sediment-water interface. Both depend on geological conditions and to a large extent on possible alterations of hydraulic conductivity i of the sediment in the vicinity of the entry path of a penetrator.Symbols c concentration ml m^–3 - c _a concentration of adsorbed solute mg kg^–1 (relative to dry weight of sorbing substance) - c _in solute concentration of source q mg m^–3 - c ₀ initial concentration mg m^–3 - ID dispersion tensorm ²s^–1 - ID ^* diffusion tensor m²s^–1 - D coefficient of dispersion m²s^–1 - d ₀ coefficient of molecular diffusion m²s^–1 - d coefficient of effective diffusion m²s^–1 - g gravity m²s^–1 - h piezometric pressure m - k hydraulic conductivity m²s^–1 - m mass kg - p pressure Pa - q source/sink m³s^–1 - S ₀ specific surface m²m^–3 - t time s - v velocity m s^–1 - x, z cartesian coordinates m - compressibiliy of sediment m²N^–1 - _L longitudinal dispersivity m - effective porosity (decimal fraction) - density kg m^–3 - _s density of sediment kg m^–3 - _w density of water kg m^–3 - decay constant per s - kinematic viscosity m²s^–1     

高放废物地质处置库选址中的水文地质调查

本文根据国内外研究经验,讨论了高放废物地质处置库选址中水文地质调查的目的、过程、方法和作用,同时对水文地质参数的获取进行了阐述,旨在为今后研究工作提供参考。     

高放废物深地质处置地下水流数值模拟方法研究进展

地下水流数值模型不仅是认识深部水动力场形成演化机制的有效工具,也是建立核素迁移数值模型的基础,因而是高放废物处置场选址和安全评价中重要的技术手段。高放废物深地质处置地下水流数值模拟方法较多,如何选择适当的方法也是值得关注的问题。针对高放废物深地质处置地下水流数值模拟技术展开研究,通过阅读大量国内外文献,文章系统阐述了目前常用的4 类地下水流数值模拟方法的研究进展、适用条件和实例应用;综述了深地质处置中常用的模型不确定性分析方法及研究成果,列表给出了适用于放射性废物地质处置的地下水流数值模拟软件及其在废物处置选择和安全评价中的应用。研究结果表明:等效连续介质模型适用于大区域、长序列、裂隙发育程度较高或较均匀的地区,该类模型方法成熟、所需的数据和参数易于获得,但是不能精确刻画裂隙介质中地下水的流动特征。离散裂隙网络模型适合解决处置场地、储罐尺度等需要精细刻画的地下水流问题,但由于需要大量裂隙及其连通性数据、相关参数等,该方法存在着工作量大、耗时多的缺点。双重介质模型主要用于解决区域尺度裂隙水流问题,但并不能表现出裂隙介质的各向异性、不连续性等特征,因而适用范围存在一定的限制。等效-离散耦合模型可以通过区域分解法对裂隙密度大的区域采用等效连续介质模型,对于裂隙密度较小的地区采用离散裂隙网络模型,从而更符合一般地质条件下裂隙渗流的特征,但也存在交换量难以确定、模型耦合技术问题。通过灵敏度分析,将不同敏感因子对模型敏感指标的影响程度进行排序,提高模型精度、减少参数不确定性分析的工作量。蒙特卡罗法是目前常用的一种模型不确定性方法,原理简单、易于实现。文章展望了数值模型在仿真性、不确定性分析、预测和多介质耦合等方面的研究前景。     

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.     

我国高放废物深地质处置战略规划探讨

本文探讨我国高放废物地质处置的战略规划，提出我国高放废物处置库的开发可参考采用“三步曲”式的技术路线，即处置库选址和场址评价一特定场址地下实验室研究一处置库建设。处置库的选址和场址评价工作可与地下实验室研究的相关工作结合。以2040年前后建成处置库为目标，把工作划分为4个阶段，即选址和场址评价阶段、场址确认和地下实验室建设阶段、现场实验和示范处置阶段及处置库建设阶段，规划出各阶段的工作目标和工作内容。论证各项工作内容之间的逻辑关系，指出选址和场址评价是基础、基础研究和地下实验室研究是支撑、性能评价是“指挥棒”、设计并建造出符合标准的处置库是目标。将选址工作划分为预选地段对比、预选场址对比和场址确认3个阶段。按此规划设想，我国将在2015年以前确定处置库和地下实验室的场址，并开始建造地下实验室。2025年左右建成地下实验室，2040年建成处置库。     

热-力耦合条件下高放废物处置室间距研究

根据我国高放废物地质处置研究开发规划指南、北山预选区实测地质环境参数和处置库围岩和缓冲材料实测性能参数,采用有限差分计算程序FLAC2D模拟不同处置室间距时高放废物处置库热-力耦合条件下的温度场、应力场和位移场规律。经多方案比较,处置室中线间距为2倍处置室直径时,缓冲材料中的最高温度在95℃左右;由于高放废物释热,引起岩体膨胀,产生热应力,在处置室下角部产生应力集中,但均小于处置库围岩的长期强度;处置库围岩产生的膨胀位移,均在其弹性应变范围内,不会引起处置库围岩破坏。该方案可作为北山预选区建造高放废物处置库的参考依据。     

Glass-iron-clay interactions in a radioactive waste geological disposal: An integrated laboratory-scale experiment

Glass-iron-clay setups were reacted at 90 °C for 6-18 months to investigate the coupled interactions between glass alteration, Fe corrosion and clay transformation. The reacted interfaces were probed at the microscopic level using complementary characterization methods (scanning electron microscopy coupled with energy-dispersive X-ray analysis, micro-Raman spectroscopy, micro X-ray diffraction, micro X-ray fluorescence spectroscopy, and micro X-ray absorption near-edge structure spectroscopy). The 10-μm thick Fe foil was fully corroded within 10 months, exposing glass to the pore solution. Iron corrosion led to the formation of a layer containing mostly magnetite, siderite and Fe-rich phyllosilicates with one tetrahedral and one octahedral sheet (TO) or two tetrahedral and one octahedral (TOT) sheet per layer. The clay in contact with this corrosion layer was enriched in siderite (FeCO₃). Glass alteration resulted in the formation of a gel layer whose thickness increased with reaction time (from 20 μm after 6 months to 80 μm after 18 months) and a thin layer of secondary precipitates that concentrated lanthanides, P, and Mo. Assuming conservative behavior of Zr, the Si molar concentration in the gel is about 57% that in the glass. Glass dissolution remained at a rate close to the initial dissolution rate r₀. The data are consistent with glass dissolution sustained by the uptake of dissolved Si and charge-compensating cations on secondary (corrosion) products, thus maintaining the gel porosity open and facilitating the leaching of easily soluble elements.     

高放废物地质处置新场岩体三维地质模型构建与应用

我国高放废物地质处置经过30多年研究,已初步确定新场为地下实验室推荐场址。开展新场岩体三维地质建模研究,一方面能够充分利用已有资料准确地表达各种地质现象,直观再现地质单元的空间展布及其相互关系;另一方面可以挖掘隐含的地质信息,方便地质分析和工程决策,这在地下实验室研究阶段非常重要。在对新场已有资料综合分析和解译的基础上,建立了岩体的三维地质模型,直观地再现了新场岩体地质环境特征;并基于地质建模结果,开展了钻孔设计等工程应用,取得了较好的效果。本研究可为我国高放废物地质处置后续工作中地质分析与工程设计提供有益参考和技术支持。     

高放废物地质处置元数据设计与编辑模块开发

在对元数据的含义、作用、组成及对高放领域研究的数据内容和特点分析的基础上,对元数据编辑模块的设计进行了详细描述,并在VS2005平台下应用C#语言进行了模块的实际开发,最终实现了便于进行数据共享的XML格式存储功能。元数据编辑模块的应用将促进我国高放废物地质处置信息化工作的进一步深入。     

高放废物深地质处置地下水数值模拟应用综述

开展地下水数值模拟研究是高放废物处置场地安全评价的重要组成部分,然而深地质处置介质类型的复杂性、基岩深部资料的相对匮乏性导致模拟结果存在不确定性,如何刻画深部地下水动力场并评估可能引起的风险已成为高放废物处置安全评价中重点关注的问题。在大量文献调研的基础上,综述了世界典型国家高放废物深地质处置场地的地下水数值模拟与不确定性分析应用,并归纳总结该领域研究经验,得到以下认识：（1）深地质处置场深部构造、裂隙的发育与展布决定了地下水循环条件,探究适用于基岩裂隙地区新的水文地质试验方法是提高地下水数值模型仿真性的基础;（2）不同尺度模型融合是解决深地质处置地下水模拟的有效技术方法,区域尺度多采用等效连续介质法,场地尺度使用等效连续多孔介质和离散裂隙网络耦合模型,处置库尺度使用离散裂隙网络方法,其次需重点关注未来大时间尺度下放射性核素在地质体中的迁移转化规律,模拟预测场址区域地下水环境长期循环演变对核素迁移的潜在影响;（3）考虑到不同的处置层主岩岩性以及在多介质中发生的THMC（温度场—渗流场—应力场—化学场）过程,目前国内外常用的地下水模拟软件有：Porflow、Modflow、GMS及MT3DMS等用于模拟孔隙或等效连续介质,Connectflow、Feflow及FracMan等用于模拟地下水和核素在结晶岩、花岗岩等裂隙中的迁移,TOUGH系列软件主要应用于双重介质的水流、溶质及热运移模拟;（4）指导开展有针对性的模型和参数的不确定性分析工作,减少投入工作量,提高模型精度,并可针对处置库长期演变、废物罐失效、极端降雨等多情景预测模拟,为处置库安全评价及设计提供基础数据支撑;（5）针对我国深地质处置地下水数值模拟研究现状,下一步应加强区域地质、水文地质、裂隙测量以及现场试验等相关的调查及监测工作,多介质耦合、多场耦合模拟及不确定性分析研究将会是未来的研究重点。