An In-situ Experimental Study of Fractures Network Identification within Bedrock by High-density Electrical Resistivity Tomography

对于以低渗透性基岩为处置围岩的核废物地质处置库来说,基岩中的裂隙网络是地下水进入处置库和放射性核素从处置库向外界环境迁移的主要通道.因此,研究基岩内裂隙网络的空间分布特征,对核废物的地质处置是否安全具有十分重要的意义.本文以花岗岩为例通过野外试验,分析研究了利用高密度电阻率成像法(ERT),对处于不同状态条件下的基岩中裂隙网络进行三维原位识别的效果及能力.结果表明在以高阻值为主要电特性的基岩中,利用ERT对基岩中三维裂隙网络进行原位识别是可行的,并且裂隙的电阻率特征能较好的反映出裂隙的空间分布位置和形态,但对裂隙网络识别的效果与测线方向有关.     

基于高密度电阻率法的土体干缩裂隙动态发育过程精细监测研究

土体在气候作用下发育干缩裂隙是一种常见的自然现象,裂隙的存在会极大弱化土体的工程性质,诱发许多岩土与地质工程问题。为了实时掌握黏性土中干缩裂隙网络的发展状态,提出了一套基于高密度电阻率层析成像技术（ERT）的土体干缩裂隙动态发展过程精细监测方法。分别开展模型试验及原位试验,利用自行研制的测定系统持续采集电流-电位差数据,随后利用自行开发的有限元法电阻率层析成像（FemERT）系统进行数据处理,获取了裂隙网络在不同发育阶段的空间分布特征。结果表明：（1）ERT可以实现土体裂隙发育过程的精细监测,具备监测三维裂隙网络几何形态的能力,裂隙宽度的识别精度达到毫米级,裂隙深度的识别精度达到厘米级;（2）ERT的感度分布特征解释了裂隙发育对于土体电阻率的影响规律,测定电阻值时程曲线因裂隙产生位置的不同而呈现不同的变化规律;（3）反演电阻率及其相对变化率（Rev）可以直观表征裂隙网络在不同阶段的空间几何形态,凸显裂隙动态发育过程对于土体导电性的影响。     

巴彦诺日公地段花岗岩微裂隙发育特征研究

花岗岩是高放射核废料地质处置库的主要围岩之一,其水力学特征优劣直接决定了花岗岩体能否有效地阻隔地下水对处置库中核废物的侵袭。围岩微裂隙结构和化学风化程度是水力学特征的直观表现,微裂隙结构量化和化学风化程度计算对高放射核废料地质处置库埋藏深度的比选有一定的科学意义。本文以阿拉善某600 m的花岗岩钻孔内不同深度的岩芯作为研究对象,得到大量的微裂隙显微照片;通过测网法和图像处理技术,获得岩芯数字化的裂隙空间分布图像,并从中提取了微裂隙特征参数（裂隙条数、隙宽、裂隙率、裂隙面密度参数等）;然后,对微裂隙特征参数进行了统计分析,描述了裂隙空间分布变异性。花岗岩样品的化学风化程度评价指标有:天然含水率、CIA。同时,利用SEM-EDS探测裂隙部位形貌特征及元素含量等数据。最后,对花岗岩完整性与核废料处置库适宜性进行综合评价。通过本研究,得到了以下结论:（1）研究区的花岗岩微裂隙发育情况随深度增大逐渐变弱。（2）花岗岩受外力作用后微裂隙首先发育在石英中,其次是在长石和黑云母中。（3）临近破碎带,裂隙率、平均隙宽、裂隙条数均增大,同时受后期热液填充的影响,CIA也会表现出高值。（4）破碎带的发育,对周边完整岩体的微观破裂影响距离可以超过10.6 m,后期热液对下方完整岩体的侵染蚀变影响距离可以超过50.27 m。（5）通过综合评价,来自该钻孔不同深度的7个样品中,取自地下598.8 m处的样品所代表的位置最适宜作为核废料处置库深埋区。     

加拿大和美国高放废物地质处置研究概况

本文主要介绍了作者2000年对加拿大和美国进行科学访问时实地考察所了解到的加拿大的核素在裂隙和孔隙介质中的迁移模式，地下实验室，白壳实验室（Whiteshell Labs)，高放废物处置库场地预选和美国拟作为高放废物处置库的尤卡山场地和已投入运行的军用超铀废物处置库的废物镉离实验工厂；着重论述了这两个国家的高放废物地质处置研究现状、经验和所取得的成果。     

裂隙岩体渗透系数确定方法研究

裂隙岩体渗透系数以及渗透主方向的确定对研究岩体渗透性大小及各向异性具有重要意义。高放废物地质处置库介质岩体的渗透性能将直接影响其使用安全性。本文运用离散裂隙网络模拟的方法对我国高放废物处置库甘肃北山预选区3#钻孔附近裂隙岩体进行了渗透性质分析。通过对3#钻孔171.5~178.0m段压水试验数据的反演,标定了离散裂隙网络渗流模型中的裂隙渗透参数(导水系数T)。利用标定的离散裂隙网络模型对场区裂隙岩体进行了渗流模拟,确定了该区域裂隙岩体的渗流表征单元体(REV)的尺寸大小以及渗透主值和主渗透方向。运用离散裂隙网络模型计算得出的渗透主值的几何均值与现场压水试验计算结果较接近,证明了计算结果的有效性。     

黏土岩饱和过程中水分运移规律试验研究

黏土岩作为高放废物地质处置库的备选介质,目前得到世界各国的高度重视。黏土岩地质处置库巷道施工过程中,一方面,围岩因开挖损伤生成裂隙使得渗透性增强,对核素的阻滞作用降低;另一方面,在应力和水的耦合作用下,黏土岩良好的裂隙渗透损伤自修复能力使得围岩的渗透性逐渐恢复接近于原始状态。基于电阻率测试,首先开展了黏土岩试样在不同条件下的饱和过程试验研究,得到了黏土岩试样饱和过程中等效电阻率的变化规律,分析了不同损伤程度试样、盐溶液对等效电阻率的影响,进而揭示黏土岩饱和过程中水分运移规律。试验结果表明：（1）等效电阻率随着含水率增加而逐渐减小,并逐渐趋于一个稳定值;（2）等效电阻率的大小不仅与含水率有关,试样内部裂隙的存在也会影响等效电阻率分布,这一发现为电阻率法可以探测试样中裂隙的存在提供了依据;（3）水流在黏土岩中扩散,内部裂隙成为优先通道,水流在裂隙中的快速扩散加快了黏土岩的饱和速度。同时,随着黏土岩中水分与黏土矿物的水化膨胀反应,内部裂隙有一定程度闭合,加深对裂隙闭合机制认识,通过电阻率测试可以有效地揭示这一过程。     

高放射性核废物深地质处置的环境问题

基于“多重屏障原理”的深地质处置是国内外公认的处置高放射性核废物的合适方法。本文从处置库建造与运行对环境的影响及环境对处置库反作用两方面探讨高放废物深地质处置中涉及的有关环境问题,并指出为确保处置库的长期安全性,必须特别注重选址及多因素耦合作用研究。     

高放射性核放心物深地质鼾的环境问题

基于“多重屏障原理”的深地质处置是国内外公认的处置高放射性核废物的合适方法，本文从处置库建造与运行对环境的影响及环境对处置库反作用两方面探讨高放废物深地质处置中涉及的有关环境问题，并指出为确保处置库的长期安全性，必须特别注重选址及多因素耦合作用研究。     

基于离散裂隙网络模型的核素粒子迁移数值模拟研究

高放废物地质处置过程中涉及的核素在围岩裂隙地下水中的迁移问题已引起广泛关注,数值模拟是研究核素粒子运移的重要方法。目前裂隙介质中渗流模型主要是等效连续介质模型、双重介质模型和离散裂隙网络模型。对于岩体尺度裂隙地下水的流动,离散裂隙网络模型能充分表现裂隙介质的各向异性、不连续性等特征。因此,针对裂隙介质准确概化及核素迁移模拟等难点,文章结合Monte Carlo随机生成裂隙方法、裂隙渗流有限元算法和高放射性核素衰变方程等方法,依据花岗岩深钻孔裂隙统计数据,采用离散裂隙网络模型对内蒙古阿拉善高放废物地质处置预选区展开了核素粒子迁移数值模拟研究,并讨论了实例预测分析结果。结果显示:针对设定的地质模型,核素粒子从中心运移到边界的迁移路径长度平均为1293.35 m,粒子运移到边界耗费的时间平均为1.70E+11 d。     

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。     

填砂裂隙岩体渗流传热模型试验与数值模拟

选取中国高放射核废物地下处置库重要预选场区--甘肃北山地区的花岗岩,加工组合成规则裂隙岩体,将垂直裂隙用粒径为0.5～0.63 mm的砂土填充,进行了裂隙水渗流传热试验;对模型试验进行了数值模拟,进而计算分析了热源温度、裂隙水流速和裂隙开度变化对裂隙岩体模型稳态温度场的影响。模型试验表明,当热源温度维持在120 ℃时,裂隙水仍无相变,裂隙岩体模型稳态温度场分布规律与热源温度为95 ℃时一致;热源温度越高,热源的水平影响距离越大,模型达到稳态需要的时间越长;裂隙填砂加强了裂隙两侧岩石之间的热传导,热源的水平影响距离和模型到达稳态需要的时间均明显大于无填充裂隙岩体模型的情况。模型试验得到的岩体模型温度场与数值计算得到的岩体模型温度场规律一致。试验过程中裂隙岩体模型在边界上存在一些热量散失,无法与数值计算中的绝热边界条件等同,致使试验数据低于数值计算值,并且热源温度越高,两者之间的差异越大。模型试验和数值计算均表明,邻近热源侧的裂隙水渗流对模型的温度场分布起控制作用,而远离热源侧的裂隙水渗流则主要影响该侧的边界温度和模型达到稳态所需要的时间。数值参数敏感性分析表明,裂隙水流速与裂隙开度越大,裂隙水对水平传热的阻滞作用越明显。     

Multi-scale groundwater flow modeling during temperate climate conditions for the safety assessment of the proposed high-level nuclear waste repository site at Forsmark,Sweden

Forsmark in Sweden has been proposed as the site of a geological repository for spent high-level nuclear fuel, to be located at a depth of approximately 470 m in fractured crystalline rock. The safety assessment for the repository has required a multi-disciplinary approach to evaluate the impact of hydrogeological and hydrogeochemical conditions close to the repository and in a wider regional context. Assessing the consequences of potential radionuclide releases requires quantitative site-specific information concerning the details of groundwater flow on the scale of individual waste canister locations (1–10 m) as well as details of groundwater flow and composition on the scale of groundwater pathways between the facility and the surface (500 m to 5 km). The purpose of this article is to provide an illustration of multi-scale modeling techniques and the results obtained when combining aspects of local-scale flows in fractures around a potential contaminant source with regional-scale groundwater flow and transport subject to natural evolution of the system. The approach set out is novel, as it incorporates both different scales of model and different levels of detail, combining discrete fracture network and equivalent continuous porous medium representations of fractured bedrock.     

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.     

Mechanisms and Models for Bentonite Erosion Used for Geologic Disposal of High Level Radioactive Waste: A Review

The performance of the bentonite buffer in nuclear waste repository concept relies to a great extent on the buffer surrounding the canister having sufficient dry density. Loss of buffer material caused by erosion remains as the most significant process reducing the density of the buffer. In the worst case, the process is assumed to last as long as the free volume between the pellets in the pellets filled regions is filled with groundwater. Erosion rate and mass erosion are calculated based on the erosion model, and the measures are presented to prevent the geological disaster due to bentonite erosion. The groundwaters may solubilise the smectite particles in the bentonite and carry them away as colloidal particles. A dynamic model is developed for sodium gel expansion in fractures where the gel soaks up groundwater as it expands. The model is based on a force balance between and on smectite particles, which move in the water. Attractive van der Waals forces, repulsive electric diffuse layer (DDL) forces, gravity and buoyancy forces and forces caused by the gradient of chemical potential of the particles act to move the particle in the water. The effect of the fracture width and the frictions between particles and water and surrouding rock is analysed based on erosion model. The DDL forces strongly depend on the type of clay minerals and the type of ion and concentration in the water surrounding the particles. In the designed safe use of nuclear waste disposal (tens of thousands of years to hundreds of thousands of years), the safety of nuclear waste disposal is affected by the hydrodynamic and chemical effects, and bentonite erosion. Due to the bentonite erosion, the buffer/backfill layers become loose, and their permeability increases, which causes the nuclear element diffusion and convection, and even the nuclear disaster. In this paper, the mechanisms, models, experiments and control measures of bentonite erosion were systematically summarized. The current deficiencies of bentonite erosion were pointed out, and new methods were put forward to carry out the research for bentonite erosion. The measures were presented to prevent the geological disaster due to bentonite erosion through changes. The project is not only academic innovation, but also has a large practical significance. The research results of this project can be widely applied to the design, construction and maintenance of the bentonite buffer in nuclear waste repository.     

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。     

Methodology for bounding calculations of nuclear criticality of fissile material accumulations external to a waste container at Yucca Mountain, Nevada

The possibility of nuclear criticality, however remote, in the vicinity of the proposed repository at Yucca Mountain, Nevada generates justified concerns and may impact the performance of the repository. A heuristic approach is presented here for determining the amount, spatial distribution and other characteristics of fissile material accumulation in the rock beneath a waste package that could contribute to such an event. This study is concerned primarily with waste packages containing special spent fuel from the Department of Energy and high-level nuclear waste glass. Mixing with less alkaline waters and the subsequent drop in pH is the mechanism that is most efficient for precipitating fissile material from the waste package internal leachate, in contrast to natural deposits in which redox changes are the main precipitation driver. External accumulation size is determined by (1) computing the chemical composition of the leachate leaving a package as its internal materials degrade (with the batch geochemical code EQ3/6), (2) determining precipitation of fissile material into mineral phases (using the 1D geochemical code PHREEQC) as the effluent mixes with percolation water, and (3) heuristically scaling results to a 3D volume and computing the criticality coefficient (using the code MCNP). Loci for accumulation are the multiple lithophysal cavities and the fracture system. A bounding conservative approach is used by necessity in Step 3. Nuclear criticality is sensitive to small variations in the distribution of fissile material and parameters of natural systems vary by orders of magnitude. Because the most likely combinations of parameters are not conducive to nuclear criticality, this study focuses on extreme values of parameter probabilistic distributions, such as limited flow into the package associated with a large percolation rate, combinations of material degradation rates favoring actinide release, and very high host-rock porosity values. By considering these combinations, most favorable to criticality but unlikely, it was concluded that external nuclear criticality is not a concern at the proposed repository.