Numerical analysis of thermal process in the near field around vertical disposal of high-level radioactive waste

For deep geological disposal of high-level radioactive waste(HLW)in granite,the temperature on the HLW canisters is commonly designed to be lower than100fiC.This criterion dictates the dimension of the repository.Based on the concept of HLW disposal in vertical boreholes,thermal process in the nearfield(host rock and buffer)surrounding HLW canisters has been simulated by using different methods.The results are drawn as follows:(a)the initial heat power of HLW canisters is the most important and sensitive parameter for evolution of temperaturefield;(b)the thermal properties and variations of the host rock,the engineered buffer,and possible gaps between canister and buffer and host rock are the additional key factors governing the heat transformation;(c)the gaps width and thefilling by water or air determine the temperature offsets between them.     

Numerical analysis of thermal process in the near field around vertical disposal of high-level radioactive waste简

For deep geological disposal of high-level radioactive waste (HLW) in granite, the temperature on the HLW canisters is commonly designed to be lower than 100 °C. This criterion dictates the dimension of the repository. Based on the concept of HLW disposal in vertical boreholes, thermal process in the near field (host rock and buffer) surrounding HLW canisters has been simulated by using different methods. The results are drawn as follows: (a) the initial heat power of HLW canisters is the most important and sensitive parameter for evolution of temperature field; (b) the thermal properties and variations of the host rock, the engineered buffer, and possible gaps between canister and buffer and host rock are the additional key factors governing the heat transformation; (c) the gaps width and the filling by water or air determine the temperature offsets between them.     

Inverse modelling of thermal conductivity from temperature measurements at the Prototype Repository, Äspö HRL

The Prototype Repository, at the Äspö HRL (Hard Rock Laboratory), is a demonstration project for the deposition of spent nuclear fuel, and provides a full-scale reference for testing predictive models relating to a spent nuclear fuel repository, both its individual components as well as the complete system. The final layout involves six deposition holes, four in an inner section and two in an outer, each fitted with an electrically heated canister. The access tunnel is backfilled with a mixture of bentonite and crushed rock. In 2001, the inner section was completed and monitoring of the heating process started. Temperature measurements in the rock mass are performed at 37 different points.In this paper, the measured thermal response in the surrounding rock is analysed by inverse modelling of the thermal conductivity of the rock mass. A three-dimensional finite difference model of the prototype repository (canisters, buffers, tunnel, etc.) is used to calculate the transient temperature increase due to the heat generation in the canisters. The value of a homogeneous rock thermal conductivity is chosen to obtain the best fit with measured data for each of the 37 temperature sensor points. The evaluation period for the fitting procedure is varied in order to study sensitivity to different time-scales.Measurements of thermal properties have been conducted within the prototype repository prior to the full-scale test. The thermal properties were predicted based on both field and laboratory measurements. These predictions are verified by comparison with thermal conductivity values calculated through inverse modelling.     

Influence of rock properties on selection of design for a spent nuclear fuel repository

In Sweden, researchers are examining designs for a repository for storage of spent nuclear fuel. By the year 2010 (when nuclear power production will cease, according to parliamentary decision), approximately 10,000 m³ of spent fuel will need to be encapsulated in the repository. One such system, the KBS-3 (abbreviation for Nuclear Fuel Safety, in English), has been evaluated over the past decade. Another design, called the Very Long Hole (VLH) system, is also being considered. The main difference between the two systems is in the shape of the rock openings in the disposal areas. This paper analyzes the properties of the rock in the near field (i.e., the rock affected by the excavation of the repository) and compares them for the two repository systems.     

Interface shear behaviour of tunnel backfill materials in a deep-rock nuclear waste repository in Finland

The need for environmental protection and safety in facilities dedicated for the final safe disposal of spent nuclear fuel is paramount. Highly engineered multi-barriers are widely used in such waste containment facilities in order to provide a tight seal for the waste they contain. In Finland, several research studies have been conducted to investigate the feasibility of the final safe disposal of spent nuclear fuel in crystalline bedrock by incorporating the KBS-3V multi-barrier repository concept. As the saturation of the tunnels in a repository progresses, the pre-compressed bentonite buffer may swell and generate very high swelling pressure in the range of 7–15 MPa. Such high swelling pressure can cause the upheaval and the compression of the tunnel backfill that would eventually decrease the density of the buffer. For various reasons, the current KBS-3V design suggests that the saturated density of the buffer should be maintained within a narrow range of 1950–2050 kg/m³ at all times. As the swelling of the buffer directly influences the saturated density of the buffer, it must be controlled by designing a tunnel backfill that possesses an adequate amount of interface shear strength to sustain any additional pressure that is exerted by the swelling of the buffer. This study presents the findings of a series of direct shear box tests conducted on various tunnel backfill interfaces. Additionally, different types of rock profiles were also tested with the selected backfill materials. Based on the results, it was observed that the interface shear behaviour of different backfill-rock interfaces varied significantly with the surface roughness of the rock, while clay blocks resulted in similar shear behaviour with all the backfill materials.     

Value of information analysis for site investigation programs accounting for variability,uncertainty and scale effects with the Äspö HRL prototype repository as an example

An important feature of underground projects is the early site investigations, performed as a means to identify and quantify hazards. A methodology is presented for identifying the most cost-effective investigation program among a set of alternatives. Methodologies are presented for both investigation of thermal conductivity in hard rock and collection of rock mechanic data for stress induced spalling problems. The cost-effectiveness of an investigation program is estimated by means of value of information analysis (VOIA). Each investigation program of thermal conductivity is associated with uncertainty due to natural variability and lack of knowledge. These uncertainties are taken into account in a simulation model with the aim to estimate the distribution of thermal conductivity values at different scales. The output is a set of thermal conductivity values from which a design parameter can be estimated. The simplest measure of the value of a site investigation is the expected reduction of uncertainty of the design parameter.The methodology is demonstrated with a case study for the prototype nuclear waste repository at Äspö Hard Rock Laboratory, Sweden. A set of four investigation programs for thermal conductivity were evaluated, and the most effective one identified. The application illustrates that an investigation program may supply very different value to a project, depending on how the objective of the investigation is defined. This is demonstrated by using two different objectives and comparing the results. Practical applications of the methodology on both thermal properties and rock mechanics are discussed, with emphasis on site investigations performed by the Swedish Nuclear Fuel and Waste Management (SKB).     

高放废物处置库花岗岩热–力耦合模拟研究

 根据典型高放废物处置库概念模型,应用FLAC3D有限差分程序模拟计算了数百年内热–力耦合(TM)条件下高放废物地质处置库围岩的温度场、应力场和变形场的变化特征,初步得到在一定废物罐热源强度及衰变函数条件下花岗岩体的热力学特征及处置巷道工程设计的合理间距。模拟结果表明：在－500 m水平建造处置库,按设定的释热强度,处置坑合理间距为8～12 m,废物罐表面温度为130 ℃,处置坑中线岩体最高温度为40 ℃,并保持数百年左右。     

缓冲材料热-水-力耦合模型试验研究

高放废物深地质处置采用多重屏障设计体系,缓冲材料是位于废物罐和围岩之间的一道重要的人工屏障。在放射性衰变热、地下水入侵和围岩应力等作用下,缓冲材料经历复杂的热-水-力耦合过程,评价其长期性能对高放废物地质处置库的稳定运行至关重要。缓冲材料模型试验是研究膨润土在多场耦合环境下性能变化的重要途径。中型实验台架是大型实验台架(China-Mock-up)的重要补充,用来模拟与大型实验台架边界相同的环境下,即热量和水分别从缓冲材料的不同侧向另一侧传递和渗透的条件下膨润土的行为特征。通过实时监测缓冲材料在长期加热和加水条件下的温度、相对湿度、力学等特征参数,揭示了在热-水-力耦合条件下缓冲材料的性能变化规律,同时对台架进行了拆解,对拆解样品的含水量、干密度和微观结构等进行了分析测试,研究结果可为高放废物处置库缓冲材料的工程设计提供参数支持。     

用单轴抗压强度试验分析瑞典Äspö 废物 罐回取试验中的岩石损伤

 废物回取试验是一个在瑞典Äspö 地下实验室完成的,历时近5 a,为全尺寸处置库模拟加热试验。试验在一个直径f 1.75 m、深度8.5 m的钻孔中进行。开挖和加热后周边岩石中的温度升高、应力改变,因此,试验中岩石中可能产生的损伤是工程设计中关心的课题之一。为此,试验结束后,在试验孔3个不同深度处沿垂直和平行于最大主应力方向施打6个深度约1.5 m的近水平取样孔,并采集了12组岩样。对这12组岩样用MTS 815 岩石力学试验系统进行了单轴抗压强度试验。从单轴抗压强度、裂隙起始应力、裂隙损伤应力、最大裂隙体积应变和最大总体变进行了对比和分析,试验结果分析表明：从最大裂隙体积应变分析,在垂直于最大主应力方向的处置孔孔壁的岩石上可能存在一些轻微的微破裂为特征的损伤。从宏观力学特性来说,岩石没有任何可测的损伤。     

Thermally induced mechanical and permeability changes around a nuclear waste repository—a far-field study based on equivalent properties determined by a discrete approach

A numerical investigation is conducted on the impacts of the thermal loading history on the evolution of mechanical response and permeability field of a fractured rock mass containing a hypothetical nuclear waste repository. The geological data are extracted from the site investigation results at Sellafield, England.A combined methodology of discrete and continuum approaches is presented. The results of a series of simulations based on the DFN–DEM (discrete fracture network–distinct element method) approach provide the mechanical and hydraulic properties of fractured rock masses, and their stress-dependencies. These properties are calculated on a representative scale that depends on fracture network characteristics and constitutive models of intact rock and fractures. In the present study, data indicate that the large scale domain can be divided into four regions with different property sets corresponding to the depth. The results derived by the DFN–DEM approach are then passed on to a large-scale analysis of the far-field problem for the equivalent continuum analysis.The large-scale far-field analysis is conducted using a FEM code, ROCMAS for coupled thermo-mechanical process. The results show that the thermal stresses of fractured rock masses vary significantly with mechanical properties determined at the representative scale. Vertical heaving and horizontal tensile displacement are observed above the repository. Observed stress and displacement fields also shows significant dependency on how the mechanical properties are characterized. The permeability changes induced by the thermal loading show that it generally decreases close to the repository. However, change of permeability is small, i.e., a factor of two, and thermally induced dilation of fracture was not observed. Note that the repository excavation effects were not considered in the study.The work presented in this paper is the result of efforts on a benchmark test (BMT2) within the international co-operative projects DECOVALEX III and BENCHPAR.     

缓冲胶结的反应-传递模型的试验校正

在KBS-3核废料储存库的早期及非恒温演变期,热梯度变化能改变膨润土缓冲层的矿物特性,且由于矿物质沉淀作用,黏土颗粒将相互胶结起来.据此,使用已发表的有关结晶岩石中缓冲黏土的水-热转换的试验结果,研究并评估一种缓冲胶结的反应-传递模型.模型的预测值能够定性地与高温下缓冲区的观测试验值相吻合,表明次生矿物的沉淀(无水石膏(无定形二氧化硅(opal-CT)(方解石)和由Na基蒙脱石变为Ca 基蒙脱石(皂石的相交替过程.与现场观测中的一种情况不同的是,模型没有预测出这一区域中石英、高岭土和长石的大量溶解,这种不一致现象可能是由于缺乏热动力基本数据造成的,使得在高岭土-蒙脱石混合地层中高电荷蒙脱石变为低电荷蒙脱石过程中的水-热转换在模型中不能加以表示.然而,在整体上模型预测与试验观测还是较为符合的.这表明,用已知的时间-温度及再饱和过程的试验数据,该模型能够用于对KBS-3近场潜在的缓冲胶结进行长至几百年的敏感性分析.     

用单轴抗压强度试验分析瑞典Aspo废物罐回取试验中的岩石损伤

废物回取试验是一个在瑞典(A)sp(o) 地下实验室完成的,历时近5 a,为全尺寸处置库模拟加热试验.试验在一个直径Φ1.75 m、深度8.5 m的钻孔中进行.开挖和加热后周边岩石中的温度升高、应力改变,因此,试验中岩石中可能产生的损伤是工程设计中关心的课题之一.为此,试验结束后,在试验孔3个不同深度处沿垂直和平行于最大主应力方向施打6个深度约1.5 m的近水平取样孔,并采集了12组岩样.对这12组岩样用MTS 815 岩石力学试验系统进行了单轴抗压强度试验.从单轴抗压强度、裂隙起始应力、裂隙损伤应力、最大裂隙体积应变和最大总体变进行了对比和分析,试验结果分析表明:从最大裂隙体积应变分析,在垂直于最大主应力方向的处置孔孔壁的岩石上可能存在一些轻微的微破裂为特征的损伤.从宏观力学特性来说,岩石没有任何可测的损伤.     

Estimating the potential for spalling around a deep nuclear waste repository in crystalline rock

Stress-induced brittle failure (spalling) is probable at a deep geological repository for nuclear waste in crystalline rock. In the early stages of repository design it is unlikely that orientation and magnitudes of the principal stresses and the rock mass strength will be accurately known. A simple methodology is developed for estimating if spalling will occur and the severity of the hazard. The methodology is calibrated to case studies and applied to a site in Sweden. Results from the methodology are expressed in terms of a factor of safety for the mean input values and the probability of spalling based on input parameter distributions. It is shown based on the calibration studies that a factor of safety of 1.25 using the mean values should be adequate to reduce the probability of yielding to negligible levels. The methodology is proposed as a screening tool in the early design stages of a project to identify potential spalling problems.     

基于损伤预测的粘土围岩深地质处置库硐距分析

粘土岩作为我国高放废物深地质处置库建设的预选围岩类型之一,其地下处置库硐距的优化对处置库的设计施工及核素迁移安全具有重大意义。本文基于塔木素粘土岩高放废物深地质处置预选地区天然地应力现场测量及室内岩石力学试验结果,采用弹性力学柯西问题理论分析与有限差分数值模拟相结合的方法,对处置库硐室开挖过程形成的损伤区范围进行预测。通过对比单硐与双硐开挖时不同硐室间距下开挖所形成的损伤区影响范围,来进行处置库硐距的优化,通过对比不同硐距下开挖损伤区的变化,分析当前应力水平下,硐室开挖后重分布应力随与中间岩柱中心点距离的变化趋势。结果表明,d=8~10 m是粘土围岩高放废物处置库的合理的硐距,此时开挖后的相邻硐室间损伤区没有产生明显交圈效应,当硐距大于10 m后,相邻硐室间的影响将逐渐减小,接近单硐开挖情况。该方案可为粘土岩预选区建造高放废物处置库提供参考依据。     

新疆阿奇山地段岩体节理特征研究

我国高放废物处置库新疆预选区的阿奇山地段是有利候选地段之一。本次在阿奇山地段进行详细的地质节理调查,以地表露头及采石坑内的节理为主要对象,通过现场节理调查取得节理数据,共测得3680条节理数据。进行节理几何特征的概率分析,对岩体赋存的结构条件进行有效、准确的描述和定量分析,准确地确定岩体节理的分布。由结果可以看出阿奇山地段的岩体节理以陡倾角的剪节理为主,且节理倾向和倾角符合正态分布规律,节理间距用负指数分布规律很好表示,阿奇山地段各区域的节理间距属于宽间距,由此表明此地段岩体完整性较好。本次研究得到阿奇山地段岩体节理基本特征参数,为高放废物深地质处置库的选址、概念库设计以及进行核素迁移规律研究提供翔实的资料。     

高放废物处置库黏土围岩温度-应力耦合测试方法展望

高放废物深地质处置库围岩在温度及应力耦合作用下的工程稳定性是关系环境安全的关键问题.本文对黏土围岩高放废物处置库在放射性物质放热温度及深部地应力环境应力耦合研究工作中采用的测试方法进行了系统的回顾和总结.从黏土岩热学性能测试方法以及温度对黏土岩力学性质影响研究方法两个方面,对目前高放废物温度-应力耦合特性测试的试验方法...     

高放废物处置库的工程设计

 高放废物处置库是一项特殊的岩石地下工程。与一般地下岩石工程相比,处置库具有许多特点,相应的处置库的设计也有别于一般地下岩石工程。总结高放处置库的若干特点,简要介绍瑞典D1阶段处置库地下岩石工程设计的主要内容,设计过程与设计时考虑的因素,初步讨论处置库的功能目标与设计年限、概念设计、总体要求、温度限制、不同围岩中处置库的主要问题、可回取的处置库设计及处置库的建设成本等问题。     

The Äspö Hard Rock Laboratory—a step toward the Swedish final repository for high-level radioactive waste

Sweden's energy consumption produces about 250 metric tons of spent nuclear fuel annually. In order to meet the country's growing need for high-level radioactive waste disposal, the Swedish Nuclear Fuel and Waste Management Company (SKB) has developed a waste management system that will ensure safe handling of all of Sweden's radioactive waste. To prepare for the siting and licensing of the final disposal site, SKB is constructing an underground research laboratory, the Äspö Hard Rock Laboratory, which will replace the underground laboratory at Stripa Mine that has been in operation since 1977. This article describes the R & D objectives of the Hard Rock Laboratory, as well as site characterization, layout and construction of the laboratory, which began in October 1990.     

Coupled thermo-hydro-mechanical analysis of a heater test in fractured rock and bentonite at Kamaishi Mine — comparison of field results to predictions of four finite element codes

Four computer codes were applied for a prediction of coupled thermo-hydro-mechanical responses during an in situ heater experiment which simulates a nuclear waste deposition hole with a waste over-pack and bentonite buffer, surrounded by fractured rock. The elevated temperature in the heater surroundings, which was maintained at 100°C for 8.5 months, generated substantial heat-driven moisture flow and swelling in the clay buffer, and thermal expansion of the surrounding fractured rock. Predicted system responses — including temperature, moisture content, fluid pressure, stress and displacement — were compared to measurements at 70 sensors located both in the clay buffer and the near-field rock. An overall good agreement between modeling and measured results indicates that most thermo-hydro-mechanical responses are fairly well represented by the coupled numerical models. Uncertainties occur for modeling of hydromechanical behavior of the swelling clay buffer at low saturation, modeling of near-field heterogeneous mechanical behavior of the low-stressed fractured rock, and modeling of the rock–buffer interface.