裂隙水流-传热对高放废物处置库近场温度影响的三维离散元分析

裂隙水流-传热是高放废物处置库行为的重要影响因素。为研究裂隙水流-传热对高放废物处置库近场温度的影响,采用3DEC离散元软件计算分析了完整岩体模型和裂隙岩体水流模型对处置库近场温度分布和演变的影响。计算分析表明:由于裂隙水流的吸热降温作用,裂隙岩体模型的废物罐表面膨润土温度低于完整岩体模型的废物罐表面膨润土温度,并缩短了达到稳态所需要的时间;裂隙水流上游区域废物罐表面膨润土温度显著低于裂隙水流下游区域废物罐表面膨润土温度;在设定条件下,裂隙岩体模型的废物罐表面膨润土最高温度约为完整岩体模型废物罐表面膨润土最高温度的75%,裂隙水流速度从0.2mm/s增大到0.5mm/s,废物罐表面膨润土最高温度降低约4%。     

高放废物处置区域尺度岩体适宜性评价研究

岩体适宜性评价是高放废物处置库选址和设计的重要工作内容,以判断场址岩体是否满足处置库长期包容和隔离核素的功能要求。依据我国的高放废物处置概念和场址条件,提出了Q_HLW岩体适宜性评价方法,但目前Q_HLW在场址尺度展开了较为深入的研究,尚未在处置区域尺度、处置巷道及处置坑尺度建立完善系统岩体适宜性评价方法。结合芬兰地下实验室研究和处置库设计经验,建立处置区域尺度岩体适宜性评价准则Q^P_HLW,提出了裂隙带影响、地下水化学条件、岩体渗透特性、岩体强度应力比值以及岩体完整性等评价指标的取值方法,并确定岩体适宜性评价分级标准。随后,利用芬兰高放废物处置ONKALO地下实验室场址数据,测试和验证处置区域尺度岩体适宜性评价准则Q^P_HLW的合理性与可行性。最后以北山地下实验室新场场址为评价对象,开展处置区域尺度岩体适宜性评价,适宜性评价结果表明：新场场址在处置深度400～450 m及550～600 m内岩体完整性高,岩体适宜性程度高,适合布置处置巷道。     

高放废物处置库岩体不同条件下缓冲材料饱和机理研究

缓冲材料作为高放废物处置库内最后一道人工屏障,其饱和过程对处置库多重屏障体系设计具有重要意义。以高放废物处置单元为研究对象,建立了相应的热-流-固耦合模型,模拟了饱和花岗岩体条件下缓冲材料饱和过程,并开展了岩体不同边界水压力、岩体裂隙对缓冲材料饱和过程影响研究。研究结果表明：处置库内缓冲材料外部、下部靠近花岗岩的区域饱和度率先增大,然后逐渐由外部向内部、下部向上部进入饱和状态。饱和花岗岩体边界水压分别为5、3和1 MPa时,缓冲材料达到饱和时间分别为28.9、45.63和110.72 a,随着花岗岩体边界水压的下降,处置单元内缓冲材料达到饱和时间将大幅增加。岩体裂隙的渗透率分别为1.098×10^-17、1.098×10^-15、1.098×10^-13和1.098×10^-11m²时,缓冲材料达到饱和时间分别为22.69、22.09、17.34和17.12 a,岩体中裂隙的存在能够加快缓冲材料的饱和过程。     

高放废物处置库近场裂隙水流-传热-处置室间距相互作用的三维数值分析

为研究高放废物地质处置库近场裂隙水流-传热-处置室间距的相互作用机理,采用3DEC软件计算裂隙水流-传热-处置室间距相互作用对处置库近场温度分布影响。结果表明:(1)在处置室间距相同条件下,流动的裂隙水显著改变了处置库近场温度场,使岩体温度降低,缩短模型达到稳态所需要的时间。(2)处置室间距增大,温度叠加效应减弱,处置库近场温度越低,并且废物罐表面膨润土温度越低,裂隙出水口水温越低,模型达到稳态所需要的时间越短。(3)水平和垂直裂隙水流共同传热使处置库近场裂隙水流下游区域温度显著高于裂隙水流上游区域。(4)处置室间距为6 m和8 m时,水平裂隙出水口水温高于垂直裂隙,处置室间距为10 m时,水平裂隙出水口水温低于垂直裂隙。     

高放废物处置库围岩开挖损伤区研究的关键问题及进展

高放废物(HLW)地质处置是将高水平放射性废物埋存于地下500～1 000 m地质体中,使放射性废物与生物圈长期隔离。地质处置库对核素的长期隔离能力是安全评价的关键课题。地下硐室的开挖将不可避免地对围岩造成损伤,形成开挖损伤区(EDZ),改变围岩的物理力学特性,对高放废物地质处置长期安全性存在潜在的影响。目前多个国家建成了高放废物处置地下实验室,并开展了大型原位开挖损伤区的研究,研究开挖损伤区的形成过程及其物理力学特性的变化。本文综述了国外结晶岩地下实验室开展的开挖损伤区研究,总结了EDZ关键研究问题;梳理了加拿大、瑞典、芬兰3个地下实验室多年来开展的系统的EDZ研究工作,对当前EDZ预测模型及模拟技术进行了总结;对我国地下实验室将开展的开挖损伤区研究工作进行了初步探讨,期望为我国的相关研究提供借鉴。同时,高放废物处置库是地下工程新实践,其EDZ的研究成果,形成的技术方法将对其他行业地下工程的建设,如引水隧洞、公路铁路隧道等也有重要的参考价值。     

北山花岗岩闭合裂隙剪切破坏机理研究

随着我国高放废物处置库研发战略的推进,处置库花岗岩围岩工程特性成为研究重点。天然裂隙是花岗岩中的薄弱环节,岩体往往沿既有结构面发生剪切滑移。因此裂隙剪切特性和剪切破坏机理研究意义重大。天然裂隙包括张开裂隙和闭合裂隙,其剪切特性和剪切破坏机理存在差异。针对北山地下实验室场址范围花岗岩闭合裂隙,开展剪切试验。将裂隙结构面扫描数据嵌入离散元颗粒流软件（PFC）,构建北山花岗岩闭合裂隙剪切颗粒模型,利用fish语言监测颗粒间接触（contact）失效并生成对应裂隙（fracture）以模拟试样剪切过程内部微裂纹的发育。试验获得了北山花岗岩典型闭合裂隙的剪切强度和闭合裂隙三维形貌。将试验与模拟结果进行对比,研究认为：构建的颗粒模型可较好地模拟裂隙三维形貌对剪切强度的影响,表征了花岗岩闭合裂隙剪切过程的微裂纹发育扩展过程,展示了闭合裂隙的剪切破坏机理。     

高放废物地质处置中的工程材料

凡人类从事于与核材料有关的许多生产、生活活动均可能产生不同活度的放射性废物.高放废物由于具有放射性水平高、发热量大、核素寿命长等特点,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.借鉴已有研究成果,我国采用多重工程屏障系统(包括废物固化体、废物罐及其外包装和缓冲/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.参照国际上该领域的研究成果,结合我国处置概念,本文就高放废物地质处置中的工程材料(废物固化体、废物罐、外包装、缓冲材料、回填材料),以及其材料选择、设计要求和研究重点等进行了总结.     

国内外核废物处置库近场温度场模拟预测

核废物处置后因所含的放射性核素衰变而产生的衰变热通过传导、对流以及辐射等方式从废物体向外传递,从而引起废物罐体、缓冲材料及近场围岩温度升高,导致废物体至近场围岩之间形成温度梯度。温度梯度随着时间的延续而变化,最终会影响地下水系统和核素迁移。本文对一些国家的处置库温度预测模式进行了调研,对源项、处置库模型简化、热传递数学模型和模拟结果做了初步总结,为我国拟建处置库的温度场预测提出了建议。     

为放射性废料处理探查花岗岩体中的近水平裂隙带

一些国家正积极调查利用结晶岩体作为放射性核废料最终处理场所的可行性。作为加拿大废核燃料处理计划方案评价的一部份,对马尼托巴东南部的Lacdu Bonnet岩体进行了一次多学科调查研究。在这个岩体中正准备处置一些密封的放射源或示踪同位素。高分辨率地震发射波测量结果和从几个深钻孔取得的资料一起显示出,直到800 m深度的岩体内部存在着有影响的水平裂隙带。综合解释地震数据和选出的测井资料后发现,有热水沿主要裂隙带上涌。在岩体中寻找合适的放射性废料处置场地时,不能不把主要裂隙带的渗透性能和     

阿奇山1号岩体现场水文地质试验及渗透性评价

岩体渗透性是高放废物处置场址评价的重要指标之一。采用双栓塞水文地质试验设备,对高放废物地质处置新疆预选区阿奇山1号岩体的钻孔进行了分层水文地质试验,并利用稳定流和非稳定流参数拟合相结合的方法对试验数据进行了解译,获得了岩体的渗透系数。结果显示:阿奇山1号岩体渗透性极低,90%的试验段渗透系数低于10~(-8)m·s~(-1),钻孔揭露岩体渗透系数呈现随深度增加而降低的趋势,从渗透性的角度判断,该岩体是有利场址之一;深部岩体渗透系数与裂隙数量呈现明显的正相关性。     

多因素耦合作用下缓冲材料对铀的长期阻滞效应数值模拟北大核心CSCD

近场环境条件下核素在缓冲材料中的迁移扩散受控于温度场、渗流场、膨胀应力场和化学吸附场的耦合作用,其对核素的阻滞特性将影响到核素随地下水向处置库围岩迁移并返回生物圈的能力,开展多因素耦合作用下缓冲材料对铀的长期阻滞效应研究,对地质处置库的长期安全性评价具有重要的意义。本研究基于混合物理论、连续介质理论、质量守恒、动量守恒、能量守恒及溶质扩散的Fick定律,推导出饱和缓冲材料中核素迁移扩散的热-水-力-化耦合控制方程,并借助于COMSOL Multiphysics软件的直接全耦合求解优势,以自主研制的缓冲材料长期阻滞性能Mock-up实验装置为几何模型,采用内置接口和添加热-水-力-化耦合控制方程中的耦合项作为源项相结合方式,实现了多物理场耦合作用下铀在饱和缓冲材料中迁移扩散行为的直接耦合分析,其长期阻滞特性数值模拟结果表明:初期阶段铀在缓冲材料中迁移扩散较缓慢,迁移距离随时间增幅在1 m左右;中后期阶段,随缓冲材料对铀的吸附容量逐渐趋于饱和后,其迁移距离较初期阶段增加更为明显,迁移距离随时间增幅为3 m左右。多因素耦合下核素在饱和缓冲材料中迁移扩散的热-水-力-化耦合控制方程构建、求解及长期阻滞性能模拟研究的方法,能够为我国高放废物深地质处置库地下实验室开展1∶1工程尺度的工程屏障设计与安全性能评价提供技术参考。     

Analyses of disposal efficiency based on nuclear spent fuel cooling time and disposal tunnel/pit spacing for the design of a geological repository

The purpose of deep geological disposal of high-level radioactive waste (HLW) including nuclear spent fuels is to isolate and to inhibit the release of radioactive material for a long time so that its toxicity does not affect the biosphere. The main requirement for the HLW repository design is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. The cooling time of the spent fuels discharged from nuclear power plants is the key consideration factor for the efficiency and economic feasibility of such a repository. We analyze the spacing of the disposal tunnels and pits, the disposal area and the uranium density for the deep geological repository layout to satisfy the thermal requirement of the disposal system. To do this, thermal stability analyses of a disposal system have been performed using varying spent fuel cooling times and spacing of the disposal tunnels and pits. The results show that the time to reach the maximum temperature within the design limit of the temperature in the disposal site is likely to be shortened as the cooling time of the spent fuel becomes shorter. Also it seems that controlling the disposal pit spacing is considered more advantageous than controlling the disposal tunnel spacing to meet the allowable thermal criteria in the repository from thermal and economical points of view. The results of these analyses can be used for a deep geological repository design and detailed analyses with exact site characteristics data will reduce the uncertainty of the results.     

Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.     

高水平放射性废物处理处置标准分析

本文阐述了我国高水平放射性废物处理处置标准的重要性,对国内外高水平放射性废物处理处置标准现状进行了阐述和分析,针对高放废物处理处置标准体系、高水平放射性废液成份分析、高放废液固化体性能要求及检验方法、高放废物处理处置工程经济及深地质处置等方面的标准化问题进行了研究分析,提出了开展高水平放射性废物处理处置标准化工作的意见和建议。     

Thermal aspects of radioactive waste burial in hard rock

One of the possible methods that has been considered for the disposal of radioactive waste is deep burial in stable rock formations. This paper reviews recent work on modelling the way in which the heat emitted by the decaying radionuclides in the waste could affect this disposal option, emphasizing both the effects on depository design and on migration by flowing groundwater. It focuses particularly on research in the U.K. into the feasibility of burying high-level waste in fractured crystalline rock. After introductory sections on the characteristics of the waste and rock, there are three major sections on the temperature field in the surrounding rock, the stresses generated in the rock, and the groundwater flow.     

高放玻璃多重屏障介质多因素耦合处置方法研究

地下水透过多重屏障介质与高放玻璃固化体直接接触后,放射性核素会从固化体中释放,因此成为高放废物处置库安全评价的源项。为更精确地预测玻璃固化体长期处置行为,本文考察了围岩、回填材料等因素对模拟高放玻璃固化体中各关键元素浸出的影响,实验处置温度为90 ℃,模拟高放玻璃固化体依据德国配方制备。结果表明,围岩对玻璃体中不同元素的阻滞作用有所差异。B、Re和U的浸出浓度在二长花岗岩中最大;膨润土含水量高时,玻璃体中元素释出量大;而含水量低时,释出量小;在膨润土中添加5%的素玻璃粉,对玻璃的腐蚀有抑制作用。     

Effects of α-radiation on a direct disposal system for spent nuclear fuel – (1) review of research into the effects of α-radiation on the spent nuclear fuel,canisters and outside canisters

The Japanese geological disposal programme has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter “direct disposal of SF”) as an alternative management option other reprocessing followed by vitrification and geological disposal of high-level radioactive waste. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with decomposition of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate of SF and the solubility of radionuclides. Focusing especially on the effects of α-radiation in safety assessment, this study has reviewed research into the effects of α-radiation on the SF, canisters and environment outside the canisters.     

Long-term lining performance - Civil engineering problem of potential retrieval of buried spent nuclear fuel

The current solution for the spent fuel, high-level and long-lived radioactive waste is to store them at surface facilities from which they will be subsequently moved to a deep repository. No such repositories are in operation currently but several such facilities are close to the construction phase. A deep repository can be situated in several types of geological conditions including clay formations, salt sediments, argillites and tuffitic and granitic rocks. The character of the host rock is the key factor determining the design and specific requirements of individual components of such a facility. The future potential retrieval of canisters containing nuclear waste from the repository is a further influential factor. The reason for retrieval of containers lies in the development of fast reactors and increased interest for spent fuel reprocessing. Naturally, the decision as to whether retrievability is technically feasible must be made before finalising the design and construction process of the repository. If the decision is made to retrieve, a design which will include all the relevant safety aspects for the potential retrieval of canisters must be determined. The lay-out of the repository, the materials to be used and the design of the various structures of the facility (e.g. access tunnels, disposal shafts, buffer and backfill) are not the only issues to be addressed. The long-term stability of the system as a whole, i.e. of all the components, is crucial. Depending on the disposal concept chosen, the thermal load generated by the waste in the disposal container, saturation by water from the surrounding environment and the loading of the host rock massif will constitute the main processes which will affect the behaviour, safety and future functioning of the repository from the civil engineering point of view. The long-term stability of the lining of disposal galleries is a basic precondition for the safe removal of spent nuclear waste from deep underground repositories. The stability problems of tunnel linings exposed to long-term thermal load have not yet been properly addressed and form the subject of the European TIMODAZ project (Thermal Impact on the Damaged Zone around a Radioactive Waste Disposal in Clay Host Rocks) and also supported by the “Complex System of Methods for Directed Design and Assessment of Functional Properties of Building Materials” project. This paper describes the design, construction and currently available results of a 1:1 scale “in situ” disposal tunnel model which has been built at the Josef Underground Educational Facility in the Czech Republic.     

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。