Medium-Deep or Very Deep Disposal of Highly Radioactive Waste？

Commonly proposed concepts, like KBS-3V, for disposal of highly radioactive waste imply construction at medium depth （400-600 m） in granitic rock, which is excellent for constructing a stable repository. VDH （very deep boreholes） represent an altemative concept with the advantage that the rock is much less permeable and that the very salt, heavy groundwater is stagnant. Both require engineered barriers in the form of canisters and waste-embedding clay but for somewhat different purposes. Canisters are the most important waste-isolating barriers for KBS-3V but are less important for VDH. The waste-embedding clay is needed for preserving the KBS-3V canisters by being tight and ductile, but plays a minor role for the VDH. The backfilled deposition tunnels in a KBS-3V repository provide very limited hindrance of radionuclides to move to the biosphere while the clay seals of VDH effectively prevent possibly released radionuclides to reach up to the biosphere. Comparison of the KBS-3V and VDH concepts indicates that the last mentioned one has several advantages but that certain issues remain to be worked on for becoming a number one candidate.

Study on the residence time of deep groundwater for high-level radioactive waste geological disposal

Residence time of deep groundwater is one of the most important parameters in safety and performance assessment for high-level radioactive waste geological disposal. In this study, we collected the deep groundwater samples of Jijicao in Gansu Beishan pre-selected region. The deep groundwater residence time at two depths estimated by Helium-4 accumulation method were 3.8 ka and 5.0 ka respectively upon measurement and calculation, which indicates that the deep groundwater is not derived from the deep crust circulation process. Hence, deep groundwater is featured with long residence time as well as slow circulation and update rate, and such features are conductive to the safe disposal of high-level radioactive waste.

The Beishan underground research laboratory for geological disposal of high-level radioactive waste in China:Planning, site selection,site characterization and in situ tests

With the rapid development of nuclear power in China, the disposal of high-level radioactive waste（HLW） has become an important issue for nuclear safety and environmental protection. Deep geological disposal is internationally accepted as a feasible and safe way to dispose of HLW, and underground research laboratories（URLs） play an important and multi-faceted role in the development of HLW repositories. This paper introduces the overall planning and the latest progress for China＇s URL. On the basis of the proposed strategy to build an area-specific URL in combination with a comprehensive evaluation of the site selection results obtained during the last 33 years, the Xinchang site in the Beishan area,located in Gansu Province of northwestern China, has been selected as the final site for China＇s first URL built in granite. In the process of characterizing the Xinchang URL site, a series of investigations,including borehole drilling,geological mapping, geophysical surveying,hydraulic testing and in situ stress measurements, has been conducted. The investigation results indicate that the geological,hydrogeological, engineering geological and geochemical conditions of the Xinchang site are very suitable for URL construction. Meanwhile, to validate and develop construction technologies for the Beishan URL, the Beishan exploration tunnel（BET）, which is a 50-m-deep facility in the Jiujing sub-area, has been constructed and several in situ tests, such as drill-and-blast tests, characterization of the excavation damaged zone（EDZ）, and long-term deformation monitoring of surrounding rocks, have been performed in the BET. The methodologies and technologies established in the BET will serve for URL construction.According to the achievements of the characterization of the URL site, a preliminary design of the URL with a maximum depth of 560 m is proposed and necessary in situ tests in the URL are planned.

Main outcomes from in situ thermo-hydro-mechanical experiments programme to demonstrate feasibility of radioactive high-level waste disposal in the Callovo-Oxfordian claystone

In the context of radioactive waste disposal,an underground research laboratory(URL)is a facility in which experiments are conducted to demonstrate the feasibility of constructing and operating a radioactive waste disposal facility within a geological formation.The Meuse/Haute-Marne URL is a sitespecific facility planned to study the feasibility of a radioactive waste disposal in the Callovo-Oxfordian(COx)claystone.The thermo-hydro-mechanical(THM)behaviour of the host rock is significant for the design of the underground nuclear waste disposal facility and for its long-term safety.The French National Radioactive Waste Management Agency(Andra)has begun a research programme aiming to demonstrate the relevancy of the French high-level waste(HLW)concept.This paper presents the programme implemented from small-scale(small diameter)boreholes to full-scale demonstration experiments to study the THM effects of the thermal transient on the COx claystone and the strategy implemented in this new programme to demonstrate and optimise current disposal facility components for HLW.It shows that the French high-level waste concept is feasible and working in the COx claystone.It also exhibits that,as for other plastic clay or claystone,heating-induced pore pressure increases and that the THM behaviour is anisotropic.

On area-specific underground research laboratory for geological disposal of high-level radioactive waste in China

Underground research laboratories （URLs）, including ＂generic URLs＂ and ＂site-specific URLs＂, are un- derground facilities in which characterisation, testing, technology development, and/or demonstration activities are carried out in support of the development of geological repositories for high-level radioactive waste （HLW） disposal. In addition to the generic URL and site-specific URL, a concept of ＂areaspecific URL＂, or the third type of URL, is proposed in this paper. It is referred to as the facility that is built at a site within an area that is considered as a potential area for HLW repository or built at a place near the future repository site, and may be regarded as a precursor to the development of a repository at the site. It acts as a ＂generic URL＂, but also acts as a ＂site-specific URL＂ to some extent. Considering the current situation in China, the most suitable option is to build an ＂area-specific URL＂ in Beishan area, the first priority region for China＇s high-level waste repository. With this strategy, the goal to build China＇s URL by 2020 mav be achieved, but the time left is limited.

Key Scientific Challenges in Geological Disposal of High Level Radioactive Waste

The geological disposal of high level radioactive waste is a challenging task facing the scientific and technical world.This paper introduces the latest progress of high level radioactive disposal programs in the world,and discusses the following key scientific challenges:(1)precise prediction of the evolution of a repository site;(2)characteristics of deep geological environment;(3)behaviour of deep rock mass,groundwater and engineering material under coupled conditions(intermediate to high temperature,geostress,hydraulic,chemical,biological and radiation process,etc);(4)geochemical behaviour of transuranic radionuclides with low concentration and its migration with groundwater;and(5)safety assessment of disposal system.Several large-scale research projects and several hot topics related with high-level waste disposal are also introduced.

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.

High-level radioactive waste disposal in China: update 2010

For geological disposal of high-level radioactive waste （HLW）, the Chinese policy is that the spent nuclear fuel （SNF） should be reprocessed first, followed by vitrification and final disposal. The preliminary repository concept is a shaft-tunnel model, located in saturated zones in granite, while the final waste form for disposal is vitrified high-level radioactive waste. In 2006, the government published a long-term research and development （R＆D） plan for geological disposal of high-level radioactive waste. The program consists of three steps： （1） laboratory studies and site selection for a HLW repository （2006-2020）; （2） underground in-situ tests （2021-2040）; and （3） repository construction （2041-2050） followed by operation. With the support of China Atomic Energy Authority, comprehensive studies are underway and some progresses are made. The site characterization, including deep borehole drilling, has been performed at the most potential Beishan site in Gansu Province, Northwestern China. The data from geological and hydrogeological investigations, in-situ stress and permeability measurements of rock mass are presented in this paper. Engineered barrier studies are concentrated on the Gaomiaozi bentonite. A mock-up facility, which is used to study the thermo-hydro-mechano-chemical （THMC） properties of the bentonite, is under construction. Several projects on mechanical properties of Beishan granite are also underway. The key scientific challenges faced with HLW disposal are also discussed.

Clays in radioactive waste disposal

Clays and argillites are considered in some countries as possible host rocks for nuclear waste disposal at great depth.The use of compacted swelling clays as engineered barriers is also considered within the framework of the multi-barrier concept.In relation to these concepts,various research programs have been conducted to assess the thermo-hydro-mechanical properties of radioactive waste disposal at great depth.After introducing the concepts of waste isolation developed in Belgium,France and Switzerland,the paper describes the retention and transfer properties of engineered barriers made up of compacted swelling clays in relation to microstructure features.Some features of the thermo-mechanical behaviors of three possible geological barriers,namely Boom clay（Belgium）,Callovo-Oxfordian clay（France） and Opalinus clay（Switzerland）,are then described,including the retention and transfer properties,volume change behavior,shear strength and thermal aspects.

High-Level Nuclear Wastes and the Environment: Analyses of Challenges and Engineering Strategies

The main objective of this paper is to analyze the current status of high-level nuclear waste disposal along with presentation of practical perspectives about the environmental issues involved. Present disposal designs and concepts are analyzed on a scientific basis and modifications to existing designs are proposed from the perspective of environmental safety. A new concept of a chemical heat sink is introduced for the removal of heat emitted due to radioactive decay in the spent nuclear fuel or high-level radioactive waste, and thermal spikes produced by radiation in containment materials. Mainly, UO2 and metallic U are used as fuels in nuclear reactors. Spent nuclear fuel contains fission products and transuranium elements which would remain radioactive for 104 to 108years. Essential concepts and engineering strategies for spent nuclear fuel disposal are described. Conceptual designs are described and discussed considering the long-term radiation and thermal activity of spent nuclear fuel. Notions of physical and chemical barriers to contain nuclear waste are highlighted. A timeframe for nuclear waste disposal is proposed and time-line nuclear waste disposal plan or policy is described and discussed.

Experiments on thermo-hydro-mechanical behaviour of Opalinus Clay at Mont Terri rock laboratory, Switzerland

Repositories for deep geological disposal of radioactive waste rely on multi-barrier systems to isolate waste from the biosphere.A multi-barrier system typically comprises the natural geological barrier provided by the repository host rock e in our case the Opalinus Clay e and an engineered barrier system(EBS).The Swiss repository concept for spent fuel and vitrified high-level waste(HLW)consists of waste canisters,which are emplaced horizontally in the middle of an emplacement gallery and are separated from the gallery wall by granular backfill material(GBM).We describe here a selection of five in-situ experiments where characteristic hydro-mechanical(HM)and thermo-hydro-mechanical(THM)processes have been observed.The first example is a coupled HM and mine-by test where the evolution of the excavation damaged zone(EDZ)was monitored around a gallery in the Opalinus Clay(ED-B experiment).Measurements of pore-water pressures and convergences due to stress redistribution during excavation highlighted the HM behaviour.The same measurements were subsequently carried out in a heater test(HE-D)where we were able to characterise the Opalinus Clay in terms of its THM behaviour.These yielded detailed data to better understand the THM behaviours of the granular backfill and the natural host rock.For a presentation of the Swiss concept for HLW storage,we designed three demonstration experiments that were subsequently implemented in the Mont Terri rock laboratory:(1)the engineered barrier(EB)experiment,(2)the in-situ heater test on key-THM processes and parameters(HE-E)experiment,and(3)the full-scale emplacement(FE)experiment.The first demonstration experiment has been dismantled,but the last two ones are on-going.

Temperature Effects on Unsaturated Hydraulic Property of Bentonite-Sand Buffer Backfilling Mixtures

The influence of temperature on the engineered properties of bentonite-sand mixtures （B/S） is of major concern in the design of engineered barriers in underground repositories for high-level radioactive waste disposal. We experimentally studied the influence of temperature on soil unsaturated hydraulic properties related to water holding capacity and permeability of GMZ B/S in China. The vapor equilibrium method and water infiltration apparatus were used to measure the soil water characteristic curve （SWCC） and unsaturated hydraulic conductivity （k）. The results show that the SWCC under different temperatures from 20℃ to 60 ℃ tends to be the same. Temperature influence on unsaturated permeability is more relevant at low suctions, no clear effect is detected below a degree of saturation of 74%, and experimental data show that temperature dependence on unsaturated permeability is small.

Influence of Heating and Water Exposure on Liquid Limit of Bentonite

Bentonite has been considered as the buffer material for embedding canisters with high-level radioactive waste(HLW)in the deep geology repositories. GMZ bentonite deposit which is located in Xinghe County,Inner Mongolia has been proposed as buffer/backfill material for HLW repository in China. Liquid limit of natural Na-bentonite GMZ01 and commercial Na-bentonite MX80 which were previously put in the oven at 80℃and 95℃,and exposed to water for different times were measured.The liquid limit of GMZ01 increased slightly at the beginning, and then decreased as the heating time increased.

Influence of heating and water-exposure on the liquid limits of GMZ01 and MX80 bentonites

Bentonite has been considered as a buffer material for embedding canisters with high-level radioactive waste（HLW） in deep geological repositories.GMZ bentonite deposit,located in Xinghe County,Inner Mongolia Autonomous Region,China was proposed as a buffer/backfill material for HLW repository in China.The liquid limits of natural Na-bentonite GMZ01 and commercial Na-bentonite MX80,which are previously heated at 80 ℃ and 95 ℃,respectively,and exposed to water for different times are measured.It is observed that the liquid limit of GMZ01 increases slightly at the beginning,and then decreases as the heating time increases,while the liquid limit of MX80 decreases with the heating time.The liquid limits of both GMZ01 and MX80 decrease with increasing water-exposure time.After the samples are heated at 80℃and 95 ℃ for several months,the mineralogical composition of GMZ01 does not exhibit evident change,whereas MX80 experiences some changes.In addition,the chemical composition,cation exchange capacity（CEC） and exchangeable cation of all the samples do not change significantly.

围压和裂隙倾角对花岗岩力学行为的影响研究

含裂隙围岩的力学行为对高放废物地质处置库的长期稳定和安全至关重要。为研究围压和裂隙倾角对花岗岩围岩力学行为的影响,以完整花岗岩和含不同倾角贯通单裂隙的花岗岩为研究对象,开展了单轴压缩和三轴压缩试验研究。结果表明:0~10 MPa围压范围内,倾角30°和60°裂隙花岗岩分别发生了穿裂隙面破坏和沿裂隙面滑移破坏;随围压升高,倾角45°裂隙花岗岩由复合破坏转变为穿裂隙面破坏,后又在10 MPa围压条件下再次转变为倾斜“Z”字型裂纹的复合破坏。围压的存在制约了预制裂隙面的劣化作用。三轴压缩条件下低倾角裂隙花岗岩的抗压强度及变形发展曲线与完整花岗岩的结果相近,试样峰值破坏时的轴向应变在围压0,2,5,10MPa条件下分别约为0.38%,0.49%,0.59%和0.75%。沿裂隙面滑移破坏试样的抗压强度以及峰值破坏时的轴向应变均显著低于穿裂隙面破坏试样的相关结果,且其抗压强度与试验前后裂隙面的分形维数差值符合幂函数增长关系。

膨润土砌体墙受压性能及其本构模型的构建

我国高放地质处置库“三步走”战略稳步推进,研究缓冲/回填材料砌体结构力学性能及其破坏特性,可为缓冲/回填材料砌体结构的设计,以及砌块的地面堆砌储存提供基础数据。本文采用合适的压制路径将高庙子钠基膨润土粉末压制成砌块,并采用干砌方法堆砌成砌体墙,设计并进行了砌块单元间界面性能试验和轴心抗压试验,得到了砌块单元间界面剪切强度与剪切位移关系,分析了砌体墙应力-应变曲线和破坏形态。试验结果表明,砌体墙变形破坏经历特性变形、局部裂纹扩展、竖向裂纹贯通、失稳破坏和残余变形几个阶段。砌体墙的第一条裂隙出现在单个砌块单元内部,反映了砌块单元的压制质量是控制砌体墙工程性能的重要影响因素之一。基于试验数据,通过对比分析施楚贤、朱伯龙、TURNSEK和HODGKINSON及仲继清的本构关系发现,在峰前上升曲线段采用试验数据对仲继清模型参数进行修正,获得的理论应力-应变曲线与试验曲线较为吻合。

高放废物处置地下实验室现场试验数据管理顶层设计

我国高放废物地质处置研发已经进入地下实验室建设的关键阶段。在地下实验室建设和运行期间,将开展大量现场试验,这些试验数据具有传输距离长、采集周期长、数据类型多和数据量大等特点,同时数据质量、安全与可追溯性要求高,数据管理难度较大。通过分析我国高放废物地质处置北山地下实验室现场试验数据特点,以iS3智慧数据服务系统为基础,提出现场试验数据管理顶层设计方法,为现场试验和科研工作顺利开展提供重要的数据管理支撑。

我国高放废物地质处置新突破

2019年5月6日,国家原子能机构批复了我国高放废物地质处置地下实验室建设工程立项建议书。2021年6月17日,地下实验室工程正式开工建设,标志着我国高放废物地质处置进入一个新的阶段——地下实验室阶段,实现了新的突破。本文介绍近五年(2019-2024)来我国高放废物地质处置在法律法规、技术标准、场址评价、缓冲材料、工程技术、岩石力学、放射性核素迁移和安全评价等方面取得的新进展,重点介绍北山地下实验室工程建设及研究开发的新进展。

甘肃北山新场深部地下水中铀的赋存形态及其影响因素的地球化学模拟研究

核能的快速发展产生了大量的高水平放射性废物(简称高放废物),对生态环境和人类健康构成了巨大的威胁,其安全处置已成为国际放射性废物管理的重难点问题。深部地质处置被普遍认为是高放废物最安全可靠、技术上非常可行的处置方法。目前,我国已启动了甘肃北山新场高放废物地质处置地下实验室的建设,放射性元素铀在深部地下水中的形态分布及其影响因素从微观层面能为北山深地质处置性能评价提供基础数据,但尚不明晰,亟需开展相关地球化学模拟研究。基于PHREEQC中的llnl.dat数据库,添加NEA-TDB和ThermoChimie数据库中U(Ⅵ)的热力学数据,使用地球化学模拟软件PHREEQC.v3和PhreePlot.v11对铀在我国甘肃北山新场高放废物地质处置地下实验室场址BS28孔深部地下水中的赋存形态及分布进行了模拟计算,旨在厘清铀在北山深部地下水中的赋存形态,探讨pH、pE以及Ca^(2+)与HCO_(3)^(-)浓度比对铀在北山深部地下水中形态分布的影响。模拟结果表明,在该深部地下水环境下,铀主要以六价的形式存在,Ca_(2)UO_(2)(CO_(3))_(3)(aq)和CaUO_(2)(CO_(3))_(3)^(2-)是主要的赋存形态,其占比分别为84.14%和15.16%。受氧化还原氛围的影响,铀的价态在pE<1时逐渐由U(Ⅵ)转变为U(Ⅳ),且在pE=-1~1时,UO_(2)(s)和可溶性碳酸铀酰络合物可同时存在。Ca^(2+)与HCO_(3)^(-)的浓度比低倾向于形成CaUO_(2)(CO_(3))_(3)^(2-),浓度比高易于形成Ca_(2)UO_(2)(CO_(3))_(3)(aq)。Ca^(2+)-UO_(2)^(2+)-CO_(3)^(2)三元体系很大程度上会影响铀在北山深部地下水的溶解度和迁移能力。本研究结果将有助于查明铀在北山花岗岩裂隙中的迁移和吸附机理。

高放废物处置库二维轴对称温度场模型构建及应用研究

高放废物处置库温度场长期演化特征是处置库缓冲层安全性能评估和设计、处置容器空间优化布局的重要依据之一。建立了处置单元二维轴对称双层热分析模型,以揭示处置库多重屏障系统近场温度演化规律。对传热控制方程应用拉普拉斯和有限傅里叶正弦变换,得到缓冲层和围岩层温度的拉普拉斯域解,借助Crump方法对拉普拉斯域解进行数值反演获得对应的时域解;通过与线热源解和数值解的对比验证了模型的正确性,分析了处置单元近场多屏障系统温度场的时空分布特征,探讨了不同参数对缓冲层峰值温度的影响;借助模型半解析解确定最小处置容器间距以及对原位试验结果进行预测。结果表明:①与线热源解相比,模型半解析解可以更准确地模拟缓冲层温度变化;②当处置隧道间距超过40m时,其对缓冲层峰值温度影响较小;③在处置隧道间距取40m时,最小处置容器间距为7.7m;④模型半解析解可以较好地预测原位试验结果。