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.

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.

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.

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.

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.

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.

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.

Disposal of high-level radioactive waste in crystalline rock: On coupled processes and site development

Safe disposal of high-level radioactive nuclear waste(HLW)is crucial for human health and the environment,as well as for sustainable development.Deep geological disposal in sparsely fractured crystalline rock is considered one of the most favorable methods for final disposal of HLW.Extensive research has been conducted worldwide and many countries have initiated their own national development programs for deep geological disposal.Significant advancements of national programs for deep geological disposal of HLW in crystalline rock have been achieved in Sweden and Finland,which are currently under site development stage,focusing on detailed site characterization,repository construction,and post-closure safety analysis.Continued research and development remain important in the site development stage to ensure long-term safety of the HLW disposal repository.This work presents an overview and discussion of the progress as well as remaining open scientific issues and possibilities related to site development for safe disposal of HLW in crystalline rock.We emphasize that developing a comprehensive and convergent understanding of the coupled thermal,hydraulic,mechanical,chemical and biological(THMCB)processes in fractured crystalline rock remains the most important yet challenging topic for future studies towards safe disposal of HLW in crystalline rock.Advancements in laboratory facilities/techniques and computational models,as well as available comprehensive field data from site developments,provide new opportunities to enhance our understanding of the coupled processes and thereby repository design for safe geological disposal of HLW in crystalline rock.

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.

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。

高放废物地质处置预选地段调查与适宜性研究

本项研究从高放废物安全处置的需求出发,针对我国高放废物处置库首选预选区(北山预选区)内算井子预选地段,综合运用地质—地球物理—水文地质—地球化学等研究方法获得了地质、水文地质、未来自然变化、地球化学、建造与工程、环境保护和社会经济等方面的数据和资料,评价了预选地段岩体的安全性;从工程建造和工程安全等角度论证了预选岩体的可建造性;确认了预选岩体在运输条件、土地利用、社会经济与人文等方面的可接受性。在此基础上,建立了较为完善的处置库预选地段调查与适宜性综合分析方法,并通过地段适宜性综合分析和地段内场址圈定及其安全性综合对比,在算井子地段中筛选出1个花岗岩处置库候选场址。研究成果将直接服务于我国高放废物地质处置库场址筛选和场址特性评价,对保障我国核废物安全管理和核能可持续发展具有重要的现实意义。

高放废物地质处置新场候选场址水文地质特征研究

场址水文地质条件对高放废物处置库而言至关重要,是决定场址适宜性及长期安全的关键因素之一。以我国高放废物地质处置新场候选场址为研究对象,重点分析讨论了场址岩体的渗透性、地下水动力场、水化学与同位素特征及地下水成因机制,目的是从水文地质角度为高放废物处置库场址的适宜性评价提供技术支撑。结果表明:新场候选场址地下水主要以大气降水入渗补给为主,整体上补给能力差;受蒸发浓缩作用影响显著,地下水化学类型主要为Cl·SO_(4)-Na型;场址地下水位的空间分布整体受地形控制,呈现西高东低、中间高南北低,此外,地下水位在垂向上还表现为浅部高、深部低的趋势,具有典型的补给区特征;场址断层和岩体深部的渗透性极低,相应的处置深度渗透系数小于10^(-10)m·s^(-1);14C测年结果表明:场址浅部地下水表观年龄一般小于5.4 ka,深部年龄高达13.2 ka,说明地下水循环更新速率缓慢。新场候选场址水文地质条件总体上具有补给条件差、岩体渗透性低、垂向水力联系弱和循环更新速率缓慢等特点,这些特点对高放废物地质处置都是有利的。

Current status of the geological disposal programme and an overview of the safety case at the pre-siting stage in Japan

In Japan,high-level radioactive waste and specific low-level radioactive waste which includes long-lived radionuclides are planned to be disposed of in the geological formations at depths greater than 300 m.The disposal site will be selected through a stepwise site investigation process that consists of a Literature Survey,Preliminary Investigation,and Detailed Investigation phases.In October 2020 a Literature Survey was launched in Japan at two municipalities in Hokkaido for the first time since NUMO initiated a nationwide call for volunteer municipalities in 2002,and the outcomes are currently being compiled.To enhance the public’s understanding of how to implement safe geological disposal in Japan based on the latest scientific knowledge and technology,NUMO,as the implementing organisation,developed and published a safety case for geological disposal at the pre-siting stage.This safety case provides multiple lines of arguments and evidence to demonstrate the feasibility of the geological disposal and a basic structure for a safety case that will be applicable to any potential sites in Japan.The safety case also presented some R&D challenges to enhance the technical confidence of the project,including the R&D topics related to rock mechanics.This report presents the current status of the geological disposal programme in Japan,together with the status of the Literature Survey phase and an overview of the NUMO safety case.

低放废物处置场辐射防护设计关键问题研究

低放废物处置场辐射防护设计一般主要涉及废物处置单元和废物接收厂房。废物处置单元根据处置场接收准则、废物码放处置方案等进行辐射防护设计,废物接收厂房的辐射防护设计主要关注废物装卸区、开箱倒装区、无损检测区等。本文主要介绍了低放废物处置场辐射分区准则的制定,辐射防护设计原则的考虑,废物处置单元及废物接收厂房辐射防护设计等方面的内容。重点关注了废物处置单元废物分层、分隔断码放处置情景,不同分隔码放方式所允许码放处置的放射性废物表面剂量率限值,码放层数限制,处置单元侧墙直接照射、处置单元天空散射、水泥砂浆及封顶的辐射防护设计考虑,以及废物接收厂房开箱倒装区的辐射防护设计关键问题。介绍了低放废物近地表处置场辐射防护设计相关的问题处理及经验,为相关设施辐射防护设计提供了参考。

一种处置场地下截水盲沟设计

文章针对采取浅地表处置且地下水埋深较浅的低、中水平放射性固体废物处置场场址,结合北方某处置场的设计实践,对其场址自然条件进行分析,针对自然条件的不足,提出一种降低地下水水位的截水盲沟设计设想,并对该设计的长期有效性和长期稳定性进行论证,以实现处置场在运行时,其与地下水的安全隔离。文章结合工程实际,对提高处置场本质安全度进行研究,提出处置场设计的新思路,为后续处置场的设计提供帮助和参考。

中国高放废物深地质处置(英文)

介绍了中国高放废物地质处置计划的背景、初步技术战略和长远规划。中国的高放废物地质处置研究始于1985年,计划于21世纪中叶建成国家高放废物处置库。位于我国西北甘肃省的北山地区被选为最有远景的处置库预选区。1999～2004期间,在该区开展了初步的场址特性评价研究,包括地表地质、水文地质和地球物理调查、4口钻孔(北山1#,2#,3#及4#孔)的施工及钻孔现场试验,并获得了大量成果。在缓冲回填材料、放射性核素迁移以及天然类比等方面也取得了进展。

高放废物深地质处置及国内研究进展

核能在产生电力造福社会的同时 ,也留下了放射性废物。我们有责任对这些废物实施安全和正确的管理。本文阐述了高放废物深地质处置的一般概念及处置库选址研究中的若干问题 。

中国高放废物地质处置研究进展:1985～2004

如何安全处置高放废物是核工业可持续发展面临的挑战性问题.我国的高放废物深地质处置研究从1985年开始,提出的计划目标是:于21世纪中叶建成我国高放废物地质处置库,处置的对象是玻璃固化块、超铀废物和部分乏燃料,处置库为竖井-坑道型,候选围岩为花岗岩,位于饱和带中.在1985～2004的20 a中,我国高放废物地质处置研究取得了进展,已确定我国高放废物最终处置走'深地质处置',并且是'三步曲'式的技术路线,即处置库选址和场址评价-地下实验室-处置库.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,该区地处戈壁,地壳结构完整,地壳稳定,人烟稀少,地质条件和水文地质条件均有利.20世纪90年代初期,开展了地下实验室的选址工作,初步选择了北京郊区2处地点为我国高放废物地质处置'普通地下实验室'的场址.已确定使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.对膨润土的矿物学、岩土力学、物理力学性质和热学性质进行了研究.已获得一批放射性核素(主要是Np、Pu、Tc)在北山花岗岩和膨润土上的吸附分配比、扩散系数和弥散系数等参数,建立了低氧手套箱和模拟处置库温度、压力和氧化-还原条件的小型实验装置.高放废物中的关键核素的化学行为研究也取得进展.花岗岩接触带核素迁移、铀矿床中超铀元素迁移、青铜器腐蚀等天然类比研究取得了成果.还开展了高放废物地质处置系统总性能评价源项和生物圈模式的调研.概念设计研究仅在20世纪90年代初开展了部分研究.从1999年开始,与国际原子能机构开展了2期高放废物地质处置技术合作项目,极大地提高了我们的技术水平.20 a的科研工作为我国在21世纪完成高放废物地质处置奠定了一定基础.

高放废物处置库预选场址包气带土壤渗透性研究

在高放废物地质处置库选址和性能评价过程中，水文地质条件是其中最要的因素之一，而包气带土壤渗透性又是水文地质研究中的核心问题之一。甘肃北山地区是我国高放废物地质处置库场址首选预选区，新场一向阳山地段是北山预选区的重点候选场址之一。为查明该地区包气带土壤渗透性特征，本文选用双环法和Guelph入渗仪法两种试验方法开展包气带土壤渗透性的研究并综合分析了影响土壤渗透性的主要因素。结果表明：（1）区内包气带土壤渗透性能差异较大。总体来看，沟谷地带的土壤饱和渗透系数（心）大于缓坡地带，缓坡地带玩大于平滩地带；（2）构造影响带如很大，不同构造影响带K。存在一定的差异；（3）研究区包气带土壤渗透性的最主要影响因素是粒径〉0．5mm的粗砂与粒径〈0．1mm的极细砂、粉粒及黏粒含量，而粒径为0．5～0．1mm的中砂、细砂含量影响最小。