Effect of drying cracks on swelling and self-healing of bentonite-sand blocks used as engineered barriers for radioactive waste disposal

Experiments were conducted to evaluate the healing of drying cracks in air-dried bentonite-sand blocks after hydration and swelling in groundwater,providing justifications to simplify the protection of blocks prior to installation in a high-level radioactive waste repository.Synthetic groundwater was prepared to represent the geochemistry of Beishan groundwater,and was used to hydrate the blocks during the swelling pressure and swelling strain measurements,as Beishan is the most promising site for China's repository.Healing of the surface cracks was recorded by photography,and healing of the internal cracks was visualized by CT images and hydraulic conductivity of air-dried blocks.The results indicate that the maximum swelling pressure and swelling strain are primarily affected by the geochemistry of Beishan groundwater,but not affected by the drying cracks.The maximum swelling pressure and swelling strain of air-dried blocks are comparable to or even higher than the pressure and strain of fresh blocks.The maximum swelling pressure measured in strong(i.e.high ion strength)Beishan groundwater was 44%of the pressure measured in deionized(DI)water,and the maximum swelling strain was reduced to 23%of the strain measured in DI water.Nevertheless,the remained swelling of the blocks hydrated in strong Beishan groundwater was sufficient to heal the surface and internal drying cracks,as demonstrated by the pictures of surface cracks and CT images.The hydraulic conductivity of the air-dried block permeated with strong groundwater was comparable(3.7×higher)to the hydraulic conductivity of the fresh block,indicating the self-healing of drying cracks after hydration and swelling in groundwater.A simplified method of protecting the block with plastic wraps before installation is recommended,since the remained swelling of the block hydrated in Beishan groundwater is sufficient to heal the drying cracks.

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.

Design and development of large-scale in-situ PRACLAY heater test and horizontal high-level radioactive waste disposal gallery seal test in Belgian HADES

In Belgium,the Boom clay was selected as a potential host formation for the disposal of high-level radioactive waste（HLW）.To demonstrate the suitability of Boom clay for bearing thermal load induced by the HLW,a large-scale in-situ heater test,called PRACLAY heater test,will be conducted in the underground research laboratory（URL） in Mol.Owing to the limitations of the test（a short period of time compared with that considered in a real repository,different boundary conditions,etc.）,the test is designed to simulate,in a conservative way,the most critical state and phenomena that could occur in the host rock.The PRACLAY gallery was excavated at the end of 2007;the heating phase will begin in 2010 and will last for at least 10 years.The PRACLAY gallery itself leaves an opportunity to study the possibilities of sealing a disposal drift in Boom clay and testing the feasibility of hydraulic cut-off of any preferential pathway to the main access gallery through the excavation damage zone（EDZ） and the lining with a seal in a horizontal drift（horizontal seal）.Indeed,this is a generic problem for all deep geological disposal facilities for HLW.An annular seal made of compacted swelling bentonite will be installed in the front of the heated part of the PRACLAY gallery for these purposes.This paper provides detailed considerations on the thermo-hydro-mechanical（THM） boundary conditions for the design of the PRACLAY heater test and the seal test with the support of numerical calculations.It is believed that these important items considered in the PRACLAY heater test design also constitute key issues for the repository design.The outcome of the PRACLAY heater test will be an important milestone for the Belgian repository design.

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.

Experimental study on the physico-mechanical properties of Tamusu mudstone — A potential host rock for high-level radioactive waste in Inner Mongolia of China

Tamusu mudstone, located in Bayin Gobi Basin in Inner Mongolia of China, has been selected as a potential host rock for high-level radioactive waste(HLW) disposal in China. A series of tests has been carried out, including X-ray diffraction(XRD) tests, scanning electron microscopy(SEM) tests, disintegration tests, permeability tests and triaxial compression tests, to estimate the physico-mechanical properties of Tamusu mudstone in this work. The mineral composition of Tamusu mudstone was analyzed and it was considered as a stable rock due to its low disintegration rate, i.e. approximately 0.11%after several wet/dry cycles. Based on the results of permeability test, it was found that Tamusu mudstone has a low permeability, with the magnitude of about 10—20m^(2). The low permeability makes the mudstone well prevent nuclide migration and diffusion, and might be influenced by temperature.The triaxial tests show that Tamusu mudstone is a stiff mudstone with high compressive strength, which means that the excavation disturbed zone would be smaller compared to other types of mudstone due to construction and operation of HLW repositories. Finally, the properties of Tamusu mudstone were compared with those of Opalinus clay, Callovo-Oxfordian(COx) argillite, and Boom clay to further discuss the possibility of using Tamusu mudstone as a potential nuclear waste disposal medium.

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.

Simulation of groundwater and nuclide transport in the near-field of the high-level radioactive waste repository with TOUGHREACT

In order to know the mechanism of groundwater transport and the variation of ion concentrations in the near-field of the high-level radioactive waste repository,the whole process was simulated by EOS3 module of TOUGHREACT.Generally,the pH and cation concentrations vary obviously in the near-field saturated zone due to interaction between groundwater and bentonite.Moreover,the simulated results showed that calcite precipitation could not cause obvious variations in the porosity of media in the near-filed if the chemical components and their concentrations of groundwater and bentonite pore water are similar to those used in this study.

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.

Clearing of the Radioactive Liquid Waste from Oils and Oil Products by UV-Radiation at NPPs

The basic methods of concentration and purification of liquid radioactive waste on the nuclear power plant are distillation and ionic exchange. During vaporization of oil waste products and the fulfilled washing solutions the part of oil passes into a condensate. Clearing of such condensate on ion-exchanged filters results to oiling of ion-exchanged materials and to decrease number of filter cycles. Due to the often regeneration of ion-exchanged filters additional volumes of waste products as the fulfilled reclaiming solutions, washing and loosening waters are formed. The attention of scientists was involved with methods of clearing of water environments from the organic substances, based on deep oxidizing transformations of hydrocarbons into carbonic gas and water. From processes of oxidation of hydrocarbons up to CO2 and H2O, sold at moderate conditions, our attention has involved photochemical oxidation with the help of UV-radiation.

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.

Crystalline Silicotitanate: a New Type of Ion Exchanger for Cs Removal from Liquid Waste

The research and developments of a new type of inorganic ion exchanger, crystalline silicotitanate (CST) are reviewed. This material is stable against radiation, and the CST has high selectivity for Cs over Na, K and Rb. It performs well in acidic, neutral, and basic solutions. The results of ion exchange tests show that CST is an excellent candidate for Cs removal from high-level liquid waste.

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.

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.

医用放射性废液取样容器尺寸的最优化研究

医用放射性废液中核素放射性活度必须经监测满足相关标准后,该放射性废液方可排放,采用LaBr3(Ce)晶体探测医用放射性废液的核素活度,废液样品体积及其在探测器周围的分布情况直接影响探测效率,为此利用Geant4工具建立LaBr3(Ce)晶体探测模型,探索最优马林杯样品盒尺寸参数的变化规律,并在实验室利用?25.4 mm×25.4 mm的LaBr3(Ce)探测器和3D打印光敏树脂样品盒作验证实验。结果显示:单纯增加样品体积不能提高探测效率,随着样品体积增大,探测效率随样品盒环状部分深度方向的变化趋势趋于平缓。研究得到马林杯最优尺寸比例:环状部分深度(h2)和顶部半径(r)分别约为探测器晶体长度和空心腔直径的两倍,顶部半径(r)和样品盒高度(H)之比约为0.5。在最优化尺寸下,全能峰探测效率实验值与模拟结果一致,相对偏差优于2.5%。研究结果将为后续医用放射性废液监测装置在探测器选择、采样容器设计、加工及放射性废液监测量值溯源方法等方面提供重要的技术借鉴。

核电厂放射性废液环境对设备衬胶耐老化性能的影响

国内某核电厂放射性废液收集、处理、贮存、排放系统容器类设备采用内衬橡胶防腐设计,随着服役时间增长,内衬橡胶层性能会逐渐降低,乃至老化失效,影响设备运行安全,甚至导致放射性液体泄漏。通过橡胶热老化实验,推算对比内衬胶板在辐照、辐照浸渍等不同条件下的使用寿命;通过红外测试、热重测试、冲击强度测试、硬度测试,对比原始胶板、辐照胶板、辐照浸渍胶板等各项理化性能的变化,最终确定核电厂放射性废液环境下橡胶老化的主要影响因素。

沸石在中、低水平放射性废液水泥固化中的应用研究

对比研究了不同掺量(0、1%、3%、5%、7%)的斜发沸石和菱沸石对中、低水平放射性废液水泥固化体的流动度、抗压强度、抗浸泡性和抗浸出性的影响。结果表明:适量的斜发沸石或菱沸石均能有效提高水泥固化体的流动度、抗压强度和抗浸泡性;随着沸石掺量的增加,沸石对水泥固化体中^(90)Sr、^(137)Cs的吸附效果随之增强;斜发沸石对^(90)Sr的吸附效果优于菱沸石,而对^(137)Cs的吸附效果弱于菱沸石。

放射性废物焚烧炉设置泄爆装置的必要性分析

放射性废物焚烧过程中,有机气体积聚等工况下可能发生燃爆,通过调研国内外放射性废物焚烧设施,以及国内核设施、锅炉安全等方面的法律法规,针对不同源项下的放射性废物焚烧系统,分析设置泄爆装置的必要性。分析认为,放射性固体废物焚烧设施非常有必要设置泄爆装置,放射性有机废液焚烧设施设置泄爆装置可以进一步提高系统安全性。放射性废物焚烧系统应采取严密的防爆设计,尽可能避免燃爆事故的发生。

后处理污溶剂精馏回收工艺优化

乏燃料后处理厂运行过程中会产生大量的污溶剂,对污溶剂进行精馏处理,不仅可减少污溶剂的贮存和处理体积,还可以回收复用其中的有机溶剂,对后处理厂溶剂管理以及实现放射性废物最小化有重要意义。本文结合国内以往的工程工艺流程并对其进行优化设计,优化后的方案提高了系统的安全性和稳定性,并大大降低了能源的损耗,可为后续项目提供设计参考。