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.     

Shale as a repository for radioactive waste: The evidence from Oklo

Study of the 1.8 billion-year-old “fossil nuclear reactor” zones at the Oklo Mine in the Republic of Gabon shows that many of the elements produced by fission have been almost completely retained, as evidenced by proper budgets of stable daughter elements. Plutonium, ruthenium, the rare earth elements, zirconium, and palladium have been effectively retained while most chalcophile elements exhibit some degree of remobilization. The alkali and alkaline earth elements have migrated to varying degrees but their presence in gangue affected by younger periods of alteration suggests redistribution not far removed from sites of formation. More important, such migration may not have started until some 25,000,000 years after the reactor shut down. The noble gases xenon and krypton escaped with apparent ease during the 500,000 years the reactor was operative, and iodine seems to have been mobile. The Oklo reactor ores, significantly, occur in shale infilled into a fracture system in organo-argillaceous sandstone. So many of the fission-produced elements retained in this shale along with evidence that most others may have been only locally redistributed lends support to considering shales in geologically stable areas for radioactive waste disposal.     

Analysis of the hydration of a bentonite seal in a deep radioactive waste repository

A deep geological repository for nuclear waste requires the backfilling and sealing of shafts and galleries to block any preferential path for radioactive contaminants. The paper presents the coupled hydromechanical analyses of an in situ test carried out in the HADES underground laboratory in Mol, Belgium. The test examines the effectiveness of an expansive clay seal in a horizontal borehole specifically drilled for this purpose. The analysis covers the phase of seal hydration up to saturation and subsequent pore pressure equilibration. Hydraulic and mechanical constitutive laws suited to expansive clay materials have been chosen for the analyses with all the parameters determined independently of the in situ test. A quite good agreement has been found between test observations and computed results suggesting that the numerical formulation employed is able to reproduce the main features of a real sealing system.     

The potential effect of cementitious colloids on radionuclide mobilisation in a repository for radioactive waste

Colloid-facilitated transport of contaminants could enhance the release rate of radionuclides from the cementitious near field of a repository for radioactive waste. In the current design of the planned Swiss repository for intermediate-level radioactive waste, a gas-permeable mortar is employed as backfill material for the engineered barrier. The main components of the material are hardened cement paste (HCP) and quartz aggregates. The chemical condition in the backfill mortar is controlled by the highly alkaline cement pore water present in the large pore space. The interaction of pore water with the quartz aggregates is expected to be the main source for colloids. Colloid transport is facilitated due to the high porosity of the backfill mortar. Batch-type studies have been performed to generate colloidal material in systems containing crushed backfill mortar or quartz in contact with artificial cement pore water (ACW) at pH 13.3. The chemical composition of the colloidal material corresponds to that of calcium silicate hydrates (CSH). Batch flocculation tests show that, after about 20 days reaction time, the concentration of the CSH-type colloids is typically below 0.1 mg l⁻¹ due to reduced colloid stability in ACW. Uptake studies with Cs(I), Sr(II) and Th(IV) on a CSH phase (initial C:S ratio=1.09) have been carried out to assess the sorption properties of the colloidal material. The influence of uptake by colloids on radionuclide mobilisation is expressed in terms of sorption reduction on the immobile phase (HCP). Sorption reduction factors can be estimated on the assumption that the sorption properties of the colloidal material are either similar to those of the CSH phase or HCP, and that sorption is linear and reversible. A scaling factor accounts for the higher specific surface area of the colloidal material compared to the CSH phase and HCP. At colloid concentration levels typically encountered in highly alkaline cement pore waters, colloid-induced sorption reduction is predicted to be negligibly small even for strongly sorbing radionuclides, such as Th(IV). Thus, no significant impact of cementitious colloids on radionuclide mobilisation in the porous backfill mortar is anticipated.     

花岗岩中高放废物地质处置库系统分析方法研究

首先以甘肃北山预选区花岗岩场址为例,提出该场址中高放废物地质处置库概念设计和结构设计,然后以系统分析方法论为基础,描述处置库的系统功能、结构、环境及其演化过程,并以模拟软件GoldSim为工具,建立该处置库演化过程的计算机模型,最后以该计算机模型为模拟实验平台,模拟处置库中辐射毒性时空分布,分析模型中的参数灵敏度,优化设计参数,并预测和评价处置库性能。其研究成果可为合理配置资源和有效协调各研究项目之间关系提供技术支持。     

新疆地区选建高放射性核废物处置库的可能性

高放废物选址是一项万年大计工程,需要多个预选场地的综合比较,决策废物处置库的具体选建场地。论述了高放废物处置库选址过程中需要考虑的主要问题,然后针对新疆的自然环境,水资源、花岗岩体和缓冲/回填材料的分布,以及区域地壳稳定性等因素,讨论了在新疆选建高放废物地质处置库的可能性,并提出了选址方法。     

新疆地区选建高放射性核废物处置库的可能性

高放废物选址是一项万年大计工程,需要多个预选场地的综合比较,决策废物处置库的具体选建场地.论述了高放废物处置库选址过程中需要考虑的主要问题,然后针对新疆的自然环境,水资源、花岗岩体和缓冲/回填材料的分布,以及区域地壳稳定性等因素,讨论了在新疆选建高放废物地质处置库的可能性,并提出了选址方法.     

新疆地区选建高放射性核废物处置库的可能性

高放废物选址是一项万年大计工程,需要多个预选场地的综合比较,决策废物处置库的具体选建场地。论述了高放废物处置库选址过程中需要考虑的主要问题,然后针对新疆的自然环境,水资源、花岗岩体和缓冲/回填材料的分布,以及区域地壳稳定性等因素,讨论了在新疆选建高放废物地质处置库的可能性,并提出了选址方法。     

Fluorescence spectroscopy of U(VI)-silicates and U(VI)-contaminated Hanford sediment

Time-resolved U(VI) laser fluorescence spectra (TRLFS) were recorded for a series of natural uranium-silicate minerals including boltwoodite, uranophane, soddyite, kasolite, sklodowskite, cuprosklodowskite, haiweeite, and weeksite, a synthetic boltwoodite, and four U(VI)-contaminated Hanford vadose zone sediments. Lowering the sample temperature from RT to ∼ 5.5 K significantly enhanced the fluorescence intensity and spectral resolution of both the minerals and sediments, offering improved possibilities for identifying uranyl species in environmental samples. At 5.5 K, all of the uranyl silicates showed unique, well-resolved fluorescence spectra. The symmetric O = U = O stretching frequency, as determined from the peak spacing of the vibronic bands in the emission spectra, were between 705 to 823 cm⁻¹ for the uranyl silicates. These were lower than those reported for uranyl phosphate, carbonate, or oxy-hydroxides. The fluorescence emission spectra of all four sediment samples were similar to each other. Their spectra shifted minimally at different time delays or upon contact with basic Na/Ca-carbonate electrolyte solutions that dissolved up to 60% of the precipitated U(VI) pool. The well-resolved vibronic peaks in the fluorescence spectra of the sediments indicated that the major fluorescence species was a crystalline uranyl mineral phase, while the peak spacing of the vibronic bands pointed to the likely presence of uranyl silicate. Although an exact match was not found between the U(VI) fluorescence spectra of the sediments with that of any individual uranyl silicates, the major spectral characteristics indicated that the sediment U(VI) was a uranophane-type solid (uranophane, boltwoodite) or soddyite, as was concluded from microprobe, EXAFS, and solubility analyses.     

高放废物地质处置库北山预选区地下水的形成和分布

本文论述北山地区不同类型地下水的形成和赋存规律。松散岩类孔隙水主要分布于沟谷和洼地，主要为潜水。沟谷潜水的形成以基岩裂隙水的侧向补给及洪流渗漏补给为主，洼地潜水可由沟谷潜水、断裂带或基岩裂隙水、古洪积扇潜水侧向补给，也可由地表洪流渗入补给；碎屑岩类孔隙-裂隙水主要分布在由自垩系或第三系砂砾石构成的山间沉积盆地中，以承压水为主，洪水下渗和盆地周围基岩裂隙水的侧向径流是此类地下水的主要补给来源；基岩裂隙水主要赋存于变质岩、岩浆岩、碎屑岩、碳酸盐岩节理、裂隙中，以潜水为主。降水垂直渗入、侧向补给或洪流渗漏补给为其主要来源。     

MapGIS在东天山高放废物地质处置库选址中的应用

根据高放废物选址的要求,利用MapGIS对东天山地区不同时代岩体的空间位置、分布形态、出露面积,以及岩体与断裂构造、地震和矿床点等的空间关系进行了研究,对海量资料信息进行分析处理,可起到事倍功半的效果。利用MapGIS分析结果,初步筛选出阿奇山1号岩体、雅满苏北岩体作为高放废物地质处置库选址的有利岩体。     

Scoping analyses of the coupled thermal-hydrological-mechanical behaviour of the rock mass around a nuclear fuel waste repository

Scoping calculations were performed in order to assess the influence of radiogenic heat on the performance of the rock mass around a nuclear fuel waste repository. The full coupling between the thermal, mechanical and hydrological processes involved was considered by using the finite element code, FRACON, developed through an extension of Biot's classical theory of soil consolidation. By considering the full T---H---M coupling, several important safety features, which would otherwise be omitted in uncoupled analyses, were detected in the present study. In particular, it was shown that the heat-induced pore pressure increase around the repository has the potential to significantly increase the rate of groundwater flow, and affect the structural integrity of the rock mass.     

高放废物处置库北山预选区泉水成因分析

文章通过前人资料的分析,结合水文地质调查工作,探讨了北山地区泉水的成因,认为该区分布的泉大部分是下降泉,接受潜水补给;少部分为上升泉,接受盆地地下水或断裂带地下水补给;属于北山内部的地下水循环交替方式之一,与深大断裂关系不明显。     

ArcGIS在北山高放废物处置库预选区地学数据管理中的应用

以甘肃北山高放废物处置预选区所积累的数据为基础,研究地学信息数据管理平台的建立,实现对各种空间数据及其他非空间数据统一管理与应用,并提出了基于ArcGIS产品的预选区数据管理与共享的解决方案。     

高放废物处置库甘肃北山预选区地下水位动态特征

阐明了甘肃北山预选区地下水位动态特征及其与当地大气降水的关系,并对其动态类型进行了剖析,由此得出北山地区地下水源自当地大气降水补给的认识。为处置库预选区场址评价提供了水文地质依据。     

考虑细观等效热学参数的高放废物处置库围岩应力−渗流−温度耦合模拟方法

高放废物处置库中乏燃料持续释放的热量对围岩的应力场和渗流场及其长期稳定性具有重要影响。围岩的热学参数依赖于岩石矿物组成、孔隙率和孔隙流体等因素,准确取值是进行高放废物地质处置库多场耦合分析的前提。通过细观力学分析,建立了围岩等效热学参数（热容、热传导系数、热膨胀系数）取值方法,并基于Biot孔隙介质理论,建立应力?温度?渗流三场耦合模型,进而提出了高放废物处置库围岩应力?渗流?温度耦合数值模拟方法。最后通过COMSOL Multiphysics多场耦合软件,利用瑞士Mont Terri高放废物地下试验室围岩温度?渗流?应力多场耦合现场试验数据对数值模拟方法进行验证,并探讨了温度?渗流?应力耦合过程的演化规律。研究表明,模拟结果和试验值吻合良好。研究结果可为我国高放废物处置库的安全评估和选址提供科学依据。     

Reduction of U(VI) to U(IV) on the surface of magnetite

To increase the understanding of uranium transport in the environment and in the presence of steel corrosion products, the interaction of U(VI) with natural magnetite has been studied. Sorption studies have been carried out using X-ray photoelectron spectroscopy (XPS) and secondary ion mass spectrometry (SIMS). The XPS results clearly indicate the reduction of U(VI) to U(IV) on the surface of magnetite facilitated by electron transfer between the Fe and U, leading to a coupled oxidation of Fe(II) to Fe(III).     

The effect of U(VI) bioreduction kinetics on subsequent reoxidation of biogenic U(IV)

Microbially mediated in situ reduction of soluble U(VI) to insoluble U(IV) (as UO₂) has been proposed as a means of preventing the migration of that radionuclide with groundwater, but preventing the oxidative resolubilization of U has proven difficult. We hypothesized that relatively slow rates of U(VI) bioreduction would yield larger UO₂ precipitates that would be more resistant to oxidation than those produced by rapid U(VI) bioreduction. We manipulated U(VI) bioreduction rates by varying the density of Shewanella putrefaciens CN32 added to U(VI) containing solutions with lactate as an electron donor. Characterization of biogenic UO₂ particles by extended X-ray absorption fine-structure spectroscopy and transmission electron microscopy revealed that UO₂ nanoparticles formed by relatively slow rates of U(VI) reduction were larger and more highly aggregated than those formed by relatively rapid U(VI) reduction. UO₂ particles formed at various rates were incubated under a variety of abiotically and biologically oxidizing conditions. In all cases, UO₂ that was formed by relatively slow U(VI) reduction was oxidized at a slower rate and to a lesser extent than UO₂ formed by relatively rapid U(VI) bioreduction, suggesting that the stability of UO₂ in situ may be enhanced by stimulation of relatively slow rates of U(VI) reduction.     

我国高放废物地质处置库场址筛选总体技术思路探讨

综合对比瑞典、加拿大、芬兰和美国等国高放废物地质处置库场址筛选技术思路,分析国外高放废物地质处置库场址筛选过程中取得的经验、教训,总结了我国处置库选址工作取得的成果和存在的问题。在综合研究基础上,分析提出我国高放废物地质处置库场址筛选总体技术思路,包括应遵循的原则、工作范围、目标和总体技术步骤等,以利于今后处置库选址工作更系统、规范和统一。