The Effect of Montmorillonite Partial Density on the Role of Colloid Filtration by a Bentonite Buffer

Japan's concept for high-level radioactive waste disposal includes a barrier of compacted bentonite (montmorillonite) that, based on its microstructure, buffers the waster package and prevents migration of colloids with sizes between 1 and 1,000 nm. However, results of experimental studies indicate that colloid migration through the barrier (breakthrough) depends upon the partial density of the montmorillonite buffer. To assess the possibility of colloid breakthrough, interstitial pore size as a function of montmorillonite partial density is estimated through a homogenization analysis. Homogenization analysis indicates that interstitial pores sizes of <1 nm may be achieved for montmorillonite partial densities in excess of 1.2Mg/m³. It should be noted that this estimation depends upon the specific structural/geometric parameters of the compacted bentonite under consideration.     

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。     

Recent Waste Management Related Developments in Germany

The Federal Republic of Germany intends to dispose of all types of radioactive waste in deep geological formations. This waste comprises spent fuel elements, vitrified fission products, nuclear power plant operational and decommissioning waste as well as spent sealed radiation sources and miscellaneous waste originating from small waste generators. The Federal Government made a pronounced change in energy policy since 1998, the most important feature of which is the abandoning or phasing out of nuclear energy, finally set in force by the April 2002 amendment of the Atomic Energy Act. According to the new approach to waste management and disposal, further sites in various host rocks shall be investigated for their suitability. A special expert group was set up to develop repository site selection criteria and respective procedures on a scientifically sound basis. The objective of the site selection procedure is to identify—with public participation—potential disposal sites in a comprehensible and reliable way. Step by step and based on the said criteria, those areas, site regions and eventually sites shall be selected, which offer particularly favourable conditions for the later demonstration of the final site's suitability and its confirmation in a licensing procedure.     

PhreePlot绘制某些可变价核素Eh-pH图

U、Np、Pu、~(99)Tc、~(79)Se等放射性元素或核素对氧化还原条件敏感,存在多种价态,且锕系元素(U、Np、Pu)具有丰富的配位化学,其在生物圈和地质圈中的化学行为较为复杂。电势(Eh)-pH图对解决在水溶液中发生的一系列反应及平衡问题具有广泛的应用。本工作采用PhreePlot软件绘制了这些元素或核素在人体血清、唾液、胃液、DMEM培养基和北山地下水中的Eh-pH图以研究它们的主要氧化态和赋存形态。结果表明,U在生物圈中的主要氧化态是Ⅵ,Np是Ⅳ,Pu是Ⅳ和Ⅲ,99 Tc是Ⅶ和Ⅳ,79Se是0和Ⅳ。相比其他元素或核素,pH对铀的赋存形态影响更大。在地质圈中,根据Eh-pH图来看,北山处置库有利于Np、Pu的长期储存而对于U、99 Tc、79Se的储存不利。本工作的对比研究也表明,在热力学数据库一致的情况下,采用免费软件PhreePlot绘制的元素Eh-pH图与Geochemist’s Workbench(GWB)程序绘制的Eh-pH图具有高度的一致性。     

放射性废物处置安全若干进展

结合2005年放射性废物处置安全国际大会反映出来的放射性废物处置安全领域的最新进展,介绍了全球放射性废物安全框架、废物处置安全战略、安全方案、地质处置设施安全、近地表处置设施安全、中等深度废物处置方案和公众沟通等方面的若干新进展和新观点。     

Building technical and social confidence in the safety of geological disposal in Japan

Geological disposal has been adopted as the most feasible option for the method of long-term management of high-level radioactive waste (HLW) in every country in the world, regardless of the pros and cons of the nuclear power generation. Building stakeholders’ confidence in safety of geological disposal is indispensable to reach the point where the implementation of geological disposal is accepted by the current generation. The safety case is a key input to build confidence in geological disposal stepwise as the program progresses and regarded to play an important role as a common platform in the communication among stakeholders.

The aim of this paper is to review arguments relevant to building technical and social confidence in the progress of Japanese research and development activities as well as international discussions.     

新疆天山地区活动构造特征及区域地壳稳定性分析

区域地壳稳定性是高放废物处置库选址的重要指标之一。基于我国新疆天山地区的活动构造特征及其与地震空间分布关系的研究,对研究区地壳稳定性进行分析,选出相对稳定的"安全岛"地区,为高放废物处置库选址提供依据。     

对法国高放废物处置安全审评的分析研究

介绍了法国高放废物处置研究现状和规划,对法国高放处置场的审评技术单位法国核与辐射安全研究院（IRSN）所开展的高放处置安全研究和审评工作及其提出的审评原则和审评要点进行了分析研究,并对我国的高放处置安全审评工作提出了建议。     

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.     

高放废物处置库选址中的水文地质特性评价

根据国内外研究经验,综述高放废物地质处置库选址不同研究阶段水文地质特性评价的方法和作用,并将评价方法分为区域、地区和场址3个阶段水文地质调查来讨论。同时,介绍了我国高放废物地质处置场地水文地质特性评价研究现状。     

放射性废物处置安全全过程系统分析中的指标

以往放射性废物处置的安全评价中通常使用放射性安全指标(即剂量和危险),随着放射性废物处置安全全过程系统分析这一概念的提出,辅助指标已成为评价中的一个重要组成部分。本文介绍了安全全过程系统分析中所使用指标的发展、分类和相应标准等。依评价对象不同,辅助指标通常分为安全指标和性能指标,有些组织还提出了安全功能指标;与上述指标相对应的用于比较的标准分别为参考值、指标标准和安全功能指标标准。将处置系统划分为不同库室时,指标还可分为“包容物和浓度”相关指标、“通量”相关指标和“屏障状态”相关指标三类。建议我国尽快开展放射性废物处置的安全全过程系统分析工作,建立完善的指标体系,选取适当的评价指标,并基于我国放射性废物处置的场址特性确定相应的标准,以期实现安全和防护的最优化。     

微生物对高放废物地质处置库的影响

介绍了国内外微生物对高放废物地质处置安全性影响的研究现状,并对我国本领域未来的研究工作提出了相应的建议。     

高放废物地质处置安全全过程系统分析研究进展

高水平放射性废物(以下简称“高放废物”)地质处置安全全过程系统分析是对处置库长期安全的综合性系统分析。本文论述了安全全过程系统分析的概念及近20年发展情况,强调了其在高放废物地质处置库规划、选址、设计、建设、运行、关闭和关闭后等各阶段的重要性。以国际原子能机构(IAEA)和经合组织核能署(NEA)技术文件为主,阐述了安全全过程系统分析的背景、安全策略、评价基础、安全评价及综合等核心组成部分,列举了其应用中管理和技术方面的12个关键点,分析了芬兰、瑞典、美国和法国的研究应用现状,并对我国放射性废物地质处置安全全过程系统分析的发展及应用现状进行了探讨。     

2013—2018年山西省放射性废物库库区环境γ辐射水平监测

为了进一步加强对放射性废物库的安全管理,确保辐射环境安全,对2013—2018年山西省放射性废物库库区环境γ辐射剂量率的监测结果进行了分析。结果表明,库区环境γ辐射水平满足《核技术利用放射性废物库选址、设计与建造技术要求(试行)》中有关库房内源坑盖板上方0.5 m处γ辐射剂量率不超过20μGy/h、源库墙外表面0.2 m处γ辐射剂量率不超过2.5μGy/h的规定要求,各年度间环境电离辐射水平处于本底涨落范围内,未对周围环境产生辐射影响,辐射环境质量总体良好。此外,健全的库区安全防范也为促进我省核技术利用和安全、健康、可持续发展提供了坚强保障。     

Preliminary and Scoping Analysis for Uncertainty of HLW Geologic Repository Performance with Multiple Canister Repository Model by Earth Simulator

An uncertainty analysis of repository performance has been made by the VR code, which incorporates interference effects of multiple canisters in a repository. A problem of the previous study with the VR code is that the number of connected canisters was determined arbitrarily. In this study, first, the probability distribution functions for the number of canisters connected by the flow-bearing fracture clusters have been determined by the FFDF code, and then uncertainty associated with the peak fractional release rate of ²³⁷Np to the far field resulting from uncertainties with the buffer sorption coefficient, solubility, and the number of connected canisters has been numerically evaluated by the Latin-Hypercube Sampling method. The effects of uncertainty with the total number of connected canisters become less important as the number increases because the radionuclide concentration saturates in downstream compartments. Uncertainty with the buffer sorption and solubility shows an important contribution to that with the nuclide release rate.     

安全评价软件Ecolego解法器选取研究

Ecolego软件已广泛应用于放射性废物处置场的安全评价中,其提供了十余种解法器供用户使用,但不同解法器的性能相差较大,在安全评价中如何选取适宜的解法器给用户造成了困扰。本研究在对各解法器特点进行论述的基础上,利用Ecolego 6.5提供的解法器对瑞典高放废物处置安全评价模型进行测试,结果表明NDF解法器在地质圈及总模型计算耗时最少,BDF及RADAU5解法器分别在近场模型及生物圈模型耗时最少;Rosenbrock与TR-BDF2解法器计算结果稳定。综合计算耗时与求解精度考虑,在类似瑞典高放废物处置场地的安全评价模拟中,推荐使用TR-BDF2解法器。     

Impact of Partitioning and Transmutation on High-Level Waste Disposal for the Fast Breeder Reactor Fuel Cycle

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large storage capacity because the number of waste packages produced is significantly reduced by a factor of 5 from that of the glass waste package in the FBR fuel cycle without PT.