Mechanical Properties of Ductile Cast Iron and Cast Steel for Intermediate Level Waste Transport Containers

Abstract

UK Nirex Ltd is developing Type B re-usable shielded transport containers (RSTCs) in a range of shielding thicknesses to transport intermediate level radioactive waste (ILW) to a deep repository. The designs are of an essentially monolithic construction and rely principally on the plastic flow of their material to absorb the energies involved in impact events. Nirex has investigated the feasibility of manufacturing the RSTCs from ductile cast iron (DCI) or cast steel instead of from forgings, since this would bring advantages of reduced manufacturing time and costs. However, cast materials are perceived to lack toughness and ductility and it is necessary to show that sufficient fracture toughness can be obtained to preclude brittle failure modes, particularly at low temperatures. The mechanical testing carried out as part of that programme is described. It shows how the measured properties have been used to demonstrate avoidance of brittle fracture and provide input to computer modelling of the drop tests.     

Boxes for the Transport and Disposal of Low Level and Decommissioning Intermediate Level Radioactive Wastes

Abstract

UK Nirex Ltd is developing specifications for standard containers for the packaging of low and intermediate level radioactive wastes for disposal in a deep underground repository. The methodology used for selection of the dimensions and payloads of boxes for LLW and decommissioning ILW are described, and design features for handling the boxes are also briefly discussed. It must be emphasised that the results of this study are as yet preliminary, and may change in the light of the evolving system design for waste packaging, storage, transport and disposal.     

Preliminary Work Related to the Transport of Low and Intermediate Level Radioactive Waste in the Republic of Croatia

Abstract

In the Republic of Croatia, there is a project for a repository for low level (LL) and intermediate level (IL) radioactive waste disposal. Among many preliminary proceedings related to the proposed construction of the repository (such as site selection, repository project design, public acceptance, and the like), are the problems related to the transport of LL/IL radioactive waste from the place of its generaiion or storage to the location of the final disposal site. In this phase of the preliminary works— prior to site selection and further working out of project documentation for the facility, it has only been possible to commence with some study papers related to the problem. During 1992/1993 the first version of the generic study related to transport was worked out with the aim of preparing for a more detailed study, and technical and investment support for further working out of the project. The study is firstly a literature abstrac,, with an analysis of the transport problem for Croatia and recommendations for a transport system, type of transport, transport equipment and dynamics for the requirements of the Repubiic of Croatia.     

Study of Standard Test Method for the Characterization ofSolidified Waste of Low and Intermediate Level Radioactive Waste

Thecharacterizationofsolidifiedwasteoflowandintermediatelevelradioactivewastebynearsurfacedisposaldirectlyaffectourlivingenvironment.It'snecessarytostudythestandardtestmethodforthecharacterizationofsolidifiedwaste.Thestandardtestmethodforthecharacterizati…     

An Environmental Impact Assessment for Sea Transport of High Level Radioactive Waste

Abstract

This work was carried out to study the safety evaluation in a hypothetical submergence accident onto the seabed, prior to the international maritime transpott between Europe and Japan in 1995. In this study, inadmissibly conservative assumptions were omitted in order to construct adequate accident scenarios from the engineering aspect. Input data of source terms of high level vitrified wastes, various flow coefficienss in the sea, and other factors were thoroughyy examined and, finally, a new concept of a solution method for radioactive nuclides concentration was proposed with regard to oceanography.     

放射性废物中等深度处置

放射性废物中等深度处置是介于近地表处置和地质处置的一种处置方式。本文介绍了国际上放射性废物中等深度处置的发展概况及我国可能适于中等深度处置的废物源项,提出我国实施放射性废物中等深度处置需考虑的问题及今后开展相关工作的建议。     

Conceptual Design of Large Boxes for the Packaging of Intermediate and Low Level Radioactive Waste

Abstract

United Kingdom Nirex Limited (Nirex) is developing standard containers for the packaging of radioactive waste for disposal in a deep underground repository. Waste Package Specifications have been produced for each standard package to provide the essential link between waste package design and the design of the deep repository. Previous studies carried out by Nirex identified the dimensions and key features of standard boxes for decommissioning intermediate level waste and for low level waste: the 4 m ILW box, 4 m LLW box·and 2 m LLW box. Nirex has now produced conceptual designs for these boxes.     

低中放固体废物处置场关闭后的景象分析

在低中放固体废物处置场关闭后的一定时期内仍将继续监护。当监护期结束以后,处置场场址实现无限制或有限制开放,公众可能通过某种途径与处置场填埋废物近距离接触,从而受到填埋废物中的放射性核素的辐射影响。为估算处置场关闭后对环境造成的影响,需对处置场关闭后影响核素释放和迁移的特征、事件以及过程进行分析和筛选。本文依据IAEA研究成果,对国内某低中放固体废物处置场关闭后的景象进行了分析和筛选。分析和筛选结果表明,处置场关闭后景象的选择应在充分调查的基础上,合理确定评价场景。     

Transport of Radioactive Waste in Poland

Abstract

General information is given about the regulations and limits concerning radioactive waste in Poland. Radwaste—being in 95% low level—comes at present from one research reactor and over 2000 smaller producers; there is no high level waste. The responsibility for collecting, handling and disposing of all radwaste is delegated to one organisation partially supported by the state. The frequency of transport to the Central Repository is about 50 times a year. The total volume of conditioned radwaste is about 200 m³.     

西北处置场低、中放固体废物处置实践

介绍了我国西北处置场的基本情况,西北处置场低、中放固体废物接收、处置的组织机构及管理体系,工艺技术,辐射防护及环境监测,以期为我国放射性废物处置工程提供借鉴和支持。     

Computer Modelling of the Impact Performance of Containers for the Transport of Radioactive Materials

Abstract

ABA Technology is an acknowledged world expert in the structural dynamics area through its research programmes in support of nuclear power. Initially, the impact and blast response of metal and reinforced concrete structures relevant to nuclear power plants were studied. Since the mid 1980s these studies have been expanded to include containers for transporting radioactive materials. These programmes are continuing and new areas of interest are being addressed. Experimental transport performance studies are conducted using large gas guns and drop test facilities to achieve impact velocities at, and above, regulatory conditions. Instrumentation allows monitoring of transient pressures, accelerations, displacements, strains and loads. The data are used to predict full-scale structural response from scale models, and to verify calculation methods which range from PC-based mechanistic models to non-linear finite element codes on AEA Technology's CRAY-2 computer. Emphasis is placed on verifying procedures to predict margins to failure in containers. For many structures this involves identifying appropriate failure criteria and the requirement for including strain rate effects. Where necessary special developments of computer codes are made. To date three types of container have been studied: a cylindrical flask for spent fuel, and waste packages consisting of a ruggedised ISO container and a steel clad concrete container. The combination of experimental testing and analysis provides an optimum solution for design and assessment problems and the proven techniques are being transferred to many other applications in other industries.     

Radioactive Waste Transport in the Russian Federation

Abstract

According to the national radwaste management programme, low and intermediate level wastes originating from nuclear power plant (NPP) operation, from the nuclear fuel cycle and from the application of radioisotopes in medicine, industry and research are centrally collected and transported for disposal at 16 special repositories. Packagings and conveyances for these purposes are discussed. A concept of regional repositories and transport of wastes from NPPs and the Navy is also considered together with transport packagings and transport methods.     

低水平放射性废物的混合

介绍了美国低水平放射性废物混合的现状及其4种混合方案,提出了在我国实施放射性废物混合的建议。     

模拟非α低中放废液水合陶瓷固化体的水化产物和力学性能

以模拟非α低中放废液为固化处理对象,用水热法合成了非α低中放废液中和处理后沉淀的＂碱-矿渣-粉煤灰-偏高岭土＂水合陶瓷固化体。采用X射线衍射仪分析了固化体的水化产物,确定了水化产物的组成,并测试了固化体的抗压强度。研究结果表明：温度为150～180℃、废液沉淀与固化原材料的质量比值（即盐灰比）为0.10～0.30时,固化体水化产物的主要物相为方沸石,随着温度升高和反应时间延长,水化产物中方沸石的衍射峰不断增加。固化体抗压强度测试结果表明,该固化体具有较高的抗压强度,但盐灰比由0.10增加至0.30时,固化体抗压强度由26.33 MPa下降到8.46 MPa。     

放射性物料的清洁解控

放射性物料的清洁解控是放射性废物最小化的一项重要手段。为了进一步指导物料解控,本文介绍了与清洁解控水平相关的三个国家标准,明确了解控标准的适用条件,说明了使用标准进行物料解控过程中应该注意的问题,并结合解控审管实践对解控过程中几个重要的审管要求进行了讨论,最后提出了申请解控文件应包括的主要内容。     

高放废物嬗变研究发展及前景

高放废物的处置关系到核能的发展和环境保护。分离与嬗变作为高放废物处置的可能途径之一，在世界上已有近３０年的研究发展历史，目前，越来越受人们的关注。本文分别介绍了利用轻水堆、液态金属燃料堆、聚变堆（聚变－裂变混合堆）以及加速器驱动的次临界系统嬗变高放废物的原理、研究发展及前景。为我们开展高放废物嬗变研究提出一些建议。     

基于蒙特卡罗法的放射性物质运输容器的失效概率分析

考虑放射性物质运输容器的主要评定参数(如机械应力水平、火烧温度)的随机性,应用重要抽样蒙特卡罗方法在Matlab环境下编写计算机程序,求得在一定置信度下的失效概率。结果表明,蒙特卡罗方法反映了评定参数客观存在的不确定性,是历史记录和实验的有效补充。