Safety Aspects of Long-term Dry Interim Storage Of Type-b Spent Fuel and Hlw Transport Casks

Abstract

Based on the German decision to minimise transport of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities were licensed until the end of 2003. Because of the large amount of Type B(U) transport casks which are going to be used for long-term interim storage the question of timelimited Type B(U) licence maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of qualification of transport casks for interim storage is the long-term behaviour of the metallic seal–lid system. Here results are presented from current long-term experimental tests with metallic 'Helicoflex' seals in which pool water is enclosed. This series of tests has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an exchange of experience, know-how and state-of-the-art between authorities and technical experts with regard to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called 'coordinating institution for cask dispatching information' ('KOBAF') which entails management of an online database of cask-specific documents and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the coming decades.     

Burnup in Criticality Safety for Spent Fuel Casks

Abstract

Criticality safety margins must be based upon the combination of the best available prediction of the margin and all uncertainties in the prediction. Inclusion of the effects of burnup in the evaluation of spent fuel shipping or storage casks must be based upon a thorough understanding of the prediction of the effects of burnup and the uncertainties in the measurements (or predictions) of burnup and predictions of the effects. A preliminary estimate of the effects of burnup and its uncertainties is presented. This will serve as a first step in the effort to develop acceptance criteria that assure public safety. An assembly average burnup of 20,000 MWD/MTU represents an increase in the criticality safety margin of about 20% (δk/k), and the current estimate of the uncertainty in this value is close to 4% (δk/k). The uncertainties in the components of the effects of burnup were based upon relevant literature citations and — where no other information was available — upon estimates. Consequently, the margins and uncertainties in the margin presented here should be considered as initial estimates upon which more refined analyses should build to develop a defensible basis for predicting and reviewing the criticality safety margins which include the effects of burnup.     

Long-Term Containment Performance Test Facilities for Spent Fuel Transport/Storage Casks (CRIEPI,Japan)

Abstract

Long-term containment tests of full-scale transport/storage cask models have been in progress since 1990 at CRIEPI, Japan. The results demonstrate and confirm the very reliable containment pelformance of the cask lid structure with metallic gaskets.     

Design Assessment for Transport and Storage Casks

Abstract

The design assessment concerning the mechanical behaviour of transport and storage casks for radioactive material to fulfil nuclear safety criteria has to be based on two essential considerations: (1) Effective analysis of the stress–strain state of the cask components under both normal operational and test conditions including hypothetical accident scenarios with suitable accepted methods. (2) Economic estimation of the required properties and the structural state of the cask components with sufficient exactness. In an overview of the codes which are available at GNS/GNB for cask impact strength analyses (ANSYS, ADINA, VDI Codes), procedures and aspects of benchmarking and validation of calculation codes are described. The results of experimental full size cask drop test programs (CASTOR, POLLUX) and corresponding pre-test calculational analyses show the suitability of the codes used. The influence of dynamic effects on the mechanical properties of material (ductile cast iron, wood) has been investigated experimentally. By consideration of these dynamic values in strength analyses of casks at impact a good agreement between experimental and calculational results has been achieved.     

Numerical Assessment of Spent Fuel Casks Impacting on Real Targets

Abstract

A reference container of high capacity was analysed for loads beyond those it has to withstand during a 9 m IAEA drop test onto an unyielding target. In doing this a lid-end drop with shock absorber onto a real target was simulated. This is a possible accident for the rail transport of such casks. In this case the most critical components of the containment system are the primary lid bolts. The behaviour of the lid system and its sealing function were investigated with finite element (FE) analysis. To correlate the findings with a corresponding impact velocity onto real targets an analytical method was used. Despite the conservative assumptions made in this study a two-fold safety factor compared to the 9 m drop tests onto the unyielding target could be shown. The quantification of the additional safety the cask might provide requires further basic investigations on the behaviour of the real targets considered as well as the reduction of the conservatism included in the assumptions made up to now.     

压水堆乏燃料干式贮存的安全考量

通过对压水堆乏燃料干式贮存的准备、运输和贮存三个工艺区域的设备设施、操作过程和环境及其潜在的危险因素的研究, 重点针对作用于干式贮存设施的临界安全、放射性物质的包容、衰变热的移除和辐射防护的影响因素开展分析, 提出了针对性的防控对策;同时总结了秦山第三核电厂重水堆乏燃料干式贮存实施过程安全操作和管理的经验。研究成果可为解决压水堆乏燃料干式贮存设施的规划、设计、建设和运行过程有关安全问题提供思路。     

高密度乏燃料贮存格架临界安全设计

基于第三代先进非能动压水堆核电厂的设计特点、运行方式及其复杂的燃料组件设计,考虑各种能谱硬化因素,研究组件燃耗计算的运行条件组合,获得指定燃耗深度下的核素密度。建立乏燃料贮存格架的临界计算模型,并对临界安全分析中的关键因素(如末端效应、可信事故工况等)进行详细研究,最终初步设计出满足我国最新(临界)标准和要求的、可应用于实际工程的高密度乏燃料贮存格架。     

AP1000乏燃料贮存格架临界安全分析

基于SCALE6程序包对西屋公司采用燃耗信任制技术的AP1000核电厂乏燃料贮存格架(SFSRs)临界安全分析过程进行了复现,在此基础上结合AP1000核电厂堆芯反应性控制特性,分析了轴向燃耗分布对系统反应性的影响。结果表明,高燃耗下采用机械补偿(MSHIM)轴向燃耗分布计算得到的系统反应性更保守,同时临界安全分析中需考虑吸收体在组件燃耗过程中对反应性的影响,且不应信任可溶硼。     

LWR Spent Fuel Storage Behaviour

KWU and DWK performed a joint program on LWR Spent Fuel Storage Behaviour covering all essential aspects of the fuel behaviour for wet and dry storage both for intact and fuel with operational defects. The program comprises theoretical work, laboratory experiments and performance tests with spent fuel. From the results we can conclude that wet storage is without any practical limitations in time. A suitable decay time in a pond — which may be less than 1 year — will assure also safe and reliable dry storage under inert conditions.     

AP1000核电厂乏燃料贮存安全性分析

福岛事故暴露出了二代沸水堆乏燃料组件贮存的安全问题。本文比较了三代AP1000核电技术与二代沸水堆技术在乏燃料贮存方面的差异。AP1000核电厂乏燃料水池冷却系统运用先进的非能动设计,通过多种补水方式和补水水源以及沸水蒸汽排放控制等措施可有效地解决福岛事故中存在的问题,保障了乏燃料组件贮存的安全性。     

飞机撞击下高温气冷堆乏燃料厂房损伤研究

本文针对典型高温气冷堆乏燃料厂房在双发商用飞机撞击载荷下的响应及结构完整性开展研究,并探讨结构特性对撞击损伤的影响。对乏燃料厂房及飞机分别建立有限元模型,通过弹体-目标相互作用分析模拟了飞机撞击过程,综合IAEA与NRC的评价准则对乏燃料厂房在飞机撞击下的损伤程度进行评估。数值结果表明：厂房上对应于机身及发动机的撞击位置发生可接受的局部损伤;乏燃料贮存井墙体对于提高构筑物抗飞机撞击能力有重要作用。此外,构筑物外形对损伤有很大影响,圆柱形壳体的抗飞机撞击能力显著强于方形厂房,是核电厂厂房设计的优化方向之一。     

International Working Group for Sabotage Concerns of Transport and Storage Casks

Abstract

The International Working Group for Sabotage Concerns of Transport and Storage Casks (IWGSTSC), gathers multiple organisations from different countries (for US party Department of Energy, Nuclear Regulatory Commission, and Sandia National Laboratories; for German party Gesellschaft für Anlagen- und Reaktorsicherheit and Fraunhofer Institut; for the French party Institut de Radioprotection et de Sûreté Nucléaire). The goal of the IWGSTSC is to continue cooperation to improve the analytic capabilities, through information sharing and collaborative research and development plus modelling, to understand the potential adverse public health effects and environmental impacts of radiological sabotage directed at or associated with the transport and storage of civilian nuclear material or other civilian radioactive materials. The Parties may also undertake collaborative research and development in other areas of the physical protection of civilian nuclear materials or other radioactive materials. Since 2000, the IWGSTSC has conducted an extensive test programme for the assessment of the aerosol source term produced in the case of spent fuel transport sabotage by a high energy density device, after having examined several scenarios. The major goal of this programme is to produce an accurate estimate of the so called spent fuel ratio in the domain of respirable, aerosol particles produced. All the reports prepared by Sandia National Laboratories have precisely emphasised the important efforts they have made from the beginning and the amount of work already accomplished. In parallel, the International Atomic Energy Agency (IAEA), assisted by technical experts from different countries, has provided a draft document promised to become guidance for the security of radioactive or nuclear materials during transport. The IAEA document contains general guidance addressed to anyone who intends to implement or improve the security of material transports, but the text is, as of today, limited to rather general recommendations. Based on all the knowledge accumulated from past experiments and also based on the work carried out in Vienna at the IAEA, the IWGSTSC members have decided to work on the development of a method for the evaluation of the vulnerability and the source term. So for doing that, joint projects for the research, development, testing and evaluation of the consequences of the malevolent actions during transport are being pursued and are described in this paper.     

200MW供热堆乏燃料的贮存

本文结合２００ＭＷ供热堆的结构特点，介绍了供堆的乏燃料贮存方式，并讨论这一贮存方式所带来的优点。     

Safety Concerning Transportation of Spent Fuel

The Central Research Institute of Electric Power Industry (CRIEPI) has been conducting, under contract with the Science and Technology Agency of Japan, the spent fuel transport cask reliability demonstration test since 1977 to verify the safety and reliability of spent fuel transport casks. The first phase of this test was completed in 1987.

In this demonstration test, both 50 t and 100 t class of casks, designed and manufactured by current techniques, were subjected to tests to verify the integrity and adequacy of the design and manufacturing techniques through observation of behavior of the cask under test conditions. The casks were subjected to tests under normal conditions and under the accident conditions specified in the Japanese regulations and the IAEA regulations, and also to pressure tests, which were performed from the viewpoint of safety in shipping, although by sea, this is not specified in the Japanese regulations.

From the test results, it was confirmed that the 1001 class cask maintained its integrity and characteristics in conformity with regulations even after accident condition tests. It is clear that the design concept and manufacturing procedure employed for this cask is adequate.     

乏燃料贮存格架吸收体布置方案对临界安全的影响研究

基于燃耗信任制的方法,采用APOLLO和MCNP5程序对格架中可溶硼浓度对中子有效增殖因子(keff)的影响进行研究,并对中子毒物类型和布置方式对keff的影响进行了分析。结果表明:格架中可溶硼的浓度变化引起keff变化的速率随着富集度升高而变慢,近似线性变化。格架内中子毒物间的互相干涉效应是影响其毒物价值的主要原因,中子毒物价值与硼不锈钢(BSS)板间距呈线性关系。根据乏燃料组件外中子能谱的分布改进中子毒物的布置方案,可以提高乏燃料贮存系统的临界安全性和经济性。     

Actinide-Only Burnup Credit for Spent Fuel Transport

Abstract

A conservative methodology is described that would allow taking credit for burnup in the criticality safety analysis of spent nuclear fuel packages. Requirements for its implementation include isotopic and criticality validation, generation of package loading criteria using limiting parameters, and assembly burnup verification by measurement. The method allows credit for the changes in the ²³⁴U, ²³⁵U, ²³⁶U, ²³⁸U, ²³⁸Pu,²³⁹Pu,²⁴⁰Pu,²⁴¹Pu,²⁴²Pu,and ²⁴¹Am concentrations with burnup. No credit for fission product neutron absorbers is taken. Analyses are included regarding the methodology's financial benefits and conservative margin. It is estimated that the proposed actinide-only burnup credit methodology would save 20% of the transport costs. Nevertheless, the methodology includes a substantial margin. Conservatism due to the isotopic correction factors, limiting modelling parameters, limiting axial profiles and exclusion of the fission products ranges from 10 to 25% k.     

Small Gas Leakage Rate Measuring and Monitoring System for Spent Fuel Transport/Storage Cask

A containment function of transport and/or storage casks of radioactive materials is essential to prevent the materials from being released excessively into the environment. It is not practical for containment tests to measure directly the radioactivity release so that gas volumetric leakage rates are usually assessed and gas pressure decrease or increase method is usually applied. As gas flow model for evaluation, the ISO standards has deleted the concept of choked flow which is adopted by ANSI N14.5. Provided that the choked flow is not adopted to the leakage rate evaluation, the criteria of the test should be severer, and a new leakage rate measuring system with high accuracy and reasonable measuring time is required. Transport casks are often inspected in a temporary cask-storage facility where simultaneous measurement of many casks is required. In a storage cask system, multiple casks are monitored on their containment function during a storage period, and the method for simultaneous monitoring at many points for long term is required. In this study, two kinds of small gas leakage rate measuring systems are developed. One is to measure gas leakage rates directly and is called “flow measuring system”, which can measure gas leakage rate of 10^?4 to 10^?2 cm³/s with high accuracy and short measuring time. The other is to measure the pressure decreasing rate and is called “pressure decreasing rate measuring system”, which can monitor the pressure change at many points simultaneously.     

田湾核电站乏燃料干式贮存临界安全计算研究

以CASTOR 1000/19干式贮存容器装载田湾核电站六角形乏燃料组件为例,研究六角形乏燃料干式贮存的临界安全问题。基于新燃料假设,应用MONK9A程序对贮存容器满装载乏燃料进行不同工况下k_eff的计算。计算结果表明：正常工况下,k_eff远小于临界安全限值,是临界安全的;事故工况下,当²³⁵U富集度大于3.15%时,系统存在临界安全风险,须减少乏燃料装载量来确保临界安全。考虑燃耗信任制后,采用相同的模型计算得出贮存容器满装载的参考装载曲线,按此曲线要求装载能确保所有工况下的系统临界安全。采用燃耗信任制技术提高了贮存容器的利用率。该研究可为田湾核电站采用乏燃料干式贮存方案提供依据。     

核燃料工艺运输系统水下自动定位装置研究

针对核燃料工艺运输系统人工标定效率低、精度差的不足,对一种应用于水下燃料格架自动定位的装置进行了研究。阐述了系统构成及原理,设计实现了视频引导初步定位、基于STM32单片机与电感式接近传感器完成偏差测量与自动调整的精确定位功能。并对包括系统的软硬件设计、自动控制策略及初始参数标定进行论述。经现场样机试验表明,该装置可显著提高定位精度并缩短系统调试周期,还可应用于核电站在役运行期间的系统维修后再定位及定期校准。