The BNFL Vitrified Residue Transport and Storage Flask

Abstract

A highly active liquor residue arises from the reprocessing of spent nuclear fuel. Arisings from the Sellafield reprocessing operations will be vitrified into a solid form within special stainless steel containers to enable long term safe storage. BNFL are committed to the return of these residues to their overseas customers. This has necessitated the provision of a customised transport flask design to safely transport the vitrified residues safely back to their countries of origin. Residues from a number of fuel reprocessing waste streams will be vitrified at Sellafield, both from Magnox fuel arisings and oxide fuel arisings. BNFL operating plans require that the feed to the vitrification process is in general a mixture of the various waste streams. However, customers have required that any flask design should accommodate residues containing varying mixtures of the streams ranging from a Magnox dominated mix with high gamma activity to an oxide dominated mix with high neutron activity. This requirement, together with the constant full length source strength of vitrified residue, has imposed unusual and severe constraints for the flask shielding design. A flask design has been evolved capable of transporting 21 vitrified packages whilst satisfying the stringent shielding requirements within the overall UK weight and size limitations. This has necessitated the fine optimisation between shielding requirements, heat transfer requirements and overall size and weight limitations. Liaison between the UK and France has resulted in the harmonising of certain key handling dimensions and features between the UK and French flask designs.     

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.     

Flask Tiedown Design and Experience of Monitoring Forces

Abstract

The design requirements of fuel transport flasks for containment integrity is well covered by international regulations. Less well defined are the requirements for restraint, or tiedown, as a means of securing the transport flask to its prime mover. This paper refers specifically to the NTL range of LWR fuel flasks, though many principles are applicable to a wider range of transport packages. The tiedown system is defined, and different aspects discussed in detail: the practical requirements, for example, where operation, maintenance and inspection are considered, and the need for the tiedown to harmonise with the flask so that performance is not impaired. A review of regulations and guidelines appropriate to tiedowns is included, together with a statement on their applicability. The derivation of the standards applied by NTL is described, in the context of transport by rail, sea and road. Aspects of detail design of various components of the tiedown system are described, with specific reference to the influence of this design on package performance. NTL has conducted a number of practical trials to evaluate typical values of accelerations encountered by flasks and their tiedown systems in the various transport modes. Results of these practical trials are available as input to future designs. In conclusion, the paper serves to highlight the high degree of care and consideration paid to this peripheral area of irradiated fuel transport.     

Application of Lefm Design to Nuclear Material Transport Casks

Abstract

Fracture safe design can be assured by proper application of fracture mechanics. Linear elastic fracture mechanics (LEFM) can generally be applied to a broad range of cask designs and component materials. The use of LEFM is straightforward when the linear elastic plane strain fracture toughness (K_1c) of the cask material can be directly measured. When the plane strain fracture toughness cannot be directly measured, a special form of LEFM can be used. The fracture toughness can be equivalently determined through measurement of the J_1c (elastic–plastic) fracture toughness. While this LEFM approach can only be used under specific conditions, such conditions are generally met by heavy-walled casks under severe loading conditions. The regulatory drop test, in which a subcritical flaw has been intentionally introduced into a prototype cask, can be used to demonstrate the suitability of applying LEFM design to a specific cask. This paper describes the LEFM design approach as applied to cask design for a broad class of materials. The advantages and limitations of the LEFM approach are also discussed with respect to existing regulatory acceptance criteria.     

The Elements of a Brittle Fracture Avoidance Plan for Transport Flask Body Steels

Abstract

New flask design regulations and guidelines have been introduced which require that freedom from brittle fracture must be demonstrated. This paper presents aspects of a fracture avoidance plan to meet the new requirements. Manufacturing control, non-destructive testing and materials selection are all considered and the method of assessment is discussed in detail. The problem of obtaining valid test data for very thick material is identified and a method of testing based on the ‘master curve’ is proposed.     

A Response Map for use with Nuclear Electric's Irradiated Fuel Transport Flask Emergency Plan

Abstract

A response map to be used in the initial phase of Nuclear Electric's Emergency Plan for irradiated fuel transport has been produced. The Emergency Plan itself is outlined and the role of the map in identifying the most appropriate source of health physics expertise is described. With the recent change in Alert Centre the opportunity was taken to revise this map in order to improve its usefulness and confirm its accuracy. The revised map was prepared using route planning computer software to determine the boundaries of the response area of each Power Station and other agencies involved in the Emergency Plan. The potential for future development in emergency response is discussed, examining in particular the role of route finding and mapping software, including databases of specialist materials and equipment, and in-car systems to guide those attending the emergency.     

The Application of Fracture Mechanics to The Safety Assessment of Transport Casks For Radioactive Materials

Abstract

BAM is the responsible authority in Germany for the assessment of the mechanical and thermal design safety of packages for the transport of radioactive materials. The assessment has to cover the proof of brittle fracture safety for package components made of potentially brittle materials. This paper gives a survey of the regulatory and technical requirements for such an assessment according to BAM's new 'Guidelines for the application of ductile cast iron for transport and storage casks for radioactive materials'. Based on these guidelines, higher stresses than before will be permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof using the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA's advisory material which are based on the prevention of crack initiation. Examples of BAM's research and safety assessment practices are given. Recommendations for further developments towards package designs with higher acceptable stress levels will conclude the paper.     

放射性物品运输后果评价软件CRAMTRA的程序设计

放射性物品运输后果评价软件是放射性物品运输后果评价的重要工具。我国于20世纪80年代开始陆续引进相应程序,但至今未有独立知识产权的国产软件。本文介绍了放射性物品运输后果评价软件CRAMTRA的程序设计,包括物理模型、流程图、UML图等内容,为下一步的代码开发提供基础保障。     

Reliability of the Spent Fuel Identification for Flask Loading Procedure Used by Cogema for Fuel Transport to La Hague

Abstract

The Spent Fuel Identification for Flask Loading (SFIFL) procedure designed by COGEMA is analysed and its reliability calculated. The reliability of the procedure is defined as the probability of transporting only approved fuel elements for a given number of shipments. The procedure describes a non-coherent system. A noncoherent system is the one in which two successive failures could result in a success, from the system mission point of view. A technique that describes the system with the help of its maximal cuts (states) is used for calculations. A maximal cut contains more than one failure which can split into two cuts (sub-states). Cuts splitting will enable us to analyse, in a systematic way, non-coherent systems with independent basic components.     

Application of the Transport System Concept to the Transport of LSA Waste

Abstract

Transport Regulations rely essentially on the packaging and do not take into account the contribution to safety which may be made by other features of the operation. In some situations, mainly for routine transports not fully complying with the Regulations, it would be beneficial to envisage the possibility of using a package which does not meet all the Type B requirements, complemented by additional safety measures put in place to compensate for these shortfalls. The ‘Transport System’ concept will take into account the contributions to safety from these additional measures. It will ensure that the proposed system is at least as safe as a reference operation complying fully with the Regulations. If this equivalent safety level can be properly demonstrated, the Competent Authority will provide a ‘Transport System Approval’ for well defined shipments over a specific period. Two examples are presented. In the first case, a thermally insulated ISO container is envisaged for the transport of drums containing combustible LSA material having a total activity per conveyance up to 600 A₂. In the second one, two dedicated trucks transporting conditioned waste in drums has been shielded so as to comply with the regulatory dose rate limits. These examples show the benefits of the TS concept. Nevertheless, the full requirements of the Regulations should be implemented as far as reasonably practicable, and the TS concept should be applied only to particular difficulties and is not suitable to all situations. Therefore, some general restrictions (applicable to every TS) have to be set by IAEA. Depending on the case, complementary ones may be required by the CA. Bearing in mind possible restrictions presented in this paper, the TS concept will be useful in solving some of the current problems of the transport of waste without needing a fundamental change in the Regulations.     

基于AT＆MCA的SSRF低能输运线上层应用软件设计

AT和MCA都是基于MATLAB的介于上层应用软件和底层加速器控制系统之间的加速器物理软件工具包.文章详细介绍了基于这两个软件包的SSRF低能输运线上层应用软件设计.离线及在线仿真测试以及SSRF低能输运线调束结果表明,该软件所有功能基本运行正常,软件通过EPICS与真实电源之间数据读写准确,计算结果与SSPF低能输运线各参数相符,匹配及轨道校正等功能在SS-RF低能输运线的调束工作中也起到了一定的作用.     

Application of IAEA TECDOC 717 to the Assessment of Brittle Fracture in Transport Containers with Plastic Flow Shock Absorbers

Abstract

UK Nirex is developing re-usable shielded transport containers (RSTCs) in a range of shielding thicknesses (from 70 nun to 285 nun) to transport immobilised intermediate level radioactive waste (ILW) to a deep repository. The RSTCs are being designed to meet the requirements of the IAEA Transport Regulations for Type B packages, including the requirement to maintain shielding and containment following a drop of 9 m onto an unyielding surface. The RSTCs are essentially monolithic in construction and the heaviest version weighs up to 65 tonnes when loaded with contents. They rely principally on plastic flow of the material of construction to absorb the high energies involved in impact events. Specific features of the designs, such as the solid metal comer shock absorbers and side ribs have been optimised for this purpose. 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. In this paper the methodology set out in IAEA-TECDOC-717 is applied to the Nirex RSTC, including the application of elastic plastic fracture mechanics methods.     

应用模糊综合评定法评定反应堆操纵台人因设计水平

反应堆操纵台的人因设计水平是每个设计者必须考虑的问题。本文应用模糊数学中的模糊综合评定法对岷江堆和高通量工程试验堆临界装置操纵台上的开关,按钮,表盘和指示灯排列组合的人因设计水平进行了模糊综合评定,从评定结果中发现了两操纵台各自的优缺点,并提出了一些改进建议。     

INES Scale: French Application to Radioactive Material Transport

Abstract

After gaining control of radioactive material transport in June 1997, the French Nuclear Safety Authority (ASN) decided to apply the International Nuclear Event Scale (INES scale) to transport events. The Directorate General for Nuclear Safety and Radioprotection (DGSNR) requests that radioactive material package consignors declare any event occurring during transport, and has introduced the use of the INES scale adapted to classify transport events in order to inform the public and to have feedback. The INES scale is applicable to events arising in nuclear installations associated with the civil nuclear industry andevents occurring during the transport of radioactive materials to and from them. The INES scale consists of seven levels. It is based on the successive application of threetypes of criterion (off-site impact, on-site impact and degradation of defence in depth) and uses the maximum level to determine the rating of an accident. As the transport in questiontakes place on public thoroughfares, only the off-site impact criteria and degradation ofdefence in-depth criteria apply. This paper deals with DGSNR's feedback during the past 7 years concerning the French application of the INES scale. Significant events that occurred during transport are presented. The French experience was used by the International Atomic Energy Agency (IAEA) to develop a draft guide in 2002 and the IAEA asked countries to use a new draft for a trial period in July 2004.     

小波尺度函数展开在连续能量中子输运方程计算中的应用

小波展开能够很好地拟合剧烈变化的函数,近年来已被应用于模拟中子角注量率随角度剧烈变化的问题,并取得了令人满意的结果.中子能谱在共振区具有剧烈震荡的特性,本文介绍了利用能群与小波尺度函数展开相耦合来离散连续能量中子输运方程中能量自变量的方法.对中子注量率在共振区关于能量用小波尺度函数进行拟合,而在快中子区和热中子区利用分群计算的方法.初步的数值结果表明,该方法使有效增殖系数计算精确,并能够得到中子注量率在共振区随能量的精细分布,对共振自屏蔽的精确计算具有重要意义.     

Probabilistic Safety Assessment for the Transport of Radioactive Waste to a UK Repository at Sellafield

Abstract

A study has been undertaken to provide a detailed understanding of the radiological and non-radiological risks associated with the transpott of radioactive waste from the sites at which waste is produced in the UK to a proposed deep repository at Sellafield, and to ensure that these risks meet the design targets specified by Nirex. The routine transport collective dose to members of the public was assessed to be 0.2 man.Sv per year, which is only about 0.004% of the natural background dose. Accident frequencies were calculated using event tree methodology. The radiological consequences of accidents were assessed using the probablistic computer code CONDOR. The risk expectation value was calculated to be 1.5 × 10^?5 ? 8.6 × 10^?6 latent cancer fatalities per year (depending on the transport mode scenario). These values are significantly lower than the corresponding prediciions for non-radiological accident fatality rates, 0.05 ? 0.035 fatalities per year. The radiological accident risk for the most exposed individual member of the public was assessed to be 5 × 10^?11 ? 1.7 × 10^?11 per year, very much less than the Nirex target of 5 × 10^?7 per year. Plots of societal risk were shown to lie in the region of ‘negligible risk’, as defined by the UK Health and Safety Commission for non-radioactive dangerous goods transport.     

MCNPX, MONK, and ERANOS analyses of the YALINA Booster subcritical assembly

This paper compares the numerical results obtained from various nuclear codes and nuclear data libraries with the YALINA Booster subcritical assembly (Minsk, Belarus) experimental results. This subcritical assembly was constructed to study the physics and the operation of accelerator-driven subcritical systems (ADS) for transmuting the light water reactors (LWR) spent nuclear fuel. The YALINA Booster facility has been accurately modeled, with no material homogenization, by the Monte Carlo codes MCNPX (MCNP/MCB) and MONK. The MONK geometrical model matches that of MCNPX. The assembly has also been analyzed by the deterministic code ERANOS. In addition, the differences between the effective neutron multiplication factor and the source multiplication factors have been examined by alternative calculational methodologies. The analyses include the delayed neutron fraction, prompt neutron lifetime, generation time, neutron flux profiles, and spectra in various experimental channels. The accuracy of the numerical models has been enhanced by accounting for all material impurities and the actual density of the polyethylene material used in the assembly (the latter value was obtained by dividing the total weight of the polyethylene by its volume in the numerical model). There is good agreement between the results from MONK, MCNPX, and ERANOS. The ERANOS results show small differences relative to the other results because of material homogenization and the energy and angle discretizations.The MCNPX results match the experimental measurements of the ³He(n,p) reaction rates obtained with the californium neutron source.     

放射性物质运输评价系统INTERTRAN2的移植

介绍了将放射性物质运输评价系统ＩＮＴＥＲＴＲＡＮ２软件包向微机移植过程中所做的工作,通过对源程序及数据库的修改,软件包各个软件能在微机上运行。经过多个便题的运算和计算结果的分析,证明该软件包的移植是成功的。     

Application of a Chemical ion Exchange Model to Transport Cask Surface Decontamination

Abstract

Radionuclide contamination of stainless steel surfaces occurs during submersion in a spent fuel storage pool. Subsequent release or desorption of these contaminants from a nuclear fuel transportation cask surface under varying environmental conditions occasionally results in the phenomenon known as contamination ‘weeping’. Experiments have been conducted to determine the applicability of a chemical ion exchange model to characterise the problem of cask contamination and release. Surface charge characteristics of Cr₂O₃ and stainless steel (304) powders have been measured to determine the potential for ion exchange at metal oxide-aqueous interfaces. The solubility of Co and Cs electrolytes at varying pH and the adsorption characteristics of these ions on Cr₂O₃ and stainless steel powders in aqueous slurries have been studied. Experiments show that Co ions do reversibly adsorb on these powder surfaces and, more specifically, that adsorption occurs in the nominal pH range (pH=4–6) of a boric acid moderated spent fuel pool. Desorption has been demonstrated to occur at pH≤3. Cs⁺ ions also have been shown to have an affinity for these surfaces although the reversibility of Cs⁺ bonding by H⁺ ion exchange has not been fully demonstrated. These results have significant implications for effective decontamination and coating processes used on nuclear fuel transportation casks.