IP-2 package containing plutonium contaminated slag pots

Abstract

Croft Associates were approached by PC Richardson to carry out the licensing of an industrial package, IP-2 package containing plutonium contaminated slag pots. This project provided several unique challenges for all those involved. The slag pots belonged to Outokumpu a steel company based in Sheffield. During use, sampling indicated a high level of plutonium in the form of Pu-238 was present in a slag pot. Further sampling identified that four slag pots were contaminated with Pu-238 and required disposal. The source of the Pu-238 contamination was believed to be a single heart pacemaker present in scrap that was melted down for reuse. All four slag pots weighed between 50 and 65 t and three of the pots had large patched cracks. The size and nature of the contamination prevented the slag pots being broken up therefore they had to be packaged and transported intact. REMAC designed a steel box to package the slag pots. It was the responsibility of Croft to license this package. Several challenges were faced when licensing: first, the weight, size and contents prevented any physical drop testing of the package; second, the box was to be assembled around the slag pot, limiting the leak testing that could be carried out. All of the evaluation of the ability of the package to withstand the regulatory drop tests was therefore carried out by finite element analysis by AMEC NNC. This approach was also used to check the tie down and lifting systems. Leak testing was carried out via the soap bubble method on the assembled box before grouting. The transportation box, once assembled and licensed as an IP-2 package, was transported to Drigg from Sheffield by road.     

The French Nuclear Safety Authority's Experience with Radioactive Transport Inspection

Abstract

About 300,000 radioactive material packages are transported annually in France. Most consist of radioisotopes for medical, pharmaceutical or industrial use, but the nuclear industry deals with the transport of fuel cycle materials (uranium, fuel assemblies, etc.) andwaste from power plants, reprocessing plants and research centres. France is also a transit country for shipments such as spent fuel packages from Switzerland or Germany, which are bound for Sellafield in the United Kingdom. The French nuclear safety authority(DGSNR, Directorate General for Nuclear Safety and Radioprotection) has since 1997 been responsible for the safety of radioactive material transport. This paper presents DGNSR's experience with transport inspection: a feedback of key points based on 300 inspections achieved during the past 5 years is given.     

Pilot study on return of reprocessed wastes from Dounreay,Scotland to Germany

Abstract

The German Federal Ministry of Education and Research (BMBF) assigned DBE Technology GmbH with a project to review the prerequisites and contractual boundary conditions for the return of cemented residues from the reprocessing plant at Dounreay to Germany. For this purpose, the bilateral contracts between the German research facilities and the operator of the reprocessing plant at Dounreay, the UK Atomic Energy Authority (UKAEA), were examined. Possible interim storage sites in Germany were sought, flasks suitable for transport and casks suitable for interim storage and final disposal were researched, and transportation options were explored. Based on the results of theses investigations, strategies for the return of the drums containing cemented residues were developed, including time and effort estimates.     

New generation standard nuclear material container

Abstract

Nuclear Filter Technology (NucFil) is working with the Los Alamos National Laboratory (LANL) to design a nuclear material storage container that complies fully with the requirements of DOE M 441·1-1. LANL provided NucFil with a specification that outlines requirements to comply with the manual, as well as to satisfy specific needs of their own. NucFil has taken this specification and designed a container known as the new generation standard nuclear material container (NG SNMC). The premise of the design is a simple, robust container that is easy to use. The sealing mechanism is a single large cross-section, low durometer o-ring. The large cross-section provides a tight seal that has enough elastic rebound to compensate for any distortion of the sealing faces after a potential drop. The low durometer keeps the force required to open and close the container low. Once compressed, the seal is kept in place by a bayonet style closure that is locked in place by a positive mechanical engagement. The components of the container exposed to the load are manufactured of corrosion resistant 316L stainless steel. The container has a filter made of a heat resistant ceramic fibre to retain particles after a fire, and a water resistant membrane to keep moisture out of the container. Pewter shielding can be attached and is latched in place. These features are present in all seven sizes of the NG SNMCs, including 1, 3, 5, 8 and 12 quart and 5 and 10 gallon.     

Package plutonium metal and oxide in flowformed canisters

Abstract

Plutonium metal and oxide are typically handled and processed in isolated enclosures (gloveboxes) and then transferred to storage packing systems. The United States Department of Energy (DOE) outlines general requirements relevant to the planned DOE activities to stabilise and package plutonium materials in containers, and place them in long term storage. Inner and outer containers have been qualified as the primary barrier to the environment during storage of the plutonium materials. The design and fabrication of the containers are consistent across the DOE complex. The flowform manufacturing process has been used for over 20 years to fabricate the weld free, precise containers for several DOE sites meeting ASME section III code, subsection NCA-3851.2 (a)(1).     

Ten years of transport of vitrified high level waste (HLW) from Areva La Hague

Abstract

Areva has extensive experience in reprocessing irradiated fuel from both gas cooled and water cooled reactors and has legal obligations to return residues to respective foreign owners, usually in the form of vitrified waste. Such returns are subject to appropriate quality assurance controls. Appropriate transport flasks have been designed and certified. Transport operations are routinely conducted to Belgium, Germany, Japan, Switzerland and the Netherlands. The present paper reviews the experiences over a ten year period from 1995 to 2005.     

Mechanical design assessment approaches of actual spent fuel and HLW transport package designs

Abstract

In recent years, BAM Federal Institute for Materials Research and Testing finalised the competent authority assessment of the mechanical and thermal package design in several German approval procedures of new spent fuel and high level waste package designs. The combination of computational methods and experimental investigations in conjunction with materials and cask components testing is the most common approach to mechanical safety assessment. The methodology in the field of safety analysis, including associated assessment criteria and procedures, has evolved rapidly over the last years. The design safety analysis must be based on a clear and comprehensive safety evaluation concept, including defined assessment criteria and constructional safety goals. In general, for new package designs, the implementation of experimental package drop tests in the approval process should be obligatory. Additionally, pre- and post-test calculations as well as components or material testing could be important. The extent to which drop tests are necessary depends on the individual package construction, the materials used and identified safety margins in the design.     

An overview of 20 years' Romanian experience on packaging and components testing according to national and international regulations

Abstract

More than 20 years ago, the Institute for Nuclear Research (INR) Pitesti in Romania, through its Reliability and Testing Laboratory, was licensed by the Romanian Nuclear Regulatory Body – CNCAN to carry out qualification tests for packages intended for the transport and storage of radioactive materials. The radioactive material is placed in packaging which are designed in accordance with national and the International Atomic Energy Agency's (IAEA's) Regulations for safe transport to the disposal centre. A broad range of verification and certification tests are performed at INR on radioactive material packages or component sections, such as packages used for the transport of radioactive sources to be used for industrial or medical purposes. This paper describes some of the various tests, which have been performed, and how they relate to normal conditions and minor mishaps during transport. Quality assurance and quality control measures taken in order to meet technical specification provided by design there are also presented.     

Catalytic mitigation of hydrogen risk during wet transportation of radioactive materials

Abstract

Major challenges in the area of wet transportation of radioactive materials are reliability and safety of transportation casks. In most cases, the bottom part of the cask is filled with water whereas a gaseous mixture is contained in the upper part. In such a configuration, water radiolysis leads to the formation of hydrogen and oxygen, which continuously enriches the gaseous mixture. Among the functions to be satisfied, wet transportation systems shall thus allow the control of the hydrogen content below its flammability limit. This is currently achieved by limiting the transportation duration so as to reopen the cask before the critical hydrogen concentration is reached. Development of new technologies that would mitigate the hydrogen risk is all the more motivated because it would allow an extension of the transportation duration. AREVA-TN International and the Institut de recherches sur la catalyse et l'environnement de Lyon have developed a catalytic system which aims at buffering the hydrogen concentration far below the flammability limit. The principle of this catalyst is to recombine the hydrogen with the oxygen formed by water radiolysis. The present paper gives an overview of the development of this catalytic recombining system. It describes the laboratory qualification tests undertaken for the evaluation of the recombining efficiency. Particular attention is placed on the recombining efficiency after immersion of the catalyst in borated water, which would occur in a nuclear reactor pool during loading of used fuel. Laboratory investigations, carried out in an autoclave simulating a transportation cask, showed that, after immersion in borated water, the catalytic system allows the recombination of 3% hydrogen in less than 24 h at temperatures as low as 35°C.     

Transportation safety risk for source recovery versus consequence of leaving radioactive sources in place and vulnerable due to limited transport options or denial of shipment

Abstract

Potential risks associated with transportation safety of recovered radioactive sources in normal commerce are rhetorically compared to the latent risk of not recovering disused radioactive sources due to limited transport options or outright denial of shipment. It is essential, during each phase of the recovery process, to ensure secure, timely, cost effective and reliable means to return vulnerable radioactive sources to safe and protected locations by land, sea and/or air transport. In some cases, only limited transport options exist or denials of shipment may occur that impede the recovery process. Risks associated with normal transportation of recovered sources are considered less significant than the risks related to leaving disused radioactive sources at their original location.     

Specific transport and storage solutions for residues return programmes

Abstract

The treatment of used nuclear fuel, performed at AREVA's La Hague plant, allows recovering uranium 95% and plutonium 1% for recycling, the remaining 4% being considered as ultimate waste that can be sorted into two categories: high level activity waste (HLW) which is vitrified, and long-lived intermediate level waste (ILW) composed of structural elements of used nuclear fuel which is compacted. Whether vitrified or compacted, the waste is conditioned in the same universal and multipurpose container, named the Universal Canister. The resulting residue is named CSD-V for vitrified waste and CSD-C for compacted waste; they both remain property of the utilities and must be returned to countries of origin. In order to transport Universal Canisters in the best technical and economical conditions, TN International designs two kinds of cask solutions for its customers, either for transport only or for dual purpose, storage and transport, depending on the facility. Since the mid-1990s, TN International has transported CSD-V residues to Belgium, the Netherlands, Switzerland, Germany and Japan and is now starting the CSD-C return program. The purpose of this paper is to explain how the experience gained during the CSD-V return program has been used to optimize the CSD-C return program, in terms of cask design and licensing and of transport logistics. In some cases, casks initially developed for CSD-V transports have been adapted and in other cases, new casks are being designed specifically for CSD-C transport to increase the cask capacity and reduce the number of shipments.     

Lessons learned from the West Valley spent nuclear fuel shipment within the United States

Abstract

This paper describes the lessons learned from the US Department of Energy (DOE) transport of 125 DOE-owned commercial spent nuclear fuel (SNF) assemblies by railway from the West Valley Demonstration Project to the Idaho National Laboratory (INL). On 17 July 2003, the DOE made the largest single shipment of commercial SNF in the history of the United States. This was a highly visible and political shipment that used two specially designed Type B transport and storage casks. This paper describes the background and history of the shipment. It discusses the technical challenges for licensing Type B packages for hauling large quantities of SNF, including the unique design features, testing and analysis. This paper also discusses the pre-shipment planning, preparations, coordination, route evaluation and selection, carrier selection and negotiations, security, inspections, tracking and interim storage at the INL.     

BWR spent fuel transport and storage with TN9/4 and TN24BH casks

Abstract

In 2001 the Swiss nuclear utilities started to store spent fuel in dry metallic dual purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd, as the owner of the Mühleberg nuclear power plant, is involved in this process and has selected to store the spent fuel in a new high capacity dual purpose cask, the TN24BH. For the transport Cogema Logistics has developed a new medium size cask, the TN9/4, to replace the NTL9 cask, which has performed numerous shipments of BWR spent fuel in past decades. Licensed by the IAEA 1996, the TN9/4 is a 40 t transport cask, for seven BWR high burnup spent fuel assemblies. The spent fuel assemblies can be transferred to the ZWILAG hot cell in the TN24BH cask. These casks were first used in 2003. Ten TN9/4 shipments were made, and one TN24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN24BH high capacity dual purpose cask and the TN9/4 transport cask and describe in detail their characteristics and possibilities.     

Spent fuel transport using cask loaded outside pool

Abstract

During the last year, Sogin (the Italian company in charge for decommissioning of Italian nuclear power plants) had to implement an accelerated decommissioning plan of a EUREX spent fuel pool due to finding a water leakage into the environment from the pool. EUREX is no longer operating a pilot reprocessing plant, which some years ago became the responsibility of Sogin. There were 52 spent fuel assemblies from the Trino Vercellese PWR nuclear power plant, 48 irradiated pins from a Garigliano BWR fuel assembly, and 10 plates from an irradiated MTR fuel assembly stored in the EUREX pool, so the first step of the accelerated decommissioning plan consisted in the evacuation of this spent fuel. Considering the necessity to start the evacuation as soon as possible, Sogin decided to use an already existing cask (AGN-1) used in the past for the transport of Trino and Garigliano fuel assemblies. This cask was requalified in order to obtain a transport licence for the fuel assemblies stored in the EUREX pool according to ADR 2005 regulation. The transport license for the AGN-1 cask loaded with EUREX fuel assemblies was released by APAT (the Italian Safety Authority) in the spring of 2007. Owing to the limited capacity of the EUREX pool crane (27 t for nuclear loads) and limited dimensions of pool operational area, it was not possible to transfer the AGN-1 cask (50 t) into the pool for fuel assemblies charging. The solution implemented to overcome this problem was the loading of the cask outside the pool. A special shielding shuttle was developed and used to allow safe spent fuel transfer between the pool and the cask. This procedure avoided also the problem of excessive contamination of cask surfaces that could have occurred due to very high level of contamination of EUREX pool water if the cask had been immersed in the pool. Additional shielding devices were developed and used to reduce dose rate during cask loading operations. Although the evacuation of spent fuel assemblies from the EUREX pool was a very challenging activity due to the short time available, unfavourable space conditions inside the pool building and handling tool limitations; all loading and transport operations were performed successfully and without particular problems. Ten transports were carried out to evacuate all of the spent fuel stored in the EUREX pool. Spent fuel was transferred to the Avogadro Deposit pool. The first loading sequence started on 2 May 2007 and the first transport was performed on 6 May 2007. The tenth and last transport was performed on 21 July 2007. A dose less than 50 μSv (neutron + gamma) was measured for the most exposed operator during a complete cask loading sequence.     

Guidance for Package Approvals in the United Kingdom

Abstract

Approval is required under the transport regulations for a wide range of package designs and operations, and applications for competent authority approval and validation are received from many sources, both in the UK and overseas. To assist package designers and applicants for approval, and to promote consistency in applications and their assessment, the UK Department for Transport issues guidance on the interpretation of the transport regulations and the requirements of an application for approval and its supporting safety case.The general guidance document, known as the Guide to an Application for UK Competent Authority Approval of Radioactive Material in Transport, has been issued for many years and updated to encompass the provisions of each successive edition of the IAEA transport regulations. The guide has been referred to in a number of international fora, including PATRAM, and was cited as a 'good practice' in the report of the IAEA TRANSAS appraisal of the UK in 2002. Specialist guides include the Guide to the Suitability of Elastomeric Seal Materials, and the Guide to the Approval of Freight Containers as Types IP-2 and IP-3 Packages. This paper discusses the guidance material and summarises the administrative and technical information required in support of applications for approval of package designs, special form and low-dispersible radioactive materials, shipments, special arrangements, modifications and validations.     

Evaluation of safety of French type B package designs in severe accident environments other than regulatory

Abstract

Packages used to transport radioactive materials in France must be designed to meet the safety performance requirements when subject to the test conditions set forth in the International Atomic Energy Agency (IAEA) Regulations. During actual use, the packages may be subject to quite different accident conditions. The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) has evaluated the behaviour of typical packages designed to transport spent fuel, high activity waste, fresh mixed oxide (MOX) fuel and plutonium oxide powder under realistic conditions of mechanical impact and fire. The studied designs remain safe after impact onto targets present in the real environment of transport. The energy absorption by the package ancillary equipment (transport frame) compensates for the kinetic energy increase by comparison to the energy expended during the regulatory tests. New software was developed to correctly simulate the thermal behaviour of the neutron shielding materials. The studied package designs exhibit large margins of safety concerning resistance to fire. The results obtained have been used to develop tools in support of the appraisal of emergency situations.     

停堆氧化运行中主回路活化腐蚀产物的迁移与控制

核电站主回路中腐蚀产物的氧化净化是核电站大修的源项控制中最重要的一环.本文综合分析氧化净化过程中pH值、H2O2和溶解氢浓度对主回路活化腐蚀产物迁移的影响,结合大亚湾核电站的实践经验,提出了低pH值、低H2和集中引入适量氧的控制手段.     

超大剂量6MeV电子束照射对大鼠脑Ca2+、Mg2+含量的影响

观察超大剂量6 MeV电子束照射对大鼠脑Ca2+、Mg2+含量的动态变化,探讨其在放射性脑损伤发病机制中的作用.对SD大鼠用6MeV电子线进行10Gy、20Gy和30Gy全脑单次照射,应用等离子直读光谱仪定量分析大鼠脑放射性损伤后不同时间、不同剂量脑组织中Ca2+、Mg2+含量的动态变化,用干-湿重法测定脑组织含水量.大鼠全脑受照射后,均存在脑组织中Ca2+含量升高、Mg2+含量下降和脑水肿的发生.其中20 Gy照射后24 h脑组织中Ca2+含量与对照组相比有显著升高(p＜0.05),Mg2+含量在照后7天与对照组相比有显著下降(p＜0.05).在照后7天,上述各指标变化的幅度为30Gy组＞20Gy组＞10Gy组.基于大鼠放射性脑损伤后脑组织中Ca2+、Mg2+含量发生的变化,探讨了电离辐射所致脑损伤的机理.