Safety Margins of Spent Fuel Transport and Storage Casks Considering Aircraft Crash Impacts

Abstract

The safety of spent fuel transport casks in severe accident conditions is always a matter of concern. This paper surveys German missile impact tests that have been carried out in the past to demonstrate that German cask designs for transport and interim storage are safe even under conditions of an aircraft crash impact. A fire test with a cask beside an exploding propane vessel and temperature calculations concerning prolonged fires also show that the casks have reasonably good safety margins in thermal accidents beyond regulatory fire test conditions.     

Moving the Largest Capacity PWR Dual-Purpose Cask in the World from Gösgen NPP to the Zwilag Interim Storage Site

Abstract

The Swiss Gösgen nuclear power plant (NPP) has decided to use two different methods for the disposal of its spent fuel. (1) To reprocess some of its spent fuel in dedicated facilities. Some of the vitrified waste from the reprocessing plant will be shipped back to Switzerland using the new COGEMA Logistics, TN81 cask. (2) To ship the other part of its spent fuel to the central interim storage facility at Zwilag (Switzerland) using a COGEMA Logistics dual-purpose TN24G cask. The TN24G is the heaviest and largest dual-purpose cask manufactured so far by COGEMA Logistics in Europe. It is intended for the transport and storage of 37 pressurised water-reactor (PWR) spent fuel assemblies. Four casks were delivered by COGEMA Logistics to Gösgen NPP. Three transports of loaded TN24G casks between Gösgen and Zwilag were successfully pelformed at the beginning of 2002 using the new COGEMA Logistics Q76 wagon specifically designed to transport heavy casks. This article describes the procedure of operations and shipments for the first TN24G casks up to storage at Zwilag. The fourth shipment of loaded TN24G was due to take place in October 2002. The TN24G cask, as part of the TN24 cask family, proved to be a very efficient solution for Kemkraftwerk Gösgen spent fuel management.     

German approach and feedback on experience of transportability of SNF packages after interim storage

In Germany, the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently revised ‘Guidelines for dry interim storage of irradiated fuel assemblies and heat-generating radioactive waste in casks’ by the German Waste management Commission. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties, which satisfy the proofs for the compliance of the safety objectives at that time. In recent years, the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependent system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report, including evaluation of long term behaviour of components and specific operating procedures of the package. The present research and knowledge concerning the long term behaviour of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behaviour of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period. Assessment methods for the material compatibility, the behaviour of fuel assemblies and the aging behaviour of shielding parts are issues as well. This paper describes the state of the art technology in Germany, explains recent experience on transport preparation after interim storage and points out arising prospective challenges.     

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.     

Radiological aspects and behavior of spent fuel considering long-term storage

For spent nuclear fuel management in Germany, the concept of dry interim storage in dual purpose casks before direct disposal is applied. Current operation licenses for storage facilities have been granted for a storage time of 40 years. Due to the current delay in site selection, an extension of the storage time seems inevitable. In consideration of this issue, GRS performed burnup calculations, thermal and mechanical analyses as well as particle transport and shielding calculations for UO₂ and MOX fuels stored in a cask to investigate long-term behavior of the spent fuel related parameters and the radiological consequences. It is shown that at the beginning of the dry storage period, cladding hoop stress levels sufficient to cause hydride reorientation could be present in fuel rods with a burnup higher than 55 GWd/tHM. The long-term behavior of the cladding temperatures indicates the possibility of reaching the ductile-to-brittle transition temperature during extended storage scenarios. Surface dose rates are 3 times higher when a cask is partially loaded with 4 MOX fuel assemblies. Due to radioactive decay, long-term storage will have a positive impact on the radiological environment around the cask.     

Safety Analysis of Casks Under Extreme Impact Conditions

Abstract

The determination of the inherent safety of casks under extreme impact conditions has been of increasing interest since the terrorist attacks of 11 September 2001. For nearly three decades BAM has been investigating cask safety under severe accident conditionslike drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. One of the most critical scenarios for a cask is the centric impact of a dynamic load onto the lid-seal system. This can be caused, for example, by a direct aircraft crash (or just its engine) as well as by an impact due to thecollapse of a building, e.g. a nuclear facility storage hall. In this context BAM is developing methods to calculate the deformation of cask components and — with respect to leak-tightness — relative displacements between the metallic seals and their counterparts. This paper presents reflections on modelling of cask structures for finite-element analyses and discusses calculated results of stresses and deformations. Another important aspect is the behaviour of a cask under a lateral impact by aircraft or fragments of a building. Examples of the kinetic reaction (cask acceleration due to the fragments, subsequent contact with neighbouring structures like the ground, buildings or casks) are shown and discussed in correlation to cask stresses which are to be expected.     

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.     

Safety of transport of radioactive Material: 2004 in retrospect

Abstract

There are basically two main technologies for the intermediate storage of spent nuclear fuel in Europe: dry storage in casks or vaults and wet storage in pools. The advantage of casks is their modularity and hence investment can be phased to suit the planned dates of loading individual casks, pools and vaults usually provide longer term capacity and thus require a greater initial investment for operators. Transnucléaire has developed a range of modular dry cask solutions for customers and more than 100 examples of the TN 24 type cask have been licensed for transport and storage in Belgium, Switzerland, Italy, Germany, the United States of America and Japan. This paper compares the requirements for cask licensing in Europe and the USA and shows how two particular BWR cask designs were developed by Transnucléaire. (1) The TN 97 L cask was designed primarily for the European market and the first use is foreseen at the Leibstadt nuclear power station in Switzerland. (2) The TN 68 cask was designed by Transnuclear Inc. and its first use is foreseen at the Philadelphia Electric Company's Peach Bottom Atomic Power Station.     

Development of new transportable storage casks for interim dry storage

Abstract

We have started a programme to design a new type of transportable storage cask (Hitz casks) for both boiling water reactor (BWR) and pressurised water reactor (PWR) fuels for use in the new interim dry spent fuel storage project in Japan. The basic policy of this development is to use proven technology to realize a safe and cost-effective design with a high transport and storage capacity and a low fabrication cost. Since it is not permissible to change the lid gaskets at the storage facility, the double-lid system is designed to be able to use double metallic gaskets as the containment boundary for transport after the storage period; this is one of the new design features used in the casks. With the basket design we tried to achieve a capacity of 69 fuel assemblies for BWR fuel and 26 fuel assemblies for PWR fuel. Further details about these and other topics are discussed.     

Transport of MOX fuel: continuous challenge

Abstract

For 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the spent nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. This paper focuses on the casks used to transport the fresh and used mix oxide (MOX) fuel. To transport the fresh MOX boiling water reactor and pressurised water reactors fuel, TN International has developed two designs of casks: the MX 6 and the MX 8. These casks are and have been used to transport MOX fuel for French, German, Swiss and in a near future Japanese nuclear power plants. A complete set of baskets have been developed to optimise the loading in terms of integrated dose and also of course capacity. Mixed oxide used fuel has now its dedicated cask: the TN 112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in September 2008 in the Electricité de France nuclear power plant of Saint-Laurent-des-Eaux. By its continuous involvement in the nuclear transportation field, TN International has been able to face the many challenges linked to the radioactive materials transportation especially talking of MOX fuel. TN International will also have to face the increasing demand linked to the nuclear renaissance.     

Adaptation of Approved Transport and Storage Cask Lines to Individual Customer Requirements

Abstract

Within the decommissioning programmes of the Italian nuclear power plants, the Italian multi-utility company ENEL decided to rely on on-site dry storage while waiting for the availability of the national interim storage site. SOGIN (Società Gestione Impianti Nucleari SpA, Rome, Italy), now in charge of all nuclear power plant (NPP) decommissioning activities was created in the ENEL group but is now owned by the Italian government. In 2000 it ordered 30 CASTOR® casks for the storage of its spent fuel not covered by existing or future reprocessing contracts. Ten CASTOR X/A17 casks will contain the Trino pressurised water reactor (PWR) fuel and the Garigliano boiling water reactor (BWR) fuel currently stored in pools at the nuclear power plant Trino and the Avogadro nuclear facility at Saluggia. Additionally 20 CASTOR X/B52 casks will contain the BWR fuel assemblies, which are stored in the pool at the Caorso nuclear power plant. GNB (Gesellschaft fuer Nuklear-Behaelter mbH, Essen, Germany) has completed detailed studies for the design of both types of cask. The tailored cask design is based on the well-established and proven design features of CASTOR reference casks and is responsive to the needs and requirements of the Italian fuel and handling conditions. The design of the CASTOR X/A17 for up to 17 Trino PWR fuel assemblies or 17 Garigliano BWR fuel assemblies and the CASTOR X/B52 cask holding up to 52 Caorso BWR fuel assemblies is suitable for the following conditions of use: loading of the casks in the fuel pools of the nuclear installations at Trino, Caorso and Avogadro; no upgrading of the Current on-site crane capacities; transport of the fuel assemblies, which are currently stored at the Saluggia facility to the nuclear power plant Trino; on-site storage in a vertical or horizontal position with the possibility of transfer to another temporary storage or a final repository, even after a number of years; the partial loading of mixed oxide (MOX) and failed fuel; loading and drying of bottled Garigliano fuel assemblies. On the basis of the CASTOR V/19 and CASTOR V/52 cask lines, the design of the CASTOR X/A17 and X/B52 casks aims at optimising safety and economics under the given boundary conditions. The long time for which fuel is kept in intermediate wet storage results in a reduced shielding and thermal-conduction requirement. This is used to meet the tight mass and geometry restrictions while allowing for the largest cask capacity possible.     

Drop and fire testing of spent fuel and HLW transport casks at ‘BAM Test Site Technical Safety’

Abstract

BAM, as a competent German government institute for the mechanical and thermal testing of radioactive material transport and storage containers, operates unique drop and fire test facilities for experimental investigations on the open air BAM Test Site Technical Safety. To be able to perform even drop tests with full scale spent fuel or HAW casks (i.e. the German CASTOR cask designs), BAM constructed in 2004 a large drop test facility capable to handle 200 ton test objects, and to drop them onto a steel plate covered unyielding target with a mass of nearly 2600 ton. Drop test campaigns of the 181 ton GNS CONSTOR V/TC, the 129 ton MHI MSF-69BG and a 1∶2 scale model of the GNS CASTOR HAW28M (CASTOR HAW/TB2) have been performed since then. The experimental BAM drop testing activities can be supported also by drop testing of smaller packages (up to 2 ton) in an in-house test facility and by dynamic, guided impact testing of package components and material specimen inside a new drop test machine. In May 2008, a new modern fire test facility was put into operation. The facility provides two test stands fired with liquid propane. Testing in every case has to be completed by computational investigations, where BAM operates appropriate finite element modelling on appropriate computer codes, e.g. ABAQUS, LS-DYNA, ANSYS and other analytical tools.     

Description of the German Concept of Final Storage Cask

Abstract

The German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t^?1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.     

Methodological aspects for finite element modelling of lid systems for Type B(U) transport packages

Abstract

The regulatory compliance of the containment system is of essential importance for the assessment process of Type B(U) transport packages. The requirements of the International Atomic Energy Agency safety standards for transport conditions imply high loading on the containment system. The integrity of the containment system has to be ensured in mechanical and thermal tests. The containment system of German spent nuclear fuel and high level waste transport packages usually includes bolted lids with metal gaskets. The finite element (FE) method is recommended for the analysis of lid systems according to the guideline BAM-GGR 012 for the assessment of bolted lid and trunnion systems. The FE analyses provide more accurate and detailed information about loading and deformation of such kind of structures. The results allow the strength assessment of the lid and bolts as well as the evaluation of relative displacements between the lid and the cask body in the area of the gasket groove. This paper discusses aspects concerning FE simulation of lid systems for type B(U) packages for the transport of spent nuclear fuel and high level waste. The work is based on the experiences of the BAM Federal Institute for Materials Research and Testing as the German competent authority for the mechanical design assessment of such kind of packages. The issues considered include modelling strategies, analysis techniques and interpretation of results. A particular focus of this paper is on the evaluation of the results with regard to FE accuracy, influence of the FE contact formulation and FE modelling techniques to take the metallic gasket into account.     

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.     

Mobile system for the management of fuel assemblies from research reactors

Interim storage in transport and storage casks of the CASTOR type, and later the final storage of these casks are planned for the management of spent fuel assemblies from German research reactors.A mobile transfer unit is used for loading the casks with fuel assemblies on the reactor sites. Key components of the mobile transfer unit are a transfer cask, the recharging lock, and an air-cushion transport system. By means of the air-cushion transport system, the whole equipment, as well as the CASTOR casks, is transported into the reactor building. Thus, handling of the 16 t CASTOR casks is possible even on reactor sites within sufficient crane capacity. A 20 ft container accommodates the mobile transfer unit and all accessories so that the whole equipment can be transported to the reactor sites by truck.     

Assessment of integrity of a dual-purpose metal cask after long-term interim storage: seal performance under transport conditions

Abstract

Spent fuels generated in nuclear power plants (NPPs) must be stored until they can be reprocessed into new energy sources in Japan. The quantity of spent fuel stored at each NPP site is increasing, and early realisation of a method of interim storage is expected. Dual-purpose metal casks will be used which will not be reopened until they are delivered to a reprocessing plant, in order to minimize radiation exposure to personnel. The Japan Nuclear Energy Safety Organization (JNES) was established on 1 October 2003 with the mission of ensuring public safety from the potential hazards of nuclear energy. JNES have also been conducting a study of dry storage technology for interim storage. The study of verification of metal cask storage technology, the results of which from a storage and subsequent transport safety point of view are presented in this paper, was originally initiated with government funds in 1999 at the Nuclear Power Engineering Corporation, before being transferred to and conducted by JNES up to the end of 2003.     

Fracture Toughness of DCI Casks and Prediction by Small Specimen Test

Abstract

Spent nuclear fuel transport and/or storage containers (casks) must maintain their structural integrity even when subjected to hypothetical accidents during transport or handling accidents at storage facilities. For ductile cast iron (DCI) to be used as a cask containment boundary material, adequate fracture toughness must be demonstrated at service temperatures and Impact loading conditions of concern. In Japan, comprehensive studies of the fracture toughness of heavy section DCI have been undertaken by a number of research organisations to provide the safety assurance for the DCI casks. In the present study, the fracture toughness data were used to develop a lower bound trend curve for heavy section DCI and to examine the prediction methods by small specimen tests. The fracture toughnesses K_Ic, K_IIc and K_IIIc were also obtained to study the safety assessment of DCI casks under different loading mode conditions.