Internal cask content collisions during drop test of transport casks for radioactive materials

Abstract

In transport casks for radioactive materials, significantly large axial and radial gaps between cask and internal content are often present because of certain specific geometrical dimensions of the content (e.g. spent fuel elements) or thermal reasons. The possibility of inner relative movement between content and cask will increase if the content is not fixed. During drop testing, these movements can lead to internal cask content collisions, causing significantly high loads on the cask components and the content itself. Especially in vertical drop test orientations onto a lid side of the cask, an internal collision induced by a delayed impact of the content onto the inner side of the lid can cause high stress peaks in the lid and the lid bolts with the risk of component failure as well as impairment of the leak tightness of the closure system. This paper reflects causes and effects of the phenomenon of internal impact on the basis of experimental results obtained from instrumented drop tests with transport casks and on the basis of analytical approaches. Furthermore, the paper concludes the importance of consideration of possible cask content collisions in the safety analysis of transport casks for radioactive materials under accident conditions of transport.     

Spent Fuel Transport Associated With Other Dangerous Goods in Regular Train Units — Assessment of Hypothetical Explosion Impacts

Abstract

Continental railway transport regulations (RID) do not exclude the transport of spent fuel casks in a regular train unit that also contains wagons with other hazardous materials. In the case of a train accident the release or reactions of those dangerous goods could potentially give significant accidental impacts on to the spent fuel casks. The assessment of fires from inflammable liquids and the explosion impacts from pressurised inflammable gases (like LPG) is well known from other studies which have usually revealed sufficient safety margins to the robust spent fuel cask designs. A new problem to be assessed is the potential impact from a detonation blast wave from explosives transported in the same train unit as a spent fuel cask. BAM is assessing this problem by developing a numerical model to calculate the effect of the dynamic pressure of a external shockwave on the cask construction. The calculation results show that the integrity of a robust monolithic cask with a screwed lid closure system is preserved after the effect of a 21 tonne (equivalent weight of TNT) explosive detonation in the regular transport configuration with a distance of 25 m between the centre of the explosion and the front of the cask.     

Thermomechanical Finite-element Analyses Of Bolted Cask Lid Structures

Abstract

Analysis of complex bolted cask lid structures under mechanical or thermal accident conditions is important for the evaluation of cask integrity and leak tightness in package design assessment according to the transport regulations or in aircraft crash scenarios. In this context BAM is developing methods based on finite elements (FE) to calculate the effects of mechanical impacts onto the bolted lid structures as well as effects caused by severe fire scenarios. In the case of fire it might not be enough to perform only a thermal heat transfer analysis. A complex cask design together with a severe hypothetical time–temperature curve representing an accident fire scenario will create a strong transient heating up of the cask body and its lid system. This causes relative displacements between the seals and their counterparts that can be analysed by a so-called thermomechanical calculation. Although it is currently not possible to directly correlate leakage rates with results from deformation analyses, an appropriate finite element model of the considered type of metallic lid seal has been developed. For the present it is possible to estimate the behaviour of the seal based on the calculated relative displacements at its seating and the behaviour of the lid bolts under the impact load or the temperature field, respectively. Except for the lid bolts, the geometry of the cask and the mechanical loading is axisymmetric which simplifies the analysis considerably, and a two-dimensional finite-element model with substitute lid bolts may be used. The substitute bolts are modelled as one-dimensional truss or beam elements. An advanced two-dimensional bolt submodel represents the bolts with plane stress continuum elements. This paper discusses the influence of different bolt modellings on the relative displacements at the seating of the seals. The influence of bolt modelling, thermal properties and the detailed geometry of the two-dimensional finite-element models on the results are discussed.     

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.     

Dynamic impact characteristics of KN-18 SNF transport cask – Part 1: An advanced numerical simulation and validation technique

Domestic and international regulations for the transportation of radioactive materials strictly prescribe the design requirements for spent nuclear fuel (SNF) transport casks. According to the applicable codes, a transport cask must withstand a free-drop impact of 9 m onto an unyielding surface and a free-drop impact of 1 m onto a mild steel bar. However, the structural performance of a transport cask is not easy to evaluate precisely because the dynamic impact characteristics of the cask, which includes impact limiters to absorb the impact energy, are so complex.     

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.     

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.     

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.     

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.     

Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions

Abstract

Transport packages for spent fuel have to meet the requirements concerning containment, shielding and criticality as specified in the International Atomic Energy Agency regulations for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are, among others, important inputs for the evaluation of correspondent package capabilities under these conditions. The kind, accuracy and completeness of such information depend upon purpose of the specific problem. In this paper, the mechanical behaviour of spent fuel assemblies under accident conditions of transport will be analysed with regard to assumptions to be used in the criticality safety analysis. In particular the potential rearrangement of the fissile content within the package cavity, including the amount of the fuel released from broken rods has to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods of each fuel assembly among themselves as well as between fuel assemblies, basket, and cask body or cask lid, the exact mechanical analysis of such phenomena under drop test conditions is nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context a simplified analytical methodology for conservative estimation of fuel rod failures and spent fuel release is described. This methodology is based on experiences of BAM acting as the responsible German authority within safety assessment of packages for transport of spent fuel.     

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.     

Optimised Concrete Containers for Waste Management

Abstract

Over the last 8 years, Siempelkamp has tested different types of containers, partly with concrete shielding according to the requirements of German transport regulations as well as to the acceptance criteria for the planned German final storage site, Schacht Konrad. After the process of testing, Siempelkamp produced more than 1000 containers of different cubic and cylindrical shapes with concrete shielding. In the course of using the containers, certain possibilities for improvement have become obvious in order to fulfil the handling and transport requirements. These improvements are based on: protection against improper operational handling on site; improvement of surface protection; and optimisation of the lid attachment. Furthermore, Siempelkamp has developed a concept to adjust non-qualified containers to achieve the requirements for transport and storage. Considering long-term interim storage of the containers, improving the containers not only allows better handling but also provides an adequate basis for safe deposition in interim storage as well as later final storage.     

Effect of irradiation on the impact and seismic response of a spent fuel storage and transport cask

The spent fuel storage and transport cask must withstand various accident conditions such as fire, free drop and puncture in accordance with the requirement of the IAEA and domestic regulations. The spent fuel storage and transport cask should maintain the structural safety not to release radioactive material in any condition. And also the effects of the irradiation should be considered because the spent fuels stored in the cask for a long time and be possible to change the mechanical properties of the cask.In this study, the changed mechanical properties of the cask after irradiation for the 30 years storage periods are assumed and applied to the impact analysis using ABAQUS/Explicit code and seismic analysis using ANSYS code. The stress intensity on each part of the cask is calculated and the effects of irradiation are studied and structural integrity of the package is evaluated.     

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.     

Drop test program with half scale model CASTOR® HAW/TB2

Abstract

Federal Institute for Materials Research and Testing (BAM) is the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste in Germany. In context with package design approval of the new German high level waste cask CASTOR® HAW28M, BAM performed several drop tests with a half scale model of the CASTOR® HAW/TB2. The cask is manufactured by Gesellschaft für Nuklear Service mbH and was tested under accident transport conditions on the 200 tons BAM drop test facility at the BAM Test Site Technical Safety. For this comprehensive test program, the test specimen CASTOR® HAW/TB2 was instrumented at 21 measurement planes with altogether 23 piezo resistive accelerometers, five temperature sensors and 131 triaxial strain gauges in the container interior and exterior respectively. The strains of four representative lid bolts were recorded by four uniaxial strain gauges per each bolt. Helium leakage rate measurements were performed before and after each test in the above noted testing sequence. The paper presents some experimental results of the half scale CASTOR® HAW/TB2 prototype (14?500 kg) and measurement data logging. It illustrates the extensive instrumentation and analyses that are used by BAM for evaluating the cask performance to the mechanical tests required by regulations. Although some of the quantitative deceleration, velocity and strain values cannot be shown because of confidentially issues, they are provided qualitatively to illustrate the types of measurements and methodologies used at BAM.     

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.     

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.     

Aging management and post-storage transport of dual purpose cask for spent fuel

Aging management of spent fuel storage facility may follow lessons learned from literature for nuclear power plant and a review for spent fuel dry cask storage system by US NRC, DOE, by German BAM, that by Japan NISA, etc. Namely, the essence of systematic approach to aging management includes Understanding aging, Plan (Development and optimisation of activities for aging management), Do (Managing aging mechanisms), Check (Monitoring, inspection and assessment), and Act (Maintenance). The PDCA cycle will optimise the systematic approach to the aging management. An aging management programme (AMP) for the storage system over the period of extended storage will address uncertainties in the safety relevant functions of the system that may otherwise be impaired by aging mechanisms. The AMP identifies system, structure and components (SSCs) that need specific actions to mitigate aging and ensures that no aging effects result in a loss of their intended function of the SSCs, during an intended licensed period. AMPs generally include Prevention, Mitigation, Monitoring, Inspection, and Maintenance programmes. Aging management plans should ensure compliance with transportation requirements after extended storage. Potential issue would be a significant change of the transport regulations in the future. If the regulations changed significantly, a gap analysis should be performed to identify any impact to the cask safety. Compensating arrangements, if necessary, should be proposed at that time. Assuming that the regulations will not change significantly after long term storage, we will be able to renew the license both for transport and storage of the cask during the storage period. For example, in Japan, a holistic approach was established for the license of a 50 year storage and transport. In this approach, we can evaluate integrity of spent fuel, basket, etc. with respect to chemical, thermal, mechanical, and radiation factors. With this approach we will not have to open the cask lid for visual inspection of the spent fuel, basket, etc. prior to the post-storage transport.     

Thermal analysis of proposed transport cask for three advanced burner reactor used fuel assemblies

Abstract

Preliminary studies of used fuel generated in the US Department of Energy's Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide advanced burner reactor used fuel assemblies with a burn-up of 160 MWD kg^–1. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its centre to its exterior surface, is determined by two-dimensional computational fluid dynamics simulations of conduction, convection and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.