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.     

Expansion of the capabilities of the GA-4 legal weight truck spent fuel shipping cask

Abstract

General Atomics has developed the model GA-4 legal weight truck spent fuel cask, a high-capacity cask for the transport of four pressurised water reactor (PWR) spent fuel assemblies, and obtained a certificate of compliance (CoC, No. 9226) in 1998 from the US Nuclear Regulatory Commission (NRC). The currently authorised contents for this CoC, however, are much more limiting than the actual capability of the GA-4 cask to transport spent PWR fuel assemblies. The purpose of this paper is to show how the authorised contents can be significantly expanded by additional analyses without any changes to the physical design of the package. Using burn-up credit as outlined in US NRC Interim Staff Guidance 8, Revision 2, the authorised contents can be significantly expanded by increasing the maximum enrichment as the burn-up increases. Use of burn-up credit eliminates most of the criticality imposed limits on authorised package contents, but shielding still limits the use of the cask for higher burn-up, short-cooled fuel. By reducing the number of assemblies transported (downloading) to two and using shielding inserts, even high-burn-up fuel with reasonable cooling times can be transported.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

Verification of activity release compliance with regulatory limits within spent fuel transport cask assessment

Abstract

Admissible limits for activity release from type B(U) packages for spent fuel transport specified in the International Atomic Energy Agency regulations (10^?6 A₂ h^?1 for normal conditions of transport and A₂ per week for accidental conditions of transport) have to be kept by an appropriate function of the cask body and its sealing system. Direct measurements of activity release from the transport casks are not feasible. Therefore, the most common method for the specification of leak tightness is to relate the admissible limits of activity release to equivalent standardised leakage rates. Applicable procedure and calculation methods are summarised in the International Standard ISO 12807 and the US standard ANSI N14·5. BAM as the German competent authority for mechanical, thermal and containment assessment of packages liable for approval verifies the activity release compliance with the regulatory limits. Two fundamental aspects in the assessment are the specification of conservative design leakage rates for normal and accidental conditions of transport and the determination of release fractions of radioactive gases, volatiles and particles from spent fuel rods. Design leakage rates identify the efficiency limits of the sealing system under normal and accidental transport conditions and are deduced from tests with real casks, cask models or components. The releasable radioactive content is primarily determined by the fraction of rods developing cladding breaches and the release fractions of radionuclides due to cladding breaches. The influence of higher burn-ups on the failure probability of the rods and on the release fractions are important questions. This paper gives an overview about methodology of activity release calculation and correlated boundary conditions for assessment.     

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.     

Safety Concerning Transportation of Spent Fuel

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

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

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

Analysis of the Castor RBMK Cask Cooling Preceding Discharge of Spent Nuclear Fuel

Abstract

An important problem of the handling of casks intended for spent nuclear fuel transport and storage is providing safety during all operations. In particular the safety requirements should be fulfilled during the cask cooling that precedes the discharge of spent nuclear fuel from the cask. An analysis has been performed for the CASTOR RBMK cask heat removal system. This provides forced cooling of the cask with the spent fuel assemblies in it, by water delivery into the cask inner cavity. As a result of analyses performed for the different flow rates of the cooling water, the maximum pressure in the cask cavity caused by water evaporation has been estimated and compared with the maximum permissible value and the time taken by the cask in cooling to the given temperature limit has been determined. On the basis of the analysis results the most preferable regime for CASTOR RBMK cask cooling is suggested.     

Vulnerabilities of legacy fuel casks when evaluated using modern structural analysis tools

Abstract

Cylindrical fuel casks often have impact limiters surrounding the ends of the cask shaft in a typical 'dumbbell' arrangement. The primary purpose of these impact limiters is to absorb energy to reduce loads on the cask structure during impacts associated with a severe accident. Impact limiters are also credited in many packages with protecting closure seals and reducing peak temperatures during fire events. For this credit to be taken in safety analyses, the impact limiter attachment system must be shown to retain the impact limiter following normal conditions of transport (NCT) and hypothetical accident conditions (HAC) impacts. Large casks are often certified by analysis only because of the cost associated with testing. Therefore, some cask impact limiter attachment systems have not been tested in real impacts. A recent structural analysis of the T-3 spent fuel containment cask found problems with the design of the impact limiter attachment system. Assumptions in the original safety analysis for packaging (SARP) concerning the loading in the attachment bolts were found to be inaccurate in certain drop orientations. This paper documents the lessons learned and their applicability to impact limiter attachment system designs.     

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.     

Operational experience of on-site transport of spent nuclear fuel in Korea

Abstract

With the rapid development of the nuclear power programme in Korea, the amount of accumulated spent nuclear fuel has inevitably increased year by year. The spent nuclear fuel is being stored in on-site storage pools at the nuclear power plants. As the current storage capacity for spent nuclear fuel is insufficient, at-reactor storage is being expanded at each site with regard to optimisation of technical and economic factors. On-site transport between neighbouring reactors has been necessary to secure sufficient storage capacity for pressurised water reactor spent nuclear fuel assemblies. A complete on-site transport system has been developed, and so far more than 800 spent nuclear fuel assemblies have been transported using two kinds of transport cask.