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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

Topics of research and development on concrete cask storage of spent nuclear fuel

This paper addresses topics of research and development (R&D) being challenged for realization of concrete cask storage of spent nuclear fuel in Japan. Comparison between metal cask storage and concrete cask storage is addressed. Background of these R&D and current status of technology on spent fuel storage are described. Need and design concepts of concrete cask storage technology, tests and evaluation of integrity of spent fuel, materials, concrete casks under normal and accident conditions, monitoring technology, etc. are systematically arranged and introduced. Topical problems of these R&D are described.     

Safety Aspects of Long-term Dry Interim Storage Of Type-b Spent Fuel and Hlw Transport Casks

Abstract

Based on the German decision to minimise transport of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities were licensed until the end of 2003. Because of the large amount of Type B(U) transport casks which are going to be used for long-term interim storage the question of timelimited Type B(U) licence maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of qualification of transport casks for interim storage is the long-term behaviour of the metallic seal–lid system. Here results are presented from current long-term experimental tests with metallic 'Helicoflex' seals in which pool water is enclosed. This series of tests has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an exchange of experience, know-how and state-of-the-art between authorities and technical experts with regard to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called 'coordinating institution for cask dispatching information' ('KOBAF') which entails management of an online database of cask-specific documents and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the coming decades.     

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.     

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.     

Design Assessment for Transport and Storage Casks

Abstract

The design assessment concerning the mechanical behaviour of transport and storage casks for radioactive material to fulfil nuclear safety criteria has to be based on two essential considerations: (1) Effective analysis of the stress–strain state of the cask components under both normal operational and test conditions including hypothetical accident scenarios with suitable accepted methods. (2) Economic estimation of the required properties and the structural state of the cask components with sufficient exactness. In an overview of the codes which are available at GNS/GNB for cask impact strength analyses (ANSYS, ADINA, VDI Codes), procedures and aspects of benchmarking and validation of calculation codes are described. The results of experimental full size cask drop test programs (CASTOR, POLLUX) and corresponding pre-test calculational analyses show the suitability of the codes used. The influence of dynamic effects on the mechanical properties of material (ductile cast iron, wood) has been investigated experimentally. By consideration of these dynamic values in strength analyses of casks at impact a good agreement between experimental and calculational results has been achieved.     

Extended Drop Tests Of DCI Casks with Artificial Flaws Demonstrating the Existing Safety Margins

Abstract

The use of spent fuel shipping and storage casks made of ductile cast iron (DCI) has been common practice for about 15 years when the development of such casks started in Germany where qualified foundries are able to produce these heavy section castings at the high quality level needed for this kind of application. To promote the discussion on safety against brittle failure a lot of research had been carried out in different countries. The two test programmes in Germany on casks with big artificial flaws under severe impact conditions is summarised in this paper. The first test object was a thick walled DCI ‘pipe’ (150 mm wall thickness) with dimensions equivalent to a 1:2.5 scale cask model. It was dropped with a 40 mm deep laser sharpended flaw from heights of up to 9 m onto rails. As a second test object a full scale CASTOR VHLW cask was used. This specimen had a flaw with a depth of 120 mm in a 260 mm thick wall. With increasing drop heights (up to 14 m) and stress intensity factors (up to material fracture toughness) this object was also dropped onto rails. For both cases the measured data (decelerations, crack opening displacement, strains, material properties) are presented. No brittle failure occurred, although in the 14 m drop of the CASTOR VHLW Cask the impact was 6.5 times higher than the impact measured in the mechanical test of the type B package design. The results demonstrate that DCI casks have significantly high safety margins even in the hypothetical case of an impact beyond type B package design requirements.     

Consideration of aging mechanism influence on transport safety of dual purpose casks for spent nuclear fuel or HLW

When storage of spent nuclear fuel or high level waste is carried out in dual purpose casks (DPC), the effects of aging on safety relevant DPC functions and properties have to be managed in a way that a safe transport after the storage period of several decades is capable and can be justified and certified permanently throughout that period. The effects of aging mechanisms (e.g. radiation, different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) on the transport package design safety assessment features have to be evaluated. Consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components that cannot be directly inspected or changed without opening the cask cavity, like the inner parts of the closure system and the cask internals, like baskets or spent fuel assemblies. The design criteria of that transport safety case have to consider the operational impacts during storage. Aging is not the subject of technical aspects only but also of ‘intellectual’ aspects, like changing standards, scientific/technical knowledge development and personal as well as institutional alterations. Those aspects are to be considered in the management system of license holders and in appropriate design approval update processes. The paper addresses issues that are subject of an actual International Atomic Energy Agency TECDOC draft ‘Preparation of a safety case for a dual purpose cask containing spent nuclear fuel’.     

Innovative fast breeder reactor concept ‘RAPID’ for improvement of reactor performance and proliferation resistance

The 60 MWe metal fueled fast breeder reactor concept ‘RAPID’ to improve reactor performance and proliferation resistance has been demonstrated. The reactor can be operated without refueling for up to 5 years. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly (IFA) instead of conventional fuel subassemblies. RAPID concept enables quick and simplified refueling by replacing an IFA in which all the core and blanket fuel elements are comprised. An on-site storage cask achieves on-site decay heat removal of an IFA. After 3 years of on-site storage, an IFA together with an on-site storage cask can be transported directly to the reprocessing plant without any intermediate steps. Significant improvement of inherent safety features and plant availability has been discussed. Decay heat removal capability, safety consideration on criticality of the IFA and shielding design of the on-site storage cask has been confirmed. The RAPID refueling concept possesses high resistance to state-supported removal of plutonium for nuclear weapons production.     

Heat removal verification tests using concrete casks under normal condition

Heat removal tests using two types of full-scale concrete casks were conducted. This paper describes the results under a normal condition of spent fuel storage. In the tests, data on heat removal performance and integrity of cask components were obtained for different storage periods. The change of decay heat of spent fuel was simulated using electric heaters. Reinforced concrete cask (RC cask) and concrete filled steel cask (CFS cask) were the specimen casks. Decay heat at the initial period of storage 60 years of storage, the middle period (20 years of storage), and the final period (40 years of storage) correspond to 22.6 kW, 16 kW and 10 kW, respectively. Quantitative temperature data of the cask components were obtained as compared with their limit temperature. In addition, heat balance data requested for heat removal analyses were obtained.     

Implementation of burn-up credit approach for transport and storage cask

Abstract

TN International currently uses burn-up credit methodology for the design of casks dedicated to the transport of pressurised water reactor uranium oxide spent fuel assemblies. As long as the fuel enrichment of the pressurised water reactor fuel assemblies was sufficiently low, a burn-up credit methodology based on the sole consideration of actinides and the use of a partial burn-up was satisfactory to cover the needs without necessity to design new casks. Nevertheless, the continuous increase in the fuel enrichment during the last decade has led TN International to continue the investigations on the burn-up credit methodology to limit both the increase in the neutron poison content in the new basket designs and the burn-up constraints attached to the acceptability of the fuel assemblies for transport. The strategy of TN International was then to take benefit of the large negative reactivity reserves, which might be gained by the consideration of the fission products coming from the fuel irradiation. A big step forward has recently been reached by TN International on this field with the definition of an advanced burn-up credit methodology based on the consideration of relevant fission products recommended by OECD. In the meantime, TN International has taken the opportunity to use such burn-up credit approach in the design of the TN 24 E transport and storage cask developed for the German nuclear power plants. The relevant task has been carried out according to the German standard DIN 25712 for burn-up credit application. The present paper will describe the basic principles of the burn-up credit methodology implemented by TN International such as:

(i) the current state of the art concerning the burn-up credit application in the criticality assessment

(ii) the basic approach used for the implementation of the advanced burn-up credit methodology (bounding axial burn-up profiles, fuel irradiation parameters, fission products, etc.)

(iii) the area of validity of the TN International burn-up credit approach with fission products

(iv) example of application of the burn-up credit methodology for the design of the TN 24 E transport and storage cask under licensing in Germany

(v) the perspectives of development of the burn-up credit methodology.     

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.     

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.