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.     

Finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel storage cask

A spent fuel storage cask is required to prove the safety of its canister under a hypothetical accidental drop condition which means that the canister is assumed to be free dropped on to a pad of the storage cask during the loading of the canister into a storage cask. In this paper, finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel dry storage cask were carried out to improve the structural integrity of the canister under a hypothetical accidental drop condition. The pad of the storage cask was originally designed as cylindrical steel structure filled with concrete. The pad was modified by using the structure composed of steel and polyurethane-foam instead of the quarter of the upper concrete as an impact limiter. The effects of the shape and the thickness of the steel structure and the density of the polyurethane-foam which was used in between steel structures were studied. As the optimized pad of a spent fuel dry storage cask, the quarter of the upper concrete was replaced with 12 mm thick circular steel structure and polyurethane-foam whose density was 85 kg/m³. The drop tests of a 1/3 scale model for the canister on to the original pad and the optimized pad were conducted. The effect of the pad structure was evaluated from the drop tests. The optimized pad has a greater shock-absorbing effect than the original pad. In order to verify the analysis results, strains and accelerations in the time domain by the analytical methods were compared with those by a test. The numerical method of simulating the free drop test for a dry storage cask was verified and the numerical results were found to be reliable.     

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.     

Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

Abstract

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.     

核电厂乏燃料组件厂内转运关键设备研制

乏燃料组件厂内转运是解决核电厂燃料水池贮存空间不足问题的方法之一。本文分析了乏燃料组件厂内转运的设计准则、安全风险,介绍了用于运输容器内破损组件检测和运输容器内组件冷却用设备的工作原理及其应用情况。应用结果表明:破损检测设备可以快速有效地检测乏燃料运输容器内是否存在破损组件;乏燃料组件冷却设备可以较为安全地冷却装有乏燃料组件的运输容器。      

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.     

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.     

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.     

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.     

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’.     

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.     

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.     

A comparison of dose rate calculations for a spent fuel storage cask by using MCNP and SAS4

This paper presents a detailed comparison of the surface dose rate calculations for the NAC-UMS spent fuel storage cask by using MCNP and SAS4 computer codes. Their accuracy and computation efficiencies are compared. For such a real world deep penetration and streaming problem, effective variance reduction techniques are indispensable for a Monte Carlo simulation to obtain results of small statistic errors within reasonable computing time. The TORT-coupled MCNP calculation based on the CADIS methodology has been used in this study. The main differences between MCNP and SAS4 calculations are the underlying cross-section libraries and the adjoint functions used for variance reduction in Monte Carlo simulations. The cross-section libraries and their formats should be the root cause for some significant discrepancies between the MCNP and SAS4 results. In addition, limited by the 1D adjoint biasing scheme, SAS4 is inefficient in calculating the dose rates near inlet/outlet apertures. Considering all the computer time spent and the statistical errors of results obtained, the overall computation efficiency by using the TORT-coupled MCNP is better than SAS4 in the shielding calculations of spent fuel storage casks. More specifically, although the SAS4 efficiency is better when the cask side calculation is the only concern, the TORT-coupled MCNP technique is more efficient for the gamma-ray transport in cask top configurations and almost all the vent-streaming problems.     

A study on dynamic impact of vertical concrete cask tip-over using explicit finite element analysis procedures

This work presents a study on dynamic impact of a vertical concrete cask (VCC) tip-over, using explicit finite element analysis (FEA) procedures. The VCC presented in this paper is made of reinforced concrete casted with a steel liner for accommodating a canister containing spent nuclear fuels. An explicit FEA code, LS-DYNA, is employed to treat the highly nonlinear problems encountered in postulated tip-over events. The plasticity and fragmentation of concrete are respectively treated by the pseudo-tensor material model and the element erosion technique. The interface de-bonding between VCC concrete and steel liner, contact/impact between VCC and target pad are all considered in order to investigate the reasonable impact load for cask design. Four cases with various analysis assumptions are respectively implemented and compared one another for ease of getting design load. The significance of interface de-bonding and concrete fragmentation in VCC to spent fuel cask design is highlighted in the reported numerical results.     

Analysis for seismic response of dry storage facility for spent fuel

Most of the dry storage systems for spent fuel are freestanding, which leads to stability concerns in an earthquake. In this study, as a safety check, the ABAQUS/Explicit code is adopted to analyse the seismic response of the dry storage facility planned to be installed at Nuclear Power Plant #1 (NPP1) in Taiwan. A 3D coupled finite element (FE) model was established, which consisted of a freestanding cask, a concrete pad, and underneath soils interacting with frictional contact interfaces. The scenario earthquake used in the model included an artificial earthquake compatible to the design spectrum of NPP1, and a strong ground motion modified from the time history recorded during the Chi-Chi earthquake. The results show that the freestanding cask will slide, but not tip over, during strong earthquakes. The scale of the sliding is very small and a collision between casks will not occur. In addition, the differential settlement of the foundation pad that takes place due to the weight of the casks increases the sliding potential of the casks during earthquakes.     

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.     

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.