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.     

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.     

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.     

A methodology for the evaluation of fuel rod failures under accident conditions

Abstract

Recent studies on the long-term behaviour of high-burnup spent fuel have shown that, under normal conditions of storage, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride cracking, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar regulatory rules have not yet been developed to address failures of fuel rod cladding that could potentially lead to reconfigured fuel geometry under hypothetical transport accidents. At issue is the effect on cladding ductility of potential changes in zirconium hydride morphology during dry storage. Recent studies have shown that above a certain level of cladding hoop stress, the decaying temperature history during dry storage can cause the hydrogen in solid solution to precipitate in the form of radial hydrides, which, depending on their relative concentration, can induce brittle failures in the cladding. From a US regulatory perspective such cladding failures, if they were to cause fuel reconfiguration, could invalidate the cask's criticality and shielding licensing analyses, which are based on coherent geometry. This paper describes a methodology for high-burnup spent fuel to determine the frequency of cladding failure and failure modes under drop accidents, considering end-of-storage spent fuel conditions. The degree to which spent fuel reconfiguration could occur during handling or transport accidents would depend to a large extent on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there are no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, this paper focuses on the development of a methodology for modelling and analysis that deals with this general problem on a generic basis. First, consideration is given to defining accident loading that is equivalent to the bounding hypothetical transport accident of a 9 m drop onto an essentially unyielding surface. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A model of material behaviour, with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite-element code. The hydride precipitation model, which describes the hydride structure of the cladding at the end of dry storage, and the hydride-dependent properties of high-burnup fuel cladding form the main input to the constitutive model. The third element in the overall process is to utilise this material model and its host finite-element code in the structural analysis of a transport cask subjected to bounding accident loading to calculate fuel rod failures and failure mode configurations. This requires detailed modelling of the transport cask and its internal structure, which includes the canister, basket, fuel assembly grids and fuel rods. The overall methodology is described.     

一种干式外圆机械切割装置的研制

在乏燃料后处理中，需要回取已封装在乏燃料贮存容器中的乏燃料。根据热室使用环境及乏燃料贮存容器的特点，从耐辐射设计、乏燃料贮存容器固定、切割进给、切割刀具及刀具更换、放射性废物最少化等方面进行设计响应，研制了一种在热室内开启乏燃料贮存容器的干式外圆机械切割装置。功能性试验验证了该装置满足设计和使用要求。      

Validation of the thermal calculation method and the applicability of confinement monitoring with a full-scale mock-up test of a concrete cask

Abstract

The Mitsui Engineering & Shipbuilding Co. Ltd (MES) has designed and fabricated a full-scale mock-up system that can be used to store spent nuclear fuel (SNF). The system is made up of two parts; a concrete shield that is vented and an inner steel canister that provides containment of the SNF. A benchmark analysis of this storage system was carried out using a combined thermal calculation method. Initially airflows and temperatures outside the canister were calculated using a three-dimensional thermal flow analysis method. The results from this analysis were used as the boundary conditions to calculate the maximum temperatures inside the canister using a two-dimensional heat transfer method. The calculated results agreed well with the measurements and the validity of the combined method of analysis was confirmed. Since all measured temperatures were within their acceptable limits, it was also confirmed that the concrete cask storage system has sufficient heat removal capability. MES has also proposed a new canister confinement monitoring system. It is based on the relationship between the inner pressure of the canister and the temperature of the canister lid and the pedestal. The validity and applicability of the system are confirmed by the full-scale mock-up experiment results. The conceptual design of the monitoring system is considered, and the system can realised at low cost, with high reliability and easy maintenance.     

Development of the detecting method of helium gas leak from canister

In the concrete cask storage system, spent fuel is installed and weld-sealed in a cylindrical container called a canister. The canister is filled with helium gas and its containment shall be maintained and inspected during storage. The helium gas enhances heat removal from spent fuel. When the helium gas leaks, the effect of helium gas convection is weakened in the canister. Thereof, the temperature on the canister surface changes.In present tests, it was found that temperatures of the center of the top and the bottom on the canister surface change remarkably during the helium gas leak. Therefore, we defined the temperature difference as ΔT_BT. And one can detect helium gas leak using the change of ΔT_BT. ΔT_BT increases monotonously toward a constant value during helium gas leak, even if the inlet air temperature drops. The helium gas leak can be detected at the early stage of the leak by observing both ΔT_BT and inlet air temperature.     

乏燃料密封容器开盖及其内容物回取技术研究

对于采用干湿法贮存的乏燃料而言，其后处理时面临的最大问题是如何安全高效地将乏燃料等内容物从封焊的密封容器中取出。针对这一问题，基于乏燃料密封容器及其内容物的结构特点，开展了乏燃料密封容器开盖及内容物回取技术研究，综合考虑切割热室使用环境、内容物回取后的收集和转移以及产生废物的收集处理等因素，制定了合理可行的开盖及回取工艺，研制了用于开盖和筒体分段切割的解体装置以及回取和吊装工具，并通过试验验证了工艺的可行性以及研制的工装具的可用性。      

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.     

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.     

Heat removal characteristics of a concrete cask by a simplified test model

The present study is concerned with the characteristic cooling flow in the annular gap of a concrete cask used to store spent nuclear fuel. The concrete cask is cooled by a natural convection flow of air passing through an internal annular gap between the outside of a canister and the inside of the concrete vessel. The heat transfer coefficient and friction loss coefficient of such a flow could not be fully estimated even if we used existing handbooks, because the airflow has unique features. Simulation experiments using a simplified model for the cooling path have been conducted to estimate the heat transfer coefficient and friction loss coefficient. It was found by this study that the heat transfer coefficient well agreed with the prediction by an empirical formula applied to the free convection, and the friction loss coefficient was 2–2.5 times the value of an isothermal flow.     

Analyses to demonstrate structural performance of KN18 in hypothetical drop accident scenarios

Abstract

The KN18 is a new cask design by KONES for KHNP for the dry or wet transportation of up to 18 PWR spent nuclear fuel assemblies in South Korea. The containment vessel consists of a cylindrical thick-walled forged steel body, closed by a stainless steel lid with bolts. Spent fuel assemblies are located in a basket which consists of a tube disc system. Two pairs of trunnions are attached for lifting, manoeuvring and tie-down. A pair of impact limiters manufactured from wood and encased in steel cladding provide impact energy absorption during the hypothetical accident conditions. The package complies with the requirements of 10 CFR Part 71 for Type B(U)F packages. It received its transport license from the Korean Competent Authority KINS in early 2010 and is expect to enter service in 2011. Structural performance of the package in the normal and accident conditions were demonstrated against the requirements of 10 CFR Part 71 by analysis including extensive calculations by state-of-the-art finite element methods, and confirmed by tests carried out on a one-third scale test model which were also used to verify the numerical tool and methods used in the analyses. For the analyses of the hypothetical accident drop conditions, the models consisted of the complete package, including the impact limiters, the containment structure and the basket, which was modelled explicitly in detail and in three dimensions, to take into account the complex interaction between the components and the non-linearities in the geometry, the material behaviour and overall behaviour. The analyses were carried out using the explicit transient finite element method so that the transient behaviour could be robustly simulated. This paper presents two of the analyses from the suite of analyses for demonstrating the performance of the package in the hypothetical accident drop scenarios, discussing the analyses methodology, modelling technique and evaluation methodology, as well as analyses results and package response. The one-third scale model drop testing and benchmarking of the model to the scale model tests are the subject of a separate paper.     

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.     

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.     

Spent fuel transportation risk assessment: cask fire analyses

Abstract

The US Nuclear Regulatory Commission has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. This assessment considered the response of three certified casks to a range of fire accidents in order to determine whether or not they would lose their ability to contain the spent fuel or maintain effective shielding. The casks consisted of a lead shielded rail cask that can be transported either with or without an inner welded canister, an all steel rail cask that is transported with an inner welded canister, and a DU shielded truck cask that is transported with directly loaded fuel. For the two rail casks, large pool fires that were concentric (fully engulfing), offset from the casks by 3 m, and offset from the cask by 18 m were analysed using the computational fluid dynamics CAFE-3D fire modelling code coupled with the finite element analysis PATRAN-Thermal heat transfer code. All of the fires were assumed to last for 3 h. In addition to these extraregulatory fires, the regulatory 30 min fire was analysed using both the regulatory uniform 800°C boundary condition and the more realistic CAFE-3D fire modelling code. For the truck cask, only the engulfing fire case was analysed using a 1 h fire duration. In all of the fire analyses, the seal region of the cask stayed below the failure temperature; therefore, there would be no release of radioactive material. In addition, the temperature of the fuel rods stayed below their burst rupture temperature, providing another barrier to release. For the lead shielded cask, very severe fires cause some of the lead to melt. There is no leak path for this molten lead to exit the shield region, but its expansion during the melting and subsequent contraction due to solidification during cool down results in a reduction in gamma shielding effectiveness.     

Spent fuel transportation risk assessment: cask impact analyses

Abstract

The Nuclear Regulatory Commission (NRC) has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. This assessment considered the response of three certified casks to a range of impact accidents in order to determine whether or not they would lose their ability to contain the spent fuel or maintain effective shielding. The casks consisted of a lead shielded rail cask that can be transported either with or without an inner welded canister, an all-steel rail cask that is transported with an inner welded canister, and a DU shielded truck cask that is transported with directly loaded fuel. Finite element analyses were performed for impacts at speeds of 48, 97, 145 and 193 kilometres per hour into a rigid target. Impacts in end-on, side-on, and CG-over-corner orientations were analysed for each cask and impact speed. Calculations were performed to equate these impacts onto rigid targets with higher speed impacts onto the yielding targets that exist in the real world. These analyses indicated that a cask with an inner welded canister or a truck cask would not release radioactive material in any impact accident and that only very high-speed impacts onto hard rock targets could result in either release of material or significant degradation of shielding for rail casks without an inner canister. Impacts other than those onto flat unyielding targets were also considered. Analyses show that an impact that bypasses the impact limiters on the ends of the casks does not result in seal failure and neither does an impact by a locomotive also between the impact limiters.     

Analysis of thermal problems in development of spent fuel transport cask for research reactors

A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper, both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled.     

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.