Safety Analysis of Casks Under Extreme Impact Conditions

Abstract

The determination of the inherent safety of casks under extreme impact conditions has been of increasing interest since the terrorist attacks of 11 September 2001. For nearly three decades BAM has been investigating cask safety under severe accident conditionslike drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. One of the most critical scenarios for a cask is the centric impact of a dynamic load onto the lid-seal system. This can be caused, for example, by a direct aircraft crash (or just its engine) as well as by an impact due to thecollapse of a building, e.g. a nuclear facility storage hall. In this context BAM is developing methods to calculate the deformation of cask components and — with respect to leak-tightness — relative displacements between the metallic seals and their counterparts. This paper presents reflections on modelling of cask structures for finite-element analyses and discusses calculated results of stresses and deformations. Another important aspect is the behaviour of a cask under a lateral impact by aircraft or fragments of a building. Examples of the kinetic reaction (cask acceleration due to the fragments, subsequent contact with neighbouring structures like the ground, buildings or casks) are shown and discussed in correlation to cask stresses which are to be expected.     

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

Abstract

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

Methodological aspects for finite element modelling of lid systems for Type B(U) transport packages

Abstract

The regulatory compliance of the containment system is of essential importance for the assessment process of Type B(U) transport packages. The requirements of the International Atomic Energy Agency safety standards for transport conditions imply high loading on the containment system. The integrity of the containment system has to be ensured in mechanical and thermal tests. The containment system of German spent nuclear fuel and high level waste transport packages usually includes bolted lids with metal gaskets. The finite element (FE) method is recommended for the analysis of lid systems according to the guideline BAM-GGR 012 for the assessment of bolted lid and trunnion systems. The FE analyses provide more accurate and detailed information about loading and deformation of such kind of structures. The results allow the strength assessment of the lid and bolts as well as the evaluation of relative displacements between the lid and the cask body in the area of the gasket groove. This paper discusses aspects concerning FE simulation of lid systems for type B(U) packages for the transport of spent nuclear fuel and high level waste. The work is based on the experiences of the BAM Federal Institute for Materials Research and Testing as the German competent authority for the mechanical design assessment of such kind of packages. The issues considered include modelling strategies, analysis techniques and interpretation of results. A particular focus of this paper is on the evaluation of the results with regard to FE accuracy, influence of the FE contact formulation and FE modelling techniques to take the metallic gasket into account.     

Spent fuel transport cask thermal evaluation under normal and accident conditions

The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario.The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers.Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself.In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the ‘80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are discussed.     

Long term performance of metal seals for transport and storage casks

Dual purpose casks for the transportation and storage of spent nuclear fuel and other radioactive materials require very high leak tightness of lid closure systems under accident conditions as well as in the long term to prevent activity release. For that purpose metal seals of specific types with an inner helical spring and outer metal liners are widely used and have shown their excellent performance if certain quality assurance requirements for fabrication and assembling are satisfied. Well defined surface roughness, clean and dry inert conditions are therefore essential. No seal failure in a loaded cask happened under these conditions until today. Nevertheless, the considered and licensed operation period is limited and all safety assessments have been performed and approved for this period of time which is 40 years in Germany so far. However, in the meantime longer storage periods might be necessary for the future and therefore additional material data will be required. BAM is involved in the qualification and evaluation procedures of those seals from the early beginning. Because long term tests are always time consuming BAM has early decided to perform additional tests with specific test seal configurations to gain a better understanding of the long term behaviour with regard to seal pressure force, leakage rate and useable resilience which is safety relevant mainly in case of accidental mechanical loads inside a storage facility or during a subsequent transport. Main test parameters are the material of the outer seal jacket (silver or aluminium) and the temperature. This paper presents the BAM test program including an innovative test mock-up and most recent test results. Based on these data extrapolation models to extended time periods are discussed, and also future plans to continue tests and to investigate seal behaviour for additional test parameters are explained.     

Similarity aspects for closure systems in reduced scale package drop testing

Abstract

An exact scaling of all structural components of a package for radioactive materials and their mechanical characteristics is not always possible in drop tests with reduced scale models. This has to be especially considered for bolted closure systems. On the one hand, the sizes of the bolts cannot be scaled with the same geometrical scale factor. On the other hand, the possibilities of an accurate scaling of seal characteristics are very limited. Owing to non-linearity of the force–compression relationship of a gasket, it is, for instance, impossible to simulate the maximum compression force and permissible elastic decompression of a metallic gasket simultaneously on the same scale model. Additional problems can also result from a dispersion of friction conditions in the bolted joints (at threads and under bolt heads). This dispersion, as well as an imprecise bolt tightening technique, leads to more or less considerable scatter of the bolt pretension. The minimum pretension creates more severe conditions in a drop test with regard to the seal function (higher probability of the lid opening and sliding). The maximum pretension is usually conservative for the total bolt stress (the sum of the initial tension and additional load due to the drop test). These circumstances should be considered in planning drop tests as well as regarding the interpretation and transfer of test results to the original package design. In this paper some recommendations are described concerning the modelling of closure systems based on Federal Institute for Materials Research and Testing (BAM) experience in the approval assessment of transport casks for radioactive materials.     

Temperature tests on package lid seal materials over one year duration

Abstract

The design of lid seal system of any package transporting irradiated nuclear fuel is a major aspect of the containment safety case under normal and accident conditions. Consequently, when BNFL, now known as British Nuclear Group Sellafield, BN-GS, decided on a change in lid seal material, year long proving trials on the new material were conducted, these simulating actual service and accident temperature conditions. Several years ago, BN-GS with the Rubber and Plastic Research Association (RAPRA) developed an elastomer called EPDM 30H which could perform at temperatures below ?40°C and had excellent resistance to radiation, as later proven by long term testing. EPDM 30H material has been used in many site applications, with its first transport package application being on the Excellox 6 type, commissioned in 1991. Towards the end of the 1990s, BN-GS became aware that the fluorocarbon grade of lid seal material, used on several other packages, may be discontinued by the manufacturers and so they took the decision to replace the seals on the NTL 11 type packages with EPDM 30H. Consequently, BN-GS embarked on an extensive programme of seal testing which went beyond any it had previously carried out. Comprehensive data were available on radiation resistance and performance at low temperatures, but additional data were needed on its behaviour at elevated temperature over periods of about one year. A number of test sets were assembled comprising seals in representative lid seal grooves. Most of the test sets were fitted with EPDM 30H seals but another grade of EPDM seal was also included. Two test sets were continuously maintained at temperatures of 120 and 150°C, for one year. Other test sets were maintained at 60, 90, 120, 140, 160 and 180°C, but periodic inspection and compression set measurements were taken. At the end of one year the continuously heated seals were subjected to a thermal cycle corresponding to the thermal accident safety cases studies. These tests demonstrated that the EPDM 30H material had lower compression set characteristics than the other grade of EPDM and that EPDM 30H was a suitable seal material for all irradiated LWR fuel transport packages currently operated BN-GS.     

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

Abstract

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

Investigation of seal effects according to axial compression variation of metal seals for transport and storage casks

Abstract

Casks for the transport and storage of heat generating radioactive waste in Germany are normally provided with screwed lid systems, which are in most cases equipped with double jacket metal seals with an inner spring wire to provide long term resistance to the seal compression force. Preservation of the high sealing quality of those seals under operational and accidental stress conditions is essentially important to the safety of those casks. Relative displacements of the lid system surfaces caused by specific impact scenarios cannot be excluded and have to be evaluated with respect to a possible increase in the leakage rate.

To get representative data for such metal sealed lid systems, BAM has developed a special conceptualised flange system placed in an appropriate testing machine for relevant mechanical loading of the metal seals under static and cyclic conditions. Furthermore, the flange system enables continuous measurement of the standard helium leakage rate during each test.

The primary aim of the investigation is to identify the correlation between variation of installation conditions (axial displacements) caused by external loads and the standard helium leakage rate. An essential parameter in this case is the useable resilience r_u of a metal seal under relevant stress conditions. The useable resilience r_u is the vertical difference in the cross-section between the seal's assembling status and the point where the leakage rate, by means of external load relieving, exceeds the quality criterion of 10^–8 Pa m³ s^–1. Load relieving can instantly occur due to modification of the seal groove dimension caused by accident impacts and deformation of the lid system. Furthermore, component specific basis data for the development of finite element calculation models should be collected. In the tests, seals are subjected to static and cyclic loads. All tests are performed at ambient temperature.

This paper presents the test configuration, different test series and results of the current experiments. Typical load–displacement–leakage rate correlations are presented and discussed.     

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 Competent Authority Guidance in Finite-element Methods Applications for Package Design Assessment

Abstract

The development of new methods in analysing package designs using the finite-element method is of increasing importance. Package designers are increasingly applying the growing opportunities afforded by numerical methods to perform safety assessments for their products; this also requires suitable methods for competent authorities like BAM to verify applicants' results. This paper gives a topical overview of experiences and trends within the complex field of finite-element design testing. First some general and more formal aspects are described concerning the selection of the correct finite-element program and documentation of modelling, material properties, boundaries and calculation results, including their interpretation. To give a reliable basis for applicants in Germany BAM has recently drawn up and published a Finite Element Guideline. Secondly, actual technical questions are discussed: these are of wide interest and range from mechanical reflections on cask drop and extreme impact scenarios to thermal reflections on the removal of decay heat and fire scenarios. Examples from BAM's work on finite-element development activities are given to demonstrate the great opportunities as well as the difficulties of using finite-element methods for package safety analysis and design testing.     

Leakage potential of LWR containment penetrations under severe accident conditions

This paper is an overview of a Sandia National Laboratories, Albuquerque (SNLA) study of the performance of mechanical penetrations in light-water reactor (LWR) containment buildings that are subjected to severe accident environments. The study is concerned with modes of failure as well as the magnitude of leakage. The following tests have been completed, are under way, or are planned: (a) seals and gaskets have been tested to register the effects of radiation aging, thermal aging, seal geometry, and squeeze on seal and gasket materials in severe accident environments; (b) the performance of a full-scale airlock will be evaluated at severe accident temperature and pressure levels; (c) personnel airlock and equipment hatch tests were made on a model of a steel containment building; and (d) tests of mechanical penetrations are planned as part of a test on a model of a reinforced concrete building. This program is part of an overall US Nuclear Regulatory Commission (USNRC) effort to evaluate the integrity of LWR containment buildings.     

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.     

Numerical Assessment of Spent Fuel Casks Impacting on Real Targets

Abstract

A reference container of high capacity was analysed for loads beyond those it has to withstand during a 9 m IAEA drop test onto an unyielding target. In doing this a lid-end drop with shock absorber onto a real target was simulated. This is a possible accident for the rail transport of such casks. In this case the most critical components of the containment system are the primary lid bolts. The behaviour of the lid system and its sealing function were investigated with finite element (FE) analysis. To correlate the findings with a corresponding impact velocity onto real targets an analytical method was used. Despite the conservative assumptions made in this study a two-fold safety factor compared to the 9 m drop tests onto the unyielding target could be shown. The quantification of the additional safety the cask might provide requires further basic investigations on the behaviour of the real targets considered as well as the reduction of the conservatism included in the assumptions made up to now.     

主螺栓断裂对压力容器密封性能、应力及疲劳的影响分析

在理论分析和数值仿真技术基础上,研究并提出了一种主螺栓断裂对反应堆压力容器（RPV）密封性能、螺栓应力及疲劳的影响分析方法,采用该方法对主螺栓断裂影响进行了评价分析。结果表明,该方法适用于分析1根或多根主螺栓断裂情况对压力容器安全性能的影响,可以用于核电厂在运行中发生类似问题时判断反应堆能否继续运行。      

主螺栓断裂对压力容器密封性能、应力及疲劳的影响分析

Base on the mechanics theory and numerical simulation technique, a method to analyze the effect of the main bolt break on the stress, fatigue and seal is studied in this paper, and is adopted to evaluate and analyze the fracture influence of main bolt. The results show that this method is applicable for the analysis of the RPV safety performance induced by one bolt break or several bolts break accident, and for the determination if the nuclear reactor can be operated when similar problems occur.     

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.     

Plans for the future for the Packaging,Transport, Storage and Security of Radioactive Material Journal

Abstract

The regulatory driven design of radioactive material transportation packages leads package vendors to perform analyses that demonstrate the ability of packages to meet the regulatory requirements. For risk assessment and communication, the analysis of package response to thermal environments that are more severe than those described in the regulations is required. In general, experimental and analytical assessments of casks exposed to thermal insults other than the regulatory environment are performed in the USA by the Department of Energy national laboratories. This paper provides a brief summary of some recent thermal analyses of spent fuel transportation packages exposed to thermal environments different from regulatory standards. The analyses were performed by Sandia National Laboratories under several different projects for multiple customers. These analyses examined the response of spent fuel packages exposed to severe thermal environments different from the regulatory hypothetical accident condition. One assessment determined the response of four generic casks to very long duration engulfing fires. The results from these analyses included fire durations necessary to reach critical temperatures of the fuel and seals. In another assessment, two certified spent fuel casks were analysed for exposure to 1 h pool fires. The height of the cask above the pool was varied to study the effect of the vapour dome on the heating of the casks. Another assessment investigated the effect of offset long duration fires on rail cask performance, which showed that casks can withstand offset fires of much longer duration than the regulatory fire. Other assessments examined the response of packages to thermal environments resulting from propane fires and realistic liquid hydrocarbon fires that included various positions of the transportation rail car in the simulation.     

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.     

Transport Package Response to Severe Thermal Events,Part 2: Legal Weight Truck Cask

Abstract

The response of intact and damaged versions of the GA-4 Legal Weight Truck Cask to a range of severe thermal events is simulated using finite element computer analysis. The minimum fire durations that cause the containment seals and fuel cladding to reach their respective temperature limits are evaluated for a range of hydrocarbon fire temperatures. Containment seals reach their temperature limit in shorter duration fires as compared to the cladding, for both an undamaged package and a cask whose impact limiter is destroyed moments before the fire begins. However, if the neutron shield is destroyed, the cladding reaches its limit first in high temperature fires. A margin of safety exists between the conditions of the IAEA regulatory fire test and all of the performance envelopes calculated in this work.