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.     

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.     

Use of impact analysis in licensing industrial container for transporting new fuel

Abstract

The finite element (FE) method is a powerful tool for the simulation of mechanical and thermal behaviour of structures. In recent years, the explicit FE method has increasingly been used in the development of transport packages and as part of approval applications to demonstrate the performance of packages. Testing and analysis are the two methods specified in the IAEA Regulations for the Safe Transport of Radioactive Material for demonstrating the structural and thermal performance of a transport package against the requirements of the Transport Regulations. The roles of testing and analysis, and the relative prominence of the two, may vary between Competent Authorities in different countries. This can range from analysis being regarded as the primary mode of demonstration with testing as confirmatory, to testing being the primary mode of demonstration supplemented by analysis. This paper describes the use of the non-linear FE code LS-DYNA in the licensing of a new container for the transport of new nuclear fuel. The package was classified as an Industrial Package (Fissile) in accordance with the IAEA Regulations, and hence it was necessary, among other things, to demonstrate that criticality criteria were satisfied under postulated impact conditions. Physical drop tests were carried out and the results are compared with LS-DYNA computer calculations using the same FE models developed to support the design of the new container. The analyses and tests clearly demonstrate the novel use of polyurethane foam as the container main energy absorber. The FE predictions are compared for accelerations, bolt loadings and global deformations of the container. In general good correlation was obtained between predictions and tests and the differences, which did occur, particularly for accelerations, are discussed and reconciled. The paper concludes that explicit analysis codes are now so reliable for container impact calculations that minimal test work should be pursued basically for key confirmatory impact scenarios.     

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.     

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.     

Mechanical characteristics of fuel rod claddings in transport conditions

Abstract

A synthesis on the mechanical characteristics of unirradiated and irradiated fuel rod claddings was performed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to have reference data for the assessment of the safety demonstrations in normal and accident conditions of transport required by the procedure of package licensing. Indeed, the transport conditions correspond to a range of cladding temperatures (200–550°C) which is only partly covered by the data acquired within the framework of the safety demonstration relative to the reactor normal operating conditions, especially beyond 400°C. This work concerned Zircaloy-4 cladding material (Zry-4) and M5^TM. Data about mechanical properties (elastic and ductile properties, creep behaviour), oxidation (in reactor and under air during transport), hydrides and fracture toughness have been collected and synthesised. The laws presented in the document can be used to obtain orders of magnitude of oxide layer thickness, hydrogen content and creep deformation rate. The following phenomena which could influence the mechanical behaviour of the cladding were more particularly studied: oxidation which could become very important during transport in case of cladding temperatures of ～500°C; creep for which only a few data ～500°C are available and which depends in particular on the internal pressure of the rods, the cladding oxidation and the presence of the hydrides; and recrystallisation of Zry-4 at ～500°C, which could have consequences on the mechanical properties of the cladding after cooling during the storage. For other topics of interest for the study of the mechanical behaviour of the cladding, such as the fracture toughness for example, it was identified that the data available is scarce.     

Multiple water barriers: alternative to assessment of fuel assemblies during accident conditions of transport

Abstract

Packages for the transport of radioactive material have to comply with national and/or international regulations. These regulations are widely based on the requirements set forth by the International Atomic Energy Agency (IAEA) in the 'Regulations for the safe transport of radioactive material'. In this framework, packages to transport fuel assemblies (including spent fuel assemblies) have to meet the requirements for packages containing fissile material. In accident conditions of transport, the applicant for the package design approval has to show that the package remains subcritical taking due account of the status of the contents in these conditions. In most cases, considering water ingress in the package, it is not possible to assume that the fissile material included in the fuel assemblies is dispersed in the package with the most severe conceivable distribution regarding criticality. In order to alleviate this difficulty, during the last years, we have provided a significant better knowledge of the conditions of the fuel assemblies to be transported. This was part of the Fuel Integrity Project, whose progress was regularly reported during PATRAM 2001 and PATRAM 2004 Symposia. However, for packages which encounter a large g-load during accident conditions of transport and/or which contain spent fuel assemblies with very high burn-up, it can be difficult to demonstrate that the fuel assemblies are not significantly damaged. Then, to make the criticality assessment considering water inleakage into the flask and a large release of fissile material within its cavity will not allow meeting the subcriticality criteria. For that reason, for our package designs, which use a gas and not water as an internal coolant and which fall into that category, the author has decided to take credit of the possibilities provided by the subparagraph 677 (b) of the Regulations. This paragraph allows not taking into account water in the package, provided that the package exhibits 'multiple high standard water barriers'. The paper describes the author's experience with the implementation of this paragraph. Two different cases are considered: either a double vessel, or a double lid. It will be explained when each of these solutions is implemented, and give examples of package designs with such features, as well as the approvals which were granted for these designs in various countries.     

Criticality risks during transportation of spent nuclear fuel

Abstract

A probabilistic risk assessment (PRA) quantifies the frequency of criticality accidents during railroad transport of spent nuclear fuel casks (SFCs) in the USA. It evaluates the likelihood that undetected errors in fuel selection and/or fuel handling could result in a misloaded SFC susceptible to a criticality event following an accident during rail transport of the cask. The PRA shows that existing fuel burnup records and formal procedures for loading a SFC make the likelihood of shipping a misloaded SFC on the order of 2·6 × 10^–6 per SFC. When combined with historical evidence regarding train accidents and an estimate of the likelihood that an accident could breach and submerge a SFC, the calculated frequency of criticality is below 2 × 10^–12 over the 11 000 shipments that would be required to ship the spent fuel inventory generated by the current US fleet of nuclear reactors, assuming that they each operate for 60 years.     

Buckling sensitivity analysis of cracked thin plates under membrane tension or compression loading

Thin-walled structural components such as plates and shells are commonly used in practical applications such as aerospace, naval, nuclear power plant, pressure vessels, mechanical and civil engineering structures and so on, and the safety assessment of such structures must carefully consider all the phenomena which can decrease the bearing capacity of such elements. Among them, the presence of cracks in thin-walled structures can heavily affects their safety factor with respect to the more common modes of failure such as buckling or fracture. For very thin plate, buckling collapse under compression or even under tension, apart fracture or plastic failure in this last case, can easily take place: the presence of flaws such as through-the-thickness cracks can sensibly modify such ultimate loads. In the paper the effects of cracks’ length and orientation on the buckling loads of rectangular elastic thin-plates - characterised by different boundary conditions and by various Poisson's ratio - under tension and compression, is considered. For tensioned flawed plates a fracture-buckling and a plastic-buckling collapse maps are obtained. After a short explanation of the buckling phenomena in plates, several FE numerical parametric analyses results are presented in terms of critical load multiplier in compression or in tension in cracked plates. The obtained results are discussed and some interesting and useful conclusions regarding the sensitivity to cracks’ presence of buckling loads of thin plates under compression or tension (or fracture in this last case) are explained. The interesting case of tensioned cracked plates is considered by studying the easiest collapse between fracture, plastic flow and buckling: in such cases some failure-type maps are finally determined.     

Mechanical design assessment approaches of actual spent fuel and HLW transport package designs

Abstract

In recent years, BAM Federal Institute for Materials Research and Testing finalised the competent authority assessment of the mechanical and thermal package design in several German approval procedures of new spent fuel and high level waste package designs. The combination of computational methods and experimental investigations in conjunction with materials and cask components testing is the most common approach to mechanical safety assessment. The methodology in the field of safety analysis, including associated assessment criteria and procedures, has evolved rapidly over the last years. The design safety analysis must be based on a clear and comprehensive safety evaluation concept, including defined assessment criteria and constructional safety goals. In general, for new package designs, the implementation of experimental package drop tests in the approval process should be obligatory. Additionally, pre- and post-test calculations as well as components or material testing could be important. The extent to which drop tests are necessary depends on the individual package construction, the materials used and identified safety margins in the design.     

Characterization of crud deposited on fuel rods under HWC environment in Kuosheng Nuclear Power Plant

Under normal water chemistry conditions, the oxygen and hydrogen peroxide produced by water radiolysis in the coolant of boiling water reactors(BWRs) can lead to intergranular stress corrosion cracking in the constituent materials of plant components. This fact has led to the wide-scale adoption of hydrogen water chemistry(HWC) in the nuclear industry to counteract these effects.This study seeks to characterize the metallic composition and the surface properties of the constituent materials of plant components in order to determine their effects on the accumulation of chalk river unidentified deposits(crud) on fuel rods in the BWR Unit-1 of the Kuosheng Nuclear Power Plant in Taiwan. Inductively coupled plasma-atomic emission spectroscopy was used to calculate the concentrations of surface crud and gamma spectrometry was used to determine the radioactivity of the corrosion products, as well as their axial distribution across the surface of the fuel rods. X-ray diffraction analysis and scanning electron microscopy/energy-dispersive X-ray spectroscopy were used to identify the crystalline phase and morphology of the crud as irregular shapes and flakes. The amount of crud deposited during the fourth fuel cycle exceeded that of the third fuel cycle due to extended burn-up time. Our analytical results indicate that the implementation of HWC had no significant effect on the characteristics of subsequent crud.     

Row of fuel assemblies analysis under seismic loading: Modelling and experimental validation

The aim of this study was to develop a numerical model for predicting the impact behaviour at fuel assembly level of a whole reactor core under seismic loading conditions. This model was based on a porous medium approach accounting for the dynamics of both the fluid and structure, which interact. The fluid is studied in the whole reactor core domain and each fuel assembly is modelled in the form of a deformable porous medium with a nonlinear constitutive law. The contact between fuel assemblies is modelled in the form of elastic stops, so that the impact forces can be assessed. Simulations were performed to predict the dynamics of a six fuel assemblies row immersed in stagnant water and the whole apparatus was placed on a shaking table mimicking seismic loading conditions. The maximum values of the impact forces predicted by the model were in good agreement with the experimental data. A Proper Orthogonal Decomposition analysis was performed on the numerical data to analyse the mechanical behaviour of the fluid and structure more closely.     

Assessment for radiological impact of irradiated nuclear fuel package sunk in marine environment

Abstract

To gain the public acceptance for the transport of irradiated nuclear fuel, environmental impact assessments have been made by assuming that a transport package may be sunk into the sea. General circulation models for the Japan Sea and for the world ocean were employed to simulate radionuclide concentration in the ocean near shore and in deep sea areas respectively. Radionuclide concentrations were simulated for the hypothetical release from a submerged transport package of fresh mixed oxide fuel in the Japan Sea and the global ocean by general circulation models. Simulated concentrations were quite small compared to the background concentration by the fallout. The effective doses of radiation exposure to the public for both cases were much less than the effective dose limit (1 mSv year^–1) by the ICRP recommendation.     

WTRLGD—A computer program for the transient analysis of waterlogged fuel rods under the RIA condition

A computer code WTRLGD has been developed to describe the transient internal pressure of a waterlogged fuel rod during power burst and also to predict the possibility of the rod failure in the mode of cladding rupture. The code predicts transient thermal behavior of the fuel rod on the basis of an assumption of axisymmetry, and thermal-hydraulic transients of the internal water on the basis of a homogeneous volume-junction model modified so as to involve the cladding deformation. Calculated transients of the rod pressure are in fairly good agreement with those measured in the NSRR experiments, simulating the fuel rod behavior under an RIA condition. The comparison between calculation and experiment verifies that the code is an effective tool for the prediction of the failure of a waterlogged fuel rod.     

在线加钚模式下的熔盐堆钍铀燃料可持续性研究

基于FLiBe载体盐,Th/~(233)U启堆,仅通过在线添加反应堆级钚,以实现熔盐堆~(233)U的自持和焚烧反应堆级钚的能力。采用单栅元模型,分析其在不同熔盐体积比、不同中子损失率下233U的自持和钚的利用性能。研究发现:在熔盐体积占比为10%～85%的较大范围内都可以实现~(233)U自持,其中约43%熔盐体积比下~(233)U增殖效果最佳。与此同时,43%熔盐占比下对钚的依赖最大,在熔盐体积比较小和较大时对钚的依赖较小;在熔盐体积比较小时更有利于钚的利用,其中在熔盐体积比为10%～15%时钚的焚烧率最大,约为75%。此外,中子损失率与钚的依赖近似呈正比关系,对~(233)U自持性能影响较小。     

高燃耗下M5合金包壳燃料棒稳态和瞬态性能分析

修改并验证了分析程序FEMAXI-IVM,增加了程序的适用范围。对采用M5合金包壳的FA300-4高性能燃料组件中的燃料棒在稳态和瞬态运行工况下的燃料性能进行了分析。结果表明,此种燃料棒在稳态和瞬态工况下都能保持其完整性,能保证反应堆的安全运行。     

Modeling of gap conductance for LWR fuel rods applied in the BISON code

ABSTRACT

A new gap conductance model is proposed in this study as a combination of Toptan’s model and the Ross-Stoute model. A variance-based sensitivity analysis is performed to understand how simulation results depend on all input parameters of the proposed model. Additionally, new modeling options (e.g. fill gas thermal conductivity, temperature jump distance, thermal accommodation coefficient, etc.) are added into the nuclear fuel performance code, BISON. The need for further investigation of the gap heat transfer between fuel and cladding in BISON motivated this study to evaluate its impact on the code’s predictions. New gap conductance modeling is proposed. A series of integral-effects validation tests is performed: (1) to demonstrate the impact of the proposed model on the code’s fuel temperature predictions at the beginning of life and through the reactor’s life; (2) to ensure that the proposed model is capable of accurately modeling gap heat transfer characteristics in real-world problems; and (3) to investigate the impact of the estimation of fission gas release on the fuel temperature predictions with the proposed model. The results indicate that the proposed gap conductance model improves BISON’s predictions.     

含有可燃毒物的压水堆核电站堆芯装料设计优化研究

含可燃毒物的压水堆装料优化是燃料管理优化研究中的难点,应用通常的脱耦方法和优化算法效率低、全局性差。研究提出局部脱耦方法用以简化问题规模、缩小搜索空间,选择特征统计算法进行优化方案的搜索。利用局部脱耦方法结合特征统计算法研制出压水堆核电站堆芯LP和BP耦合装料优化程序CSALPBP。使用该程序对大亚湾第10循环和第12循环进行了装料优化计算。结果表明CSALPBP程序在求解含可燃毒物的压水堆装料优化问题方面具有很高的搜索效率和很好的全局性,能够较好地解决含可燃毒物的压水堆堆芯装料优化难题。