Free drop impact analysis of shipping cask

The WHAMS-2D and WHAMS-3D codes were used to analyze the dynamic response of the RAS/TREAT shielded shipping cask subjected to transient leadings for the purpose of assessing potential damage to the various components that comprise the cask. The paper describes how these codes can be used to provide an intermediate level of detail between full three dimensional finite element calculations and hand calculations which are cost effective for design purposes.Three free drops were addressed: (1) a thirty foot axialdrop on either end; (2) a thirty foot oblique angle drop with cask having several different orientations from the vertical with impact on the cask corner; and (3) a thirty foot side drop with simultaneous impact on the lifting trunnion and the bottom end. Results are presented for two models of the side and oblique angle drops; one model includes only the mass of the lapped sleeves of depleted uranium (DU) while the other includes the mass and stiffness of the DU. The results of the end drop aalyses are given for models with and without imperfections in the cask. Comparison of the analyses to hand calculations and simplified analyses are given.     

Topical BAM cask design evaluation using drop tests and numerical calculations: accidental cask drop without impact limiters onto storage building foundation

Abstract

Cask impacts without impact limiters onto unyielding targets result in totally different mechanical reactions from those of relatively smooth impacts using impact limiters. During the licensing procedure of the new GNS CASTOR HAW 28M design for vitrified high activity waste, BAM therefore decided to perform an additional drop test with a 1 : 2 scale test cask (CASTOR HAW/TB2). In spite of a small drop height of only 0˙3 m onto the unyielding target of the BAM drop test facility, which conservatively covers any storage building foundation, the impact caused considerable stresses to the cask structure with high stress and strain rates. This paper presents the evaluation strategy of BAM including the drop test results and the development and qualification of appropriate finite element modelling to achieve sufficient agreement between test and calculation results. Further steps include mechanical analyses of reduced and full scale cask designs to determine the most critically stressed areas of the structure, verify scaling factors and demonstrate safety with respect to cask integrity and tightness.     

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.     

Finite element mesh design of a cylindrical cask under puncture drop test conditions

Abstract

Transport casks for radioactive materials have to withstand the 9 m drop test, 1 m puncture drop test and dynamic crush test with regard to the mechanical requirements according to the IAEA regulations. The safety assessment of the package can be carried out on the basis of experimental investigations with prototypes or models of appropriate scale, calculations, by reference to previous satisfactory safety demonstrations of a sufficiently similar nature or a combination of these methods. Computational methods are increasingly used for the assessment of mechanical test scenarios. However, it must be guaranteed that the calculation methods provide reliable results. Important quality assurance measures at the Federal Institute for Materials Research and Testing are given concerning the preparation, run and evaluation of a numerical analysis with reference to the appropriate guidelines. Hence, a successful application of the finite element (FE) method requires a suitable mesh. An analysis of the 1 m puncture drop test using successively refined FE meshes was performed to find an acceptable mesh size and to study the mesh convergence using explicit dynamic FE codes. The FE model of the cask structure and the puncture bar is described. At the beginning a coarse mesh was created. Then this mesh was refined in two steps. In each step the size of the elements was bisected. The deformation of the mesh and the stresses were evaluated dependent on the mesh size. Finally, the results were extrapolated to an infinite fine mesh or the continuous body, respectively. The uncertainty of the numerical solution due to the discretisation of the continuous problem is given. A safety factor is discussed to account for the uncertainty.     

Finite element modelling of the one meter drop test on a steel bar for the CASTOR cask

This paper presents a numerical analysis of the 1 m drop test on a steel bar of a CASTOR AVR cask where the impact is in a region with cooling fins as well as in a region where the fins have been locally removed. The paper consists of two parts: (i) a parameter study with an overall objective to derive an analysis methodology and (ii) comparison with experimental data. The parameter study includes parameters that can not be, or were not, defined directly from the experimental data as well as parameters linked to the numerical procedures within the finite element procedure. The parameters are validated by their influence on the model responses and effort needed for the assessment of their appropriate values. Then the model with the “best” parameter set is verified against the experimental results. The agreement between experimental and simulation results are very good.     

Beneficial influence of the weldment rupture of cask impact limiter case on the impact absorbing behavior

In the cask impact limiter design, the functions of steel case should be evaluated for the protection of the filler materials and the impact energy absorption by the buckling deformation in both the fire and impact accidents. The objective of this paper is to evaluate the beneficial influence of the intermittent weldment of the cask impact limiter case on the cask impact behavior. This paper describes the test results for the joint strength evaluation of intermittent weldment and the cask impact analysis considering the weldment rupture of the impact limiter case. The weldment rupture of the impact limiter case causes to lose their constraining effect for the wood blocks, which are filled into the metal incasement between the case and the gussets. The weldment rupture of the impact limiter case reduces the impact force which acts on the impact target significantly in vertical and horizontal drop impacts. Therefore the beneficial effect of weldment rupture should be considered in the impact limiter design and the cask impact analysis.     

Numerical and experimental analysis of the impact of a nuclear spent fuel cask

This paper deals with the numerical and experimental analyses of a shell type shock absorber for a nuclear spent fuel cask. Nine-meter free drop tests performed on reduced scale models are described. The results are compared with numerical simulations performed with FEM computer codes, considering reduced scale models as well as the prototype. The paper shows the results of a similitude analysis, with which the data obtained by means of the reduced scale models can be extrapolated to the prototype. Small discrepancies were obtained using large-scale models (1:2 and 1:6), while small-scale models (1:12) did not give reliable results. A 1:9 scale model provided useful information with a less than 20% error.     

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.     

Full-scope simulation of a dry storage cask using computational fluid dynamics

A computational fluid dynamics (CFD) analysis of a TN24P cask was performed through a full-scope simulation using FLUENT. In order to establish the analysis methodology while minimizing the computational burden, the sensitivities of various parameters were investigated by constructing a small-scale model. The full-scale CFD predictions of the TN24P cask were compared with the experimental data and COBRA-SFS results. There was good agreement between the FLUENT predictions and the experimental data. FLUENT showed similar temperature predictions to COBRA-SFS, while there were deviations between FLUENT and COBRA-SFS in the velocity predictions. By conducting sensitivity studies for the application uncertainties using a full-scale simulation, it was found that the basket gap size was the most sensitive parameter in the analysis.     

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.     

Structural analysis of the basket for a CASTOR transport cask

The cask CASTOR 440/84 is designed to be used for transportation and storage of 84 spent PWR fuel elements from Soviet VVER 440 reactors. The fuel basket of the CASTOR 440/84 is subjected to the highest loading under type B test conditions, i.e. in the 9 m horizontal drop orientation. By quasistatic calculations using the FEM code ANSYS, the maximum stresses for a deceleration of 82g have been calculated and it has been proven that the criticality safety of the fuel arrangement is guaranteed under these severe conditions. Additional calculations for an even higher deceleration of 115g demonstrate sufficient safety reserves of the design.     

A study on technique to estimate impact location of loose part using Wigner-Ville distribution

Presented in this paper is a method to estimate impact location of a loose part using the Wigner-Ville distribution. The method uses dispersion characteristics of bending waves propagated in a plate. The power propagation velocity and arrival time difference of bending waves related to the dispersion characteristics can be obtained through the transformation of impact signals using the Wigner-Ville distribution. The distance from the impact location to the signal measuring point can be estimated using the information on the power propagation velocity and the arrival time difference of two bending waves. The experimental results show that the proposed method estimates the impact location with relative percentage error within 10% compared with the actual impact location.     

A mathematical model of a control rod drop

A mathematical model of a control rod drop for a pressurized water reactor is presented. A hybrid computer program is used to solve the non-linear differential equations which describe the model. The computer solution is briefly described. The rod drop time computed by using the model is compared with available data from experimental results for a rack and pinion control system. Additional applications of the mathematical model are described and presented.     

A study on the method of impact mass estimation of loose parts

It is a very difficult problem to realize the mass estimation of loose parts in the mechanical equipment. The result of mass estimation will influence the fault diagnosing of equipment, especially in the loose part monitoring system of nuclear power station which can provide important guidance for the type classification of loose parts. This paper is based on experiments, by wavelet energy spectrum method to make estimation for different impact mass, and by using linear interpolation method to establish the scale peak function. The results show that the method has characteristics of small estimation errors and good consistency, strong anti-interference capacity, and it has better actual application value.     

A benchmark exercise on the use of CFD codes for containment issues using best practice guidelines: A computational challenge

In the framework of the 5th EU-FWP project ECORA the capabilities of CFD software packages for simulating flows in the containment of nuclear reactors was evaluated. Four codes were assessed using two basic tests in the PANDA facility addressing the transport of gases in a multi-compartment geometry. The assessment included a first attempt to use Best Practice Guidelines (BPGs) for the analysis of long, large-scale, transient problems. Due to the large computational overhead of the analysis, the BPGs could not fully be applied. It was thus concluded that the application of the BPGs to full containment analysis is out of reach with the currently available computer power. On the other hand, CFD codes used with a sufficiently detailed mesh seem to be capable to give reliable answers on issues relevant for containment simulation using standard two-equation turbulence models. Development on turbulence models is constantly ongoing. If it turns out that advanced (and more computationally intensive) turbulence models may not be needed, the use of the BPGs for ‘certified’ simulations could become feasible within a relatively short time.     

退役核燃料干式贮存混泥土护箱受高温作用之效应研究

退役核燃料干式贮存设施主体由混凝土构成,混凝土得在长时期内承受残余核燃料释出的衰变热,加上台湾地区特殊的环境气候条件,混凝土材料可能产生劣化.依据核能安全混凝土结构物的材料规定的配比,我们制作了混凝土试样,用实验室模拟法研究干式贮存混凝土护箱在高温环境作用下可能出现的损害或劣化,甚至耐久性变差等.利用非破坏性检测方法(...     

A comparison of recent results from HONDO III with the JSME nuclear shipping cask benchmark calculations

The results of two calculations related to the impact response of spent nuclear fuel shipping casks are compared to the benchmark results reported in a recent study by the Japan Society of Mechanical Engineers Subcommittee on Structural Analysis of Nuclear Shipping Casks. Two idealized impacts are considered. The first calculation utilizes a right circular cylinder of lead subjected to a 9.0 m free fall onto a rigid target, while the second calculation utilizes a stainless steel clad cylinder of lead subjected to the same impact conditions. For the first problem, four calculations from graphical results presented in the original study have been singled out for comparison with HONDO III. The results from DYNA3D, STEALTH, PISCES, and ABAQUS are reproduced. In the second problem, the results from four separate computer programs in the original study, ABAQUS, ANSYS, MARC, and PISCES, are used and compared with HONDO III. The current version of HONDO III contains a fully automated implementation of the explicit-explicit partitioning procedure for the central difference method time integration which results in a reduction of computational effort by a factor in excess of 5. The results reported here further support the conclusion of the original study that the explicit time integration schemes with automated time incrementation are effective and efficient techniques for computing the transient dynamic response of nuclear fuel shipping casks subject to impact loading.     

非能动加速的驱动线落棒特性仿真研究

模块式浮动反应堆在倾斜和摇摆的海洋环境条件下,驱动线随时面临比竖直固定条件下更加恶劣的落棒环境。为了保证控制棒的落棒时间满足安全要求,设计上考虑在驱动线上增设弹簧,实现非能动加速落棒。在现有驱动线落棒仿真模型的基础上,考虑弹簧瞬态作用力的影响,对驱动线的落棒行为进行仿真研究,得到竖直和倾斜状态下控制棒落棒过程中的位移、速度、水力作用力随时间变化的曲线。结果表明:弹簧对控制棒落棒时间有一定的改善作用,有利于克服倾斜和摇摆条件下的落棒阻力。