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.     

A simplified model for concrete at low confining pressures

The paper describes a model for the response of concrete that is subjected to essentially monotonic straining at low confining pressures. We assume that, under these conditions, the response of the concrete is dominated by cracking when the stress state is predominantly tensile, and by gross inelastic deformation under compressive stress. The model uses a “crack detection surface” in stress space to determine when cracking takes place and the orientation of the cracking at a point, together with a damaged elasticity approach to describe the post-failure behavior of the concrete with open cracks. A yield/flow surface (associated flow) model is used to define the concrete's response in compressive states of stress. The model is simple enough that it can be implemented so as to operate effectively in an implicit finite element code: modeling accuracy is sacrificed for this purpose. Preliminary studies with the model indicate that it can give useful predictions in cases of interest.     

Heat removal verification tests using concrete casks under normal condition

Heat removal tests using two types of full-scale concrete casks were conducted. This paper describes the results under a normal condition of spent fuel storage. In the tests, data on heat removal performance and integrity of cask components were obtained for different storage periods. The change of decay heat of spent fuel was simulated using electric heaters. Reinforced concrete cask (RC cask) and concrete filled steel cask (CFS cask) were the specimen casks. Decay heat at the initial period of storage 60 years of storage, the middle period (20 years of storage), and the final period (40 years of storage) correspond to 22.6 kW, 16 kW and 10 kW, respectively. Quantitative temperature data of the cask components were obtained as compared with their limit temperature. In addition, heat balance data requested for heat removal analyses were obtained.     

Finite element 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.     

Heat removal verification tests of full-scale concrete casks under accident conditions

Heat removal verification tests using two kinds of full-scale concrete casks under accident conditions were performed. One is reinforced concrete cask and the other is concrete filled steel cask. From the test results, their safety on heat removal performance under accident conditions was confirmed. Accident conditions for the tests were partial (50%) and complete (100%) blockage of the air inlets. Because the shape of air flow area in the concrete casks are different between two types of the cask, it was found that the change of the temperature distribution and air flow pattern were different for each accident condition.     

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(...     

Stress corrosion cracking of stainless-steel canister for concrete cask storage of spent fuel

Resistance to external stress corrosion cracking (ESCC) and crevice corrosion were examined for various candidate canister materials in the spent fuel dry storage condition using concrete casks. A constant load ESCC test was conducted on the candidate materials in air after deposition of simulated sea salt particles on the specimen gage section. Highly corrosion resistant stainless steels (SS), S31260 and S31254, did not fail for more than 46 000 h at 353 K with relative humidity of 35%, although the normal stainless steel, S30403 SS failed within 500 h by ESCC. Crevice corrosion potentials of S31260 and S31254 SS became larger than 0.9 V (SCE) in synthetic sea water at temperatures below 298 K, while those of S30403 and S31603 SS were less than 0 V (SCE) at the same temperature range. No rust was found on S31260 and S31254 SS specimens at temperatures below 298 K in the atmospheric corrosion test, which is consistent with the temperature dependency of crevice corrosion potential. From the test result, the critical temperature of atmospheric corrosion was estimated to be 293 K for both S31260 and S31254 SS. Utilizing the ESCC test result and the critical temperature, together with the weather station data and the estimated canister wall temperature, the integrity of canister was assessed from the view point of ESCC.     

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.     

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.     

A CFD analysis of thermal behaviour of transportation cask under fire test conditions

Thermal design of transportation cask for shipping radioactive waste needs strict compliance with the guidelines of the regulatory bodies. Lead shielding is usually provided in these casks for arresting gamma rays and reducing hazardous emissions to the environment below permissible limits. During transportation, accidental fire may break out and cause melting of lead for a prescribed duration. The present analysis reports, for the first time, a comprehensive CFD analysis of the thermal behaviour of melting of lead under high Rayleigh number convection during the fire test. The study reveals a substantial influence of natural convection on the thermal state and melting behaviour of lead which may have a great bearing on the safety and security of public during transportation of cask.     

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.     

Heat removal characteristics of vault storage system with cross flow for spent fuel

Since the amount of spent fuel to be stored is expected to steadily increase in Japan, a use of large-scale dry storage facilities is considered to be a promising method in practice at reasonable economic cost. The present study is concerned with the heat removal experiment of vault storage system adopting cross flow with passive cooling, using a 1/5 scale model. The results show that the flow pattern of air in the storage module strongly depends on the ratio of the buoyancy to the inertia force. A simple method to estimate air flow patterns in the storage module was proposed, where the Ri (Richardson) number was considered as the most representative parameter. Then the heat transfer rate from a storage tube to cooling air was estimated, which could apply to the design of a full-scale vault storage system with cross flow, in which dozens of storage tubes were placed. The acquired information was also used to optimize heat removal design of the vault storage system in the present study.     

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.     

A simplified model for the dynamics of a BWR

A model has been developed to derive the dynamic characteristics of a BWR with natural circulation. The model is based on the basic physical processes that govern reactor dynamics. The actual values for the model parameters are estimated from experimental and theoretical data. The model enables the computation of transfer functions of reactivity and steam flow to power and pressure. The sensitivity of these transfer functions to changes in model parameters is discussed.     

Construction of a concrete containment model

This paper discusses the features and construction of a reinforced-concrete containment model that has been built at Sandia National Laboratories in Albuquerque, New Mexico. The model Light-Water-Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc. The containment model will be tested to failure to determine its response to static internal overpressurization. The results from testing the heavily instrumented containment will be used to assess the capability of analytical methods for predicting the performance of containments subject to severe accident loads as part of the US Nuclear Regulatory Commission's program on containment integrity.The scaled dimensions of the cylindrical wall and hemispherical dome are typical of a full-size containment. Features representative of a prototypical containment and included in the heavily reinforced model are equipment hatches, personnel airlocks, several small piping penetrations, and a thin steel liner attached to the concrete by headed studs.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 1: Test model and pressure test

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses.     

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.     

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.