Testing of large-scale concrete containment structural elements

The tests described in this paper are part of an Electric Power Research Institute (EPRI) program (Research Project 2172-2) to provide a test-verified analytical method of estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents.Experimental study in Phase 2 of the investigation, on which this paper is based, includes tests of five large-scale specimens with steel liner plates representing structural elements of prestressed concrete containment buildings. Four square wall element specimens and one specimen representing the wall/basemat junction region were tested.This experimental work indicates that under internal overpressurization or other accident conditions, highly localized strains in the steel liner plate can result in liner tearing and subsequent containment leakage. These results support the theory of leak before break where liner tearing occurs in a controlled manner and leakage and depressurization occur rather than global failure.     

Tension tests of concrete containment wall elements

Tension tests of concrete containment wall elements were conducted as part of a three-phase research program sponsored by the Electric Power Research Institute (EPRI). The objective of the EPRI experimental/analytical program is twofold. The first objective is to provide the utility industry with a test-verified analytical method for making realistic estimates of actual capacities of reinforced and prestressed concrete containments under internal over-pressurization from postulated degraded core accidents. The second objective is to determine qualitative and quantitative leak rate characteristics of typical containment cross-sections with and without penetrations. This paper covers the experimental portion the the EPRI program.The testing program for Phase 1 included eight large-scale specimens representing elements from the wall of a containment. Each specimen was 60-in (1525-mm) square, 24-in (610-mm) thick, and had full-size reinforcing bars. Six specimens were representative of prototypical reinforced concrete containment designs. The remaining two specimens represented prototypical prestressed containment designs.Various reinforcement configurations and loading arrangements resulted in data that permit comparisons of the effects of controlled variables on cracking and subsequent concrete/reinforcement/liner interaction in containment elements.Subtle differences, due to variations in reinforcement patterns and load applications among the eight specimens, are being used to benchmark the codes being developed in the analytical portion of the EPRI program.Phases 2 and 3 of the test program will examine leak rate characteristics and failure mechanisms at penetrations and structural discontinuities.     

Mechanical and leakrate predictions for nuclear power plant containments in accidental conditions

In 1996, EDF decided to build a containment model at the scale 1:3, the Maquette echange vapeur/air (MAEVA) mock-up, in order to check and study the behavior of a prestressed concrete containment vessel without liner in terms of mechanical strength and leaktightness, for loadings corresponding to its design and beyond design conditions. In parallel to the construction and testing of the mock-up, predictive calculations of the mechanical and leaktightness behaviour of the mock-up were performed in the framework of a cost shared R&D action supported by the European Union, the Containment Evaluation under Severe Accidents (CESA) project. The strategy of EDF concerning the R&D performed on leaktightness of concrete and concrete structures is first explained and the goal of two interesting programs is shortly presented and discussed in the first part of the paper. Then, the emphasis is made on the predictive calculations performed on the MAEVA mock-up by means of finite elements (FE) calculations. A summary of the main achievements is then given and the interest of FE calculations is discussed for describing both the mechanical and leaktightness behaviour of a concrete structure as a function of crack development.     

Current state of knowledge on the behavior of steel liners in concrete containments subjected to overpressurization loads

In the US, concrete containment buildings for commercial nuclear power plants have steel liners that act as the internal pressure boundary. The liner abuts the concrete, acting as the interior concrete form. The liner is attached to the concrete by either studs or by a continuous structural shape (such as a T-section or channel) that is either continuously or intermittently welded to the liner. Studs are commonly used in reinforced concrete containments, while prestressed containments utilize a structural element as the anchorage. The practice in some countries follows the US practice, while in other countries the containment does not have a steel liner. In this latter case, there is a true double containment, and the annular region between the two containments is vented.This paper will review the practice of design of the liner system prior to the consideration of severe accident loads (overpressurization loads beyond the design conditions).An overpressurization test of a 1:6 scale reinforced concrete containment at Sandia National Laboratories resulted in a failure mechanism in the liner that was not fully anticipated. Post-test analyses and experiments have been conducted to understand the failure better. This work and the activities that followed the test are reviewed. Areas in which additional research should be conducted are given.     

Mechanical simulations of SANDIA II tests OECD ISP 48 benchmark

This paper illustrates the work carried out by EDF within the framework of ISP 48 post-test analysis of NUPEC/NRC 1:4-scale model of a prestressed pressure containment vessel of a nuclear power plant [Hessheimer, M.F., Klamerus, E.W., Rightly, G.S., Lambert, L.D., Dameron, R.A., 2003. Overpressurization test of a 1:4-scale prestressed concrete containment vessel model. NUREG/CR-6810, SAND2003-0840P. Sandia National Laboratories, Albuquerque, NM]. EDF as a participant of the International Standard Problem no. 48 [Mathet, E., Hessheimer, M., Ali, S., Tegeler, B., 2005. An international standard problem: analysis of 1:4-scale prestressed concrete containment vessel model under severe accident conditions. Proceedings of 18th International Conference on Structural Mechanics in Reactor Technology. SMiRT 18. Beijing, China] has participated in this program, within the framework of its research and development program on the simulation of non-linear behaviour of nuclear power plant prestressed concrete pressure containment vessels. EDF performed several simulations to determine the ultimate response of the scale model. To determine the most influent parameters in such an analysis several studies were carried out. A full 3D mesh of the entire structure was then created. The mesh was built using a parametric tool to measure the influence of discretization on results. To represent the cracking of concrete, two material laws were then used. The purpose of this paper is to illustrate the ultimate behaviour of SANDIA II model obtained by Code_Aster^® finite element platform, with comparison to tests records, and also to share the lessons learned from the parametric computations and underline precautions that must be taken in such studies.     

Results of large-scale test of discontinuity region in a prestressed concrete containment

The test described in this paper is part of an Electric Power Research Institute (EPRI) program (Research Program RP2172-2) to provide a test-verified analytical method of estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents.Phase 2 of the EPRI program, on which this paper is based, includes tests of five large-scale specimens with steel liner plates. The specimens represent structural elements of prestressed concrete containment buildings. Four full-scale square wall element specimens and one specimen representing the wall/basemat junction region were tested. This paper describes results of the wall/basemat junction region test.Results of this experimental work indicate that highly localized strains in the steel liner plate caused by internal overpressurization or other accident conditions can result in liner tearing and subsequent containment leakage. It appears that this liner tearing occurs in a controller manner. Extrapolating from these test results, leakage and depressurization is more likely to occur than global failure.     

Analytic study of the steel liner near the equipment hatch in a 1:4 scale containment model

A containment scale-model test, performed at Sandia National Laboratories, was loaded by overpressurization and the first leak was concluded to be caused by tears in the steel liner found near the equipment hatch. These tears were located in the vicinity of the vertical fold in between the general curved part and the embossment (vertical bend line). A 3D finite element analysis of the region near the equipment hatch, shows that high localized strains will develop in the vicinity of the bend line. It is shown that the liner separates from the concrete wall near the bend line when the containment expands. The tensioned liner will be in contact with the surface of the concrete wall in general, but near the vertical bend line the liner tends to be straightened out. This flexural behaviour cause high strains in the weld located in the bend line. The actual peak strain level is depending on the detailed geometry in the bend line and the failure strain level of a welded biaxial stressed zone is difficult to define. However, the analysis presented in this paper shows that the flexural behaviour in the bend line most likely contributed to the liner tears found in the scale-model test. A general conclusion from the study presented in this paper is that, the non-linear plastic behaviour of the liner is very sensitive to the detailed design and the interaction between the liner and the concrete.     

秦山核电厂预应力安全壳的结构设计

本文介绍了秦山核电厂安全壳的设计概况,作为后张法预应力壳体结构,本文从选型考虑到结构概貌、设计依据、应力分析以及预应力钢束及钢衬里的设计考虑等各个重要环节都作了较为详细的叙述。它不仅对核结构的设计具有较大的参考价值,而且对某些预应力筒仓,贮罐和水池结构等的设计也有一定的参考价值。     

Concept study for a combined reinforced concrete containment

A variety of different types of steel and concrete containments have been designed and constructed in the past. Most of the concrete containments had been pre-stressed, offering the advantage of small displacements and a certain leak-tightness of the concrete itself. However, considerable stresses in concrete as well as in the tendons have to be maintained during the whole lifetime of the plant in order to guarantee the required pre-stressing. The long-time behaviour and the ductility in the case of beyond-design-load cases must be verified. Contrary to a pre-stressed containment a reinforced containment will only be significantly loaded during test conditions or when needed in case of an accident. It offers additional margins which can be used especially for dynamic loads such as impacts or for beyond-design events.The aim of this paper is to show the feasibility of a so-called combined containment which means a containment capable of resisting both severe internal accidents and external hazards, mainly the aircraft crash impact as considered in the design of nuclear power plants in Germany.The concept is based on a lined reinforced containment without pre-stressing. The mechanical resistance function is provided by the reinforced concrete and the leak-tightness function is provided by a so-called composite liner made of non-metallic materials. Some results of tests performed at Siemens laboratories and at the University of Karlsruhe which show the capability of a composite liner to bridge over cracks at the concrete surface will be presented in the paper.The study shows that the combined reinforced concrete containment with a composite liner offers a robust concept with high flexibility with respect to load requirements, beyond-design events and geometrical shaping (arrangement of openings, an integration of adjacent structures). The concept may be further optimized by partial pre-stressing at areas of high concentration of stresses such as at transition zones or at disturbances around large openings.     

SNR 300 fast breeder reactor: steel containment - design, erection, testing

The Kalkar Nuclear Power Plant which is equipped with an 300 MW fast breeder reactor is being built by a Consortium mainly comprising German, Belgian and Dutch companies.The components of the fast breeder reactor are enclosed in a concrete containment which is designed to withstand severe external and internal loading.The concrete enclosure is surrounded by a steel containment which is designed to prevent the release of radioactivity following a postulated accident involving the nuclear components inside the concrete containment.The paper describes the solutions adopted for the different parts of the steel containment, the calculations verifying the suitability of the designs, the erection and the steel containment pressure and leak tests. The tests were performed with successful results in 1984.     

某核电厂LOCA下预应力混凝土安全壳响应规律初探

核电厂LOCA发生后,预应力混凝土安全壳结构内温度场分布具有明显的非线性特征,但现行的混凝土安全壳设计规范未对LOCA下温度和应力的组合作用提出具体的计算方法。基于用ANSYS程序建立的包含预应力钢束的混凝土安全壳结构的有限元模型,本文计算了LOCA下不同时刻安全壳壳壁内的温度场分布,并与理论值进行了比较,验证了计算模型的正确性。初步分析了高温、高压作用下安全壳结构变形的规律,总结了混凝土温度效应和预应力系统的作用,可为安全壳结构设计提供参考。     

核电厂严重事故后安全壳压力的测量方法

在核电厂设计早期,安全壳大气监测系统仅考虑了设计基准事故。而与设计基准事故相比,在严重事故工况下的安全壳内压力会有较大增长,现有的安全壳压力测量仪表不能满足严重事故工况下对安全壳压力的监测。为采取有效的事故缓解对策,需考虑严重事故下的安全壳压力监视措施。目前的技术条件下,在安全壳外增设一个安全壳压力测量通道用于严重事故后的安全壳压力测量是一可考虑的方案。大亚湾核电厂实施了这种改进。通过此改进,可推迟严重事故时安全壳的排放时间,提高核电厂的安全水平。经论证,这种方案是安全和可行的。     

严重事故下预应力混凝土安全壳非线性分析及性能评估

预应力混凝土安全壳作为核电厂重要防泄漏屏障,对保证核电厂正常运行、确保人员安全至关重要。本文基于顺序热力耦合方法对严重事故工况下预应力混凝土安全壳进行非线性有限元分析,考虑了温度和内压荷载共同的影响,分析了安全壳结构典型不连续区域和连续区域的位移响应。研究表明:安全壳混凝土不连续区域位移响应沿厚度方向上差异较为显著,而连续区域处的差异相对较小;安全壳泄漏失效模式由设备闸门位置控制,50%和95%分位水平的内压分别为1.266 MPa和1.072 MPa;破口失效模式由筒体某一位置控制,50%和95%分位水平的内压分别为2.224 MPa和1.883 MPa;本文所分析的预应力混凝土安全壳的内压承载力满足最小安全裕度不小于2.5的要求。      

Thermal-hydraulic and stress analysis of AP1000 reactor containment during LOCA in dry cooling mode

Some kinds of break in the reactor coolant system may cause the coolant to exit rapidly from the failure site,which leads to the loss of coolant accident (LOCA).In this paper,a stress analysis of an AP1000 reactor containment is performed in an LOCA,with the passive containment cooling system (PCCS) being available and not available for cooling the wall's containment.The variations in the mechanical properties of the wall's containment,including elastic modulus,strength,and stress,are analyzed using the ABAQUS code.A general two-phase model is applied for modeling thermal-hydraulic behavior inside the containment.Obtained pressure and temperature from thermal-hydraulic models are considered as boundary conditions of the ABAQUS code to obtain distributions of temperature and stress across steel shell of the containment in the accident.The results indicate that if the PCCS fails,the peak pressure inside the containment exceeds the design value.However,the stress would still be lower than the yield stress value,and no risk would threaten the integrity of the containment.     

Analytical correlation and interpretation of full scale containment specimen tests

The work presented in this paper is part of an EPRI-sponsored research program to develop experimentally verified methodology for predicting failure modes and leakage characteristics of concrete containments. This paper deals specifically with recent results of the analytical correlation and interpretation of full scale containment specimen tests. The tests under consideration are a wall/skirt-basemat specimen of a typical prestressed concrete containment, a specimen with a flawed liner to study liner crack growth, and a specimen with a typical steampipe penetration. Computational models of specimens are described, and pre-test and post-test analysis results are presented. The importance of local effects is discussed, and the role of specimen tests and analysis in failure prediction of containment structures is summarized.     

The Austrian PCRV project with a hot liner — A status report

The analysis of the problem PCRV safety leads to a new vessel concept in which the pressure transmitting insulation with its cooling system is outside the hot linear. A control and adjustment system for vessel wall temperature ensures that, under all working conditions, the liner suffers only elastic compression. Outside the thermal insulation (special concrete) a cold steel barrier is placed, which can be constructed as a second liner, so that leaks are limited and can be detected and evacuated. After preliminary development work in the field of high temperature insulating concrete, prestressed concrete technology and instrumentation at elevated temperature, a large experimental ring has been constructed and tested extensively. As part of an experimental high temperature gas loop a large scale model vessel is being built at the Research Center, Seibersdorf, Austria. Construction is nearly complete and a pressure test is scheduled for early 1975. It will be followed by tests under working conditions of PWRs and later of HTGRs. The reference design of such a PCRV with a hot liner for a 1500 MW(e) pressurized water reactor will be finished at the same time as the model vessel.     

核电厂新型预应力混凝土安全壳及其非能动冷却系统设计与分析

本文以CAP1700核电厂为例,提出了一种新的核电厂预应力混凝土安全壳及其非能动冷却方案,介绍了新式非能动安全壳冷却系统热工水力计算方法,并给出事故工况下新非能动安全壳冷却系统的运行参数。结果表明,CAP1700非能动安全壳冷却系统的设计是可行的,能满足事故工况下的冷却需求。贮水箱水量有很大的裕量,可通过计算进一步优化贮水量。     

混凝土安全壳整体性能试验峰值压力持续时间探讨

核电站建造阶段必须进行安全壳整体性能试验（CTT）,验证在设计基准事故时安全壳结构的完整性。本文针对某核电厂3号机组预应力混凝土安全壳CTT进行非线性有限元分析。结果表明：筒体闸门洞口标高附近径向变形最大,预应力钢束承担了峰值压力0.483 MPa作用下大部分设计内压,安全壳整体结构处于受压状态,与实际试验状态基本吻合。同时,对国内外法规标准关于安全壳峰值压力持续时间的规定进行总结,提出相关结论及建议,可为安全壳CTT方案设计提供参考。     

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.     

Small-scale component experiments of the penetration leak characterization test in the ALPHA program

A small-scale penetration leak characterization test has been performed as a part of the ALPHA program at Japan Atomic Energy Research Institute (JAERI). Two series of experiments were performed using test sections which simulate relevant parts of an EPA (Electrical Penetration Assembly) used in Japanese PWR containments. One of the test sections simulates an alumina module and the other includes the silicone resin portion of the EPA. The test section was heated in a leak test vessel which simulated thermal-hydraulic conditions inside and outside of the containment in a severe accident. From the experimental results, it was concluded that although the silicone resin may melt at high temperature, the alumina module will remain intact under severe accident conditions. The EPA as a whole is estimated to maintain leak-tightness during a severe accident. It was found in the experiments that heat conduction along the metal portion of the test section had a strong influence on the melt progression of the resin. It was also found that the measured strain of the alumina module was predominantly caused by the elevated temperature. Therefore, the thermal load will be more of a threat to the EPA's integrity rather than the pressure load.