Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 2: Results of shaking table tests

A 1/8-scale model was constructed of a reinforced concrete containment vessel (RCCV) used in the latest advanced boiling water reactors (ABWR). Shaking table tests were conducted on it with input motions corresponding to or exceeding a design earthquake assumed for a real Nuclear Power Plant.The objectives of the tests were to verify the structural integrity and the leak-proof functional soundness of the RCCV subjected to design earthquakes, and to determine the ultimate strength and seismic margin by an excitation that led to the model's collapse. The model, the test sequence and the pressure and leak test results were addressed in Part 1. The shaking table test method, the input motions and the test results, including the transition of the model's stiffness, natural frequencies and damping factors and the effects of vertical input motions and internal pressure on the model's characteristics and behavior, the load–deformation, the ultimate strength, the failure mode of the reinforced concrete portion and the liner plate are described here. The seismic safety margin that was evaluated by the energy input during the failure test to a design basis earthquake will be described in Part 3. The analytical results of simulation using the multi-lumped mass model will be described in Part 4.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 3. Evaluation of seismic safety margin

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), had conducted a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate an 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, was used for this test. The test model and the results of pressure and leak tests are described in Part 1. 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–deformation relationship are described in Part 2. Part 3 reports the seismic design safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 will report simulation analysis results by a stick model with lumped masses.     

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.     

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.     

Ultimate strength of reinforce concrete shear walls under multi-axes seismic loads

“The model test on multi-axes loading on RC shear walls” had been carried out as for the 10-year project aiming at comprehension of the earthquake response behavior of three-dimensional (3D) reinforced concrete (RC) shear walls under the 3D of multi-axes earthquake loading condition. The motivation of the project building-up is that the current seismic design of nuclear power plant building is carried out by applying one-dimensional (1D) dynamic earthquake load to an analytical building model in each direction independently whereas actual earthquake jolts the building in the three directions simultaneously. Therefore, there were opinions requesting some testing confirm whether or not the current seismic design methodology is reliable for the input motions exceeding the design earthquake ground motion moreover for the input motions of the 3D. The project had completed with various valuable outcomes that can reply to the opinions. Moreover, the outcomes will play an important role in evaluating seismic margins of important structures in a nuclear power plant. In this paper, based on the published documents relating to this test project, the author describes a review of the whole testing and summarizes the major outcomes extracted by the test project.     

Experimental study on RCCV of ABWR plant

This research relates to the application of a reinforced concrete containment vessel (RCCV) whose form can be freely selected to the development of an advanced BWR. The researchers on RCCVs are carried out recently, and good results are obtained. The RCCV employed in this research is different from conventional ones in structural elements and form. The purpose of this research is, therefore, to confirm the adequacy and reliability of the design method of the trial-designed RCCV, and to confirm the ultimate strengt and constructibleness. Basic, partial model and total model experiments were conducted.As a result of experimental and analytical studies, it was made sure that the trial-designed RCCV is safe and reliable at a design-load level. It was also confirmed that its ultimate strength fully ensures structural performance. Realizability was confirmed from consideration of the adequacy of the design method and constructibleness.This paper describes the results of the total model experiments planned for these researches.     

Post-test analysis of a 1:10-scale top slab model of ABWR/RCCV subjected to internal pressure

Construction of the first Advanced Boiling Water Reactor (ABWR) in Japan employing a reinforced concrete containment vessel (RCCV) was started in 1991. As RCCV itself is the first structure of its kind in Japan, thorough verification tests have been performed. This paper presents the results of simulation analysis of the Top Slab partial model of the RCCV subjected to internal pressure beyond design load. The Top Slab portion is complicated, being composed of a flat Top Slab, cylindrical wall and fuel pool girders, that its simulation analysis requires the evaluation of nonlinear structural behavior of reinforced concrete members due to membrane, bending and shear forces. This paper reports that Finite Element analysis with 3-D solid elements has given a good quantitative agreement between experimental and analysis results with respect to deformation, failure load and each nonlinear behavior.     

Probabilistic seismic risk analysis of CANDU containment structure for near-fault earthquakes

In this study, the seismic risk of a CANDU (CANada Deuterium Uranium) containment structure is estimated by performing the nonlinear seismic analysis for the near-fault earthquakes. Nonlinear seismic analysis is more effective to consider the distinct nonlinear behavior of concrete structures subjected to the near-fault ground motion which has high input energy. In Korea, the seismic fragility analysis has been performed by using the design analysis results which were obtained from a linear elastic analysis.The lumped mass model of the containment structure was used for a nonlinear dynamic time history analysis. The tri-linear skeleton curve was used for the nonlinear behavior of the prestressed concrete containment structure. In order to estimate the inelastic nonlinear response of the containment, the maximum point-oriented model was used for the hysteretic rule of the shear deformation.For the nonlinear seismic analyses, 30 set of real near-fault earthquake records were used as the input motion. The seismic fragility and risk of the containment for the near-fault ground motions are compared with those from the results based on the conventional method.     

Seismic individual plant examination of external events of US nuclear power plants: insights and implications

As part of the implementation of the severe accident policy, nuclear power plants in the US are conducting the individual plant examination of external events (IPEEE). Seismic events are treated in these IPEEEs by either a seismic probabilistic risk assessment (PRA) or a seismic margin assessment. The major elements of a seismic PRA are the seismic hazard analysis, seismic fragility evaluation of structures and equipment and systems analysis using event tree and fault tree analysis techniques to develop accident sequences and calculate their frequencies of occurrence. The seismic margin assessment is a deterministic evaluation of the seismic margin of the plant beyond the design basis earthquake. A review level earthquake is selected and some of the components that are on the success paths are screened out as exceeding the review level earthquake; the remaining ones are evaluated for their seismic capacity using information from the original plant design criteria, test data and plant walkdown. The IPEEEs of over 100 operating nuclear power plants are nearing completion. This paper summarizes the lessons learned in conducting the IPEEEs and their applicability to nuclear power plants outside of the United States.     

Lower bound earthquake magnitude for probabilistic seismic hazard evaluation

This paper presents the results of a study that develops an engineering and seismological basis for selecting a lower-bound magnitude (LBM) for use in seismic hazard assessment. As part of a seismic hazard analysis the range of earthquake magnitudes that are included in the assessment of the probability of exceedance of ground motion must be defined. The upper-bound magnitude is established by earth science experts based on their interpretation of the maximum size of earthquakes that can be generated by a seismic source. The lower-bound or smallest earthquake that is considered in the analysis must also be specified.The LBM limits the earthquakes that are considered in assessing the probability that specified ground motion levels are exceeded. In the past there has not been a direct consideration of the appropriate LBM value that should be used in a seismic hazard assessment. This study specifically looks at the selection of a LBM for use in seismic hazard analyses that are input to the evaluation/design of nuclear power plants (NPPs). Topics addressed in the evaluation of a LBM are earthquake experience data at heavy industrial facilities, engineering characteristics of ground motions associated with small-magnitude earthquakes, probabilistic seismic risk assessments (seismic PRAs), and seismic margin evaluations. The results of this study and the recommendations concerning a LBM for use in seismic hazard assessments are discussed.     

Design and ultimate behavior of RC plates and shells

An iterative numerical computational algorithm is developed to design a plate or shell element subjected to membrane and flexural forces, which is based on equilibrium considerations for the limited ultimate state of the reinforcement and cracked concrete. Equations for capacities of top and bottom reinforcements in two orthogonal directions have been derived.To verify the design algorithm on the element level several experimental examples are designed. Nonlinear inelastic analyses are performed with the designed examples using the Mahmoud-Gupta’s computer program to show the adequacy of the design equations. The calculated ultimate strengths are from 3 to 18% higher than the ultimate strength obtained from the test results, except in one example. On the global structural level, a design is performed for a hyperbolic cooling tower to check the design strength to verify the adequacy of the design algorithm. Based on the ultimate nonlinear analysis performed with the designed reinforcement, the analytically calculated ultimate loads exceed the design ultimate load from 26 to 63% for analyses with various amounts of tension stiffening effect.Even though the ultimate loads are dependent on the tensile properties of concrete, the calculated ultimate loads are higher than the design ultimate loads for the cases considered. This shows the adequacy of the design algorithm developed, at least for the structures studied. The presented design algorithm for combined membrane and flexural forces can be evolved as a general design equation for reinforced concrete plates and shells through further studies involving the performance of many more designs and analyses of different plate or shell configurations.     

An evaluation of structural failure modes for prestressed concrete containments

Prestressed Concrete Containment Vessels (PCCVs) refer to a popular type of containment used in the United States for Pressurized Water Reactors (PWRs).This paper presents analytically predicted ultimate pressures and seismic levels for PCCVs, considering various modes of failures. Results for six containments are presented, and correlated with the available test data.The analytical methods use either classical techniques or finite element analyses. The ultimate capacity calculations are based upon conservative deterministic estimates of strength of the structure, under both internal pressure and earthquake loads.The results indicate the following: internal pressure capacities of PCCVs built in the US are almost uniformly equal to 2.5 times the design pressure; seismic capacities are at least two times the design level, but they vary widely among the PCCVs depending on the foundation characteristics; seismic capacity of a PCCV decreases with internal pressure; and a PCCV is expected to contribute very little to the overall seismic risk of a nuclear power plant.     

Strength and stiffness of reinforced concrete containments subjected to seismic loading: Research results and needs

The primary purpose of this paper is to present results of an experimental investigation on the strength and stiffness of reinforced concrete subjected to combined biaxial tension and simulated seismic forces. The test specimens represent a section of a wall of a containment structure carrying combined pressurization and seismic loading. Shear stiffness and strength, and their degradation with shear cycling, are given, along with simple expressions for predicting strength and extensional stiffness. The secondary purpose of the paper is to discuss research needs for improved prediction of the response of containment structures to seismic effects.     

Probability-based design criteria for shear walls in nuclear power plants

The development of probability-based criteria for the design of reinforced concrete shear walls subjected to dead load, live load and in-plane earthquake forces in nuclear plants is described. These criteria are determined for flexure and shear limit states in a load and resistance factor design (LRFD) format. The flexure limit state is defined according to traditional principles of ultimate strength analysis, while the shear limit state is established from experimental results. Resistance factors for shear and flexure, load factors for dead and live load, and a load factor for effect of Safe-Shutdown Earthquake are determined for target limit state probabilities of 1.0 × 10⁻⁶ and 1.0 × 10⁻⁵ over a a period of 40 years. Comparisons among the proposed design criteria, ACI 349 and US NRC Standard Review Plan 3.8.4 are included.     

Numerical computation of fragility curves for NPP equipment

The seismic probabilistic risk assessment (PRA) methodology is a popular approach for evaluating the risk of failure of engineering structures due to earthquake. In this framework, fragility curves express the conditional probability of failure of a structure or component for a given seismic input motion parameter A, such as peak ground acceleration (PGA) or spectral acceleration. The failure probability due to a seismic event is obtained by convolution of fragility curves with seismic hazard curves. In general, a log-normal model is used in order to estimate fragilities. In nuclear engineering practice, these fragilities are determined using safety factors with respect to design earthquake. This approach allows to determine fragility curves based on design study but largely draws on expert judgement and simplifying assumptions. When a more realistic assessment of seismic fragility is needed, simulation-based statistical estimation of fragility curves is more appropriate. In this paper, we will discuss statistical estimation of parameters of fragility curves and present results obtained for a reactor coolant system of nuclear power plant. We have performed non-linear dynamic response analyses using artificially generated strong motion time histories. Uncertainties due to seismic loads as well as model uncertainties are taken into account and propagated using Monte Carlo simulation.     

Proving test of earthquake-resistant pipings, equipment and active components

In Japan we initiated the project of a shaking table to prove the earthquake-resistant properties of key items in nuclear power stations. This two-axial shaking table will be able to shake a 1000 ton object on a table of 15 × 15 m by 2600 tonG in horizontal force and by 1300 tonG in vertical force. In this paper, the philosophy of such projects as well as various experiences on such proven tests done in Japan will be described.The main purposes of the proven tests are to understand:

1. (1)The behavior of nuclear structures and equipment under strong earthquakes both from the viewpoint of structural dynamics and process dynamics.

2. (2)The endurance limits of structures and equipment to destructive earthquakes both from the viewpoint of structural integrity and function.

3. (3)The behavior and availability of active components under deformations and accelerations induced by destructive earthquakes.

4. (4)The margin of safety of structures and equipment under assumed destructive earthquake conditions both for society at large and related engineers.

Although we have almost eleven centuries of historical data on earthquake damage, we still learn new facts from each new destructive earthquake. We have almost no experience of earthquake damage to nuclear structures and equipment. We have endeavoured to estimate the ‘modes of failure’ of various structures and items in nuclear power stations. Therefore, we should be afraid of lack of knowledge on that, because earthquakes are natural phenomena and have a somewhat unpredictable nature. However, we can understand the behavior of structures and equipment in their ultimate condition through the shaking test.One of our uncertainties on earthquake effects is the effect of vertical ground motions. The three-dimensional response seems to cause no particular problems; however, overturning and unstable moving of a solid structure are highly non-linear problems. The behavior of free-surface water in a containment under three-dimensional excitation is not known completely also. These things can be clarified only through shaking experiments using a two- or three-dimensional shaking table, including vertical component motions.The availability of active components, such as control rod driving mechanisms, valves and pumps can be checked only through shaking and/or forced deformation tests, because troubles in such active components may be induced by mechanical friction or contact of moving parts. The malfunction of electrical components is also complicated, for example chattering of the contact of relays. To evaluate the behavior of such active components and electrical components, the so-called mathematical model is inadequate. It is almost impossible to establish an adequate model of those components including all elemental factors related to its function.In Japan, we have many experiences of shaking tests of various size models and for various purposes, not only for nuclear power stations, but also for other areas. Their philosophy, methodologies, results and remarks will be briefly described.     

大亚湾核电站反应堆厂房楼层反应谱分析评估

大亚湾核电站核岛厂房的抗震分析遵循技术输出国-法国M310型机组的土建技术规范RCC-G,采用简化的阻抗函数法计算地基岩土的作用.根据大亚湾厂址的地基岩土特点,拟采用更为精确的三维连续半空间边界子结构法来考虑地基岩土的作用,并与原设计进行对比.另外,在原设计中采用多组时程作为地震输入,取各组计算结果的平均值作为设计值的基础(称为"平均"法).在研究中基于相同的时程,拟分别采用"平均"法和更为常用的"包络"法,处理多组时程的响应.基于上述两方面,通过反应堆厂房的地震响应计算,得到核电站系统设备重要的设计基础数据-楼层反应谱(FRS),并将计算的楼层反应谱同设计谱进行比较,从而对设计方法及其结果进行评估,为电站的抗震设计裕量评估和安全管理提供可资参考的结论.     

Recent results on the evaluation of the overpressure response of concrete and steel containments

Analytical studies have been performed for the evaluation of the ultimate load capacity of concrete containment structures. In addition, analyses of steel containment models were carried out to validate computer codes for the analysis of steel containment structures. This paper reports on some of the results of these analyses, dealing first with the global ultimate load behavior of typical prestressed and reinforced concrete containment structures. The results of these analyses are described, with particular attention given to identifying local effects and failure mechanisms of concrete containment structures. On the basis of the global analysis results, local effects analyses were carried out which show clear evidence of large strain concentrations in the liner. The utility of the ABAQUS-EPGEN code is also demonstrated for three steel containment small-scale models tested by Sandia National Laboratory. The basic geometry of the models consisted of a thin cylindrical shell with a hemispherical dome. One of the models included ring stiffeners in the cylinder, and the other model included penetrations without ring stiffeners. The results of these calculations are presented without test data comparisons.     

Effect of specimen sizes, specimen shapes, and placement directions on compressive strength of concrete

The compressive strength of concrete is used as the most basic and important material property when reinforced concrete structures are designed. It has become a problem to use this value, however, because the control specimen sizes and shapes may be different from country to country.In this study, the effect of specimen sizes, specimen shapes, and placement directions on compressive strength of concrete specimens was experimentally investigated based on fracture mechanics. Experiments for the Mode I failure were carried out by using cylinder, cube, and prism specimens. The test results are curve-fitted using least square method (LSM) to obtain the new parameters for the modified size effect law (MSEL). The analysis results show that the effect of specimen sizes, specimen shapes, and placement directions on ultimate strength is present. In addition, correlations between compressive strengths with size, shape, and placement direction of the specimen are investigated.     

Sleeve-type expansion anchor behavior in cracked and uncracked concrete

A test was performed to investigate the effect of concrete cracks on the static behavior of sleeve-type expansion anchors, and to confirm the seismic and fatigue resistance capability in cracked concrete. The tensile and shear test was conducted on single anchors with three different anchor diameters. Concrete test specimens are sufficiently large to prevent the effect of the concrete edges on the anchor behavior. The types of failure, the static strength and displacement behavior of the anchors in uncracked and cracked concrete were compared to evaluate the effect of the cracks. The strength reduction rate of the anchors due to the cracks was exhibited almost less than the corresponding value specified in ACI 349-01, APP. B. Through the residual strength tests, the seismic and fatigue resistance capability of the anchors was confirmed in cracked concrete. The characteristics of the anchor shear capacity significantly vary with how the displacement failure criteria are determined.