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 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 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 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.     

Extreme value risk analysis in the structural design of reactor components

The tendency to confuse “uncertainties” associated with design assumptions and parameters and compensated by the safety factor with objective ‘risks of failure’ implicit in the design, has been characteristic of the approach to probability-based structural design on all levels. However, a clear differentiation between uncertainty and risk is required to remove the lack of correlation between design safety analysis and risk analysis implicit in the present approach to the design of major structural and mechanical components of nuclear reactors as well as other structures.In a recent paper [5] the author has used the definition of the safety factor as a random variable (distribution of a quotient) to construct a probability model that justifies the introduction of the asymptotic distributions of extreme values as the physically relevant distributions of the design parameters governing ultimate load failure on which a realistic risk assessment can be based. Realistic reliability and risk assessment of reactor components subject to fatigue and creep, i.e. design conditions that exceed in practical importance that of ultimate load failure, can be based on the use of the third asymptotic distribution of smallest values.In the case of structural components working under complicated conditions it becomes necessary to perform full-scale tests reproducing, as closely as possible, the anticipated operational and, whenever necessary, critical limiting conditions to be provided in the design as well as in the associated reliability and risk assessment. The economic necessity of keeping the number of such full-scale tests to a minimum which, in the case of larger components, is usually a single or very small number of tests, raises the problem of integration of the test results into the framework of a reliability and risk assessment.     

Post DNB Heat Transfer Experiments for PWR Fuel Assemblies

Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis.     

安全棒系统试验样机冷、热态性能试验研究

安全棒系统是空间核反应堆的关键设备之一,它具有结构紧凑、传动精度高、与反应堆容器连接接口多、工作温度高等特点。通过采用全尺寸的安全棒系统试验样机,确定了冷、热态性能试验方案,设计了专用的试验装置开展冷、热态性能试验。试验结果表明,安全棒系统试验样机运行正常,性能达到设计要求,为试验样机的抗震试验提供了条件,也为安全棒系统后续设计及试验装置的改进提供了参考依据。     

Seismic instrumentation for nuclear power plants: An interpretative review of current practice and the related standard in Germany

The German nuclear safety standard KTA 2201: “Design of nuclear power plants against seismic events”, consists of the following parts: 1. basic principles; 2. characteristics of seismic excitation; 3. design of structural components; 4. design of mechanical and electrical parts; 5. seismic instrumentation; and 6. measures subsequent to earthquakes.While Part 1 was published in June 1975, Part 5 was approved by the Nuclear Safety Standards Commission — Kerntechnischer Ausschuss (KTA) — in June 1977. The other parts are still under development. The requirements of the safety standard KTA 2201.5 deal with

1. (a) number of location (number and location of acceleration recording systems for different sites, single-block plants and multi-block plants);

2. (b) characteristics of instruments (readiness and operation of instruments, margin or errors, dynamic and operation characteristics, duration of records, seismic switch);

3. (c) triggering and information (loss of electric power, start of the acceleration recording systems, threshold of acceleration for triggers and seismic switches, optical and acoustic information); and

4. (d) documentation (results of recordings, inspection and tests).

The purpose of this paper is to present some detailed requirements of the safety standard KTA 2201.5, with its philosophy, and compare these with corresponding requirements in the US. It will be shown that with relatively few instruments, which are very reliable in operation and in triggering, an optimum of data may be available after an earthquake.     

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.     

Shaking table tests on failure characteristics of base isolation system for a DFBR plant

The building of a demonstration fast breeder reactor (DFBR) plant in Japan is planed to be base isolated in the horizontal direction. To verify the seismic safety of the isolation system, a series of shaking table tests was conducted using a reduced scale model with three types of base isolation system, natural rubber bearing with steel damper (NRB+SD), lead rubber bearing (LRB), and high damping rubber bearing (HRB). The results of these tests showed NRB+SD, LRB and HRB were within the stable domain (not hardening) at 1×S2 (maximum acceleration 3.80 m s⁻²) input, and were nearly hardening at 2×S2 input. None of the rubber bearings broke at 3×S2 input, which was the design limit. All these bearings broke at over 4×S2 input.     

Vertical seismic response of overhead crane

Vertical seismic response behavior is an important issue for the seismic design of equipments. The equipment, which is comparatively soft and unrestrained vertically, may resonate and its response is significantly magnified under vertical seismic excitation. Overhead crane is an example of equipment that is unrestrained vertically. The dynamic behavior of an 150-ton-capacity overhead crane under vertical seismic excitation was investigated by scale model excitation test and nonlinear time history analysis. The excitation tests were performed with several input levels and the vertical response with each input level was obtained. The simulation analysis approximately corresponded to the results of the excitation test.     

基于有限元模型验证的控制室盘台抗震鉴定

论文以国内某新建核电站控制室盘台抗震鉴定为例,阐述了基于有限元模型验证的盘台抗震鉴定方法。通过样机试验和模型验证分析,将盘台结构设计与有限元分析进行了有机结合,同时在盘台整体有限元分析验证过程中引入了修正因子,保证了盘台的抗震性能,并使其具有一定的安全裕度。     

Creep—fatigue test of a thick-walled vessel under thermal transient loadings

To improve the damage evaluation methods in the design code for Fast Breeder Reactors (FBRs), a series of creep—fatigue tests of structural models under thermal transient loadings are going on at Oarai Engineering Center of the Power Reactor and Nuclear Fuel Development Corporation (PNC). Test models are designed to incorporate representative structures of components and pipings used in FBRs and are subjected to severer cyclic thermal transients than those experienced in FBRs. The test is planned to be continued until failure occurs. This paper describes the creep—fatigue test results and their damage evaluation for the first test model.A 40 mm thick vessel model made of SUS304 austenitic stainless steel was subjected to cyclic thermal transients, in which sodium at 600°C and 250°C flowed repeatedly. The period of each transient was 2 h. Cracks were observed at seven test portions in the model after 1002 cycles of the thermal transients.Elastic and inelastic analyses were performed to evaluate creep—fatigue damage and crack propagation. The safety margins included in the creep—fatigue design methods based on elastic analysis as well as those based on inelastic analysis are discussed. Finally fracture mechanics analyses were performed to explain the observed crack growth.     

蒸汽发生器非线性支承系统的抗震能力分析

计算核电厂设备的高置信度低失效概率(HCLPF)抗震能力是地震概率安全评价、地震裕度评价的一个重要步骤。以蒸汽发生器支承为研究对象,建立其详细的非线性有限单元模型,通过逐步增大地面运动水平,反复计算系统的响应,最后得到蒸汽发生器支承的抗震能力,并与通过确定性失效裕度法得到的HCLPF进行比较。结果表明,两者的计算结果差别较大。本文建议对于非线性较强的设备需采用非线性时程分析方法计算设备的HCLPF。     

Probabilistic assessment of thermal fatigue in nuclear components

A probabilistic framework is set up to assess the fatigue life of components of nuclear power plants. It intends to incorporate all kinds of uncertainties, such as those appearing in the specimen fatigue strength (number-of-cycles-to-failure of specimens), design margin factors (taking into account the size, surface finish and environmental effects), mechanical model (precisely, the uncertainty on the model input parameters) and the thermal loading. This paper presents the global methodology and details the statistical treatment of the fatigue specimen test data. A first analytical example shows that the reliability of a structure submitted to a periodic stress cycle S changes significantly with respect to the value of S, although the codified (deterministic) design criterion is equally fulfilled. A more comprehensive example involving a mechanical model of a pipe submitted to a deterministic inner temperature loading is finally analysed. The use of the first-order reliability method (FORM) allows to compute the probability of failure as a function of the foreseen lifetime and to rank the input random variables according to their importance in response sensitivity.     

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.     

Seismic test on a one-fifth scale HTGR core model

This paper describes a seismic test on a one-fifth scale HTGR graphite core model. The test program included: (a) a horizontal uniaxial excitation in two orthogonal directions at accelerations up to approximately 1.5 g; (b) sinusoidal, time history (El Centro, Taft, synthesized), excitations imposed on the model; (c) damping and resonance tests; and (d) variation in lateral restraint structure, soft and hard springs.The test program also included pendulum collision test of one-fifth scale and full-scale blocks, two-dimensional array tests, and instrumentation development in support of the final test. The purpose of the test was to: (a) study collision dynamics between graphite blocks; (b) employ data to aid in verifying model scaling laws; (c) investigate model dynamic behavior and response characteristics; (d) provide specific data on block relative displacement, acceleration and strain; and measure boundary support forces; (e) provide data for correlation with analytical models; and (f) provide preliminary design data.     

Analysis of Overheating Rupture in Heat-Transfer Tubes Causing Corrosive High-Temperature Reaction

In the steam generator of a liquid metal fast breeder reactor, a defect penetrating through heat-transfer tube will cause high-pressure water/steam to spout into the low-pressure sodium filling the space outside the tube, to initiate sodium-water reactions. If the leak exceeds an intermediate level (~2kg/s), the reaction jet may rupture adjoining tubes with overheating in the event of insufficient cooling available inside the tubes. Such phenomenon of overheating tube rupture presents a serious problem to the economy and safety of steam generator. With a view to clarifying the failure behavior of steam generator heat-transfer tubes under such condition a model of the phenomenon is derived through a series of tests on sodium-water reactions making use of a test loop representing the scale model of an actual fast breeder steam generator. Comparison of actual test data with analysis based on the model has yielded the following information: The failure behavior of gas-pressurized tubes fall into two categories: (a) by creep failure—occurring upon increase of cumulative damage with tube wall wastage caused by the reaction jet and (b) by ductile failure accompanied by creep—upon tube heating with the reaction jet to the extent of lowering tube wall strength below the hoop stress exerted by tube pressure. Analysis of the two categories of failure results in estimation of the percentage difference between analyzed and measured times to failure of 35–50% in the case of creep failure and of 20–50% in the case of ductile failure accompanied by creep. In practical application to steam generators in order to provide a safety margin a time factor—i.e., the safety factor indicating multiple of actual time to failure—of 3 is adopted against 1.5–2 indicated from test to be the actually applicable value.     

中国实验快堆全堆芯流量分配计算与试验

针对中国实验快堆(CEFR)堆芯和一回路的设计特点,开发水力特性计算程序DAEMON,完成不同工况下的全堆芯流量分配计算,给出流量分配不均匀性等参数。在反应堆调试阶段,进行全堆芯流量分配试验。结果表明,程序计算值与试验值符合较好。在此基础上,验证了CEFR堆芯的流体力学设计,并为反应堆调试和运行提供了基础数据。