内阻尼对柔性连接转子系统稳定性的影响

工作在高速柔轴下的柔性连接转子容易发生转子动力学失稳,柔性连接处内阻尼是引起失稳的主要因素。采用Kelvin-Voigt黏弹性理论建立转子柔性连接的复刚度内阻尼模型,通过Kelvin-Voigt黏弹性理论损耗因子与转子模态测试得到复刚度模型中代表内阻尼的虚部系数,应用拉格朗日方法建立柔性连接转子系统的动力学方程。结合试验结果验证了模型的合理性,研究了连接处内阻尼和减振器动力学参数对转子系统稳定性的影响。结果表明,增大柔性连接处内阻尼会极大降低转子系统的稳定性,增大减振器中的刚度系数和减小折合质量可有效提高稳定性,系统的稳定性需从连接处和减振器两方面协调设计。     

Analysis and test correlation of flexible and stiff piping systems

An in situ pipe test program was conducted to provide a basis for evaluating piping analysis methodologies and design philosophies. In this program, a 20.3-cm boiler feedwater line with two fundamentally different support systems was tested and analyzed. One system employed hanger supports and was very flexible. The second system employed strut and snubber supports and was relatively stiff. Snapback and forced vibration tests were performed on the piping systems. The test results were used to determine piping damping values and to correlate with analyses. These analyses were used to evaluate current piping analysis methodologies and their analytical models. Also, parametric studies were performed with the analytical models to evaluate the effect of different support systems on the pipe behavior for thermal and seismic loads. In addition, the seismic analysis results were compared to quantify the differences between direct time integration and response spectra analysis methods.     

浮动核电站系统典型用泵动力吸振器的设计

浮动核电站系统典型用泵（典型泵）在系统运行期间振动线谱突出,加大了整个系统的振动水平。本文以典型泵在49 Hz处的振动特征线谱为控制对象,开展动力吸振器的设计研究。结合系统运行环境、吸振器吸振原理、安装方式等多方面因素,初步提出吸振器设计参数,并探讨吸振器几何参数、质量、阻尼对吸振频率的影响;建立典型泵的动力吸振器有限元模型,验证动力吸振器的吸振效果,并分析吸振器质量、阻尼、安装个数对吸振性能的影响。结果表明:动力吸振器满足在8~400 Hz的频率范围内可调,加装动力吸振器后典型泵在49 Hz处的振动线谱控制效果可达到7.3 dB。      

核电厂主蒸汽管道阻尼减振与抗震分析

增设阻尼器是处理核电厂主蒸汽管道振动与地震冲击问题的主要方法。本文利用Sap2000软件建立核电厂主蒸汽管道的有限元模型,分析出了管道的固有频率、振型等动态特性。分析结果表明,平动是主要的影响振型。本文应用非线性动力时程分析计算蒸汽管道在33 Hz频率下的振动及地震响应,得到了管道加设阻尼器前后的振动位移和振动速度数据,并进行了比较,探讨了阻尼器在管道减振与抗震中的应用效果。结果表明,在不改变管道原有结构、不影响管道正常工作的前提下,安装液体黏滞阻尼器可以对主蒸汽管道产生减振与抗震的效果。     

Response of piping system on friction support to bi-directional excitation

Installation of friction devices between a piping system and its supporting medium is an effective way of energy dissipation in the piping systems. In this paper, seismic effectiveness of friction type support for a piping system subjected to two horizontal components of earthquake motion is investigated. The interaction between the mobilized restoring forces of the friction support is duly considered. The non-linear behavior of the restoring forces of the support is modeled as an elastic-perfectly plastic system with a very high value of initial stiffness. Such an idealization avoids keeping track of transitional rules (as required in conventional modeling of friction systems) under arbitrary dynamic loading. The frictional forces mobilized at the friction support are assumed to be dependent on the sliding velocity and instantaneous normal force acting on the support. A detailed systematic procedure for analysis of piping systems supported on friction support considering the effects of bi-directional interaction of the frictional forces is presented. The proposed procedure is validated by comparing the analytical seismic responses of a spatial piping system supported on a friction support with the corresponding experimental results. The responses of the piping system and the frictional forces of the support are observed to be in close agreement with the experimental results validating the proposed analysis procedure. It was also observed that the friction supports are very effective in reducing the seismic response of piping systems. In order to investigate the effects of bi-directional interaction of the frictional forces, the seismic responses of the piping system are compared by considering and ignoring the interaction under few narrow-band and broad-band (real earthquake) ground motions. The bi-directional interaction of the frictional forces has significant effects on the response of piping system and should be included in the analysis of piping systems supported on friction supports. Further, it was also observed that the velocity dependence of the friction coefficient does not have noticeable effects on the peak responses of the piping system.     

Nuclear piping system damping data studies

A program has been conducted at the Idaho National Engineering Laboratory to study structural damping data for nuclear piping systems and to evaluate if changes in allowable damping values for structural seismic analyses are justified. The existing pipe damping data base was examined, from which a conclusion was made that there were several sets of data to support higher allowable values. The parameters which most influence pipe damping were identified and an analytical investigation demonstrated that increased damping would reduce the required number of seismic supports. A series of tests on several laboratory piping systems was used to determine the effect of various parameters such as types of supports, amplitude of vibration, frequency, insulation, and pressure on damping. A multiple regression analysis was used to statistically assess the influence of the various parameters on damping, and an international pipe damping data bank has been formed.     

Evaluation of flawed piping under dynamic loading

This paper describes analytical studies of several of the large-scale flawed pipe experiments conducted for the International Piping Integrity Research Group (IPIRG), including detailed discussion of the test with the longest loading duration. Dynamic excitation with increasing load amplitude leads to failure of the piping at a predesignated test section containing a large manufactured flaw. Here, elastic analysis is shown to describe the system dynamic response reasonably well, provided that an appropriate value of structural damping can be selected. A simplified two degree-of-freedom model displays sensitivity to damping and is used to help select the optimal damping value for use in subsequent finite element calculations. The total damping for the IPIRG piping system is caused by a combination of structural damping from the support conditions and from plastic deformation at highly-stressed locations, such as at the degraded cross section itself or at the long-radius elbows. Effective, calculated damping values for the IPIRG tests varied by an order of magnitude, with low values of 0.5 to 1% associated with short-duration dynamic response and 5% or more for the long-duration test. The discussion includes comparisons of the calculated IPIRG results with ASME Code-suggested analysis damping values.     

Piping research overview

The USNRC Piping Review Committee (PRC) was formed in 1983 with a charter to review NRC piping criteria, to recommend changes to this criteria, and to identify areas that would benefit from future research. This overview will outline the NRC-sponsored research being conducted to address those PRC recommendations concerning the design of nuclear piping systems to withstand dynamic loads. A key element of this research is the joint EPRI/NRC “Piping and Fitting Reliability Research Program.” This program consists of dynamic capacity testing of piping at the system, component, and specimen levels, plus analyses needed to support recommendations for changes to the ASME Code. As part of NRC's contribution to the EPRI/NRC program, a pipe system capacity test will be conducted at ETEC. The “Nonlinear Piping Response Prediction” project at HEDL is evaluating nonlinear response prediction techniques with differing degrees of complexity and will compare the various analytical results both with each other and with physical benchmarks such as the ETEC test. An ORNL project is developing nozzle design guidance that will provide a more realistic basis for evaluating the higher nozzle loads that will result from the more flexible piping systems design that are being considered. INEL will evaluate high frequency damping by considering the existing high frequency data and by conducting high frequency/high stress tests on two piping systems. LLNL is now conducting studies to more completely assess the uncertainties in the seismic response of building structures and piping systems. As a follow-on to the research efforts reported in NUREG/CR-3811, BNL will conduct additional studies to improve combinational procedures for piping response spectra analyses.     

塑性变形对管系地震动力响应的影响

在评述线弹性分析方法的基础上,阐明了在管系特别是核管系动力响应分析中考虑塑性变形影响的重要性,介绍了现有考虑塑性影响的方法及其存在的问题.指出要降低现行规范的保守性,提出合理的管系抗震设计方法,     

Preface

This study reviews past projects, experimental techniques, instrumentation requirements, safety considerations, and the benefits of performing vibration tests on nuclear power plant containments and internal components. The emphasis is on testing to improve seismic structural models although the methods are applicable to any form of dynamic excitation. Established techniques for testing and for identification of resonant frequencies, damping, and mode shapes are presented. The benefits of testing with regard to verifying increased damping values and establishing more accurate computer models are outlined. Finally, a forced vibration test project planned to realize these benefits is presented for a typical nuclear power plant.     

Comparative methods for analysis of piping systems subjected to seismic motion

The dynamic analysis of a three-dimensional piping system of a nuclear power plant is conveniently performed through a finite element method. When the modal analysis is used, only the first few modes of vibration are computed for practical purposes. In this paper is proposed a method of residues which evaluates the neglected modes and combines them with the first calculated modes to estimate the total seismic response of the piping. This methods emphasizes the importance of the selected modes. When the approach is made through a time history input function, this latter is usually characterized by a combination of several recorded accelerograms, e.g. El Centro, San Francisco and Taft. The response of a particular piping has been evaluated by means of these two methods: the use of the modal approach will be strongly recommended due to its inherent advantage of economy and also computation time and reliability.     

Impact of structural steel flexibility and restraints gaps on the dynamic behaviour of piping

The effect of gaps present in the seismic supports of nuclear piping systems and of the flexibility of the steel structure to which intermediate supports are attached, is studied in this paper. An actual piping system is used to investigate the impact of structural steel and mechanical snubber gaps on the dynamic behaviour of piping. An evaluation is thus performed of the finite element modeling techniques employed by the designers in the dynamic analysis of piping systems.     

Influences of the dynamic strain aging on the J–R fracture characteristics of the ferritic steels for reactor coolant piping system

The leak-before-break (LBB) design of the piping system for nuclear power plants has been based on the premise that the leakage due to the through-wall crack can be detected by using leak detection systems before a catastrophic break. The piping materials are required to have excellent J–R fracture characteristics. However, where ferritic steels for reactor coolant piping systems operate at the temperatures where dynamic strain aging (DSA) could occur, the fracture resistance could be reduced with the influence of DSA under dynamic loading. Therefore, in order to apply the LBB design concept to the piping system under seismic loading, both static and dynamic J–R characteristics must be evaluated.Materials used in this study are SA516 Gr.70 for the elbow pipe and SA508 Cl.1a for the main pipe and their welding joints. The crack extension during the dynamic and the static J–R tests was measured by the direct current potential drop (DCPD) and the compliance method, respectively. This paper describes the influences of the dynamic strain aging on the J–R fracture characteristics with the loading rate of the pipe materials and their welding joints.     

Philosophy and practice of the aseismic design of nuclear power plants — Summary of the guidelines in Japan

Aseismic design is considered to be one of the most important factors for the safety of the nuclear power plants built in zones of high seismicity such as Japan. All structures, equipment and piping are classified in accordance with the importance of their radioactive safety to the plant, and the dynamic analysis and/or factored seismic coefficient analysis are applied accordingly. Site and ground conditions, as well as seismicity, should be studied thoroughly in order to estimate the intensities of the design earthquake and the safety margin check earthquake. Dynamic analyses of buildings and structures are performed using the multi-lumped-mass-system supported by soil springs with time history analysis conceptions. This idea is also applied to the design of equipment and piping by coupled system to the major structure or by the floor response spectra criteria. Tolerances are applied to damping factors although some experiments show more realistic results. Allowable stresses of ferrous metals for equipment and piping during earthquakes are more scientifically precise.

This report summarizes a guideline for aseismic design of nuclear power plants. The guideline was prepared by the Japan Electric Association in April, 1970, after three years laborious work.

In sect. 1, the philosophy and criteria are described. All components of a plant should be classified into three classes in accordance with their contributions to reactor safeties. Design to earthquake loadings should be based on “design basis earthquake” which is decided in consideration of local seismicity.

In sect 2, site selection and review for ground are described in the sense of seismic aspects.

In sect 3, deciding the earthquake motion for design is discussed. In Japan, semi-statistical approaches are used in normal practice.

In sect. 4, design philosophy and practice of building structures and containment vessels are described. They are designed under statical seismic forces, and the design of the class “A” structures should be checked by a dynamic response technique.

In sect. 5, design philosophy and practice of piping, vesels and equipment are described. Those which belong to class “A” items should be designed in a dynamic sense. Several programs for dynamic analyses of these items are prepared. Allowable stress under earthquake conditions is discussed in relation to other codes, for example, ASME Section III.

The greater part of the philosophy and design criteria have been adopted to all nuclear power plants which have been and are currently being built in Japan.     

Development of aseismic design of piping, vessels and equipment in nuclear facilities

This report involves the development of aseismic design procedures of piping, vessels and equipment in Japan. These mechanical structures show their various characteristics of vibration. Pressure boundaries, a containment vessel and safety systems belong to such structures. The vital components of nuclear power plants are classified to “A” class according to the classification for the aseismic design in Japan. All components in “A” class are required to be based on dynamic earthquake-resistant design, of which level is decided in consideration of local seismisity.

For dynamic design purposes, the following processes are the most important: 1. estimating eigenfrequencies and modes of the system; 2. estimating its damping characteristics; 3. estimating the behavior of the system during strong earthquakes; 4. deciding the design criteria, especially the allowable stresses to earthquake loadings.     

Coupling effect between liquid sloshing and flexible fluid-filled systems

Current practice in seismic design of flexible liquid-filled systems is reviewed. A coupled fluid-structure finite element method which considers the sloshing effect is developed for the seismic analysis of liquid-filled systems of various geometries with and without internal components. An analysis of the dynamic interaction between the structural vibration and liquid sloshing is also presented. Both rigid and flexible fluid-tank systems of different configurations are considered. Results demonstrate that tank flexibility can affect the amplitude of the free surface wave and hence the sloshing pressure and structural response. This result is consistent with the perturbation analysis. The dynamic interaction depends on (1) the ratio of natural frequency between fluid sloshing and the fluid-tank system and (2) the ratio of the effective areas of the fluid-structure interface and free surface of the fluid. Hence it is expected that in analyzing tanks with flexible internal components, this coupling effect can be more pronounced.     

Piping dynamic reliability and code rule change recommendations

The conservative nuclear piping design criteria for seismic and dynamic loads have led to piping systems with excessive numbers of snubbers. To improve this undesirable situation, a Piping and Fitting Dynamic Reliability Program was initiated by the Electric Power Research Institute (EPRI) in 1985 with cooperation from the U.S. Nuclear Regulatory Commission (NRC). The objective of the program is to develop improved, realistic, and defensible ASME design rules by taking advantage of the inherent dynamic margins in the nuclear piping system. The research results have demonstrated that piping systems have large reserve dynamic capacity and the dynamic failure mode is due to fatigue or fatigue-ratcheting rather than plastic collapse. Based on such physical evidence, a set of code rule change recommendations is suggested in its preliminary form.     

Dynamic load effects on gate valve operability and snubber performance

The Idaho National Engineering Laboratory (INEL) participated in an internationally sponsored seismic research program conducted at the decommissioned Heissdampfreaktor (HDR) located in the Federal Republic of Germany. An existing piping system was modified by installation of 200-mm, naturally aged, motor-operated gate valve from a U.S. nuclear power plant and a piping support system of U.S. design. Using various combinations of snubbers and other supports, six other piping support systems of varying flexibility from stiff to flexible were also installed and tested. Additional valve loadings included internal hydraulic loads and, during one block of tests, elevated temperature. The operability and integrity of the aged gate valve and the dynamic response of the various piping support systems were measured during 25 representative simulations of seismic events.     

采用变化附加阻尼的快堆堆芯组件抗震分析方法的研究

对快堆堆芯组件进行的抗震分析需要考虑冷却剂与堆芯组件之间的流固耦合作用。在之前的分析中,大多数人将流体附加阻尼处理为定值。实际上冷却剂对组件的作用还随着组件间的间隙变化而变化,其带来的附加阻尼应为变量。为更准确地模拟堆芯组件的振动,本文采用变化附加阻尼对快堆堆芯组件的抗震分析方法进行了研究。建立了快堆堆芯单排（5根）堆芯组件的抗震分析计算模型,对该模型进行了附加阻尼为定值和随间隙变化两种情况下的抗震分析,结果显示了考虑变化附加阻尼的堆芯组件抗震分析方法的可行性与有效性。本文所使用的模拟方法更为贴近堆芯组件的振动情况,为更为真实地模拟快堆堆芯组件的地震响应打下基础,这也有助于减少结构设计的保守性,具有一定的工程价值。     

Earthquake response of nuclear reactor building deeply embedded in soil

This paper is concerned with experimental and analytical studies to investigate dynamic behavior of deeply embedded structures such as nuclear reactor buildings. The principal points studied are as follows: (1) Examination of stiffness and radiation damping effects according to embedded depth, (2) verification for distributions of earth pressure according to embedded depth, (3) differences of response characteristics during oscillation according to embedded depth, and (4) proposal of an analytical method for seismic design. Experimental studies were performed by two ways: forced vibration test, and earthquake observation against a rigid body model embedded in soil. Three analytical procedures were performed to compare experimental results and to examine the relation between each procedure. Finally, the dynamic behavior for nuclear reactor buildings with different embedded depths were evaluated by an analytical method.