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.     

Effects of ductility on seismic response of piping systems and their implication on design and qualification

This study is concerned with the inelastic seismic response of nuclear power plant piping systems. Two systems are examined. The first one is an idealized four-equal-span pipe run and the second one consists of two configurations modified from an existing pipe run. Detailed finite element seismic time history analyses are performed using the computer program. By varying the various geometrical and physical parameters, calculations are made for a total of 76 cases. The results show that ductility generally contributes to reducing the response of piping systems. An empirical relation between the support load reduction factor and support ductility demand is given and a chart and simple procedures are suggested for the design and qualification of piping supports taking ductility into consideration.     

Seismic stops vs. snubbers, a reliable alternative

The Seismic Stops methodology has been developed to provide a reliable alternative for providing seismic support to nuclear power plant piping. The concept is based on using rigid passive supports with large clearances. These gaps permit unrestrained thermal expansion while limiting excessive seismic displacements. This type of restraint has performed successfully in fossil fueled power plants.A simplified production analysis tool has been developed which evaluates the nonlinear piping response including the effect of the gapped supports. The methodology utilizes the response spectrum approach and has been incorporated into a piping analysis computer program RLCA-GAP.Full scale shake table tests of piping specimens were performed to provide test correlation with the developed methodology. Analyses using RLCA-GAP were in good agreement with test results. A sample piping system was evaluated using the Seismic Stops methodology to replace the existing snubbers with passive gapped supports. To provide further correlation data, the sample system was also evaluated using nonlinear time history analysis. The correlation comparisons showed RLCA-GAP to be a viable methodology and a reliable alternative for snubber optimization and elimination.     

Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the “force equivalent area” (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques.     

Piping extreme dynamic response studies

Results are presented for a series of high-amplitude dynamic tests of a simple pressurized piping system excited through various multiple piping supports. The four-inch diameter piping achieved response levels above yield when subjected to earthquake-like time history inputs and withstood — without leakage or gross distortion — dynamic inputs that were factors of three to five times greater than those inputs required to just exceed the ASME Class 2 stress limit for Service Level D, the Safe Shutdown Earthquake condition. Despite intentionally induced support failures in several tests, piping pressure integrity was maintained, and no plastic collapse occurred. Selected snubber hardware likewise exhibited large design margins under transient loads.     

Energy absorbers used against impact loading

In the WWER-440 reactor the primary piping consists of six horizontal loops going rapidally from the pressure vessel, each loop having a horizontal steam generator. In this reactor type the relatively long primary piping with many curved sections requires special attention in order to successfully eliminate the consequences of the design basis accident. Emergency supports are located in appropriate places to restrict the movements of the pipe. Under normal conditions there is a gap of some centimeters between the pipe and a support so that in the pipe can be deformed freely under changing loads. This paper deals with those energy-absorbing structures used at the Loviisa Nuclear Power Plant for protection against impact loading. Places and circumstances where energy-absorbing structures are employed are specified. Development and design of impact absorber elements are discussed and impact tests are described.     

Decreasing snubber inservice inspection costs through snubber reduction and improved test limits

Snubber inservice inspection (ISI) requirements, along with a history of snubber malfunctions, has made inspection and maintenance of snubbers a significant part of a nuclear power plant's ISI budget. These expenses can be minimized through snubber reduction and the use of improved test limits for snubber functional testing. This paper presents a snubber overview and reviews snubber ISI requirements. Examples are given of the high cost that maintaining a snubber in an operating nuclear plant represents.Snubber reduction refers to reducing a plant's snubber population by eliminating snubbers shown not to be required to restrain piping for design basis dynamic loadings, and by replacing snubbers with other types of restraints, such as rigid struts. Snubber reduction is discussed in terms of what makes removing snubbers practical along with approaches to, and results of recently implemented snubber reduction programs.Improved or increased test limits for snubber functional testing are discussed along with an approach to, and results of an Electric Power Research Institute sponsored program to develop improved limits that would not significantly affect piping response. Improved piping acceptance criteria can be used to justify the use of increased test limits provided by snubber manufacturers. An additional use is to justify the operability of piping on which faulty snubbers were found.     

Fatigue analysis for analytically overloaded piping components and valves in nuclear power plants

Lately, in connection with life extension aspects of power plants, an increasingly accurate determination of the lifetime of components in nuclear stations is being required. In order to assess reliably current fatigue levels in piping systems, variables such as pressure, temperature, and resultant force and moment transients as well as analytical methods which take into account the real operational history must be considered. This paper presents a method for analyzing the transient heat transfer between fluid and pipe wall in order to investigate effects which until now have been assumed conservatively to be caused by a sudden jump in temperature. Further, an example is given showing that the K_e factor approach in current design codes for performing simplified elastic-plastic fatigue analyses is conservative.     

Seismic behaviour of piping systems with and without defects: experimental and numerical evaluations

Based on important experimental and analytical programs, CEA has developed a wide expertise in the domain of the seismic studies of piping. The specific behaviour of piping systems, with and without flaw, under seismic loading, has been analysed. CEA has evaluated the margins coming from present design procedures, that are shown to be greatly conservative, and has developed analytical methods devoted to a better evaluation of the global behaviour (internal moments in the piping system or reactions in supports, displacements, rotations). Non-linear time–history procedures have been built that allow for accurate modelisation, with and without defect. For industrial purposes or for sensitivity analyses, simplified methods have been proposed that are much less time consuming than non-linear time–history calculations, but much more accurate than linear methods.     

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.     

Demonstration of piping integrity with SMA technology

The safe function of a new pipe whip restraint device has been demonstrated in a full scale test. The restraint is based on using a shape memory alloy to protect a pipe and its environment in the event of a double-ended-guillotine-break. The evaluation test has been performed at boiling water reactor (BWR) operating pressure and temperature using a pipe representing BWR primary piping.     

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.     

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.     

Reliability analysis for stiff versus flexible piping

The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis.A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. We then investigated a couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design. We concluded that these changes substantially reduce calculated piping responses and allows piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements. Furthermore, the more flexible piping redesigns are capable of exhibiting reliability levels equal to or higher than the original stiffer design.An investigation of the malfunction of pipe whip restraints confirmed that the malfunction introduced higher thermal stresses and tended to reduce the overall piping reliability. Finally, support and component reliabilities were evaluated based on available fragility data. Our result indicated that the support reliability usually exhibits a moderate decrease as the piping flexibility increases. Most on-line pumps and valves showed an insignificant reduction in reliability for a more flexible piping design.     

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.     

Evaluation of the influence of bypass flow gap distribution on the core hot spot in a prismatic VHTR core

Core bypass flow in VHTR is one of the key issues for core thermal margins and efficiency. The bypass flow in the prismatic core varies during core cycles due to the irradiation shrinkage/swelling and thermal expansion of the graphite blocks. A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. The influence of the core restraint mechanism on the bypass flow gap was evaluated. The predicted gap size is as much as 8 mm when the graphite block is exposed to its allowable limit of fast neutron fluence. The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. The results indicate that the bypass gap and flow distributions are closely related to the local hot spot and its location and the core restraint mechanism preventing outward movement of the graphite block by a fastening device reduces the bypass gap size, which results in the decrease of maximum fuel temperature not less than 100 °C, when compared to the case without it.     

Seismic resistant design of safety class structures and equipment

Some commonly encountered problems in the seismic resistant design of nuclear power plant facilities are discussed. The topics included here are ground input motions, local geology versus source mechanism and travel path, three components inputs, torsional responses, floor response spectra, seismic resistant design of heavy equipment, the application of component mode synthesis technique, seismic resistant design of piping systems, equipment qualification by testing, the effects of close modes, underground pipe design, and soil structure interaction.     

Effect of plastic deformation of piping on fluid-transient propagation

A simple computational model is developed for incorporating the effect of elastic-plastic deformation of piping on pressure transient propagation in a fluid system. A computer program (PLWV) is described which incorporates this structural interaction model into a one-dimensional method-of-characteristics procedure for fluid hammer analysis. Computed results are shown to be in good agreement with available experimental data. The most significant effect of plastic deformation is to limit the peak pressure of a pulse leaving a pipe to approximately the yield pressure of the pipe, if the pipe is sufficiently long.     

核安全一级高温管道系统结构分析与安全评估方法研究

为解决复杂核安全一级高温管道系统结构分析与评定工程问题，在管道分析软件与核级高温评定规范ASME-NH之间建立了一座桥梁。首先，对管道结构（直管及弯管）在不同载荷作用下的应力状态解析解进行了详细推导分析，并且与有限元数值解进行了误差分析。结果显示，给出的直管及弯管结构应力状态解析解具有很好的准确性。随后，将一维管线力学分析模型与截面三维应力状态解析解相结合，给出了高温管道系统结构分析、评定方法及应用步骤，将ASME-NH-3650规范内容明确化。      

Structural evaluation of a piping system subjected to thermal stratification

Piping systems in nuclear power plants are often designed for pressure, mechanical loads originating from deadweight and seismic events and operating thermal transients using the equations in the ASME Boiler and Pressure Vessel Code, Section III. In the last few decades a number of failures in piping have occurred due to thermal stratification caused by the mixing of hot and cold fluids under certain low flow conditions. Such stratified temperature fluid profiles give rise to circumferential metal temperature gradients through the pipe leading to high stresses causing fatigue damage. A simplified method has been developed in this work to estimate the stresses caused by the circumferential temperature distribution from thermal stratification. It has been proposed that the equation for the peak stress in the ASME Section III piping code include an additional term for thermal stratification.