Dynamic analysis program for hydraulic piping systems in nuclear power plants

1989 ◽  
Vol 114 (1) ◽  
pp. 79-90
Author(s):  
Junichi Tanji
Author(s):  
Robert A. Leishear

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.


Author(s):  
Nicolas d’Udekem ◽  
Philippe Art ◽  
Jacques Grisel

Nowadays, the usefulness of RTR (Reinforced Thermosetting Resin) for pressure retaining equipment does not need further proof: they are lightweight, strong, with low thermal elongation and highly corrosion resistant. The use of RTR piping makes all sense for piping systems circulating raw water such as sea water at moderate pressure and temperature for plants cooling. However, this material is rarely used for safety related cooling systems in nuclear power plants. In Belgium, Electrabel and Tractebel have chosen to replace the existing carbon steel pipes of the raw water system by GRE (Glassfiber Reinforced Epoxy) pipes, in accordance with the Authorized Inspection Agency, applying the ASME Code Case (CC) N-155-2 defining the specifications and requirements for the use of RTR pipes, fittings and flanges. After a challenging qualification process, Class 3 GRE pipes are now installed and operating for raw water cooling systems in two Belgian nuclear units and will soon be installed in a third one. The paper will address the followed qualification processes and the implementation steps applied by Electrabel/Tractebel and relate the overcome obstacles encountered during manufacturing, erection and commissioning of Class 3 GRE piping in order to ensure quality, reliability and traceability required for safety equipment in nuclear power plants.


Author(s):  
Gabriel Ogundele ◽  
Guylaine Goszczynski ◽  
Darcy VanSligtenhorst

The issues over the integrity of buried piping in Nuclear Power Plants (NPPs) have received significant attention over the past few years. These piping systems have been in operation for over 30 years. Leaks from buried piping have the potential to raise safety, radiological, environmental, and financial concerns. Buried piping are subject to degradation mechanisms from the outside (soil side) as well as from the inside (fluid side) and they are primarily protected from external corrosion by applying coating on the pipe and then using cathodic protection to protect any bare areas or holidays in the coating. However, over a period of time the coating may lose its integrity and fail to provide the protection for which it was intended. As this happens, the amount of cathodic current needed for adequate protection increases. In some instances, the coating will disbond from the pipe and shield the cathodic protection from the pipe surface. Because of the economic, environmental, and safety consequences of a failure, NPPs embarked on inspection programs to determine the pipe’s condition and its suitability for continued service. This paper presents some of the observations made during the indirect and direct inspections of buried piping. In addition, the challenges encountered are reported.


2003 ◽  
Author(s):  
J. Guillou ◽  
L. Paulhiac

Several vibration-induced failures at the root of small bore piping systems occurred in French nuclear power plants in past years. The evaluation of the failure risk of the small bore pipes requires a fair estimation of the bending stress under operating conditions. As the use of strain gauges is too time-consuming in the environmental conditions of nuclear power plants, on-site acceleration measurements combined with numerical models are easier to handle. It still requires yet a large amount of updating work to estimate the stress in multi-span pipes with elbows and supports. The aim of the present study is to propose an alternate approach using two accelerometers to measure the local nozzle deflection, and an analytical expression of the bending stiffness of the nozzle on the main pipe. A first formulation is based on a static deformation assumption, thus allowing the use of a simple analog converter to get an estimation of the RMS value of the bending stress. To get more accurate results, a second method is based on an Euler Bernoulli deformation assumption: a spectral analyzer is then required to get an estimation of the spectrum of the bending stress. A better estimation of its RMS value is then obtained. An experimental validation of the methods based on strain gauges has been successfully performed.


Author(s):  
Yukio Takahashi ◽  
Yoshihiko Tanaka

It is essential to predict the behavior of nuclear piping system under seismic loading to evaluate the structural integrity of nuclear power plants. Relatively large stress cycles may be applied to the piping systems under severe seismic loading and plastic deformation may occur cyclically in some portion of the systems. Accurate description of inelastic deformation under cyclic loading is indispensable for the precise estimation of strain cycles and accumulation potentially leading to the failure due to fatigue-ratcheting interaction. Elastic-plastic constitutive models based on the nonlinear kinematic hardening rule proposed by Ohno and Wang were developed for type 316 austenitic stainless steel and carbon steel JIS STPT410 (similar to ASTM A106 Gr.B), both of which are used in piping systems in nuclear power plants. Different deformation characteristics under cyclic loading in terms of memory of prior hardening were observed on these two materials and they were reflected in the modeling. Results of simulations under various loading conditions were compared with the test data to demonstrate the high capability of the constitutive models.


2018 ◽  
Vol 4 (4) ◽  
pp. 243-249
Author(s):  
Artem Sobolev ◽  
Pavel Danilov ◽  
Aleksandr Zevyakin ◽  
Sergej Kurkov

Results of calculation seismic resistance analysis of light equipment of nuclear power plants performed on the example of a ventilation unit using two most common analytical techniques - linear spectral analysis and direct dynamic methods - are discussed. The basic concepts, assumptions and limitations of the linear spectral method are described. Examples are given of specific calculation cases when the method in question is not applicable in the generally accepted formulation. In particular, the phase difference and, possibly, accelerations (displacements) must be taken into consideration in the calculations of extended spatial structures for mutually remote boundary conditions. Another example are the reservoirs not completely filled with liquids. In such case waves may be formed in the liquid and taking them into account is not possible in the linear spectral method. Specific features are examined of application of the dynamic analysis method including the input data, approaches and methodologies required for synthesizing the calculated accelerograms. A sequence of operations performed during synthesizing calculated accelerograms is provided, materials are provided containing the description of the mathematical apparatus applied for deriving the final mathematical relations for calculating response spectra and the calculation relations as such are given. The concept of the damping coefficient is explained, its influence on the calculated results and the approaches to its determination are demonstrated. Options with complete absence of damping and with absolute damping are discussed. A real ventilation set applied in active ventilation systems of nuclear power plants was accepted as the test model. Results calculated for the detailed finite-element model of the ventilation unit using the Zenith-95 software package are presented. These results include the distribution of the calculated reduced stresses. Analysis of the results obtained using the two methods demonstrated overestimation of calculated results by the linear spectral method as compared to those obtained by the dynamic analysis method, which means that the former method underestimates the equipment’s resistance to seismic effects. In addition, the dynamic method shows additional areas in the ventilation unit where significant reduced stresses are found while the linear spectral method ignores these areas.


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