Feasibility Study for Multi-Branches and Elbow in Stub Pipe System due to Flow-Induced Vibration Using Computational Fluid Dynamics

2020 ◽  
Author(s):  
Khac-Ha Nguyen ◽  
Won-Tae Kim ◽  
Seung-Pyo Hong ◽  
Haein Lee ◽  
Ahram Lee
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Oscillation ◽  
Acoustic Resonance ◽  
Piping System ◽  
Relief Valve ◽  
Branch Number ◽  
Pipe System

Abstract Acoustic-induced vibration in piping system and other devices leads to premature wear and failure. Especially, in nuclear power plants, very high velocity and temperature gas flows inside pipe systems. Moreover, if a frequency due to the vibration in the piping system is overlapped with a natural frequency of the stud pipe, the magnitude of the amplitude will be increased resulting in severe failure. For example, damage can be considered as flow-induced acoustic resonance at the branch pipes of the safety relief valve in the main steam lines. Specially, the pipe system not only has multi-branches but also includes the elbow that the resonance could occurs making pressure oscillation stronger than that of a single branch because of the interaction between the branches and the elbow. This study has investigated a Computational Fluid Dynamics (CFD) analysis methodology to predict and quantify the vortex shedding frequencies and the pressure pulsation magnitude in the dead-end pipe. The influence of the pressure fluctuation amplitude between each branch, number of branch, and elbow is also investigated.

Feasibility Study for Flow-Induced Vibration in Stud Pipe Using Computational Fluid Dynamics

2020 ◽  
Author(s):  
Khac-Ha Nguyen ◽  
Won-Tae Kim ◽  
Seung-Pyo Hong ◽  
Haein Lee ◽  
Ahram Lee
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Vortex Shedding ◽  
Pressure Pulsation ◽  
Computing Time ◽  
Acoustic Resonance ◽  
Piping System ◽  
Turbulent Model ◽  
Relief Valve ◽  
Pipe Length

Abstract Piping systems in a nuclear plant can be damaged by high-cycle fatigue due to acoustic-induced vibration. Moreover, if the frequency of the vibration in the piping system is overlapped with a natural frequency of the piping, the magnitude of the amplitude will be increased resulting in many problems. For example, the damage is considered as flow-induced acoustic resonance at the branch pipes of the safety relief valve in the main steam lines. This study has investigated the Computational Fluid Dynamics (CFD) analysis methodology to predict and quantify the vortex shedding frequencies and the pressure pulsation magnitude in the dead-end piping system. In order to estimate the vortex shedding vibration, a high level turbulent model should be applied. Such a turbulent model, however, requires a substantial amount of computing time. Therefore, the purpose of the study is to investigate the effects of the main pipe length and the sublayer inflation rate on the vortex shedding frequency and pressure pulsation magnitude. The results for the effects will be able to reduce the size of the fluid domain so that the computing time can be significantly decreased in using the high resolution turbulent models.

Evaluation of Jet Impact Region and Fluid Force Generated From Ruptured Pipes: 2 — Evaluation of Fluid Force Using Computational Fluid Dynamics Analysis

2016 ◽  
Author(s):  
Shiro Takahashi ◽  
Qiang Xu ◽  
Noriyuki Takamura ◽  
Ryo Morita ◽  
Yuta Uchiyama ◽  
...  
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Nuclear Power ◽  
Power Plants ◽  
Jet Impingement ◽  
Cfd Analysis ◽  
Fluid Force ◽  
Free Jet ◽  
Jet Impact ◽  
Computational Fluid Dynamics Analysis

Nuclear power plants are designed to avoid damage to their safety installations because of jet impingement when a pipe is ruptured. We have investigated evaluation methods for the design basis of protection of plants against effects of postulated pipe rupture using computational fluid dynamics (CFD) analysis. The steam jet tests obtained using particle image velocimetry (PIV) were conducted in order to verify the CFD analysis. Spread of steam jets could be visualized and the shapes of the steam jets obtained by analysis were almost the same as those by tests. The spread angle of free jet was investigated using CFD analysis. We also measured jet fluid force when a cylindrical structure was installed downstream from the jet nozzle. Steam jet fluid force obtained by analysis was almost the same as that by tests. We judged the CFD analysis to be applicable to evaluation of jet fluid force generated from ruptured pipes.

Control of Steam Venting Noise in Power Plants

1978 ◽  
Vol 100 (2) ◽  
pp. 369-373 ◽  
Author(s):  
J. K. Floyd
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Safety Valve ◽  
Piping System ◽  
Noise Emission ◽  
Relief Valve ◽  
Design Considerations ◽  
Noise Measurements ◽  
Valve Operation ◽  
Effective Use

Steam venting to atmosphere from piping system pressures as high as 2500 psig, as may occur during safety valve and power relief valve operation or during initial steam piping clean up, is one of the most intense sources of noise emitted by fossil and nuclear power plants. This paper discusses characteristics of sonic and turbulent vent noise. Analytical methods to estimate vent noise intensity and frequency characteristics are presented for use where unsilenced noise measurements are not feasible. Design considerations in the effective use of silencers and acoustic lagging materials are discussed. Power plant noise measurements illustrating the severity of the noise emission and its control are presented.

Numerical Investigation of Thermal Flow Field of Air-Cooled Condensers for Yongshou Plant

2012 ◽  
Vol 248 ◽  
pp. 391-394
Author(s):  
Wen Zhou Yan ◽  
Wan Li Zhao ◽  
Qiu Yan Li
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Wind Speed ◽  
Power Plant ◽  
Flow Field ◽  
Wind Direction ◽  
Rotational Speed ◽  
Power Plants ◽  
Thermal Flow ◽  
Air Cooled Condensers

By using the computational fluid dynamics code, FLUENT, Numerically simulation is investigated for Youngshou power plant. Under the constant ambient temperature, the effects of different wind speed and wind direction on the thermal flow field are qualitatively considered. It was found that when considering about the existing and normally operating power plants, the thermal flow field is more sensitive to wind direction and wind speed. Based on the above results, three improved measures such as: increasing the wind-wall height and accelerating the rotational speed of the fans near the edge of the ACC platform and lengthen or widen the platform are developed to effectively improving the thermal flow field, and enhanced the heat dispersal of ACC.

Integrated Pipe Stress Analysis/Support Pattern Selection/Support Design CAE System

1992 ◽  
Author(s):  
M. Chatterjee ◽  
A. Unemori ◽  
A. Kakaria ◽  
D. Jain
Keyword(s):  
Stress Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Automatic Process ◽  
Pattern Selection ◽  
Flow Process ◽  
Support Design ◽  
Pipe System ◽  
Analysis And Design ◽  
Cae System

Abstract This paper describes the organization and features of the AUTO-PIPE CAE System. AUTO-PIPE is a fully integrated software package which allows the User to perform the entire sequence of piping analysis and design in a streamlined work flow process. Major tasks in this automatic process includes: (1) Pipe Stress Analysis (2) Pipe Support Location Optimization (3) Stress Isometric Drawing Generation (4) Pipe Support Pattern Selection and Member Design (5) 3D Interference Detection for Support At the core of the System is the AUTO-PIPE (Relational) Database which contains all static (project-specific) and dynamic (model-specific) data required for all of the major tasks listed above. The AUTO-PIPE CAE System has been used, and is currently being used, for pipe system design for Nuclear Power Plants in Japan to achieve substantial manpower reduction and cost savings.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

Sensitivity Analyses for a PWR SCC Case Using the Probabilistic Loss-of-Coolant-Accident (Pro-LOCA) Software

2016 ◽  
Author(s):  
Bruce A. Young ◽  
Sang-Min Lee ◽  
Paul M. Scott
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Design Criterion ◽  
Mitigation Strategies ◽  
Sensitivity Analyses ◽  
Base Case ◽  
Piping System ◽  
Loss Of Coolant Accident ◽  
Piping Systems

As a means of demonstrating compliance with the United States Code of Federal Regulations 10CFR50 Appendix A, General Design Criterion 4 (GDC-4) requirement that primary piping systems for nuclear power plants exhibit an extremely low probability of rupture, probabilistic fracture mechanics (PFM) software has become increasingly popular. One of these PFM codes for nuclear piping is Pro-LOCA which has been under development over the last decade. Currently, Pro-LOCA is being enhanced under an international cooperative program entitled PARTRIDGE-II (Probabilistic Analysis as a Regulatory Tool for Risk-Informed Decision GuidancE - Phase II). This paper focuses on the use of a pre-defined set of base-case inputs along with prescribed variation in some of those inputs to determine a comparative set of sensitivity analyses results. The benchmarking case was a circumferential Primary Water Stress Corrosion Crack (PWSCC) in a typical PWR primary piping system. The effects of normal operating loads, temperature, leak detection, inspection frequency and quality, and mitigation strategies on the rupture probability were studied. The results of this study will be compared to the results of other PFM codes using the same base-case and variations in inputs. This study was conducted using Pro-LOCA version 4.1.9.

Effects on Pipe Vibrations of Cavitation in an Orifice and in Globe-Style Valves

2006 ◽  
Author(s):  
Se´bastien Caillaud ◽  
Rene´-Jean Gibert ◽  
Pierre Moussou ◽  
Joe¨l Cohen ◽  
Fabien Millet
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Frequency ◽  
Flow Regimes ◽  
Piping System ◽  
Low Frequencies ◽  
Choked Flow ◽  
Single Hole ◽  
Piping Systems ◽  
Large Amplitude Vibrations

A piping system of French nuclear power plants displays large amplitude vibrations in particular flow regimes. These troubles are attributed to cavitation generated by single-hole orifices in depressurized flow regimes. Real scale experiments on high pressure test rigs and on-site tests are then conducted to explain the observed phenomenon and to find a solution to reduce pipe vibrations. The first objective of the present paper is to analyze cavitation-induced vibrations in the single-hole orifice. It is then shown that the orifice operates in choked flow with supercavitation, which is characterized by a large unstable vapor pocket. One way to reduce pipe vibrations consists in suppressing the orifices and in modifying the control valves. Three technologies involving a standard trim and anti-cavitation trims are tested. The second objective of the paper is to analyze cavitation-induced vibrations in globe-style valves. Cavitating valves operate in choked flow as the orifice. Nevertheless, no vapor pocket appears inside the pipe and no unstable phenomenon is observed. The comparison with an anti-cavitation solution shows that cavitation reduction has no impact on low frequency excitation. The effect of cavitation reduction on pipe vibrations, which involve essentially low frequencies, is then limited and the first solution, which is the standard globe-style valve installed on-site, leads to acceptable pipe vibrations. Finally, this case study may have consequences on the design of piping systems. First, cavitation in orifices must be limited. Choked flow in orifices may lead to supercavitation, which is here a damaging and unstable phenomenon. The second conclusion is that the reduction of cavitation in globe-style valve in choked flow does not reduce pipe vibrations. The issue is then to limit cavitation erosion of valve trims.

Non-Markovian model for the study of pitting corrosion in a water pipe system

2015 ◽  
Vol 26 (10) ◽  
pp. 1550119 ◽  
Author(s):  
A. C. P. Rosa ◽  
P. Vaveliuk ◽  
M. A. Moret
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Extreme Values ◽  
Tail Probability ◽  
Markovian Model ◽  
Nonextensive Statistical Mechanics ◽  
Long Tail ◽  
Pipe System ◽  
Proposed Model ◽  
State Regime

The main studies on pitting consist in proposing Markovian stochastic models, based on the statistics of extreme values and focused on growing the depth of wells, especially the deepest one. We show that a non-Markovian model, described by a nonlinear Fokker–Planck (nFP) equation, properly depicts the time evolution of a distribution of depth values of pits that were experimentally obtained. The solution of this equation in a steady-state regime is a q-Gaussian distribution, i.e. a long-tail probability distribution that is the main characteristic of a nonextensive statistical mechanics. The proposed model, that is applied to data from four inspections conducted on a section of a line of regular water service in power water reactor (PWR) nuclear power plants, is in agreement with experimental results.

Sign in / Sign up

Export Citation Format

Share Document