Predicting Leak Rate Through Valve Stem Packing in Nuclear Applications

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Ali Salah Omar Aweimer ◽  
Abdel-Hakim Bouzid ◽  
Mehdi Kazeminia
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Stokes Equations ◽  
Leak Rate ◽  
Flow Models ◽  
Valve Stem ◽  
Wide Range ◽  
Packing Rings ◽  
Bolted Flange ◽  
Flexible Graphite

Leaking valves have forced shutdown in many nuclear power plants. The myth of zero leakage or adequate sealing must give way to more realistic maximum leak rate criterion in design of nuclear bolted flange joints and valve packed stuffing boxes. It is well established that the predicting leakage in these pressure vessel components is a major engineering challenge to designers. This is particularly true in nuclear valves due to different working conditions and material variations. The prediction of the leak rate through packing rings is not a straightforward task to achieve. This work presents a study on the ability of microchannel flow models to predict leak rates through packing rings made of flexible graphite. A methodology based on experimental characterization of packing material porosity parameters is developed to predict leak rates at different compression stress levels. Three different models are compared to predict leakage; the diffusive and second-order flow models are derived from Naiver–Stokes equations and incorporate the boundary conditions of an intermediate flow regime to cover the wide range of leak rate levels and the lattice model is based on porous media of packing rings as packing bed (Dp). The flow porosity parameters (N, R) of the microchannels assumed to simulate the leak paths present in the packing are obtained experimentally. The predicted leak rates from different gases (He, N2, and Ar) are compared to those measured experimentally in which the set of packing rings is mainly subjected to different gland stresses and pressures.

Leak Rates of Gasses Through Packing Seals With Different Analytical Approaches

2017 ◽  
Author(s):  
Ali Salah Omar Aweimer ◽  
Abdel-Hakim Bouzid ◽  
Mehdi Kazeminia
Keyword(s):  
Flow Regime ◽  
Stokes Equations ◽  
Soft Materials ◽  
Flow Models ◽  
Micro Channels ◽  
Wide Range ◽  
Packing Rings ◽  
Analytical Approaches

Predicting leakage in packed stuffing boxes is a major engineering challenge to designers and end users. Due to the different working conditions and material products, the determination of the flow regime present in packing rings is not a straightforward task to predict. This paper presents a study on the ability of micro channel flow models to predict leak rates through packing rings made of soft materials such as graphite. A methodology based on the experimental characterization of the porosity parameters is developed to predict leak rates at different compression stress levels. Three different models are compared to predicate the leakage, where the diffusive and second order flow models are derived from Naiver-Stokes equations and incorporate the boundary conditions of an intermediate flow regime to cover the wide range of leak rate levels. The lattice model is based on porous media of packing rings as packing bed (Dp). The flow porosity parameters (Rc,Dp) of the micro channels assumed to simulate the leak paths present in the packing are obtained experimentally. The predicted leak rates from different gasses (He, N2, Ar) are compared to those measured experimentally, in which the set of packing rings is mainly subjected to different gland stresses and pressures.

scholarly journals Estimation of leak rate through circumferential cracks in pipes in nuclear power plants

2015 ◽  
Vol 47 (3) ◽  
pp. 332-339 ◽  
Author(s):  
Jai Hak Park ◽  
Young Ki Cho ◽  
Sun Hye Kim ◽  
Jin Ho Lee
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Leak Rate

scholarly journals High-Precision Noncontact Guided Wave Tomographic Imaging of Plate Structures Using a DHB Algorithm

2020 ◽  
Vol 10 (12) ◽  
pp. 4360
Author(s):  
Junpil Park ◽  
Jaesun Lee ◽  
Zong Le ◽  
Younho Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Detection Efficiency ◽  
Detection System ◽  
Automatic Detection ◽  
Guided Wave ◽  
Direct Access ◽  
Experimental Conditions ◽  
Plate Structures ◽  
Wide Range

The safety diagnostic inspection of large plate structures, such as nuclear power plant containment liner plates and aircraft wings, is an important issue directly related to the safety of life. This research intends to present a more quantitative defect imaging in the structural health monitoring (SHM) technique by using a wide range of diagnostic techniques using guided ultrasound. A noncontact detection system was applied to compensate for such difficulties because direct access inspection is not possible for high-temperature and massive areas such as nuclear power plants and aircraft. Noncontact systems use unstable pulse laser and air-coupled transducers. Automatic detection systems were built to increase inspection speed and precision and the signal was measured. In addition, a new Difference Hilbert Back Projection (DHB) algorithm that can replace the reconstruction algorithm for the probabilistic inspection of damage (RAPID) algorithm used for imaging defects has been successfully applied to quantitative imaging of plate structure defects. Using an automatic detection system, the precision and detection efficiency of data collection has been greatly improved, and the same results can be obtained by reducing errors in experimental conditions that can occur in repeated experiments. Defects were made in two specimens, and comparative analysis was performed to see if each algorithm can quantitatively represent defects in multiple defects. The new DHB algorithm presented the possibility of observing and predicting the growth direction of defects through the continuous monitoring system.

Direct Numerical Simulation of a Turbulent Flow on an Oscillating Wall Using Cut-Cell Based Immersed Boundary Method

2016 ◽  
Author(s):  
Chiaki Kino
Keyword(s):  
Turbulent Flow ◽  
Nuclear Power ◽  
Power Plants ◽  
Separated Flow ◽  
Immersed Boundary ◽  
Turbulent Structures ◽  
Flow Induced Vibration ◽  
Production Term ◽  
Large Eddy Simulation Model ◽  
Wide Range

The flow-induced vibration of a pipe is an important issue in various engineering fields, and this phenomenon is widely observed in nuclear power plants. Although turbulent structures play important roles in the velocity and pressure fields in a pipe, only a few studies have been conducted on the turbulent flow on an oscillating wall. In this study, direct numerical simulations were conducted to establish a large eddy simulation model for a turbulent flow on an oscillating wall and scrutinize the energy transfer between the grid scale (GS) and sub-grid scale (SGS). Although energy is generally transferred from the GS to SGS (forward scatter), it is likely that energy is transferred from the SGS to GS (backward scatter) under specific conditions. The present numerical results indicated that backward scatter exists in the production term in the case of a static wavy wall. On the other hand, such backward scatter could not be observed in the case of an oscillating wall. It is well known that separated flows and backward flows are generated behind the crest. Stronger backward flows accelerate the main flow and enhance the velocity gradients in a wide range behind the crest. In the case of an oscillating wall, the development of separated flow is immature because the shape of the wall is not fixed. Eventually, the backward scatter is deemed to be suppressed.

Analytical Modelling of Elastic Interaction in Bolted Flange Gasketed Joints

2017 ◽  
Author(s):  
Linbo Zhu ◽  
Abdel-Hakim Bouzid ◽  
Jun Hong
Keyword(s):  
Analytical Model ◽  
Nuclear Power ◽  
Power Plants ◽  
Experimental Tests ◽  
Elastic Interaction ◽  
Industrial Complex ◽  
Element Analysis ◽  
Minimum Number ◽  
Leakage Failure ◽  
Bolted Flange

Bolted flange joints are widely used in the fossil and nuclear power plants and other industrial complex. During their assembly, it is extremely difficult to achieve the target bolt preload and tightening uniformity due to elastic interaction. In addition to the severe service loadings the initial bolt load scatter increases the risk of leakage failure. The objective of this paper is to present an analytical model to predict the bolt tension change due to elastic interaction during the sequence of initial tightening. The proposed analytical model is based on the theory of circular beams on linear elastic foundation. The elastic compliances of the flanges, the bolts, and the gasket due to bending, twisting and axial compression are involved in the elastic interaction. The developed model can be used to optimize the initial bolt load tightening to obtain a uniform final preload under minimum number of tightening passes. The approach is validated using finite element analysis and experimental tests conducted on a NPS 4 class 900 weld neck bolted flange joint.

Systematic Evaluation of Creep-Fatigue Life Prediction Methods for Various Alloys

2009 ◽  
Author(s):  
Yukio Takahashi ◽  
Bilal Dogan ◽  
David Gandy
Keyword(s):  
Power Generation ◽  
Nuclear Power ◽  
Power Plants ◽  
Strain Range ◽  
Creep Damage ◽  
Small Strain ◽  
Systematic Evaluation ◽  
Wide Range ◽  
Creep Fatigue ◽  
Different Temperatures

Failure under creep-fatigue interaction is receiving increasing interest due to an increased number of start-up and shut-down in fossil power generation plants as well as development of newer nuclear power plants employing low-pressure coolant. These situations have promoted the development of various approaches for evaluating its significance. However, most of them are fragment and rather limited in terms of materials and test conditions they covered. Therefore applicability of the proposed approaches to different materials or even different temperatures is uncertain in many cases. The present work was conducted in order to evaluate and compare the representative approaches used in the prediction of failure life under creep-fatigue conditions as well as their modifications, by systematically applying them to available test data on a wide range of materials which have been used or are planned to be used in various types of power generation plants. The following observations have been made from this exercise. (i) Time fraction model has a tendency to be unconservative in general, especially at low temperature and small strain range. Because of the large scatter of the total damage, this shortcoming would be difficult to cover by the consideration of creep-fatigue interaction in a fixed manner. (ii) Classical ductility exhaustion model showed a common tendency to be overly conservative in many situations, especially at small strain ranges. (iii) The modified ductility exhaustion model based on the re-definition of creep damage showed improved predictability with a slightly unconservative tendency. (iv) Energy-based ductility exhaustion model developed in this study seems to show the best predictability among the four procedures in an overall sense although some dependency on strain range and materials was observed.

Experimental Investigation of Interfacial and Permeation Leak Rates in Sheet Gaskets and Valve Stem Packing

2018 ◽  
Author(s):  
Ali Salah Omar Aweimer ◽  
Abdel-Hakim Bouzid
Keyword(s):  
Gas Flow ◽  
Experimental Testing ◽  
High Stress ◽  
Design Criterion ◽  
Leak Rate ◽  
Environment And Health ◽  
Valve Stem ◽  
Packing Rings ◽  
Order Of Magnitude ◽  
Hazardous Pollutants

The quantities of leak rate through sealing systems are being regulated because of the global concern on the hazardous pollutants being released into the atmosphere and their consequences on the environment and health. The maximum tolerated leak is becoming a design criterion, and the leak rate for an application under specific conditions is required to be estimated with reasonable accuracy. In this respect, experimental and theoretical studies are being conducted to characterize the gas flow through gaskets and packing rings. The amount of the total leak that is present in a gasketed joint or a valve stem packing is the sum of the permeation leak through the sealing material and the interfacial leak at the mating surfaces between the sealing element and mechanical clamp assembly. The existing models used to predict leakage do not separate these two types of leaks. This paper deals with a study based on experimental testing that quantifies the amount of these two types of leaks in bolted gasketed joints and packed stuffing boxes. It shows the contribution of interfacial leak for low and high contact surface stresses and the influence of the surface finish as a result of a 32 and 250 micro-inch RAAH phonographic finish in the case of a bolted flange joint. The results indicate that most of the leak is interfacial reaching 99% at the low stress while the interfacial leak is in the same order of magnitude of the permeation leak at high stress reaching 10−6 and 10−8 mg/s in both packing and gaskets, respectively.

Small Diameter Effects on Internal Flow Boiling

2001 ◽  
Author(s):  
Gail E. Kendall ◽  
Peter Griffith ◽  
Arthur E. Bergles ◽  
John H. Lienhard
Keyword(s):  
Heat Transfer ◽  
Thermal Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Flow Instability ◽  
Flow Boiling ◽  
Small Diameter ◽  
Internal Flow ◽  
Two Phase ◽  
Wide Range

Abstract Since the 1950’s, the research and industrial communities have developed a body of experimental data and set of analytical tools and correlations for two-phase flow and heat transfer in passages having hydraulic diameter greater than 6 mm or so. These tools include flow regime maps, pressure drop and heat transfer correlations, and critical heat flux limits, as well as strategies for robust thermal management of HVAC systems, electronics, and nuclear power plants. Designers of small systems with thermal management by phase change will need analogous tools to predict and optimize thermal behavior in the mesoscale and smaller sizes. Such systems include a wide range of devices for computation, measurement, and actuation in environments that range from office space to outer space and living systems. This paper examines important proceses that must be considered when channel diameters decrease, including flow distribution issues in single, parallel, and split flows; flow instability in parallel passages; manufacturing tolerances effects; nucleation processes; and wall conductance effects. The discussion focuses on engineering issues for the design of practical systems.

Time Dependent Leak Rate Change in SCC Degraded SG Tubes of PWRs

2010 ◽  
Author(s):  
Seong S. Hwang ◽  
Jangyul Park ◽  
Man K. Jung ◽  
Hong P. Kim
Keyword(s):  
Stress Corrosion ◽  
Nuclear Power ◽  
Power Plants ◽  
Constant Pressure ◽  
Water Pressure ◽  
Rate Change ◽  
Time Dependent ◽  
Leak Rate ◽  
Alloy 600 ◽  
Axial Crack

Primary water stress corrosion cracking (SCC) and an outside diameter SCC have occurred in the steam generator (SG) tubes of nuclear power plants in the republic of Korea and around the world. Although high corrosion resistant alloy 690 has been replacing the alloy 600 tubings, it is important to establish the repair criteria for the remaining degraded alloy 600 tubings to reassure regarding the reactors integrity, and still maintain the plugging ratio within the limits needed for its efficient operations. For assessment and management of the degradation, it is crucial to understand the initial leak behaviors and time dependent leak rate change from SCC flaws under a constant pressure. Stress corrosion cracked tube specimens were prepared at room temperature by the contact of sodium tetrathionate solution. The initial leak rate and time dependent leak rate were measured at different pressure ranges with time. Water pressure inside the tube was slowly increased in a step like manner with a designated holding time. The leak rate was calculated by dividing the amount of water by the time. A large open and long axial crack showed an increasing leak rate with time at a constant pressure, whereas small opened cracks did not show an increase in a time dependent leak rate. Under some pressures, the leak rate did not increase with the increase of pressure due to a tightness of circumferential cracks. Throughwall axial crack of 5 mm long may exhibit the leakage of action level 1 of the EPRI leakage guideline.

Experimental Investigation on Heat Transfer for Two-Phase Flow Under Natural Convection

2012 ◽  
Author(s):  
Milan Amižić ◽  
Estelle Guyez ◽  
Jean-Marie Seiler
Keyword(s):  
Heat Transfer ◽  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Liquid Pool ◽  
Severe Accident ◽  
Bottom Plate ◽  
Interfacial Region ◽  
Two Phase ◽  
Wide Range

In the frame of severe accident research for the second and the third generation of nuclear power plants, some aspects of the concrete cavity ablation during the molten corium–concrete interaction are still remaining issues. The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat melt-through. For the purpose of experimental investigation of thermal-hydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched jointly by CEA, EDF, IRSN, GDF-Suez and SARNET. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the smallest pool configuration (50 cm × 25 cm × 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s. This paper presents some preliminary conclusions deduced from the experiments which involve a liquid pool with the gas injection only from the bottom plate. A comparison with existing models for the assessment of heat transfer has also been carried out.

Sign in / Sign up

Export Citation Format

Share Document