Managing Check Valve Flange Leaks: Technical Solutions and Challenges to Overcome

2012 ◽  
Author(s):  
L. Ike Ezekoye ◽  
William E. Densmore ◽  
William M. Turkowski ◽  
Robert E. Becse
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Check Valve ◽  
Technical Solution ◽  
Valve Body ◽  
Service Testing ◽  
Piping Systems ◽  
Maintenance Requirements ◽  
Technical Solutions ◽  
Plant Shutdown

Check valves are the simplest valves in power plants. Their simplicity and passive nature, combined with their relatively low maintenance requirements, often mask their relative importance in piping systems. Compared to power operated valves (POVs), such as motor operated valves or air operated valves, check valves have very few parts. The more parts a valve has, the more likely failures will occur. As such, power operated valves tend to have more stringent requirements that cover periodic verification of operability, in-service testing (IST), and scheduled preventive maintenance to assure functionality. Check valves, on the other hand, do not require nearly the same amount of rigor to assure operability. The passive nature of check valves often leads the user not to expect failures. Consequently, lacking of attention often results to inadvertent failures. One failure that has received significant attention from both the industry and the regulator is check valve body-to-bonnet joint leakage. In nuclear power plants this leakage can contaminate the general area where the valve is located, can lead to a plant shutdown, and pose personnel hazards. In this paper, the technical solutions that can be used to manage check valve body-to-bonnet joint leakage will be presented. The merits of each technical solution and the associated challenges will be discussed. Also, as some of the leakage containment solutions are appurtenances to the valve, the paper will address the interface between the appurtenances and the valve.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

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.

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.

Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice

2021 ◽  
Author(s):  
Abhinav Gupta ◽  
Ankit Dubey ◽  
Sunggook Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Industry ◽  
Life Extension ◽  
Seismic Loads ◽  
Thermal Loads ◽  
Loss Of Coolant Accident ◽  
Piping Systems ◽  
Recent Developments

Abstract Nuclear industry spends enormous time and resources on designing and managing piping nozzles in a plant. Nozzle locations are considered as a potential location for possible failure that can lead to loss of coolant accident. Industry spends enormous time in condition monitoring and margin management at nozzle locations. Margins against seismic loads play a significant role in the overall margin management. Available margins against thermal loads are highly dependent upon seismic margins. In recent years, significant international collaboration has been undertaken to study the seismic margin in piping systems and nozzles through experimental and analytical studies. It has been observed that piping nozzles are highly overdesigned and the margins against seismic loads are quite high. While this brings a perspective of sufficient safety, such excessively high margins compete with available margins against thermal loads particularly during the life extension and subsequent license renewal studies being conducted by many plants around the world. This paper focuses on identifying and illustrating two key reasons that lead to excessively conservative estimates of nozzle fragilities. First, it compares fragilities based on conventional seismic analysis that ignores piping-equipment-structure interaction on nozzle fragility with the corresponding assessment by considering such interactions. Then, it presents a case that the uncertainties considered in various parameters for calculating nozzle fragility are excessively high. The paper identifies a need to study the various uncertainties in order to achieve a more realistic quantification based on recent developments in our understanding of the seismic behavior of piping systems.

Computational Study of Conjugate Heat Transfer in T-Junctions

2009 ◽  
Author(s):  
Simon Kuhn ◽  
Bojan Nicˇeno ◽  
Horst-Michael Prasser
Keyword(s):  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Cold Water ◽  
Computational Study ◽  
Solid Wall ◽  
Turbulent Viscosity ◽  
Life Time ◽  
Scale Model ◽  
Piping Systems

Thermal fatigue is a relevant problem in the context of life-time extension of nuclear power plants (NPP). In many piping systems in NPPs hot and cold water is mixed, which leads to high temperature fluctuations in the region close to the solid wall and resulting thermal loads on the pipe walls that can cause fatigue. One of the relevant geometric test cases for thermal fatigue is the mixing in T-junctions. In this study we apply large–eddy simulations (LES) to the mixing of hot and cold water in a T-junction. We perform a set of simulations by using different formulations of the LES subgrid scale model, i.e. standard Smagorinsky and dynamic procedure, to identify the influence of the modelled subgrid scales on the simulation results. The results exhibit a large difference between the models, which is caused by the use of turbulent viscosity wall–damping functions when applying the standard model.

Wave Propagation Analysis of Piping Structures

2009 ◽  
Author(s):  
Akemi Nishida
Keyword(s):  
Wave Propagation ◽  
Dynamic Behavior ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Beam Theory ◽  
Spectral Element ◽  
Analysis Tool ◽  
Propagation Analysis ◽  
Piping Systems

It is becoming important to carry out detailed modeling procedures and analyses to better understand the actual phenomena. Because some accidents caused by high-frequency vibrations of piping have been recently reported, the clarification of the dynamic behavior of the piping structure during operation is imperative in order to avoid such accidents. The aim of our research is to develop detailed analysis tools and to determine the dynamic behavior of piping systems in nuclear power plants, which are complicated assemblages of different parts. In this study, a three-dimensional dynamic frame analysis tool for wave propagation analysis is developed by using the spectral element method (SEM) based on the Timoshenko beam theory. Further, a multi-connected structure is analyzed and compared with the experimental results. Consequently, the applicability of the SEM is shown.

Failure Probability Assessment of Piping Systems Affected by SCC and Pipe Wall Thinning for LBB Analysis

2011 ◽  
Author(s):  
Hiromasa Chitose ◽  
Hideo Machida ◽  
Itaru Saito
Keyword(s):  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Probability Assessment ◽  
Wall Thinning ◽  
Piping Systems ◽  
Leak Before Break ◽  
Thinning Rate ◽  
Service Inspection

This paper provides failure probability assessment results for piping systems affected by stress corrosion cracking (SCC) and pipe wall thinning in nuclear power plants. On the basis of the results, considerations for applying the leak-before-break (LBB) concept in actual plants are presented. The failure probability for SCC satisfies the target failure probability even if conservative conditions are assumed. Moreover, for pipe wall thinning analysis, pre-service inspection is important for satisfying the target failure probability because the initial wall thickness affects the accuracy of the wall thinning rate. The pipe wall thinning analysis revealed that the failure probability is higher than the target probability if the bending stress in the pipe is large.

Effects of Flow Accelerated Corrosion on Piping Steady-State Vibration Evaluations

2003 ◽  
Author(s):  
Brian J. Voll
Keyword(s):  
Steady State ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Vibration Monitoring ◽  
Accelerated Corrosion ◽  
Monitoring Programs ◽  
Wall Thinning ◽  
Piping Systems ◽  
Flow Accelerated Corrosion

Piping steady-state vibration monitoring programs were implemented during preoperational testing and initial plant startup at most nuclear power plants. Evaluations of piping steady-state vibrations are also performed as piping and component failures attributable to excessive vibration are detected or other potential vibration problems are detected during plant operation. Additionally, as a result of increased flow rates in some piping systems due to extended power uprate (EPU) programs at several plants, new piping steady-state vibration monitoring programs are in various stages of implementation. As plants have aged, pipe wall thinning resulting from flow accelerated corrosion (FAC) has become a recognized industry problem and programs have been established to detect, evaluate and monitor pipe wall thinning. Typically, the piping vibration monitoring and FAC programs have existed separately without interaction. Thus, the potential impact of wall thinning due to FAC on piping vibration evaluations may not be recognized. The potential effects of wall thinning due to FAC on piping vibration evaluations are reviewed. Piping susceptible to FAC and piping susceptible to significant steady-state vibrations, based on industry experience, are identified and compared. Possible methods for establishing links between the FAC and vibration monitoring programs and for accounting for the effects of FAC on both historical and future piping vibration evaluations are discussed.

Application and Experience Feedback of ASME Code Case N-155-2 for RTR Piping in Nuclear Plants

2013 ◽  
Author(s):  
Nicolas d’Udekem ◽  
Philippe Art ◽  
Jacques Grisel
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Sea Water ◽  
Water System ◽  
Moderate Pressure ◽  
Raw Water ◽  
Cooling Systems ◽  
Piping Systems ◽  
Asme Code

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.

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