Dynamic Pressure Data Acquisition Via Strain Gage Measurements

2005 ◽  
Author(s):  
Gyorgy Szasz ◽  
Karen K. Fujikawa ◽  
Raju Ananth
Keyword(s):  
Strain Gage ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Dynamic Pressure ◽  
Operating Conditions ◽  
Pressure Transducers ◽  
Piping Systems ◽  
Dynamic Pressure Measurements ◽  
Main Steam

Dynamic pressure measurements are often helpful in characterizing operating conditions within industrial piping. The most straight forward method to obtain this type of data is to mount pressure transducers on the piping [6]. The orifice necessary for these instruments, frequently presents an undesirable opening in the pressure boundary of the affected system. One type of pressure transducer employs a strain gage mounted internally on a membrane that is exposed to the pressure to be measured [4, 5]. The deformation of the membrane is proportional to the pressure to be measured and is reported as a pressure value. A union of these two concepts yields the idea of mounting the gages directly on the piping and thereby eliminating the need for compromising piping integrity. One of the challenges is performing this measurement in the presence of significant axial train that is not related to the internal pressure. In the recent past Structural Integrity Associates Inc. has successfully applied this innovative technique to several main steam piping systems in various nuclear power plants in the US. This paper will describe some of the considerations regarding compensation for interfering axial strains as well as provide sample results from existing installations.

On-Site Evaluation of Bending Stresses at the Root of Small Bore Pipes by Accelerometric Measurements

2003 ◽  
Author(s):  
J. Guillou ◽  
L. Paulhiac
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Numerical Models ◽  
Operating Conditions ◽  
Strain Gauges ◽  
Bending Stress ◽  
Piping Systems ◽  
Bending Stresses ◽  
Site Evaluation

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.

Study on Load Capacity of Main Steam Superpipe Nozzle Used in Nuclear Power Plant

2017 ◽  
Author(s):  
Zhou Gengyu ◽  
Liang Shuhua ◽  
Sun Lin ◽  
Lv Feng
Keyword(s):  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Load Capacity ◽  
Operating Conditions ◽  
Integrity Assessment ◽  
Inner Radius ◽  
Main Steam

The main steam super pipe used in nuclear power plant is an important safety class2 component. There are several nozzles located on it and linked with main steam safety valves. In the past two decades, the hot extrusion forming technology has been widely used to manufacture the super pipe nozzles. Comparing with traditional insert weldolet, the wall thickness of the extruded nozzle is relative small, and the nozzle inner radius is hard to control precisely in the fabrication process. Due to high temperature working condition and complicated loading conditions, the load capacity of the super pipe extruded nozzle has become an issue of concern for manufacturers and users. This paper presents a structural integrity assessment of a super pipe extruded nozzle. The nozzle stresses due to internal pressure and external loads for different operating conditions are obtained by the three-dimensional finite element analysis. The extruded nozzle is evaluated against the RCCM code Subsection C3200 Service Levels O, B and D stress limits for design, upset and faulted conditions. A parametric sensitivity analysis of the extruded nozzle inner radius size is also carried out. In addition, in order to reduce the calculation effort, an efficient calculation method is developed by using the commercial finite element program ANSYS.

Effect of Bending Load on the Failure Pressure of Wall-Thinned Pipe Bends

2010 ◽  
Author(s):  
Jin Weon Kim ◽  
Oon Young Jung
Keyword(s):  
Internal Pressure ◽  
Nuclear Power ◽  
Power Plants ◽  
Degradation Mechanism ◽  
Operating Conditions ◽  
Accelerated Corrosion ◽  
Failure Pressure ◽  
Bending Load ◽  
Piping Systems ◽  
Flow Accelerated Corrosion

Under normal operating conditions, piping systems in nuclear power plants (NPPs) are subject not only to internal pressure but also to bending loads induced by deadweight and thermal expansion [1]. Bending is thus considered to be an important factor in evaluating the integrity of defective piping components. Local wall-thinning due to flow-accelerated corrosion is a main degradation mechanism of carbon steel piping systems in NPPs [2], and the integrity evaluation of wall-thinned piping components has become an important issue [3]. This study investigated the effects of bending load on the failure pressure of wall-thinned pipe bends under internal pressure. Our previous study experimentally evaluated the bending load effects on the failure pressure of wall-thinned elbows under displacement controlled in-plane bending load [4], but the numbers of experimental data were insufficient to determine the effects of bending load on the failure pressure of wall-thinned pipe bends. Therefore, the present study systematically evaluates the effects of bending load on the failure pressure of wall-thinned pipe bends using parametric finite element analyses.

Evolution of Section XI of the ASME Boiler and Pressure Vessel Code

2000 ◽  
Vol 122 (3) ◽  
pp. 234-241 ◽  
Author(s):  
Owen F. Hedden
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
Operating Conditions ◽  
Nondestructive Examination ◽  
The Us ◽  
Regulatory Commission ◽  
Technical Basis ◽  
Year 2000

This article will describe the development of Section XI from a pamphlet-sized document to the lengthy and complex set of requirements, interpretations, and Code Cases that it has become by the year 2000. Section XI began as a set of rules for inservice inspection of the primary pressure boundary system of nuclear power plants. It has evolved to include other aspects of maintaining the structural integrity of safety class pressure boundaries. These include procedures for component repair/replacement activities, analysis of revised and new plant operating conditions, and specialized provisions for nondestructive examination of components and piping. It has also increased in scope to cover other Section III construction: Class 2, Class 3 and containment structures. First, to provide a context for the discussions to follow, the differences in administration and enforcement between Section XI and the other Code Sections will be explained, including its dependence on the US Nuclear Regulatory Commission. The importance of interpretations and Code Cases then will be discussed. The development of general requirements and requirements for each class of structure will be traced. The movement of Section XI toward a new philosophy, risk-informed inspection, will also be discussed. Finally, an annotated bibliography of papers describing the philosophy and technical basis behind Section XI will be provided. [S0094-9930(00)01703-0]

Experimental Evaluation of the Bending Load Effect on the Failure Pressure of Wall-Thinned Elbows

2008 ◽  
Author(s):  
Jin-Weon Kim ◽  
Yeon-Soo Na ◽  
Sung-Ho Lee ◽  
Chi-Yong Park
Keyword(s):  
Internal Pressure ◽  
Nuclear Power ◽  
Power Plants ◽  
Operating Conditions ◽  
Load Effect ◽  
Failure Pressure ◽  
Wall Thinning ◽  
Bending Load ◽  
Piping Systems ◽  
Plane Bending

During normal operating conditions, piping systems in nuclear power plants are subject to internal pressure and to bending loads induced by deadweight, thermal expansion, and internal pressure, and understanding the effect of bending load on the failure of wall-thinned elbows is important to evaluate the failure pressure reliably. This study includes a series of burst tests using real-scale 4-inch schedule 80 elbow specimens with local wall-thinning under combined internal pressure and in-plane bending load applied by displacement control. The results are compared with those tested under simple internal pressure only. In the tests, various circumferential thinning angles (θ/π = 0.125, 0.25, 0.5, 1.0) and thinning locations (intrados, extrados, and full circumference) were considered. Each specimen was initially subjected to an in-plane bending load, closing mode for extrados wall-thinned elbows and opening mode for intrados wall-thinned elbows, and then internal pressure was applied up to point of final failure. The results showed that the effect of in-plane bending on the failure pressure and failure mode was minor under all wall-thinning conditions. In addition, the dependence of failure pressure on the circumferential thinning angle and thinning locations was identical to that observed under simple internal pressure.

An Experimental Measurement of Toughness Locus for a Ductile Material

2005 ◽  
Vol 297-300 ◽  
pp. 2410-2415 ◽  
Author(s):  
Dong Hak Kim ◽  
Jeong Hyun Lee ◽  
Ho Dong Kim ◽  
Ki Ju Kang
Keyword(s):  
Crack Length ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Procedure ◽  
Stress Triaxiality ◽  
Crack Tip ◽  
Potential Drop ◽  
Image Correlation ◽  
Ductile Material ◽  
Main Steam

A toughness locus Jc-Q for a ductile steel, SA106 Grade C used in the main steam piping of nuclear power plants, has been experimentally evaluated. Along with the standard fracture test procedure for J-R curve, Q as the second parameter governing stress triaxiality nearby the crack tip is measured from the displacements nearby the side necking which occurs near the crack tip on the lateral surface of a fracture specimen. The displacements nearby the side necking are measured from the digital images taken during the fracture experiment based on Stereoscopic Digital Photography (SDP) and high resolution Digital Image Correlation (DIC) software. The crack length is monitored by Direct Current Potential Drop (DCPD) method and the J-R curve is determined according to ASTM standard E1737-96. The effects of crack length, specimen geometry and thickness of specimen are studied, which are included in the toughness locus Jc-Q.

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.

Structural integrity evaluation for interference-fit flywheels in reactor coolant pumps of nuclear power plants

2005 ◽  
Vol 19 (11) ◽  
pp. 1988-1997 ◽  
Author(s):  
June-soo Park ◽  
Ha-cheol Song ◽  
Ki-seok Yoon ◽  
Taek-sang Choi ◽  
Jai-hak Park
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Interference Fit ◽  
Reactor Coolant
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