Assessment of the Fatigue Life of a PWR Surge Line Including the Effect of Coolant Environment

2009 ◽  
Author(s):  
Gu¨nter Ko¨nig ◽  
Jaroslav Bartonicek ◽  
Horst Rothenho¨fer
Keyword(s):  
Boundary Conditions ◽  
Power Plants ◽  
Measurement Data ◽  
Real Life ◽  
Actual State ◽  
Optimized Design ◽  
Design Phase ◽  
Damage Mechanisms ◽  
Piping Systems ◽  
Surge Line

In Germany, the integrity concept is applied to important piping systems in most of the nuclear power plants. Regarding the framework of this concept, those damage mechanisms that cannot be controlled by analysis have to be excluded using appropriate measures. In most of the cases, these damage mechanisms are a result of local effects (like loads, medium, material characteristics) that cannot be determined exactly in advance and thus cannot be controlled by analysis, reliably. Examples are strain induced corrosion (LCF area) and corrosion fatigue (HCF area). For cases like these and given medium, suitable materials have to be chosen in combination with optimized design, appropriate manufacturing procedures (incl. welding), construction and operation. The loads and the water chemistry in operation have to be monitored and the effectiveness of the measures has to be verified, regularly, taking into account the actual state of knowledge. Regarding these boundary conditions the fatigue evaluations that have been performed until today seem to be sufficient, as experience shows with piping systems where this procedure has been applied. There are usually no significant failures (indication of failures); failures detected have been attributed to violation of the boundary conditions. With this background, there seems to be no need to change this procedure to safeguard the effect of environment. In this paper, the measures to guarantee integrity in design and operation state are discussed, first. Using the example of a surge line and the comprehensive monitoring results of this system the evaluation of fatigue usage and the assessment of the effect of coolant environment is discussed with reference to the ANL approach. Where the ANL approach is meant to be applied only in the design phase of a new reactor its relevance for the operation phase is cross-checked with real life measurement data. The conclusion summarizes where the effect of coolant environment has to be taken into account and gives advice how to find realistic transients for the design phase of new reactors.

Tackling Erosion in Nuclear Piping Systems

2007 ◽  
Author(s):  
Harold M. Crockett ◽  
Jeffrey S. Horowitz
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Training Module ◽  
Droplet Impingement ◽  
Damage Mechanisms ◽  
Accelerated Corrosion ◽  
Computerized Training ◽  
Piping Systems ◽  
Research Plan ◽  
Flow Accelerated Corrosion

Various mechanisms degrade power piping in nuclear power plants. The most important mechanism has been flow-accelerated corrosion (FAC). FAC has caused ruptures and leaks and has led to numerous piping replacements. U.S. utilities are using a combination of EPRI software and aggressive inspection programs to deal with FAC. However, current technology does not deal with erosive forms of attack including, cavitation erosion, flashing erosion, droplet impingement, and solid particle erosion. These forms of degradation have caused shutdowns and leaks have become a maintenance issue. To deal with these problems EPRI has begun a series of projects in this area. The first of these was a comprehensive report on erosion in piping systems. This work was followed with a computerized training module designed to educate utility engineers about erosive attack. Further steps are planned to deal with these forms of degradation. The first will be a meeting with knowledgeable EPRI and utility engineers to prioritize the damage mechanisms. From this meeting a research plan will be developed. This paper will present a description of erosive damage mechanisms and describe the planned R&D to deal with these mechanisms.

Forced Vibration of Continuous System With Nonlinear Boundary Conditions

1978 ◽  
Vol 100 (3) ◽  
pp. 487-491 ◽  
Author(s):  
T. Watanabe
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Nonlinear Boundary ◽  
Continuous System ◽  
Approximate Solutions ◽  
Nonlinear Boundary Conditions ◽  
Mechanical Equipment ◽  
Single Degree Of Freedom ◽  
Piping Systems

This paper deals with the nonlinear vibration problem concerning mechanical equipment-piping systems in nuclear power plants and others. An analytical method by approximate solutions is introduced for these systems as a continuous system with nonlinear boundary conditions, and some numerical examples are shown. Finally some numerical results obtained as a continuous system are compared with those of a single-degree-of-freedom system.

A Validation of RELAP on Predicting Nuclear Power Plant Phenomena

2018 ◽  
Author(s):  
Ji Soo Ahn ◽  
Michael Bluck ◽  
Matthew Eaton ◽  
Chris Jackson
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Feed Water ◽  
Multiple Channels ◽  
Horizontal Pipe ◽  
One Dimensional ◽  
Piping Systems ◽  
Surge Line ◽  
Input Model

In this study, RELAP5’s capability to simulate thermal stratification under different conditions is assessed. In nuclear power plants (NPPs), thermal stratification can occur in the following locations: pressurizer, piping systems such as hot legs, cold legs, surge lines, and cooling tanks if available. In general, thermal stratification in a horizontal pipe could not be simulated by RELAP5 due to the inherent one-dimensional setting. Moreover, RELAP5 failed to simulate turbulent penetration which was often a pre-requisite prior to thermal stratification in a pipe. This type of situation could arise in connection between hot leg and surge line, spray lines, feed water lines, etc. It is recommended that for this type of problem CFD be used. In the literature, it was found that RELAP5 was capable of simulating thermal stratification in a pool or a tank-like component if multiple channels and crossflow junctions were used. However, due to uncertainties associated with the input model, the current RELAP5 model failed to reproduce experimental data and therefore further investigation would be required to identify the sources of error.

scholarly journals Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

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.

Evaluation of Representative Piping Systems Designed by Implicit Fatigue Concept

2008 ◽  
Author(s):  
Shin-Beom Choi ◽  
Sun-Hye Kim ◽  
Yoon-Suk Chang ◽  
Jae-Boong Choi ◽  
Young-Jin Kim ◽  
...  
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Heat Removal ◽  
Environmental Water ◽  
Lessons Learned ◽  
Metal Fatigue ◽  
Aging Effects ◽  
Age Related ◽  
Piping Systems ◽  
Surge Line

NUREG-1801 provides generic aging lessons learned to manage aging effects that may occur during continued operation beyond the design life of nuclear power plant. According to this report, the metal fatigue, among several age-related degradation mechanisms, is identified as one of time-limited aging analysis item. The objective of this paper is to introduce fatigue life evaluation of representative surge line and residual heat removal system piping which was designed by implicit fatigue concept. For the back-fitting evaluation employing explicit fatigue concept, detailed parametric CFD as well as FE analyses results are used. The well-known ASME Section III NB-3600 procedure is adopted for the metal fatigue and NUREG/CR-5704 procedure is further investigated to deal with additional environmental water effects. With regard to the environmental effect evaluation, two types of fatigue life correction factors are considered, such as maximum Fen and individual Fen. As a result, it was proven that a thermal stratification phenomenon is the governing factor in metal fatigue life of the surge line and strain rate is the most important parameter affecting the environmental fatigue life of both piping. The evaluation results will be used as technical bases for continued operation of OPR 1000 plant.

scholarly journals Data Analysis and Model Validation of Natural Gas Transmission Pipeline With Compressor Station

2019 ◽  
Author(s):  
David Cheng
Keyword(s):  
Performance Analysis ◽  
Natural Gas ◽  
Model Validation ◽  
Measurement Data ◽  
Real Life ◽  
Gas Pipeline ◽  
Physical Parameters ◽  
Pressure Drops ◽  
Natural Gas Transmission ◽  
Transmission Pipeline

Abstract Data from the DCS systems provides important information about the performance and transportation efficiency of a gas pipeline with compressor stations. The pipeline performance data provides correction factors for compressors as part of the operation optimization of natural gas transmission pipelines. This paper presents methods, procedure, and a real life example of model validation based performance analysis of gas pipeline. Statistic methods are demonstrated with real gas pipeline measurement data. The methods offer practical ways to validate the pipeline hydraulics model using the DCS data. The validated models are then used as performance analysis tools in evaluating the fundamental physical parameters and assessing the pipeline hydraulics conditions for potential issues influencing pressure drops in the pipeline such as corrosion (ID change), roughness changes, or BSW deposition.

scholarly journals Characterization and Corrections for Clamp-On Fluid Temperature Measurements in Turbulent Flows

2018 ◽  
Vol 10 (3) ◽  
Author(s):  
Bijan Nouri ◽  
Marc Röger ◽  
Nicole Janotte ◽  
Christoph Hilgert
Keyword(s):  
Fluid Flow ◽  
Turbulent Flows ◽  
Power Plants ◽  
Measurement Data ◽  
Temperature Measurements ◽  
Correction Function ◽  
Acceptance Testing ◽  
Fluid Temperature ◽  
Parabolic Trough ◽  
Operation Conditions

A clamp-on measurement system for flexible and accurate fluid temperature measurements for turbulent flows with Reynolds numbers higher than 30,000 is presented in this paper. This noninvasive system can be deployed without interference with the fluid flow while delivering the high accuracies necessary for performance and acceptance testing for power plants in terms of measurement accuracy and position. The system is experimentally validated in the fluid flow of a solar thermal parabolic trough collector test bench, equipped with built-in sensors as reference. Its applicability under industrial conditions is demonstrated at the 50 MWel AndaSol-3 parabolic trough solar power plant in Spain. A function based on large experimental data correcting the temperature gradient between the measured clamp-on sensor and actual fluid temperature is developed, achieving an uncertainty below ±0.7 K (2σ) for fluid temperatures up to 400 °C. In addition, the experimental results are used to validate a numerical model. Based on the results of this model, a general dimensionless correction function for a wider range of application scenarios is derived. The clamp-on system, together with the dimensionless correction function, supports numerous combinations of fluids, pipe materials, insulations, geometries, and operation conditions and should be useful in a variety of industrial applications of the power and chemical industry where temporal noninvasive fluid temperature measurement is needed with good accuracy. The comparison of the general dimensionless correction function with measurement data indicates a measurement uncertainty below 1 K (2σ).

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