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.

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.

AP1000 Operation Procedures System and Development

2010 ◽  
Author(s):  
Fang Wen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Flow Process ◽  
Human Factor Engineering ◽  
Data Process ◽  
Operation Procedure ◽  
The Status ◽  
Procedure Classification ◽  
Plant Data ◽  
Westinghouse Electric

This paper makes a brief introduction on AP1000 operation procedure system, including procedure classification, function and composition. In addition, key points of work flow process and the advantages of AP1000 operation procedures are described, among which the application of CPS (computerized procedure system) on AP1000 operation area and human factor engineering are highlighted. CPS, as an advanced procedure system, which is relatively new to existing nuclear power plants in China, does not only have the function of electronic indication for procedures, but also have the ability to monitor plant data, process the data and then present the status of the procedure steps to the reactor operator. Moreover, based on current situation, this paper offers several suggestions on procedure development for Sanmen AP1000 nuclear power project, i.e. first, we can ensure the quality of operation procedures by preparing a precise writer’s guideline, a friendly-interfaced procedure template, an efficient work configuration and an appropriate schedule; then determine the way how we are going to use operation procedures in English version; finally realize CPS Chinesization and localization gradually by digesting and absorbing API 000 technology from Westinghouse Electric Company. This paper gives an intact and systematic discourse on AP1000 operation procedure system and its characteristics. Besides, the latter part of this paper focuses on development of AP1000 operation procedures for Sanmen nuclear power plant and it would be a worthwhile reference for newly-built AP1000 units in China.

scholarly journals Heat Pipe as a Passive Cooling System Driving New Generation of Nuclear Power Plants

2021 ◽  
pp. 30-38
Author(s):  
Ziba Zibandeh Nezam ◽  
Bahman Zohuri
Keyword(s):  
Heat Transfer ◽  
Heat Pipe ◽  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Heat Pipes ◽  
Passive Cooling ◽  
Nuclear Power Reactor ◽  
Pipe System ◽  
New Generation

The technology of the Heat Pipe (HP) system is very well known for scientists and engineers working in the field of thermal-hydraulic since its invention at Las Alamos Nation Laboratory around the 1960s time frame. It is a passive heat transfer/heat exchanger system that comes in the form of either a constant or variable system without any mechanical built-in moving part. This passive heat transfer system and its augmentation within the core of nuclear power reactors have been proposed in the past few decades. The sodium, potassium, or mercury type heat pipe system using any of these three elements for the cooling system has been considered by many manufacturers of fission reactors and recently fusion reactors particularly Magnetic Confinement Fusion (MCF). Integration of the heat pipes as passive cooling can be seen in a new generation of a nuclear power reactor system that is designed for unconventional application field such as a space-based vehicle for deep space or galaxy exploration, planetary surface-based power plants as well as operation in remote areas on Earth. With the new generation of Small Modular Reactor (SMR) in form of Nuclear Micro Reactors (NMR), this type of fission reactor has integrated Alkali metal heat pipes to a series of Stirling convertors or thermoelectric converters for power generation that would generate anywhere from 13kwt to 3Mwt thermal of power for the energy conversion system.

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.

Load Data Evaluation, Stress Analysis and Cycle Counting for the Qualified Thermal Fatigue Assessment of a NPP Component

2015 ◽  
Author(s):  
Jürgen Rudolph ◽  
José Eduardo Maneschy ◽  
Miguel Cisternas ◽  
José Luiz F. Freire ◽  
Felippe M. S. Costa ◽  
...  
Keyword(s):  
Finite Element ◽  
Stress Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Data Evaluation ◽  
Volume Control ◽  
Design Code ◽  
Fatigue Assessment ◽  
Assessment Approach ◽  
Plant Data

Qualified fatigue assessment based on realistic input data constitutes an essential part of an ageing management strategy for Nuclear Power Plants. In this context and as a continuation of a previous paper PVP2014-28716 the requirements of load data evaluation, stress analysis and cycle counting are detailed based on a real world example from a Brazilian Nuclear Power Plant. One essential prerequisite of any fatigue assessment approach is the availability of realistic load data. In the present analysis, selected operational plant data from the period 2003 to 2012 are used. One further prerequisite is the accurate component stress analysis based on a transient thermal-mechanical Finite Element Analyses. As an example, a highly loaded nozzle from the Chemical & Volume Control System (CVCS) is chosen to be analyzed. The influences on the fatigue assessment caused by the load-time histories, the stress analysis approaches and the cycle counting method are discussed in detail. The considered operational time period from 2003 to 2012 with respective selected plant data gives a consolidated background. It is one essential aim of the study to show the influence of the load-data input and the (design code conforming) stress analysis method on the resulting calculated cumulative usage factors (CUFs). In the present paper, the stress analysis employs the finite element method. Simplified elastic-plastic (application of ke plasticity factors) procedures are used in order to identify the margins and influences of design and actual loading histories on the resulting CUFs. The paper concludes with a comprehensive picture including quantification and discussion of the different influencing parameters on the resulting CUFs. This reveals margins in the fatigue design process and solutions of coping with the design code requirements.

X-Ray Residual Stress Analysis of Nickel Base Alloys

2014 ◽  
Vol 922 ◽  
pp. 274-279 ◽  
Author(s):  
Toshihiko Sasaki ◽  
Toshiyuki Miyazaki ◽  
Hamiru Ito ◽  
Takashi Furukawa ◽  
Tsuyoshi Mihara
Keyword(s):  
Residual Stress ◽  
Stress Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Ray Method ◽  
X Ray ◽  
Stress Measurements ◽  
Residual Stress Analysis ◽  
Nickel Base ◽  
Steam Generator Tubes

Nickel based alloys are widely used in steam generator tubes for nuclear power plants. 1-D X-ray stress measurements have been used for these alloys. But 1-D method requires large equipment and it is practically impossible to measure stress in power plants. In order to overcome this problem, we adopted 2-D X-ray method which requires significantly smaller equipment. In this paper we report preliminary results of 2-D X-ray stress measurements of nickel based alloys.

A Less-Conservative Approach to Simplified Seismic Analysis for Small Bore Nuclear Piping System

2014 ◽  
Author(s):  
Joon Ho Lee ◽  
In Yeung Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Response Spectra ◽  
Piping System ◽  
Successful Implementation ◽  
Conservative Approach ◽  
Analysis Method ◽  
Support Design ◽  
Floor Response Spectra

Per guidelines for piping system reconciliation such as EPRI NP-6628 (NCIG-14), simplified seismic design methods have been used by nuclear piping designers to deal with small bore piping for many years. These methods are generally based on enveloping the results of rigorous dynamic or conservative static analysis. Small bore piping is generally more flexible with larger margin in support design comparing to large bore piping. Consequently, Classes 2 & 3 piping less than 2-1/2 inch NPS that is analyzed by a conservative “cookbook method” is excluded from the as-built verification actions in IE Bulletin 79-14. The ASME code’s Subsection NF and B31.1 provide recommended pipe spans by pipe size considering only piping weight; therefore, the seismically qualified piping span must be developed for the peak acceleration of the applicable amplified or floor response spectra. Simplified seismic analysis method being considered for the Korean nuclear power plants is based on the Load Coefficient Method provided in Appendix N-1225 of ASME Section III. However, since the simplified analysis method involves conservative and enveloping approach in an effort to comply with applicable requirements and results in an excessive number of supports and unrealistically high support loads, the successful implementation largely depends on how the issues related to the excessive conservatism are resolved when determining allowable pipe spans and support design loads of the piping system. In this paper, simplified engineering equations are presented as a less-conservative approach based on a detailed computer analysis method, which is alternative to the various handbooks and design charts that are based on the conventional hand calculation method.

Experimental Investigation on the Pipe Leakage

2017 ◽  
Author(s):  
Benan Cai ◽  
Qi Zhang ◽  
Yu Weng ◽  
Hongfang Gu ◽  
Haijun Wang
Keyword(s):  
High Temperature ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Normal Operation ◽  
Pipe System ◽  
Humidity Change ◽  
Temperature And Humidity ◽  
Surge Line

Pipelines are widely used in many fields including power industry, petroleum system etc. Pipelines such as the surge line and main pipe are easily subjected to thermal stratification as a result of the non-uniform temperature distribution in the nuclear power plants. Furthermore, pipelines can suffer from thermal fatigue in virtue of long-term uneven stress distribution. When the surge line or main pipe subjected to thermal stratification and thermal fatigue keeps operating for long time, the pipe leakage may happen because of the existence of pipeline crack. The thermal pipeline crack leakage mainly appears in the region with stress concentration. As the pipe system is always covered with thermal insulation layer in the actual nuclear power plants, it is hard for workers to observe pipeline leak, which can have a bad effect on the normal operation. Since the temperature and humidity close to the pipe crack due to leakage can change compared to the normal operation, we can infer from the temperature and humidity changes that the pipe leakage occurs. Based on this idea, the temperature and humidity near the crack of the pipe need to be measured to detect the leakage fields. As the fluids with high pressure and high temperature flow in the pipe system in an actual nuclear power plant, the pipe leakage experiment was performed in the high pressure and high temperature condition. When the fluids with high temperature and pressure leak in the crack, the water will evaporate quickly, which means this process belongs to spray flash evaporation process. The temperature and humidity variations were monitored in the experiment with temperature and humidity probes which have the advantage of responding to the change of temperature and humidity sensitively. The data collection program was mainly written based on the LABVIEW platform. The collecting time step was set 1s. As the measuring position and leakage flux are two key factors for the pipe leakage, the experiment was carried out with different measuring positions and leakage fluxes conditions. The experimental results showed that the leak flux had an important influence on the temperature and humidity near the pipe crack. The temperature and humidity started to change in a very short time with large leak flux. At the same time, the velocity of the temperature and humidity change was high with large leak flux. When the pipe leakage occurred in the location near the temperature and humidity probe, the temperature and humidity responded quickly and the velocity of temperature and humidity change was large. The experiment data can be used for the prediction of the pipe leakage in the nuclear power plants.

A Study of the ASME B&PV Code for Pipe Support and Restraint Design Rules From 1974 to 2013

2015 ◽  
Author(s):  
Phillip E. Wiseman ◽  
Gaurav S. Goel
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
First Generation ◽  
Support Design ◽  
Design Rules ◽  
Linear Elastic ◽  
American Society ◽  
Support Engineering ◽  
Allowable Stress Design ◽  
Nuclear Applications

The pipe support and restraint design rules are based on the linear elastic analysis outlined in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Subsection NF. The subarticle NF-3300, primarily introduced by replicating the Allowable Stress Design (ASD) rules from the AISC Steel Construction Manual, differs from most of the ASME B&PV Code. Additionally, the Code must fulfill the Nuclear Regulatory Guide requirements, i.e. Regulatory Guide 1.124, which mandates the criteria for pipe support design in nuclear applications. The Code has been edited on multiple occasions to refine the rules for pipe support engineering since the inaugural publication of Subsection NF in the 1973 Winter Addenda. The changes in the design rules in the Code have been incorporated to meet the nuclear rules and regulations along with the general needs of the industry. A review of some of these modifications in the design rules provides a better insight on the subarticle since the institution of the first publication of Subsection NF. The changes in the design rules are essential to highlight since the first generation of commercial nuclear power plants constructed 40 years ago are approaching the end of their design life.

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