scholarly journals Implementation of a Performance Evaluation System for Nondestructive Testing Methods

2016 ◽  
Vol 2 (2) ◽  
pp. 44
Author(s):  
Daniel Algernon ◽  
Sascha Feistkorn ◽  
Michael Scherrer
Keyword(s):  
Nondestructive Testing ◽  
Evaluation System ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Analysis Tool ◽  
Performance Evaluation System ◽  
Test Pieces ◽  
Inspection Systems ◽  
Important Means

<p>Nondestructive Testing (NDT) is an important means to ensure structural integrity and safe operation of components in many industries, as for example nuclear power plants, aerospace or civil engineering. Within the qualification of Nondestructive Testing personnel as well as validation of NDT procedures, practical demonstrations on test blocks with realistic flaws play a key role. Adequate test pieces need to be designed according to specific criteria such as quantity, shape, orientation, size and position of the test flaws, depending on the requirements of the national codes and standards in the specific industries. The performance of the candidates and inspection systems is quantified and analyzed with respect to criteria, such as detection, positioning, characterization as well as length and height sizing of flaws. Statistic measures are applied to express the resulting accuracy and overall performance. The indication reports obtained from different candidates contain ample information, which might not appear evident at first sight. The complexity of the situation requires an intelligent extraction of the information from the data. An analysis tool <em>IndEva</em> was developed to handle this complexity and provide an accurate, detailed and reliable evaluation of inspection systems and personnel. Besides the plain evaluation regarding the fulfilment of the qualification requirements, critical test flaws as well as test block sections, which are likely to cause false positive indications can be identified. Statistic results display the dependency of the system performance on various parameters and parameter combinations to provide a clear picture of the performance. Country-specific evaluation standards can be applied and compared, especially with regard to the continuous improvement of the qualification methodology.</p>

A study on accumulator analysis for the valve performance evaluation system of nuclear power plants

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 39-44
Author(s):  
K. Ryu ◽  
T. Lee ◽  
D. Baek ◽  
J. Park ◽  
N. Kim
Keyword(s):  
Latent Heat ◽  
Temperature Difference ◽  
Evaluation System ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Capacity ◽  
Test System ◽  
Performance Evaluation System ◽  
High Temperature Steam

Abstract To evaluate the valves used in the nuclear power plants are working properly under the required conditions, the performance and capacity test should be performed. In the test system, the accumulator was employed to control the large amount of high pressure and high temperature steam generated in the boiler precisely. In the accumulating process, the steam is often condensed. In order to prevent condensation, it is needed to install heaters and preheat the accumulator. However, if the size of the accumulator becomes large, the installation of the heater may not be easy. Therefore, when the test is conducted, the system was preheated by the latent heat generated from the phase change. Insufficient thermal insulation may cause temperature differences and it can cause mechanical problems in the accumulator structure. If insulation is sufficient, the temperature difference is indicated by the height. As the cooled condensate moves downwards, the condensate is discharged by the drain valve control and the temperature difference of the structure can be disappeared. The results of this paper can be applied to the conceptualization of equipment that uses latent heat and for the design of high-precision steam experimental devices or the design of high-capacity steam utilization systems.

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

2004 ◽  
Author(s):  
J. C. Kim ◽  
J. B. Choi ◽  
Y. H. Choi
Keyword(s):  
Virtual Reality ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Evaluation System ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Virtual Reality Environment ◽  
Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

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

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

Electrical Power System Modeling of Atucha II Nuclear Power Plant Using etap

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Nicolás Alejandro Malinovsky
Keyword(s):  
Power System ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Electrical Power ◽  
Analysis Tool ◽  
Plant Design ◽  
Electrical Power System ◽  
Power Circuit ◽  
Electrical Systems

This work shows the introduction of the Electrical Power System Analysis (etap) software as a calculation and analysis tool for power electrical systems of the nuclear power plants (NPP) under the orbit of Nucleoeléctrica Argentina S.A (NASA). Through the use of the software, the model of the electrical power system of the Atucha II NPP was developed. To test the functionality of the modeled electrical power circuit, studies of load flow and short-circuit analysis were conducted, yielding satisfactory results, which were contrasted with the plant design values. Once the model has been validated, this will be the basis for carrying out different studies in the plant through simulation. Furthermore, with the incorporation of etap as a fundamental calculation and analysis tool for power electrical systems at the company's engineering departments, it is expected to improve the safety, operation, quality, reliability, and maintenance of both the Atucha II NPP electrical power system and the other nuclear power plants operated by Nucleoeléctrica Argentina S.A.

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.

Pre-Operational Seismic Walk-Through of NPPs in India

2002 ◽  
Author(s):  
R. S. Soni ◽  
R. K. Mishra ◽  
M. K. Agrawal ◽  
G. R. Reddy ◽  
H. S. Kushwaha ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Seismic Event ◽  
Structural Integrity ◽  
Functional Performance ◽  
Design Intent ◽  
Detailed Procedure ◽  
Construction Activity ◽  
Major Emphasis

In nuclear power plants, it is essential to design the various safety and safety related systems and components of the plant in such a manner that they maintain their structural integrity as well as serve their functional performance during a seismic event. The pre-operational seismic walk-through helps in ensuring the installation of various seismic supports as per design intent, identifying the areas where supports are inadequate, identifying the interaction concerns between the systems of various safety classes and locating the various undesired loose, untied / unanchored components, tools, etc. used during the construction activity. A detailed procedure for the pre-operational seismic walk-through of the NPPs was therefore, prepared. Since the types and locations of seismic supports for the various systems and components of the plant had been already reviewed, the major emphasis during the walk-through was laid on their proper installation.

Development of Robotic Inspection System for Small Pipelines

2000 ◽  
Author(s):  
Jinwu Qian ◽  
Yanan Zhang ◽  
Weiming Cheng ◽  
Linyong Shen ◽  
Jianliang Su ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Inspection System ◽  
Chemical Plants ◽  
Micro Robot ◽  
Potential Applications ◽  
Inspection Systems ◽  
Development Goals ◽  
Further Development

Automated probing and inspection inside small pipelines have become a hot topic among the micro-robot researchers in both universities and companies worldwide. The reason for that is the potential applications in nuclear power plants (PWR), civil engineering (gas and water) and in chemical plants and so on. This paper outlines the R & D activities on robotic inspection systems for 20mm-diameter pipelines conducted at Shanghai University in collaboration with NDT Center for Nuclear Industry. The locomotion mechanism comparison and synthesis are covered first. Several robotic inspection systems and different locomotion mechanisms are presented. Further development goals underway are briefly discussed.

Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

Overview of Recent NUGENIA Activities Particularly in Relation to Structural Integrity

2015 ◽  
Author(s):  
John Sharples ◽  
Elisabeth Keim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Nuclear Research ◽  
Joint Research ◽  
Non Profit ◽  
Integrity Assessment ◽  
Industry Research ◽  
Implementation And Dissemination ◽  
Technical Area

NUGENIA, an international non-profit association founded under Belgian legislation and launched in March 2012, is dedicated to nuclear research and development (R&D) with a focus on Generation II and III power plants. NUGENIA is the integrated framework between industry, research and safety organisations for safe, reliable and competitive nuclear power production, and is aimed at running an open innovation marketplace, to promote the emergence of joint research and to facilitate the implementation and dissemination of R&D results. The technical scope of NUGENIA consists of eight technical areas. One of these areas, Technical Area 4, is associated with the structural integrity assessment of systems, structures and components. A brief overview of recent NUGENIA activities in general is provided in this paper and a specific focus is given on developments in relation to Technical Area 4.

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