Nuclear engineering A modified Auto Associative Kernel Regression method for robust signal reconstruction in Nuclear Power Plant components

2014 ◽  
pp. 953-958 ◽  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Signal Reconstruction ◽  
Kernel Regression ◽  
Regression Method ◽  
Nuclear Engineering ◽  
Plant Components

Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Vanderley Vasconcelos ◽  
Wellington Antonio Soares ◽  
Raissa Oliveira Marques ◽  
Silvério Ferreira Silva Jr ◽  
Amanda Laureano Raso
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Human Factors ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Cooling System ◽  
Human Reliability ◽  
Non Destructive ◽  
Plant Components

Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. This inspection is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI is reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components, such as FMEA (Failure Modes and Effects Analysis) and THERP (Technique for Human Error Rate Prediction). An example by using qualitative and quantitative assessesments with these two techniques to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues, is presented.

What’s New in Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components

2004 ◽  
Author(s):  
Gary Park
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Mechanical Engineering ◽  
Nuclear Industry ◽  
Nondestructive Examination ◽  
American Society ◽  
Plant Components

The nuclear industry is a pretty dynamic industry, in that it is always on the move, changing every time we turn around. For that very reason, there is a need to keep up with the industry by providing changes to American Society of Mechanical Engineering Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” There have been many changes over the last three years. This paper addresses a few of those, but gives a feel for the number of changes from the 2000 Addenda to the 2003 Addenda, there have been a total of approximately 56 changes. Of those changes, 11 were in the repair/replacement requirements, 19 in the inspection requirements, 4 in the evaluation requirements, 18 in the nondestructive examination requirements, and 4 in the administrative requirements. The paper classifies the changes as “Technically Significant,” “Significant,” “Non-Significant,” or “Editorial.” The paper addresses only a few of those changes that were “Technically Significant.” The paper also includes some of the activities that the ASME Section XI Subcommittee is currently working on.

Reforms and Innovations in a Nuclear Engineering Course: Nuclear Power Plant Systems and Equipment

2020 ◽  
Author(s):  
Shanfang Huang ◽  
Jiageng Wang ◽  
Yisheng Hao ◽  
Guodong Liu ◽  
Minyun Liu ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Engineering Students ◽  
Teaching Strategy ◽  
Public Attitudes ◽  
Safety System ◽  
Nuclear Engineering ◽  
Tsinghua University ◽  
Industry Trends

Abstract The Fukushima nuclear accident in Japan caused a significant impact on the nuclear power industry and public attitudes towards nuclear energy. The decreased public acceptance and the regulatory authorities’ stricter requirements of nuclear safety lead to the popularity of advanced safety technologies in scientific research and engineering projects. The demand for highly qualified human resources increases by the gradual recovery of the nuclear power field in China. In order to meet this demand, a series of course innovations are taken at Tsinghua University. Focusing on the course “Nuclear Power Plant Systems and Equipment,” the paper discusses the innovations of the course stimulated by the current industry trends and demands. A brief introduction to the special commissioned-student program at Tsinghua University is given. The paper investigates the meaning and function of the course in the frame of the curriculum plan for nuclear engineering students at Tsinghua University. The personal career plan, the industry outlook, and even the public attitudes contribute to senior students’ attitudes and demands for the course, which is tied closely to the effect of teaching. The paper addresses that the objective of the innovations is to develop a course fixing different students’ demands and help them build their ability to solve practical engineering problems in their future professional careers. The selection of teaching contents and the teaching strategy are discussed. This course takes Westinghouse AP1000 as the major point. And the nuclear power plant systems are taught in a divided way. One is the operation system, and the other is the safety system. This separation is based on the different functions and roles of these two parts and could have advantages in teaching effect. The paper explains the critical points of the systems and innovations of how to deal with course difficulties. There is a corresponding part of the safety system, and this part gets more challenges due to the industry trends. Lectures, group discussions, homework, and presentation projects are discussed. Besides, the paper considers possible efforts for further development of nuclear engineering courses.

Diagnostics of Nuclear Power Plant Components Due to Thickness Measurement by Using Digital Radiography

2010 ◽  
Author(s):  
Masahito Mochizuki ◽  
Satoshi Kanno ◽  
Shunichi Shimizu ◽  
Yuichi Daitou
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Digital Radiography ◽  
Nuclear Power ◽  
Color Image ◽  
Image Intensifier ◽  
Thickness Measurement ◽  
Imaging Plate ◽  
Technical Guide ◽  
Plant Components

Technical Guide for Diagnostics of Power Plant Components Technique due to Thickness Measurement by Digital Radiography, JEAG 4224–2009, was issued from Nuclear Standard Committee of Japan Electric Association on June 2009. Two types of digital radiography are applied to thickness measurement; imaging plate (IP) and color image intensifier (Color I. I. ). Organization and detailed contents of the technical guide is introduced in this paper.

A Unified Constitutive Model With Optimized Parameters for Base and Diffusion Bonded Alloy 800H

2020 ◽  
Author(s):  
Heramb P. Mahajan ◽  
Tasnim Hassan
Keyword(s):  
Elevated Temperature ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Constitutive Model ◽  
Nuclear Power ◽  
Parameter Determination ◽  
Alloy 800H ◽  
Inelastic Analysis ◽  
Creep Fatigue ◽  
Plant Components

Abstract Current ASME Section III, Division 5 code provides elastic, simplified inelastic and inelastic analysis options for designing nuclear power plant components for elevated temperature service. These analyses methods may fail to capture the complex creep-fatigue response and damage accumulation in materials at elevated temperatures. Hence, for analysis and design of the nuclear power plant components at elevated temperature, a full inelastic analysis that can simulate creep-fatigue responses may be needed. Existing ASME code neither provides guidelines for using full inelastic analysis nor recommends the type of constitutive model to be used. Hence, a unified rate-dependent constitutive model incorporating a damage parameter will be developed, and its parameters for base metal will be determined. In addition, a full inelastic analysis methodology using this model to analyze the creep-fatigue performance of components for nuclear power applications will be developed. Base metal 800H (BM800H) data are collected from literature to determine constitutive material model parameters. The parameter determination methodology for a constitutive model is discussed. The optimized parameter set for BM 800H at different temperatures will be presented in the paper. Recommendations are provided on the constitutive model selection and its parameter determination techniques. In the future, this work will be continued for diffusion bonded Alloy 800H (DB800H) material, and obtained parameters will be compared.

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