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.

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.

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.

Application of Surface Stress Improvement Technologies to Nuclear Power Plant Components

2005 ◽  
Author(s):  
Ronald J. Payne ◽  
Stephen Levesque
Keyword(s):  
Water Stress ◽  
Stress State ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Surface Stress ◽  
Nuclear Power ◽  
Alloy 600 ◽  
Contributing Factors ◽  
Primary Water ◽  
Plant Components

The susceptibility of Alloy 600 to Primary Water Stress Corrosion Cracking (PWSCC) has proven detrimental to several nuclear power plant components. Repair, modification or replacement of the components to mitigate the effects of PWSCC on Alloy 600 has been deemed necessary. In some cases, repair or replacement of plant components can be exorbitantly expensive; therefore, modification of the components is necessary to keep the plant operable. A form of modification is surface stress improvement, which alters the stress state of the material. Changing the stress state of the material eliminates one of the contributing factors required for the propagation of PWSCC. This paper discusses the application of surface stress improvement technologies to commercial nuclear power plant components and provides insight to where these technologies can be employed in the future.

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