Introduction of JSME Rules on Fitness for Service for CASS Piping and Proposal for Improvements

2013 ◽  
Author(s):  
Kiminobu Hojo ◽  
Masayuki Kamaya
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Brittle Fracture ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Japan Society ◽  
Two Parameter

The Japan Society of Mechanical Engineers (JSME) Code Rules on Fitness for Service (FFS) for Nuclear Power Plants describe a flaw evaluation procedure for stainless steel piping including cast austenitic stainless steel (CASS) piping. It consists of three methods; limit load, elastic plastic fracture mechanics (EPFM) and two-parameter (covering failure modes from brittle fracture to limit load) methods. This paper describes a brief introduction of the flaw evaluation procedure for CASS piping in the JSME rules. Some improvements for the current rules are also proposed.

Probabilistic Models of Reliability of Cast Austenitic Stainless Steel Piping

2011 ◽  
Author(s):  
Haiyang Qian ◽  
David Harris ◽  
Timothy J. Griesbach
Keyword(s):  
Fracture Toughness ◽  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Tensile Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Probabilistic Models ◽  
Probabilistic Approach ◽  
Delta Ferrite ◽  
Steel Piping

Thermal embrittlement of cast austenitic stainless steel piping is of growing concern as nuclear power plants age. The difficulty of inspecting these components adds to the concerns regarding their reliability, and an added concern is the presence of known defects introduced during the casting fabrication process. The possible presence of defects and difficulty of inspection complicate the development of programs to manage the risk contributed by these embrittled components. Much work has been done in the past to characterize changes in tensile properties and fracture toughness as functions of time, temperature, composition, and delta ferrite content, but this work has shown a great deal of scatter in relationships between the important variables. The scatter in material correlations, difficulty of inspection and presence of initial defects calls for a probabilistic approach to the problem. The purpose of this study is to describe a probabilistic fracture mechanics analysis of the maximum allowable flaw sizes in cast austenitic stainless steel piping in commercial power reactors. Attention is focused on fully embrittled CF8M material, and the probability of failure for a given crack size, load and composition is predicted considering scatter in tensile properties and fracture toughness (fracture toughness is expressed as a crack growth resistance relation in terms of J-Δa). Random loads can also be included in the analysis, with results generated by Monte Carlo simulation. This paper presents preliminary results for CF8M to demonstrate the sensitivity of key input variables. The outcome of this study is the flaw sizes (length and depth) that will fail with a given probability when a given load is applied.

Current in-service inspections of austenitic stainless steel and dissimilar metal welds in light water nuclear power plants

1994 ◽  
Vol 151 (2-3) ◽  
pp. 539-550 ◽  
Author(s):  
Ludwig von Bernus ◽  
Werner Rathgeb ◽  
Rudi Schmid ◽  
Friedrich Mohr ◽  
Michael Kröning
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water ◽  
Dissimilar Metal

scholarly journals Parameters influencing the low-cycle fatigue life of materials in pressure water reactor nuclear power plants / Parametry ovlivòující únavové chování materiálù pro tlakovodní jaderné reaktory v režimu nízkocyklové únavy

2015 ◽  
Vol 59 (3) ◽  
pp. 91-98
Author(s):  
V. Šefl
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Carbon Steel ◽  
Austenitic Stainless Steel ◽  
Literature Review ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Cycle Fatigue

Abstract In this literature review we identify and quantify the parameters influencing the low-cycle fatigue life of materials commonly used in nuclear power plants. The parameters are divided into several groups and individually described. The main groups are material properties, mode of cycling and environment parameters. The groups are further divided by the material type - some parameters influence only certain kind of material, e.g. sulfur content may decreases fatigue life of carbon steel, but is not relevant for austenitic stainless steel; austenitic stainless steel is more sensitive to concentration of dissolved oxygen in the environment compared to the carbon steel. The combination of parameters i.e. conjoint action of several detrimental parameters is discussed. It is also noted that for certain parameters to decrease fatigue life, it is necessary for other parameter to reach certain threshold value. Two different approaches have been suggested in literature to describe this complex problem - the Fen factor and development of new design fatigue curves. The threshold values and examples of commonly used relationships for calculation of fatigue lives are included. This work is valuable because it provides the reader with long-term literature review with focus on real effect of environmental parameters on fatigue life of nuclear power plant materials.

Fracture Assessment of Austenitic Stainless Steel Piping With Multiple Flaws in Heat-Affected Zone

2010 ◽  
Author(s):  
Chihiro Narazaki ◽  
Toshiyuki Saito ◽  
Masao Itatani ◽  
Takuya Ogawa ◽  
Takao Sasayama
Keyword(s):  
Stainless Steel ◽  
Heat Affected Zone ◽  
Nuclear Power ◽  
Power Plants ◽  
Estimation Method ◽  
Limit Load ◽  
Low Carbon ◽  
Fracture Assessment ◽  
Steel Piping ◽  
Weld Heat

Stress corrosion cracking (SCC) has been observed as circumferential multiple flaws in the weld heat-affected zone of primary loop recirculation system piping and core shrouds made of low carbon stainless steel. In the Japan Society of Mechanical Engineers code, Rules on Fitness-for-Service for Nuclear Power Plants, there is no fracture assessment of piping with multiple flaws which are not subject to flaw combination rule criteria. Through fracture testing of piping with two circumferential flaws in the weld heat-affected zone, the limit load estimation method was used for fracture assessment of stainless steel piping.

Probabilistic Study of SCC Initiation Stage Based on Statistical Analysis in 316L Austenitic Stainless Steel

2013 ◽  
Author(s):  
Ryohei Ihara ◽  
Masahito Mochizuki ◽  
Shinji Fujimoto
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Probability Density ◽  
Nuclear Power ◽  
Power Plants ◽  
Crack Depth ◽  
Probability Density Distribution ◽  
Initiation Time ◽  
Density Distributions ◽  
Probability Density Distributions

Stress corrosion cracking (SCC) has been observed near the welded zones of pipes made of austenitic stainless steel type 316L in boiling water reactors (BWRs). For safety, the lifetime assessment of structures in nuclear power plants is performed on the assumption that an initial crack exists. It is said that the lifetime of a structure is greatly affected by the micro crack process of initiation, growth, and coalescence. In this process, SCC has a probabilistic feature caused by its corrosion behavior and the metal microstructure. Therefore, probabilistic evaluation of the micro crack process is useful for the lifetime assessment of nuclear power plant structures. In this study, slow strain rate testing (SSRT) was performed in a simulated BWR operating environment. Statistical analyses were conducted for micro cracks initiated during SSRT, and histograms of the probability density distributions for crack length were obtained. The probability density distributions for crack depth were estimated based on the aspect ratio of SCC. Then, the time variations of the probability density distributions for crack depth were expressed as a function of stress. As a result, the initiation time and the probability density distribution for crack depth at the initiation time of macroscopic SCC were obtained.

Studies on the Correlation Between Natural Magnetization and Mechanical Damages of an Austenitic Stainless Steel

2010 ◽  
Author(s):  
Hongmei Li ◽  
Zhenmao Chen ◽  
Tianfei Zhao
Keyword(s):  
Magnetic Field ◽  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Mechanical Damage ◽  
Magnetic Memory ◽  
Lateral Notch ◽  
Good Possibility ◽  
Measured Magnetic Field

Austenitic stainless steel has been extensively used in nuclear power plants (NPP). The nondestructive Testing (NDT) of its mechanical damage at pre-crack state is very important for the safety assessment of a NPP. Aiming at to develop a new NDT method for inspecting mechanical damage before the initiation of macro cracks, the correlation between the natural magnetization and the mechanical damage is experimentally investigated for a typical austenitic stainless steel - SUS304. In the experiments, simple tensile loads were applied to lateral notch specimens to generate states of different plastic damages, and the corresponding natural magnetic field and residual strain distribution were measured after each loading cycle. The distribution of natural magnetization was analyzed based on the measured magnetic field signals in view of the principle of the metal magnetic memory phenomenon, and the dependence of the magnetization on the mechanical damage was discussed. The experimental results reveal that there is a good possibility to detect mechanical damages by measuring the natural magnetic field.

Development of an Integrity Evaluation System for Steam Generator Tubes in a Nuclear Power Plant

2007 ◽  
Vol 120 ◽  
pp. 157-162
Author(s):  
J.C. Kim ◽  
Sang Min Lee ◽  
Yoon Suk Chang ◽  
Jae Boong Choi ◽  
Young Jin Kim ◽  
...  
Keyword(s):  
Steam Generator ◽  
Evaluation System ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Efficient System ◽  
Reference Pressure ◽  
Steam Generator Tubes

Steam generators working in nuclear power plants convert water into steam from heat produced in the reactor core and each of them contains from 3,000 to 16,000 tubes. Since these tubes constitute one of primary barriers under radioactive and high pressure condition, the integrity should be maintained carefully during the operation. The objective of this research is to introduce an integrity evaluation system for steam generator tubes as a substitute of well-trained engineers or experts. For this purpose, a couplet examination has been carried out on the complicated evaluation procedure and an efficient system named as STiES was developed employing three representative integrity evaluation methods: fracture mechanics analysis (crack driving force diagram and J-integral/Tearing modulus method) and limit load method. Exemplary analyses for steam generator tubes with various types of flaws showed good applicability of the proposed integrity evaluation system. So, it is anticipated that the system can be used for the calculation of reference pressure to decide either the continued operation or repair until next outage.

Development of Load Multipliers (Z-Factors) in Elastic-Plastic Fracture Mechanics Evaluation in Rules on Fitness-for-Service for Nuclear Power Plants

2012 ◽  
Author(s):  
Seiji Asada ◽  
Masao Itatani ◽  
Naoki Miura ◽  
Hideo Machida
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Fracture Mechanics ◽  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Elastic Plastic ◽  
Surface Flaws ◽  
Steel Piping

Not only nonmandatory Appendix C, “Evaluation of Flaws in Piping,” in ASME Boiler & Pressure Vessel Code Section XI but also Appendix E-9, “Elastic-Plastic Fracture Mechanics Evaluation,” in the JSME Rules on Fitness-for-Service for Nuclear Power Plants use the load multiplier Z-factor that is applied to elastic-plastic fracture mechanics evaluation for a circumferential flaw of austenitic stainless steel piping and ferritic steel piping. The Z-factor is defined as the ratio of the limit load to the load at fracture load. Basically, the Z-factor equations were conservatively formulated by using the Z-factors for circumferential through-wall flaws. However, the Codes require flaw evaluation for circumferential surface flaws. Accordingly, Z-factors for circumferential surface flaws should be developed to have the consistency. Therefore Z-factor equations of austenitic stainless steel piping and ferritic steel piping have been developed for circumferential surface flaws.

Analytical Study of the Relaxation of Welding Residual Stress by Excessive Loading for Austenitic Stainless Steel Piping Welds

2009 ◽  
Author(s):  
Jinya Katsuyama ◽  
Kunio Onizawa
Keyword(s):  
Residual Stress ◽  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Welding Residual Stress ◽  
Low Carbon ◽  
Three Dimensional Model ◽  
Butt Weld

Welding residual stress is one of the most important factors of stress corrosion cracking (SCC) for austenitic stainless steel in pressure boundary piping in nuclear power plants. The effect of excessive loading, such as an earthquake, on the residual stress was evaluated by three-dimensional analyses based on finite element method (FEM). The FEM analyses were performed using three-dimensional model for a 250A piping butt weld of low carbon stainless steel of Type 316L. A welding simulation method used in this work is based on the moving heat source with the double ellipsoid model and was confirmed by comparing with the experimental measurements. After conducting welding residual stress simulation, several loading patterns of bending moment and uni-axial displacements have been applied to a model by varying amount of moment and displacement. The analyses indicated that higher loading to bending and axial stresses caused higher relaxation of welding residual stress near piping welds. The difference in the effect of loading direction was observed for both cases. It is concluded that the SCC growth rate might be decreased as loading level increased.

High cycle thermal fatigue Issues in PWR nuclear power plants, life time improvement of some austenitic stainless steel components

2007 ◽  
Vol 104 (3) ◽  
pp. 156-162 ◽  
Author(s):  
J. -A. Le Duff ◽  
A. Lefrançois ◽  
Y. Meyzaud ◽  
J.-Ph. Vernot ◽  
D. Martin ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Life Time ◽  
High Cycle Thermal Fatigue
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