Study on the Influence of Strain Rate on Crack Initiation and Growth in Simulated Reactor Coolant Environment of Type 316 Stainless Steel

2018 ◽  
Author(s):  
Takahisa Nose ◽  
Takao Nakamura ◽  
Takanori Kitada
Keyword(s):  
Crack Initiation ◽  
Strain Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Crack Initiation ◽  
Crack Size ◽  
Strain Rates ◽  
3 Dimensional ◽  
Reactor Coolant ◽  
Cumulative Fatigue Damage

In order to conduct effective and rational maintenance activity of components in nuclear power plants, it is proposed to manage fatigue degradation based on crack size corresponding to an extent of cumulative fatigue damage. The purpose of this study focuses on the influence of strain rate in simulated reactor coolant environment for fatigue crack initiation and growth. 3-dimensional replica observations were conducted for environmental fatigue test specimens in different strain rates. Crack initiation and growth were observed in the experiments. It is clarified that low strain rate influences crack propagation and coalescence and increases crack growth rate that finally decrease fatigue life.

An Analytical Model for Thermal Fatigue Crack Initiation and Propagation in Mixing Zones of Piping Systems

2003 ◽  
Author(s):  
Ste´phanie Musi ◽  
Fre´de´ric Beaud
Keyword(s):  
Crack Initiation ◽  
Analytical Model ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Crack Initiation ◽  
Crack Initiation And Propagation ◽  
Thermal Fatigue Crack ◽  
Initiation And Propagation ◽  
Piping Systems ◽  
Different Temperatures

Nuclear power plants piping systems necessarily include connections of branches conveying fluids at different temperatures. Thermal-hydraulic fluctuations arising from the turbulent mixing of the flows possibly affect the inner wall of the pipes and lead to fatigue damage. This paper aims at presenting an analytical model for the mixing zone thermal-mechanical problem and consecutive crack initiation and propagation analyses.

scholarly journals Microstructure Evolution and Surface Cracking Behavior of Superheavy Forgings during Hot Forging

2018 ◽  
Vol 2018 ◽  
pp. 1-9 ◽  
Author(s):  
Zhenhua Wang ◽  
Hongpeng Xue ◽  
Deli Zhao
Keyword(s):  
Grain Boundary ◽  
Dynamic Recrystallization ◽  
Strain Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Hot Forging ◽  
Model Material ◽  
Grain Structure ◽  
High Angle ◽  
Cracking Behavior

In recent years, superheavy forgings that are manufactured from 600 t grade ingots have been applied in the latest generation of nuclear power plants to provide good safety. However, component production is pushing the limits of the current free-forging industry. Large initial grain sizes and a low strain rate are the main factors that contribute to the deformation of superheavy forgings during forging. In this study, 18Mn18Cr0.6N steel with a coarse grain structure was selected as a model material. Hot compression and hot tension tests were conducted at a strain rate of 10−4·s−1. The essential nucleation mechanism of the dynamic recrystallization involved low-angle grain boundary formation and subgrain rotation, which was independent of the original high-angle grain boundary bulging and the presence of twins. Twins were formed during the growth of dynamic recrystallization grains. The grain refinement was not obvious at 1150°C. A lowering of the deformation temperature to 1050°C resulted in a fine grain structure; however, the stress increased significantly. Crack-propagation paths included high-angle grain boundaries, twin boundaries, and the insides of grains, in that order. For superheavy forging, the ingot should have a larger height and a smaller diameter.

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

Sensitivity Analysis of Major Parameters on Piping Failure Probabilities

2004 ◽  
Vol 261-263 ◽  
pp. 821-826
Author(s):  
Sung Gyu Jung ◽  
Chang Soon Lee ◽  
In Gyu Park ◽  
Se Hwan Lee ◽  
Tae Eun Jin
Keyword(s):  
Sensitivity Analysis ◽  
Size Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Crack Size ◽  
Deterministic Approach ◽  
Inspection Interval ◽  
Different Types ◽  
Failure Probabilities

In-service inspections (ISI) of pipes in the nuclear power plants are currently performed based on mandated requirements in the ASME Section XI, which is based on deterministic approach of the critical welds. The 20 years of ISI experience in U.S.A. has revealed less correlation between the critical welds and actual failures, and much conservatism in current ISI requirements. To reduce those problems, risk-informed ISI technology has been developed and proved to be useful. This paper presented a method for predicting piping failure probabilities in an application of risk-informed ISI, and analyzed the effect of input parameters on piping failure probabilities. Results generated using this approach revealed that the calculated failure probabilities can be sensitive to the different types of stressors, crack size distribution, inspection interval, etc..

A Probabilistic Micromechanical Code for Predicting Fatigue Life Variability: Model Development and Application

2005 ◽  
Vol 128 (4) ◽  
pp. 889-895 ◽  
Author(s):  
K. S. Chan ◽  
M. P. Enright
Keyword(s):  
Fatigue Crack ◽  
Fatigue Life ◽  
Fatigue Crack Growth ◽  
Crack Initiation ◽  
Crack Growth ◽  
Model Development ◽  
Fatigue Crack Initiation ◽  
Crack Size ◽  
Crack Initiation Life ◽  
Variability Model

This paper summarizes the development of a probabilistic micromechanical code for treating fatigue life variability resulting from material variations. Dubbed MICROFAVA (micromechanical fatigue variability), the code is based on a set of physics-based fatigue models that predict fatigue crack initiation life, fatigue crack growth life, fatigue limit, fatigue crack growth threshold, crack size at initiation, and fracture toughness. Using microstructure information as material input, the code is capable of predicting the average behavior and the confidence limits of the crack initiation and crack growth lives of structural alloys under LCF or HCF loading. This paper presents a summary of the development of the code and highlights applications of the model to predicting the effects of microstructure on the fatigue crack growth response and life variability of the α+β Ti-alloy Ti-6Al-4V.

Strain Rate Effects in SA-106 Carbon Steel Pipe

1982 ◽  
Vol 104 (1) ◽  
pp. 31-35 ◽  
Author(s):  
D. Peterson ◽  
J. E. Schwabe ◽  
D. G. Fertis
Keyword(s):  
Carbon Steel ◽  
Yield Strength ◽  
Strain Rate ◽  
Tensile Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Experimental Modeling ◽  
Steel Pipe ◽  
Strain Rate Effects ◽  
Rate Effects

Experiments were performed to measure the effect of strain rate on the tensile properties of SA-106 carbon steel pipe, in support of analysis and experimental modeling of postulated pipe whip in nuclear power plants. It was observed that increasing the strain rate from 4 × 10−4 to 4 s−1 raised the yield strength by approximately 30 percent.

Enhanced ASME Section XI Appendix L Flaw Tolerance Procedure

2005 ◽  
Author(s):  
S. R. Gosselin ◽  
F. A. Simonen ◽  
R. G. Carter ◽  
J. M. Davis ◽  
G. L. Stevens
Keyword(s):  
Fatigue Crack ◽  
Crack Initiation ◽  
Nuclear Power ◽  
Damage Tolerance ◽  
Fatigue Crack Initiation ◽  
Flaw Detection ◽  
Flaw Tolerance ◽  
Detection Capabilities ◽  
American Society ◽  
Operating Plant

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code has worked since the early 1990s to develop guidelines for the nuclear power industry to evaluate the serviceability of components that are adversely subjected to fatigue stresses. Results of this work formed the basis for a non-mandatory Appendix L that became part of the 1996 Addenda to the 1995 Edition of Section XI [1]. A key part of this appendix was the introduction of a damage tolerance based examination strategy designed to assure that the component will operate reliability between subsequent inspections. Since the issuance of Appendix L, new data on the flaw detection capabilities have become available, and these data show that ultrasonic inspections can detect flaws much smaller than assumed during the original development of Appendix L. This information allows for significantly smaller sizes for initial postulated flaws. Additionally, Appendix L did not address the potential for multiple fatigue crack initiation sites for a given location on the component. In 1999 the ASME Working Group on Operating Plant Criteria (WGOPC) re-established the Task Group on Operating Plant Fatigue Assessments (TGOPFA) to address these concerns. This paper summarizes the research results, supporting computations, and technical bases for TGOPFA recommended Appendix L improvements. A detailed sample calculation is presented for a PWR charging nozzle-to-pipe weld.

Fatigue Effect of RCS Branch Line by Thermal Stratification

2003 ◽  
Author(s):  
Hag-Ki Youm ◽  
Kwang-Chu Kim ◽  
Man-Heung Park ◽  
Tea-Eun Jin ◽  
Sun-Ki Lee ◽  
...  
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Methodology ◽  
Element Analysis ◽  
Branch Line ◽  
Reactor Coolant ◽  
Stratification Effect ◽  
Fatigue Evaluation ◽  
Coolant System

Recent events reported at a number of nuclear power plants worldwide have shown that thermal stratification, cycling, and striping in piping can cause excessive thermal stress and fatigue on the piping material. These phenomena are diverse and complicated because of the wide variety of geometry and thermal hydraulic conditions encountered in reactor coolant system. Thermal stratification effect of re-branched lines is not yet considered in the fatigue evaluation. To evaluate the thermal load due to turbulent penetration, this paper presents a fatigue evaluation methodology for a branch line of reactor coolant system with the re-branch line. The locations of fatigue monitoring and supplemented inspections are discussed as a result of fatigue evaluations by Interim Fatigue Management Guideline (ITFMG) and detail finite element analysis. Although the revised CUF was increased less than 50 %, the CUF values for some locations was greater than the ASME Code limits.

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