scholarly journals Influence of Microstructural Changes’ Effects on the Linear and Nonlinear Ultrasonic Parameters of Cast Stainless Steels

2020 ◽  
Vol 10 (10) ◽  
pp. 3476
Author(s):  
Yu-Ju Lin ◽  
Che-Hua Yang ◽  
Jiunn-Yuan Huang

In this research, some nondestructive ultrasonic techniques were employed to inquire into the effect of microstructural changes induced by thermal aging and cold work on the ultrasonic response. As thermal embrittlement is a risk to the safety of nuclear power plants, a nondestructive detection method has to be developed for on-site monitoring. The austenitic stainless steel with δ-ferrite specimens were used to study the behavior of microstructural changes caused by age-treating and cold work and then examined by the velocity, attenuation, and nonlinear ultrasonic technique. The variations of the linear and the nonlinear ultrasonic parameters were related to the microstructural changes. Additionally, the experimental results suggest that the ultrasonic nonlinearity parameter of cast stainless steel is determined by the microstructure evolution caused by spinodal decomposition and the phase precipitation process.

2014 ◽  
Vol 487 ◽  
pp. 54-57 ◽  
Author(s):  
Meng Yu Chai ◽  
Li Chan Li ◽  
Wen Jie Bai ◽  
Quan Duan

304 stainless steel and 316L stainless steel are conventional materials of primary pipeline in nuclear power plants. The present work is to summarize the acoustic emission (AE) characteristics in the process of pitting corrosion of 304 stainless steel, intergranular corrosion of 316L stainless steel and weldments of 316L stainless steel. The work also discussed the current shortcomings and problems of research. At last we proposed the coming possible research topics and directions.


Author(s):  
Haiyang Qian ◽  
David Harris ◽  
Timothy J. Griesbach

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.


2013 ◽  
Vol 684 ◽  
pp. 325-329 ◽  
Author(s):  
Tian Liang ◽  
Xiao Qiang Hu ◽  
Xiu Hong Kang ◽  
Dian Zhong Li

With about equal amount of austenite and ferrite in volume fraction, duplex stainless steel (DSS) is in advantage of mechanical properties and corrosive behaviors. Hence it is widely applied to the heavy castings for nuclear power plants inshore, such as impellers, pumps and valves. However, lots of cracks usually occur in these castings during manufacturing processes, because it is susceptible to precipitate the brittle intermetallic compound of sigma phase when the castings are exposed from 600 to 1000oC. In this work, the precipitation of sigma phase was observed by optical microscope (OM) and scanning electron microscope (SEM) in a cast DSS named as MAS/6001, which aged at 850oC from 5 to 300 minutes. The effect of sigma phase on the mechanical properties was analyzed by the tensile at room temperature and impact tests at -10°C. The results show that sigma phase in the MAS/6001 steel precipitated simultaneously with the secondary austenite, which obeyed the eutectoid reaction. The interfaces between austenite or secondary austenite and sigma phase were the locations where cracks generated from the void aggregation. Cracks are susceptible to propagate along or cross these interfaces, and to promote the sigma phase breaking-off, which severely deteriorated the mechanical properties.


1994 ◽  
Vol 151 (2-3) ◽  
pp. 539-550 ◽  
Author(s):  
Ludwig von Bernus ◽  
Werner Rathgeb ◽  
Rudi Schmid ◽  
Friedrich Mohr ◽  
Michael Kröning

2012 ◽  
Vol 1475 ◽  
Author(s):  
Raul B. Rebak

ABSTRACTAll the countries that operate commercial nuclear power plants are planning to dispose of the waste in underground geologically stable repositories. The materials being studied for the fabrication of the containers include carbon steel, stainless steel, copper, titanium and nickel alloys. The aim of this work is to review results from research performed using the alloys of interest regarding their resistance to environmentally assisted cracking (EAC) under simulated repository conditions. In general, it is concluded that the environments are mild and that the studied metals may not be susceptible to cracking under the planned emplacement conditions.


Author(s):  
Wei Tang ◽  
Maxim Gussev ◽  
Zhili Feng ◽  
Brian Gibson ◽  
Roger Miller ◽  
...  

Abstract The mitigation of helium induced cracking in the heat affected zone (HAZ), a transition metallurgical zone between the weld zone and base metal, during repair welding is a great challenge in nuclear industry. Successful traditional fusion welding repairs are limited to metals with a maximum of a couple of atomic parts per million (appm) helium, and structural materials helium levels in operating nuclear power plants are generally exceed a couple of appm after years of operations. Therefore, fusion welding is very limited in nuclear power plants structural materials repairing. Friction stir welding (FSW) is a solid-state joining technology that reduces the drivers (temperature and tensile residual stress) for helium-induced cracking. This paper will detail initial procedural development of FSW weld trials on irradiated 304L stainless steel (304L SS) coupons utilizing a unique welding facility located at one of Oak Ridge National Laboratory’s hot cell facilities. The successful early results of FSW of an irradiated 304L SS coupon containing high helium are discussed. Helium induced cracking was not observed by scanning electron microscopy in the friction stir weld zone and the metallurgical zones between the weld zone and base metal, i.e. thermal mechanical affected zone (TMAZ) and HAZ. Characterization of the weld, TMAZ and HAZ regions are detailed in this paper.


2015 ◽  
Vol 59 (3) ◽  
pp. 91-98
Author(s):  
V. Šefl

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.


Author(s):  
Sarah Tioual-Demange ◽  
Gaëtan Bergin ◽  
Thierry Mazet ◽  
Luc de Camas

Abstract The sCO2-4-NPP european project aims to develop an innovative technology based on supercritical CO2 (sCO2) for heat removal to improve the safety of current and future nuclear power plants. The heat removal from the reactor core will be achieved with multiple highly compact self-propellant, self-launching, and self-sustaining cooling system modules, powered by a sCO2 Brayton cycle. Heat exchangers are one of the key components required for advanced Brayton cycles using supercritical CO2. Fives Cryo company, a brazed plates and fins heat exchangers manufacturer, with its expertise in thermal and hydraulic design and brazing fabrication is developing compact, and highly efficient stainless steel heat exchanger solution for sCO2 power cycles, thanks to their heat exchange capability with low pinch and high available flow sections. The aim of the development of this specific heat exchanger technology is to achieve an elevated degree of regeneration. For this matter, plates and fins heat exchanger is a very interesting solution to meet the desired thermal duty with low pressure drop leading to a reduction in size and capital cost. The enhancement of the mechanical integrity of plates and fins heat exchanger equipment would lead to compete with, and even outweigh, printed circuit heat exchangers technology, classically used for sCO2 Brayton cycles. sCO2 cycle conditions expose heat exchangers to severe conditions. Base material selection is essential, and for cost reasons, it is important to keep affordable heat-resistant austenitic stainless steel grades, much cheaper than a nickel-based alloy. Another advantage of high compactness of plates and fins heat exchangers is the diminution of the amount of material used in the heat exchanger manufacturing, decreasing even more its cost. The challenge here is to qualify stainless steel plates and fins heat exchangers mechanical resistance, at cycle operating conditions, and meet with pressure vessels codes and regulations according to nuclear requirements. One critical point in the development of the heat exchangers is the design of the fins. As secondary surface, they allow the maximization of heat transfer at low pressure drop. At the same time mechanical strength has to be guaranteed. To withstand high pressure, fins thickness has to be significant, which makes the implementation complicated. Efforts were dedicated to successfully obtain an optimal shape. Forming of fins was therefore improved compared to conventional techniques. Important work was undertaken to define industrial settings to flatten the top of the fins leading to a maximum contact between the brazing alloy and the fins. Consequently brazed joints quantity is minimized inducing a diminution of the presence of eutectic phase, which is structurally brittle and limits the mechanical strength of the construction. A metallurgical study brings other elements leading to the prevention of premature rupture of the brazed structure. The idea is to determine an optimized solidification path and to identify a temperature range and holding time where the brazed joint is almost free of eutectic phase during the assembly process in the vacuum furnace.


Author(s):  
Chihiro Narazaki ◽  
Toshiyuki Saito ◽  
Masao Itatani ◽  
Takuya Ogawa ◽  
Takao Sasayama

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.


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