Thermal Ageing Embrittlement of Casting Duplex Stainless Steels for Nuclear Power Plant

2010 ◽  
Author(s):  
Yuhong Yao ◽  
Jianfeng Wei ◽  
Jiangnan Liu ◽  
Zhengpin Wang ◽  
Yu Wang
Keyword(s):  
Stainless Steels ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Power Plants ◽  
Ageing Time ◽  
Thermal Ageing ◽  
Duplex Stainless Steels ◽  
Impact Properties ◽  
Α Phase ◽  
The Impact

Cast duplex stainless steels (CSS) used for PWR pipes are degraded due to thermal ageing embrittlement during long-term service at 288 °C to 327 °C. Z3CN20-09M Cast duplex Stainless Steels (CSS) made in France for domestic nuclear power plants were thermally aged at 400 °C for 100 h, 300 h, 1000 h, 3000 h and 10000 h. The tensile properties and the impact properties at different thermal aging duration were measured and the effects of the thermal aging time on the microscopic structures and substructures of Z3CN20-09M were respectively investigated by optical microscopy and transmission electron microscopy. The results showed that the tensile strengths of Z3CN20-09M CSS increased gradually with the increment of the thermal ageing time, whereas the impact properties decreased with the prolonging of the thermal ageing time. After long thermal ageing time the dislocation configurations were greatly changed in austenite, and there were precipitates along the austenite-ferrite interface. Moreover, the iron-rich α phase and the chromium-rich α phase precipitated in ferrite aged for 10000h by nucleation and growth rather than the spinodal decomposition. All of above revealed that Z3CN20-09M CSS became brittle during thermal ageing.

Effect of stress relief at 350 °C and 550 °C on the impact properties of duplex stainless steels

2007 ◽  
Vol 58 (10) ◽  
pp. 909-916 ◽  
Author(s):  
Marcelo Martins ◽  
Sergio Mazzer Rossitti ◽  
Marcio Ritoni ◽  
Luiz Carlos Casteletti
Keyword(s):  
Stainless Steels ◽  
Stress Relief ◽  
Duplex Stainless Steels ◽  
Impact Properties ◽  
The Impact

Effect of Thermal Aging on Microstructural Evolution in Ferrite of Duplex Stainless Steel in Nuclear Power Plant Applications

2017 ◽  
Vol 898 ◽  
pp. 818-825
Author(s):  
B. Zhang ◽  
F. Xue ◽  
S.L. Li ◽  
Xi Tao Wang ◽  
N.N. Liang ◽  
...  
Keyword(s):  
Aging Time ◽  
Thermal Aging ◽  
Atom Probe Tomography ◽  
Nuclear Power ◽  
Impact Test ◽  
Atom Probe ◽  
Nanoindentation Test ◽  
Α Phase ◽  
The Impact ◽  
Duplex Steels

Z3CN20-09M duplex steels are thermally aged at 400oC for up to 20,000 h. The mechanical properties have been characterized by Charpy V-notch impact test and nanoindentation test. It is found that the nanohardness in ferrite increases and the impact toughness decreases with aging time. Moreover, the distribution of alloying elements has been carefully characterized using atom probe tomography (APT). The results indicate that the ferrite decomposes into Cr-rich α' and Cr-lean α phase during the thermal aging and Ni-rich G-phase forms in ferrite. The effect of aging time on solute nanostructure has been investigated systematically.

Prediction Method of Tensile Properties and Fracture Toughness of Thermal Aged Cast Duplex Stainless Steel Piping

2005 ◽  
Author(s):  
Seiichi Kawaguchi ◽  
Takeharu Nagasaki ◽  
Koji Koyama
Keyword(s):  
Mechanical Properties ◽  
Fracture Toughness ◽  
Stainless Steels ◽  
Thermal Aging ◽  
Power Plants ◽  
True Strain ◽  
Prediction Method ◽  
Strain Curve ◽  
Duplex Stainless Steels

Cast duplex stainless steels of CF8M and CF8 are used in major components because of their superior characteristics, such as corrosion resistance, weldability, and so on. However, these stainless steels are known to have tendency of thermal aging embrittlement after long term service. Therefore, the mechanical properties have been investigated using tensile test specimens and fracture toughness specimens aged at 300 to 450°C for up to 40,000 hours. From the results, the effects of thermal aging on the mechanical properties of these cast duplex stainless steels were identified. The true stress-true strain curve prediction method (TSS model) and fracture toughness prediction method (H3T model) after long term service were established. These prediction methods are used for the evaluation on the plant life management of nuclear power plants in Japan.

Investigation of Thermal Ageing in Long Term Operated RPV Materials

2015 ◽  
Author(s):  
Charles C. Eiselt ◽  
Günter König ◽  
Hieronymus Hein ◽  
Maxim Selektor ◽  
Martin Widera
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Elevated Temperatures ◽  
Energy Levels ◽  
Thermal Ageing ◽  
Experimental Investigations ◽  
Low Alloy Steels ◽  
Term Operation ◽  
The Impact

The phenomenon of thermal ageing of low alloy steels comes more into focus in terms of long term operation of nuclear power plants (NPP). Safety-relevant components such as the RPV or the pressurizer have to bear the respective loads at elevated temperatures for longer times. However the mechanical properties of the applied materials might experience certain degradations such as a decrease of the impact energy levels and a shift in the ductile to brittle transition temperature (e.g. T41) leading to higher ductile-brittle reference temperatures and a reduction of material toughness. In terms of a safe long term operation it is important to understand in how far thermal ageing alone, meaning for the RPV without the cumulative damaging effects through neutron irradiation, has detrimental influences on the respective materials of interest. First of all an overview is provided of the current state of the art with respect to thermal ageing by describing influencing mechanisms, its implementation into different nuclear codes, standards and selected experimental investigations in this field. Following this, the test results of the thermal surveillance sets from three German PWRs are presented and discussed. The tested Charpy-V specimens, taken from representative RPV base and weld metals (22NiMoCr3-7 / NiCrMo1UP) as well as their heat affected zones, were exposed to ∼290°C for ∼30 years on the cold leg of the according plants’ main coolant loops. The obtained results are compared with the existing thermal aging data base (baseline and ∼7 years data) of the materials concerned. Finally, the role of thermal ageing particularly with respect to RPV irradiation surveillance will be assessed.

Compatibility Evaluation for Application of Lean Duplex Stainless Steels to Seawater Systems in Nuclear Power Plants

2010 ◽  
Vol 654-656 ◽  
pp. 382-385
Author(s):  
Hyun Young Chang ◽  
Heung Bae Park ◽  
Young Sik Kim ◽  
Sang Kon Ahn ◽  
Kwang Tae Kim ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Corrosion Resistance ◽  
Stainless Steels ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Strength ◽  
Duplex Stainless Steels ◽  
Stainless Steel 316L ◽  
Pitting Potential

Lean duplex stainless steels have been developed in Korea for the purpose of being used in the seawater systems of industries. The flow velocity of some part of seawater systems in nuclear power plants is high and damages of components from corrosion are severe. Therefore, this environment requires using high strength and high corrosion resistant steels. The newly developed lean duplex stainless steels STS329LD(20.3Cr-2.2Ni-1.4Mo) and STS329J3L(22.4Cr-5.7Ni-3.6Mo) are evaluated for the compatibility in seawater systems of nuclear power plants. In this study, the physical & mechanical properties and corrosion resistance of two alloys were quantitatively evaluated in comparison with commercial stainless steel 316L. Microstructures and mechanical properties of them were analyzed and the electrochemical properties related to corrosion resistance were measured such as pitting potential, passive current density, and corrosion rates from Tafel analysis. Critical pitting temperatures were measured in accordance with ASTM G48E method. The pitting initiation time and lifetime for replacement were predicted from the PRE values of test alloys and empirical equations that have been formulated from the condenser tubes of a nuclear power plant.

Thermal aging behaviors of duplex stainless steels used in nuclear power plant: A review

2021 ◽  
Vol 544 ◽  
pp. 152693
Author(s):  
Y. Fan ◽  
T.G. Liu ◽  
L. Xin ◽  
Y.M. Han ◽  
Y.H. Lu ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Stainless Steels ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Duplex Stainless Steels

Effect of Thermal Aging on the Deformation and Failure Behaviors of Cast Austenitic Stainless Steels Under Excessive Cyclic Loads

2019 ◽  
Author(s):  
Jin Weon Kim ◽  
Sang Eon Kim ◽  
Yun Jae Kim
Keyword(s):  
Stainless Steels ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Cyclic Load ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
System Structure ◽  
Cyclic Loads ◽  
Deformation And Failure ◽  
Negative Effect

Abstract This study conducts failure tests using a simulated specimen to investigate the effect of thermal aging on the deformation and failure behaviors of system, structure, and components (SSCs) of nuclear power plants (NPPs) made of cast austenitic stainless steels (CASSs) under excessive seismic loads. Both unaged and thermally aged CF8A CASSs were used for the experiment, and the large cyclic loads in the form of displacement-control and load-control were applied at a quasi-static displacement rate. Displacement-controlled tests were performed at room temperature (RT) and 316°C and load-controlled tests were performed at RT. The results show that the deformation behaviors of aged CF8A CASS under both types of cyclic load are almost the same as those of unaged CF8A CASS. The thermal aging slightly promotes the failure of CF8A CASS under displacement-controlled cyclic loads, but the failure of specimen still occurs under the cyclic load levels several times higher than the load of the design basis earthquake. Under load-controlled cyclic loads, thermal aging retards the failure of CF8A CASS. Consequently, the thermal aging has no apparent negative effect on the deformation and failure behaviors of CASSs under large cyclic loads, even if it considerably changes the strength, ductility, and fracture toughness of CASSs.

Effect of PWHT Cycles on Impact Toughness of Austenitic Stainless Steel Weldments

2009 ◽  
Author(s):  
M. M. Ibrahim ◽  
H. G. Mohamed ◽  
Y. E. Tawfik
Keyword(s):  
Impact Toughness ◽  
Impact Energy ◽  
Stainless Steels ◽  
Nuclear Power ◽  
Power Plants ◽  
Plate Thickness ◽  
Austenitic Stainless Steels ◽  
Aisi 304L ◽  
Stress Relieving ◽  
The Impact

Austenitic stainless steels are widely used welding materials in nuclear reactors and power plants because of their high strength, good ductility, excellent corrosion resistance and a reasonable weldability. These properties make austenitic stainless steels attractive candidate materials for use in the fabrication of piping systems, automotive exhaust gas systems and in a variety of equipment associated with the chemical and nuclear power industries. PWHT is a stress relieving process whereby residual stresses are reduced by typically heating to 550–650 °C for a set time depending upon plate thickness. It concerns have emerged about possible effects on the mechanical properties of the base (parent) and weld plates (PM and WM). The 6 mm AISI 304L, 316L, and 347 austenitic stainless steels were used for this work. These welds were produced by SMAW and GTAW techniques using a single vee preparation and multiple weld beads, and welded by various types of consumables. The fracture surfaces of the Charpy V-notch PM and WM (before and after PWHT) samples were examined by SEM. Scanning electron fractographs was critical in this study, in that valuable information regarding the mechanism and nature of failure could be determined. This paper reports work on the impact toughness of the three types of austenitic stainless steels. The parent and weld regions were examined for all types of steels used, and then exposed to temperature in the PWHT range. The effect of exposure to multiple PWHT cycles on these properties is discussed. A decrease in impact energy and fracture toughness with an increase in the number of heat treatments was evident in the parent metal. Similary, the weld metal showed a decrease in impact energy after two PWHT cycles.

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