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

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 Precipitate on Thermal Aging Effect of 17-4PH Martensitic Stainless Steel Used as Valve Stem in Nuclear Power Plant

2019 ◽  
Vol 944 ◽  
pp. 466-472
Author(s):  
Bing Bai ◽  
Chang Yi Zhang ◽  
Pei Pei Zhang ◽  
Wen Yang
Keyword(s):  
Stainless Steel ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Power Plants ◽  
Martensitic Stainless Steel ◽  
Volume Fraction ◽  
Aging Effect ◽  
Valve Stem

The valve stem used in the main steam system of nuclear power plant is usually 17-4PH martensitic stainless steel. When it served in 300°C for a long time, the thermal aging embrittlement of valve stem will be significant, with the performance of the ductile brittle transition temperature (DBTT) and the hardness increased, the upper stage energy (USE) decreased. It will increase the risk of brittle fracture of the valve stem, and seriously affect the safety and economic operation of nuclear power plant (NPP). Similar cases have occurred in foreign nuclear power plants. Therefore, it is important to study the thermal aging effect of the 17-4PH steel used as valves in nuclear power plant. In this work, the 17-4PH martensitic stainless steel samples served in nuclear power plant for many years were studied, and they exhibit obvious thermal aging embrittlement. By use of small angle neutron scattering (SANS) and three-dimensional atomic probe (3DAP), the nanosize precipitate in stainless steel is studied. The results show that the size of the larger cluster (~7nm) in stainless steel increases and the volume fraction of the cluster with size of ~1nm increases obviously after thermal aging. The larger nanosize precipitate was growing up during long service at high temperature, and precipitation of the smaller ones continuously occurred. Combing with the results of 3DAP, the nanosize clusters were formed by segregation of Ni, Mn and other elements with Cu-rich cluster, which are mainly in the form of Cu core and Ni-Mn shell.

Effect of Element Segregation on Thermal Aging Behavior of 17-4PH Martensitic Stainless Steel for Nuclear Power Plant

2017 ◽  
Author(s):  
Bing Bai ◽  
Hanxiao Wang ◽  
Changyi Zhang ◽  
Zhenfeng Tong ◽  
Wen Yang
Keyword(s):  
Stainless Steel ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Martensitic Stainless Steel ◽  
Lath Martensite ◽  
Optical Microscope ◽  
Valve Stem ◽  
Long Time

The valve stem used in the main steam system of nuclear power plant is usually 17-4PH martensitic stainless steel. When it served in 300 C° for a long time, the thermal aging embrittlement of valve stem will be significant, with the performance of the ductile brittle transition temperature (DBTT) and the hardness increased, the upper stage energy (USE) decreased. It will seriously affect the safety and economic operation of nuclear power plant (NPP). It is important to study the thermal aging effect of the 17-4PH steel for safe operation of nuclear power plant. In this work, Three-Dimensional Atom Probe (3DAP), Energy Dispersive X-Ray Spectroscopy (EDX), Scanning Electron Microscope (SEM) and Optical Microscope (OM) are used to analyze the element distribution in 17-4PH steel. The results show that lath martensite will grow significantly under high temperature for a long time. More δ-ferrite will be found between lath martensite, and some carbide aggregates at its interface. In addition, the number density of Cu clusters in the17-4PH steel is increased. It is found that Ni and Mn have obvious segregation with the Cu cluster.

Thermal Aging of Primary Circuit Piping Materials in PWR Nuclear Power Plant

2009 ◽  
Vol 1215 ◽  
Author(s):  
Xitao Wang ◽  
Shilei Li ◽  
Shuxiao Li ◽  
Fei Xue ◽  
Guogang Shu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Early Stage ◽  
Accelerated Aging ◽  
Austenitic Stainless Steels ◽  
Fracture Morphology ◽  
Primary Circuit ◽  
The Impact

AbstractThe reserved cast austenitic stainless steels (CASS) for primary circuit piping in Daya Bay Nuclear Power Plant were studied. The changes of microstructure, mechanical properties and fracture behavior were investigated using SEM, EPMA, TEM and nanoindentation after accelerated aging at 400°C for up to 10000 h. Microhardness of ferrite increased rapidly in the early stage and then increased slowly later. The impact energy of materials declined with the aging time and reduced to a very low level after aging for 10000 hours. Fracture morphology displayed a mixture of cleavage in ferrite along with dimple and tearing in austenite. Two kinds of precipitations were observed in ferrite by TEM after long periods of aging. The fine Cr-enriched α′ phases precipitated homogeneously in ferrite, and a few larger G phases were observed as well. The precipitation of α′ phases was considered to be the primary mechanism of thermal aging embrittlement in CASS.

Sign in / Sign up

Export Citation Format

Share Document