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.

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.

Thermal Aging Effect of 17-4PH Martensitic Stainless Steel Valves for Nuclear Power Plant

2016 ◽  
Vol 850 ◽  
pp. 96-100 ◽  
Author(s):  
Bing Bai ◽  
Chang Yi Zhang ◽  
Jia Sheng Wang ◽  
Zhen Feng Tong ◽  
Qun Xian Lv ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Martensitic Stainless Steel ◽  
Aging Effect ◽  
Fracture Morphology ◽  
Valve Stem ◽  
Long Time ◽  
Metallurgical Structure ◽  
Hardness Increase

In this work, impact toughness and tensile properties of valve stem used in NPP are obtained. Combining with microstructure analysis of fracture morphology and metallurgical structure, the thermal aging behavior of the martensitic stainless steel is studied. The results show that the thermal aging embrittlement is significant when the valve stem serves in high temperature for a long time. The ductile to brittle transition temperature (DBTT) and the hardness increase, and the upper platform energy decreases.

Lessons From an Operational Small Leakage at Tsuruga Nuclear Power Station

2015 ◽  
Author(s):  
Hiroyuki Kobayashi ◽  
Osamu Urabe ◽  
Takushi Fujino
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
The Other ◽  
Leak Test ◽  
Power Station ◽  
Long Time ◽  
Short Time ◽  
Repair Techniques

Operational small leakage is occasionally observed in a nuclear power plant, and the leak forces an operator to decide whether to shut down the plant or not. Even if the leakage is just a little, it might draw the considerable attention in the society, so that the operator sometimes gets into the situation to judge more severely than technical judgment. Furthermore, at the time of plant restart and the system leak test just after maintenance, even the operator doesn’t accept any leakage considering the long management for the leakage up to the next outage. On the other hand, once the operator shut down the plant, it sometimes takes long time to restart again because of the difficulty to obtain new pipes and valves in short time. The temporary repair techniques referred to the JSME code might be able to be applied to maintain the plant operation, however some difficulties exist in a practical process. One of the authors has faced with many cases in which the operational small leakage had to be dealt at Tsuruga nuclear power station. This paper shows some cases of them and discusses lessons which are related to the codes and standards.

Risk Based Inspection Pilot Study of Ignalina Nuclear Power Plant, Unit 2

2002 ◽  
Author(s):  
Bjorn Brickstad ◽  
Adam Letzter ◽  
Arturas Klimasauskas ◽  
Robertas Alzbutas ◽  
Linas Nedzinskas ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Pilot Study ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Damage Mechanism ◽  
Core Damage ◽  
Steel Welds ◽  
Nuclear Power Plant Unit

A project with the acronym IRBIS (Ignalina Risk Based Inspection pilot Study) has been performed with the objective to perform a quantitative risk analysis of a total of 1240 stainless steel welds in Ignalina Nuclear Power Plant, unit 2 (INPP-2). The damage mechanism is IGSCC and the failure probabilities are quantified by using probabilistic fracture mechanics. The conditional core damage probabilities are taken from the plant PSA.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

Water chemistry effect on the wear of stainless steel in nuclear power plant

2006 ◽  
Vol 39 (12) ◽  
pp. 1503-1508 ◽  
Author(s):  
D. Kaczorowski ◽  
P. Combrade ◽  
J.Ph. Vernot ◽  
A. Beaudouin ◽  
C. Crenn
Keyword(s):  
Stainless Steel ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Chemistry ◽  
Nuclear Power
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