Elastic-Plastic Fracture Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shock

2017 ◽  
Author(s):  
Huajing Guo ◽  
Zhongxian Wang ◽  
Poh-Sang Lam
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Three Dimensional ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Dimensional Finite Element ◽  
Fracture Theory ◽  
Crack Tip Constraint ◽  
Three Dimensional Finite Element ◽  
Reactor Pressure

Three-dimensional finite element models are used to analyze a reactor pressure vessel with an axial semi-elliptical surface crack subjected to pressurized thermal shock. During the thermal shock event, the J-A2 two-parameter fracture theory is used to investigate the temperature-dependent constraint effect at the deepest point and the surface point of the crack. Using the R6 methodology, a series of constraint-based crack failure assessment curves during the thermal shock can be obtained. It was found that the crack tip constraint should be considered for developing a more realistic failure criterion.

Computational studies on pressurized thermal shock in reactor pressure vessel

2021 ◽  
Vol 152 ◽  
pp. 107987
Author(s):  
Rakesh Chouhan ◽  
Anuj Kumar Kansal ◽  
Naresh Kumar Maheshwari ◽  
Avaneesh Sharma
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Computational Studies ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events

2014 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

OS1410 Evaluation of Through-wall Cracking Frequency of Reactor Pressure Vessel under Pressurized Thermal Shock events

2013 ◽  
Vol 2013 (0) ◽  
pp. _OS1410-1_-_OS1410-3_
Author(s):  
Hiroyuki NISHIKAWA ◽  
Koichi MASAKI ◽  
Kazuya OSAKABE ◽  
Jinya KATSUYAMA ◽  
Kunio ONIZAWA
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure
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