scholarly journals Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

2014 ◽  
Author(s):  
Edmund Sullivan ◽  
Michael Anderson
Keyword(s):  
Fracture Toughness ◽  
Density Distribution ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Ultrasonic Testing ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In-Service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution Per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

2011 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Michael T. Anderson ◽  
Wallace Norris
Keyword(s):  
Fracture Toughness ◽  
Density Distribution ◽  
Thermal Shock ◽  
Size Distribution ◽  
Pressure Vessel ◽  
Ultrasonic Testing ◽  
Reactor Vessel ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Reactor Pressure

The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking of a reactor pressure vessel (RPV) due to a pressurized thermal shock (PTS) event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.” The §50.61a rule, which is optional, requires licensees to analyze the results from periodic volumetric examinations required by the American Society of Mechanical Engineers (ASME) Code. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory has been working on a program to assess the ability of current inservice inspection ultrasonic testing (UT) techniques, as qualified through the ASME Code to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, “Spirit of Appendix VIII reactor vessel examination,” a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

Evaluation on Constraint Effect of Reactor Pressure Vessel Under Pressurized Thermal Shock

2013 ◽  
Author(s):  
Naoki Ogawa ◽  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Shohei Sakaguchi ◽  
Toru Oumaya
Keyword(s):  
Fracture Toughness ◽  
Thermal Shock ◽  
Surface Crack ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Constraint Effect ◽  
Elastic Plastic ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In recent years, the integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) accident has become controversial issue since the larger shift of RTNDT in some higher fluence surveillance data raised a concern on RPV integrity. Under PTS condition, the combination of thermal stress due to a temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the wall of RPV. Currently, RPV integrity is assessed by comparing stress intensity factor on a crack tip under PTS condition and a reference toughness curve based on the fracture toughness data of irradiated compact specimens. Since PTS loading is large enough to cause plastic deformation, a crack tip behavior on the inner surface of RPV can be explained by elastic-plastic fracture mechanics using the J-integral. In this study, 3D elastic plastic finite element analyses were performed to assess the crack tip behavior on surface of a RPV under Loss of coolant Accident, which causes one of the most severe PTS condition. In order to quantify the constraint effect on a surface crack, J-Q approach was applied. The constraint effect of a surface crack was compared with a compact specimen and its influence on the fracture toughness was assessed. As a result, the difference of constraint effect was clearly obtained. And it is recommended to consider constraint effects in the evaluation of structural integrity of RPV under PTS.

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.

Influence Factors of Fracture Mechanics Analysis of Reactor Pressure Vessel under Pressurized Thermal Shock

2019 ◽  
Vol 795 ◽  
pp. 333-339
Author(s):  
Juan Luo ◽  
Jia Cheng Luo
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Influence Factors ◽  
Pressurized Thermal Shock ◽  
Inner Surface ◽  
Mechanics Analysis ◽  
Reactor Pressure

When the reactor pressure vessel (RPV) is subjected to pressurized thermal shock (PTS), the cooling water injected by the emergency core cooling system (ECCS) will generate a large temperature difference in the wall thickness of the pressure vessel. On the other hand, the fracture toughness of the RPV material decreases a lot under long-term neutron irradiation. Under this condition, the PTS transient may cause a rapid growth of defects in the inner surface of the vessel, resulting in failure of the pressure vessel. In this paper, the fracture mechanics analysis method of RPV under pressurized thermal shock is studied. The thermal analysis and structural analysis of the pressure vessel are performed by finite element method. The stress intensity factor and fracture toughness are obtained through calculation. At the same time, the influence factors of fracture mechanics analysis of RPV under PTS condition are analyzed. The effects of different crack size, crack type, load transient, and neutron irradiation flux on the PTS fracture mechanics analysis results are evaluated. Results show that the larger the ratio of length to depth for axial inner surface cracks, the easier RPV crack grows. Under small break condition, the circumferential cracks are safer than axial cracks. The longer the operating time, the more severe the embrittlement of RPV materials, which will lead to the failure of RPV more easily. For the two typical PTS transients studied in this paper, the re-pressurization condition is safer than the small break condition. The results can provide basis for structural integrity assessment of RPV under PTS condition.

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