Welding Residual Stress in a Large Diameter Nuclear Reactor Pressure Vessel Nozzle

2013 ◽  
Vol 135 (2) ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Large Diameter ◽  
Reactor Pressure Vessel ◽  
Welding Residual Stress ◽  
Stress Fields ◽  
Reactor Pressure

The Atucha II nuclear power plant is a pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant was designed by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. Welding residual stresses in nuclear power plant piping can lead to cracking concerns later in the life of the plant, especially for stress-corrosion cracking. Hence, understanding the residual stress distribution from welding is important to evaluate the reliability of pipe and nozzle joints with welds. In this paper, a large-diameter reactor pressure vessel (RPV) hot-leg nozzle was analyzed. This is a nozzle from Atucha II nuclear power plant in Argentina. The main piping material is 20MnMoNi55 with Tenacito 65R weld metal, and inner diameter (ID) welded cladding at the girth weld locations is made of 309L. The special materials and weld geometry will lead interesting welding residual stress fields. In addition, postweld heat treatment (PWHT) of the girth welds and its boundary conditions could also play an important role in determining welding residual stress fields at the plant’s normal operating conditions. Sensitivity analyses were conducted and the technical observations and comments are provided.

Welding Residual Stress in a Large Diameter Nuclear Reactor Pressure Vessel Nozzle

2012 ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Large Diameter ◽  
Reactor Pressure Vessel ◽  
Welding Residual Stress ◽  
Sensitivity Analyses ◽  
Reactor Pressure

The Atucha II nuclear power plant is a pressurized heavy water reactor being constructed in Argentina. The original plant was designed by KWU in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. Many times welding residual stresses in nuclear power plant can lead to cracking concerns. Hence, understanding the residual stress distribution is important to evaluate the reliability of pipe and nozzle joints with welds. In this paper, a large diameter Reactor Pressure Vessel (RPV) hot-leg nozzle was analyzed. This is a nozzle from Atucha II nuclear power plant in Argentina. The main weld material is 20MnMoNi55 and ID welded cladding is made of 309L. The special materials and weld geometry will lead to interesting welding residual stress field. In addition, post-weld heat treatment (PWHT) region and boundary conditions could also play an important role in determining welding residual stress fields. Sensitivity analyses were conducted and the technical observation and comments will conclude the paper.

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

2004 ◽  
Author(s):  
J. C. Kim ◽  
J. B. Choi ◽  
Y. H. Choi
Keyword(s):  
Virtual Reality ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Evaluation System ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Virtual Reality Environment ◽  
Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

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.

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