scholarly journals Hydrogen embrittlement of the pressure vessel structural materials in a WWER-440 nuclear power plant

2017 ◽  
Vol 131 ◽  
pp. 379-385 ◽  
Author(s):  
J. Toribio ◽  
D. Vergara ◽  
M. Lorenzo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Hydrogen Embrittlement ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Materials

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.

The Use of the ASME PRA Standard in Pressure Vessel and Piping Work

2002 ◽  
Author(s):  
Richard A. Hill
Keyword(s):  
Risk Assessment ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Risk Information ◽  
Technical Requirements ◽  
Operation And Maintenance ◽  
Construction Operation ◽  
Standard Sets

After several years of intense labor by many industry people, ASME is about to issue its newly approved PRA standard. This standard is for probabilistic risk assessment (PRA) for nuclear power plant applications. It is not a standard on how to build a PRA model; although, that could be inferred from the standard’s technical requirements. This Standard sets forth requirements for PRAs used to support risk-informed decisions related to design, licensing, procurement, construction, operation, and maintenance. It also prescribes a method for applying these requirements depending the degree to which risk information is needed and credited.

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.

Comparison of Allowable Bending Moments for Circumferentially Flawed Pipes in ASME Boiler and Pressure Vessel Code Section XI Nonmandatory Appendices C and H

2019 ◽  
Author(s):  
Hune-Tae Kim ◽  
Ji-Su Kim ◽  
Jun-Min Seo ◽  
Yun-Jae Kim ◽  
Kuk-Hee Lee ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Bending Moments ◽  
Analytical Evaluation ◽  
Evaluation Procedures ◽  
Failure Assessment ◽  
Inner Surface ◽  
Plant Components

Abstract In this paper, allowable bending moments for circumferential inner surface cracked pipes are evaluated. ASME Boiler and Pressure Vessel code Section XI, “Rules for Inspection of Nuclear Power Plant Components” provides analytical evaluation procedures. Analytical evaluation methods based on the failure mechanism are provided in nonmandatory Appendix C and those based on failure assessment diagram are given in nonmandatory Appendix H. Allowable bending moments are evaluated using both appendices and compared with experiments. Conservativeness is compared quantitatively between both methods by normalizing allowable bending moments with experimental maximum moments.

Structural Integrity Analysis of Nuclear Power Plant Pressure Vessel Penetration Nozzle Repaired

2016 ◽  
Vol 853 ◽  
pp. 346-350
Author(s):  
Lin Wei Ma ◽  
Jia Sheng He ◽  
An Qing Shu ◽  
Xiao Tao Zheng ◽  
Yan Wang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Degradation Mechanism ◽  
Design Requirement ◽  
Original Design ◽  
Asme Code ◽  
Integrity Analysis

Primary water stress corrosion cracking (PWSCC) has been observed in CRDM nozzles, BMI nozzles and other penetration nozzles. The industry has used the repair method of replacement of nozzles fabricated of Alloy 690. After the replacement of the nozzle, the structural integrity analysis of new nozzle and welds should be performed to ensure the pressure boundary compliance with the original design requirement. In this paper, the pressurizer top head instrument nozzle of PWR nuclear power plant is evaluated as a typical pressure vessel penetration nozzle. The results showed that the repaired nozzle satisfies the ASME Code design requirement and the crack growth of the postulated flaw in 40 years of the nuclear plant life is acceptable.PWSCC degradation mechanism has been observed in CRDM nozzles, BMI nozzles and other penetration nozzles [1]. In some nuclear power plants built in China earlier, such as DAYABAY nuclear power plant and QINSHAN nuclear power plant, PWSCC degradation mechanism has been found in CRDM nozzle welds which manufactured of Alloy 600 and welded of Alloy 82/182[2]. The repair of the degraded nozzles is the popular choice for the nuclear power plant owners. After the replacement of the nozzle, the structural integrity analysis of new nozzle and welds should be performed to ensure the pressure boundary compliance with the original design requirement. In this paper, the pressurizer top head nozzle of PWR nuclear power plant is evaluated as a typical pressure vessel penetration nozzle. Stress intensities were conservatively determined for pressure and applicable thermal transients and compared to the allowable values of the ASME Code, Section III. Thermal stress of the transients was obtained from 3D finite element model (FEM). Residual stress of J-groove weld was obtained from 2D FEM analysis and used for fracture mechanics analysis. All of the analysis showed that the repaired nozzle satisfies the ASME Code design requirement and the crack growth of the postulated flaw in 40 years of the nuclear plant life is acceptable.

Effect of a Single Overload on the Fracture Behavior in Safe-End Dissimilar Metal Welded Joints in Nuclear Power Plant

2014 ◽  
Vol 1049-1050 ◽  
pp. 600-604
Author(s):  
He Xue ◽  
Yuan Kui Gui ◽  
Wei Bing Wang ◽  
Xiao Bo Li ◽  
Ying Ru Wang ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Tensile Stress ◽  
Pressure Vessel ◽  
Welded Joints ◽  
Nuclear Power ◽  
Fracture Behavior ◽  
Welded Joint ◽  
Dissimilar Metal ◽  
Single Overload

To understand the effect of a single overload on the fracture behavior in welded joints, the stress and strain field at the crack tip in a safe-end dissimilar metal welded joint in nuclear pressure vessel is simulated and analyzed by using the elastic-plastic finite element method in the paper, in which the mechanical heterogeneity in welded joint is emphatically considered. The investigating results indicate that the tensile plastic strain at crack tip increases, but the tensile stress decreases as a single overload increases, and the influence of a single overload on tensile strain is larger than one on tensile stress, which provide a theoretical basis for quantitatively estimating the crack growth rate of environmentally assisted cracking in the welded structural material of pressure vessel and piping in the nuclear power plant.

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