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Chincheng Huang - One of the best experts on this subject based on the ideXlab platform.
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Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature OverPressure Event
International Journal of Nuclear Energy, 2015Co-Authors: Hsoungwei Chou, Chincheng HuangAbstract:The probabilistic structural integrity of a Taiwan domestic boiling water Reactor Pressure Vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the Reactor Pressure Vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the Vessel wall were employed here. The low temperature overPressure transient which has been concluded to be the severest accident for a boiling water Reactor Pressure Vessel was considered as the loading condition. It is indicated that the fracture mostly happens near the fusion-line area of axial welds but with negligible failure risk. The calculated results indicate that the domestic Reactor Pressure Vessel has sufficient structural integrity until doubling of the present end-of-license operation.
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effects of fracture toughness curves of asme section xi appendix g on a Reactor Pressure Vessel under Pressure temperature limit operation
Nuclear Engineering and Design, 2014Co-Authors: Hsoungwei Chou, Chincheng HuangAbstract:According to the Code Case N-640 issued in 1999, the fracture toughness requirement of Reactor Pressure Vessel materials in ASME Section XI–Appendix G was amended to the KIC curve. In Taiwan, the present Pressure–temperature limit operation curves of normal Reactor startup (heat-up) and shut-down (cool-down) for the Reactor Pressure Vessel is still calculated per the KIa curve in 1998 or earlier editions. In the paper, the failure risks of a Taiwan domestic Reactor Pressure Vessel under various Pressure–temperature limit operations were analyzed. First, the Pressure–temperature limit curves of the Reactor Pressure Vessel based on KIa and KIC curves, and various levels of radiation embrittlement, were established. Then, the ORNL's probabilistic fracture mechanics code, FAVOR, and the PNNL's flaw model were employed to assess the failure probabilities of the Reactor Pressure Vessel under such Pressure–temperature limit transients. Further, the deterministic analyses of FAVOR code were also conducted. It is found that under the Pressure–temperature limit transients based on KIC curves, the Reactor Pressure Vessel presents higher failure probabilities, but are all below the allowable risk. The present results indicate that using the KIC curve the Pressure–temperature limits can increase the operational margin as well as maintaining the sufficient stability of the analyzed Reactor Pressure Vessel.
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Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events
Volume 7: Operations Applications and Components, 2014Co-Authors: Hsoungwei Chou, Chincheng HuangAbstract: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.Copyright © 2014 by ASME
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Effects of fracture toughness curves of ASME Section XI–Appendix G on a Reactor Pressure Vessel under Pressure–temperature limit operation
Nuclear Engineering and Design, 2014Co-Authors: Hsoungwei Chou, Chincheng HuangAbstract:According to the Code Case N-640 issued in 1999, the fracture toughness requirement of Reactor Pressure Vessel materials in ASME Section XI–Appendix G was amended to the KIC curve. In Taiwan, the present Pressure–temperature limit operation curves of normal Reactor startup (heat-up) and shut-down (cool-down) for the Reactor Pressure Vessel is still calculated per the KIa curve in 1998 or earlier editions. In the paper, the failure risks of a Taiwan domestic Reactor Pressure Vessel under various Pressure–temperature limit operations were analyzed. First, the Pressure–temperature limit curves of the Reactor Pressure Vessel based on KIa and KIC curves, and various levels of radiation embrittlement, were established. Then, the ORNL's probabilistic fracture mechanics code, FAVOR, and the PNNL's flaw model were employed to assess the failure probabilities of the Reactor Pressure Vessel under such Pressure–temperature limit transients. Further, the deterministic analyses of FAVOR code were also conducted. It is found that under the Pressure–temperature limit transients based on KIC curves, the Reactor Pressure Vessel presents higher failure probabilities, but are all below the allowable risk. The present results indicate that using the KIC curve the Pressure–temperature limits can increase the operational margin as well as maintaining the sufficient stability of the analyzed Reactor Pressure Vessel.
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Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event
Volume 7: Operations Applications and Components, 2012Co-Authors: Hsoungwei Chou, Chincheng Huang, Boyi ChenAbstract:The fracture probability of a boiling water Reactor Pressure Vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the Vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the Reactor Pressure Vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer Vessel walls. When conducting the fracture probability analyses, a transient low temperature over-Pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water Reactor Pressure Vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed Reactor Pressure Vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.Copyright © 2012 by ASME
Hsoungwei Chou - One of the best experts on this subject based on the ideXlab platform.
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Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature OverPressure Event
International Journal of Nuclear Energy, 2015Co-Authors: Hsoungwei Chou, Chincheng HuangAbstract:The probabilistic structural integrity of a Taiwan domestic boiling water Reactor Pressure Vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the Reactor Pressure Vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the Vessel wall were employed here. The low temperature overPressure transient which has been concluded to be the severest accident for a boiling water Reactor Pressure Vessel was considered as the loading condition. It is indicated that the fracture mostly happens near the fusion-line area of axial welds but with negligible failure risk. The calculated results indicate that the domestic Reactor Pressure Vessel has sufficient structural integrity until doubling of the present end-of-license operation.
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effects of fracture toughness curves of asme section xi appendix g on a Reactor Pressure Vessel under Pressure temperature limit operation
Nuclear Engineering and Design, 2014Co-Authors: Hsoungwei Chou, Chincheng HuangAbstract:According to the Code Case N-640 issued in 1999, the fracture toughness requirement of Reactor Pressure Vessel materials in ASME Section XI–Appendix G was amended to the KIC curve. In Taiwan, the present Pressure–temperature limit operation curves of normal Reactor startup (heat-up) and shut-down (cool-down) for the Reactor Pressure Vessel is still calculated per the KIa curve in 1998 or earlier editions. In the paper, the failure risks of a Taiwan domestic Reactor Pressure Vessel under various Pressure–temperature limit operations were analyzed. First, the Pressure–temperature limit curves of the Reactor Pressure Vessel based on KIa and KIC curves, and various levels of radiation embrittlement, were established. Then, the ORNL's probabilistic fracture mechanics code, FAVOR, and the PNNL's flaw model were employed to assess the failure probabilities of the Reactor Pressure Vessel under such Pressure–temperature limit transients. Further, the deterministic analyses of FAVOR code were also conducted. It is found that under the Pressure–temperature limit transients based on KIC curves, the Reactor Pressure Vessel presents higher failure probabilities, but are all below the allowable risk. The present results indicate that using the KIC curve the Pressure–temperature limits can increase the operational margin as well as maintaining the sufficient stability of the analyzed Reactor Pressure Vessel.
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Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events
Volume 7: Operations Applications and Components, 2014Co-Authors: Hsoungwei Chou, Chincheng HuangAbstract: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.Copyright © 2014 by ASME
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Effects of fracture toughness curves of ASME Section XI–Appendix G on a Reactor Pressure Vessel under Pressure–temperature limit operation
Nuclear Engineering and Design, 2014Co-Authors: Hsoungwei Chou, Chincheng HuangAbstract:According to the Code Case N-640 issued in 1999, the fracture toughness requirement of Reactor Pressure Vessel materials in ASME Section XI–Appendix G was amended to the KIC curve. In Taiwan, the present Pressure–temperature limit operation curves of normal Reactor startup (heat-up) and shut-down (cool-down) for the Reactor Pressure Vessel is still calculated per the KIa curve in 1998 or earlier editions. In the paper, the failure risks of a Taiwan domestic Reactor Pressure Vessel under various Pressure–temperature limit operations were analyzed. First, the Pressure–temperature limit curves of the Reactor Pressure Vessel based on KIa and KIC curves, and various levels of radiation embrittlement, were established. Then, the ORNL's probabilistic fracture mechanics code, FAVOR, and the PNNL's flaw model were employed to assess the failure probabilities of the Reactor Pressure Vessel under such Pressure–temperature limit transients. Further, the deterministic analyses of FAVOR code were also conducted. It is found that under the Pressure–temperature limit transients based on KIC curves, the Reactor Pressure Vessel presents higher failure probabilities, but are all below the allowable risk. The present results indicate that using the KIC curve the Pressure–temperature limits can increase the operational margin as well as maintaining the sufficient stability of the analyzed Reactor Pressure Vessel.
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Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event
Volume 7: Operations Applications and Components, 2012Co-Authors: Hsoungwei Chou, Chincheng Huang, Boyi ChenAbstract:The fracture probability of a boiling water Reactor Pressure Vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the Vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the Reactor Pressure Vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer Vessel walls. When conducting the fracture probability analyses, a transient low temperature over-Pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water Reactor Pressure Vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed Reactor Pressure Vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.Copyright © 2012 by ASME
M. Brumovsky - One of the best experts on this subject based on the ideXlab platform.
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WWER-type Reactor Pressure Vessel (RPV) materials and fabrication
Irradiation Embrittlement of Reactor Pressure Vessels (RPVs) in Nuclear Power Plants, 2015Co-Authors: M. BrumovskyAbstract:Abstract: This chapter describes requirements for speciality WWER Reactor Pressure Vessel materials in terms of their chemical composition and mechanical properties. The main principles of manufacturing technology for WWER Pressure Vessel fabrication are also discussed, including welding and cladding.
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Stresses in Reactor Pressure Vessel nozzles -- Calculations and experiments
1995Co-Authors: M. Brumovsky, H. PolachovaAbstract:Reactor Pressure Vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a Reactor Pressure Vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber`s, Hardrath-Ohman`s as well as equivalent energy ones, used in different Reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared.
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Assessment by comparison and experimental verification of VVER Reactor Pressure Vessel integrity
Theoretical and Applied Fracture Mechanics, 1995Co-Authors: M. BrumovskyAbstract:This work is concerned with assessing the lifetime integrity of VVER (water-water power Reactor) Reactor Pressure Vessel. Results based on the Soviet Code are compared with those of the ASME Code, Section III and XI. Similarities and discrepancies are illustrated and discussed in connection with numerical results of a typical RPV (Reactor Pressure Vessel) under operational conditions. Involved are data for fracture toughness, crack initiation and crack arrest conditions. Considerations are also given for possible adoption of the ASME Code for VVER Reactors.
J. Wintle - One of the best experts on this subject based on the ideXlab platform.
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Use of master curve technology for assessing shallow flaws in a Reactor Pressure Vessel material
Journal of Pressure Vessel Technology-transactions of The Asme, 2008Co-Authors: N. Taylor, P. Minnebo, B.r. Bass, Milos Kytka, Kim Wallin, Dieter Siegele, J. WintleAbstract:In the NESC-IV project, an experimental/analytical program was performed to develop validated analysis methods for transferring fracture toughness data to shallow flaws in Reactor Pressure Vessels subject to biaxial loading in the lower-transition temperature region. Within this scope, an extensive range of fracture tests was performed on material removed from a production-quality Reactor Pressure Vessel. The master curve analysis of these data is reported and its application to the assessment of the project feature tests on large beam test pieces is discussed.
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Use of Master Curve Technology for Assessing Shallow Flaws in a Reactor Pressure Vessel Material
Volume 6: Materials and Fabrication, 2006Co-Authors: N. Taylor, P. Minnebo, Milos Kytka, J. Wintle, Richard Bass, Kim Wallin, Dieter SiegeleAbstract:In the NESC-IV project an experimental/analytical program was performed to develop validated analysis methods for transferring fracture toughness data to shallow flaws in Reactor Pressure Vessels subject to biaxial loading in the lower-transition temperature region. Within this scope an extensive range of fracture tests was performed on material removed from a production-quality Reactor Pressure Vessel. The Master Curve analysis of this data is reported and its application to the assessment of the project feature tests on large beam test pieces.
C.a. English - One of the best experts on this subject based on the ideXlab platform.
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Microstructural characterisation techniques for the study of Reactor Pressure Vessel (RPV) embrittlement
Irradiation Embrittlement of Reactor Pressure Vessels (RPVs) in Nuclear Power Plants, 2015Co-Authors: Jonathan M. Hyde, C.a. EnglishAbstract:Abstract: In this chapter, the development of an understanding of the mechanisms that control Reactor Pressure Vessel (RPV) embrittlement is explored. It is shown how these insights were strongly coupled with developments in microstructural techniques and how research in this field has underpinned the development of dose–damage correlations.
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study of luders phenomena in Reactor Pressure Vessel steels
Materials Science and Engineering A-structural Materials Properties Microstructure and Processing, 2013Co-Authors: D W Beardsmore, C.a. English, Quinta J Da Fonseca, J Romero, S R Ortner, John Sharples, A H Sherry, M A WilkesAbstract:Abstract We have undertaken a combined experimental and modelling study of plasticity development in samples containing notches and cracks in a Reactor Pressure Vessel steel that exhibits the Luders phenomenon in uniaxial testing. We have used digital image correlation to study the development of plasticity in plane sided, waisted and notched samples of a mild steel that exhibits Luders behaviour. We have developed a constitutive model for use within elastic–plastic finite element analyses of structural geometries made from materials which exhibit Luders strain behaviour, and validated it against experimental data. The model is capable of describing plasticity development in smooth sided, waisted and notched specimens. With the assistance of the model we have determined that Luders behaviour affects the detailed development of plasticity not only in smooth sided and waisted specimens, which is readily observed experimentally, but also in sharp notched specimens. In sharp notched specimens the plastic zone is more constrained and stress intensification higher in the presence of Luders behaviour.
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microstructural evolution in Reactor Pressure Vessel steels
Journal of Nuclear Materials, 1993Co-Authors: W J Phythian, C.a. EnglishAbstract:Abstract The irradiation-induced changes occurring at the nanometer level in the microstructure of the Reactor Pressure Vessel have been shown to be responsible for a degradation in mechanical properties that occur in large structures with dimensions in the tens of metres. These changes can place severe restrictions on the Reactor both at the startup and under continual operation; and in the long term could compromise the safe operation of the plant. The large financial and safety related matters have given the necessary driving force to study the factors controlling the microstructural evolution, resulting in a multi disciplinary approach to the problem. This paper aims to review the current understanding of the subject, giving were possible examples of the approach and techniques used to obtain this. We also highlight areas of current research activity and indicate the type of work still required to provide information on aspects that currently lack a full understanding.
