Search Results - "Chou, Hsoung-Wei" :: K.UTB vyhledávací portál

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The structural and non–linear dynamic analysis for radioactive waste container

by Shen, Yu-Yu, Cheng, Kuei-Jen, Chou, Hsoung-Wei
Published in Nuclear engineering and technology (01.08.2023)

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Fatigue life prediction for circular rubber bearings subjected to cyclic compression

by Chou, Hsoung-Wei, Huang, Jong-Shin
Published in Journal of applied polymer science (15.02.2012)

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Effects of thermal aging on fatigue of carbon black-reinforced EPDM rubber

by Chou, Hsoung-Wei, Huang, Jong-Shin, Lin, Shih-Ting
Published in Journal of applied polymer science (15.01.2007)

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Crack initiation and propagation in circular rubber bearings subjected to cyclic compression

by Chou, Hsoung-Wei, Huang, Jong-Shin
Published in Journal of applied polymer science (05.08.2011)

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Effects of cyclic compression and thermal aging on dynamic properties of neoprene rubber bearings

by Chou, Hsoung-Wei, Huang, Jong-Shin
Published in Journal of applied polymer science (05.02.2008)

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Effects of ultraviolet irradiation on the static and dynamic properties of neoprene rubbers

by Chou, Hsoung-Wei, Huang, Jong-Shin
Published in Journal of applied polymer science (05.12.2008)

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Applicability investigation on failure assessment diagram of ASME code section XI - Appendix H for ferritic nuclear piping

by Chou, Hsoung-Wei, Yu, Jiun-Shya, Huang, Chin-Cheng
Published in The International journal of pressure vessels and piping (01.06.2024)

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International case survey and literature review of maintenance strategy for decommissioning nuclear power plant during transition period

by Lin, Sue-Ray, Chou, Hsoung-Wei, Chen, Jien-Jong
Published in The International journal of pressure vessels and piping (01.08.2022)

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$Probabilistic fracture analysis for boiling water reactor vessels considering seismic loads during decommissioning transition period$

Probabilistic fracture analysis for boiling water reactor vessels considering seismic loads during decommissioning transition period

by Chou, Hsoung-Wei, Chang, Ching-Hsin
Published in Annals of nuclear energy (01.03.2022)

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Development of pressure-temperature operation limits for a PWR vessel considering beltline shell and extended beltline nozzles

by Chou, Hsoung-Wei, Shen, Yu-Yu, Huang, Chin-Cheng
Published in The International journal of pressure vessels and piping (01.01.2020)

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$Effects of fracture toughness curves of ASME Section XI–Appendix G on a reactor pressure vessel under pressure–temperature limit operation$

Effects of fracture toughness curves of ASME Section XI–Appendix G on a reactor pressure vessel under pressure–temperature limit operation

by Chou, Hsoung-Wei, Huang, Chin-Cheng
Published in Nuclear engineering and design (01.12.2014)

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Application of Flaw Updating Process on Probabilistic Integrity Analysis for a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks

by Chou, Hsoung-Wei, Huang, Pin-Chiun, Ferng, Yuh-Ming
Published in Nuclear technology (04.05.2021)

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Large thermal gradients on structural integrity of a reactor pressure vessel subjected to pressurized thermal shocks

by Huang, Pin-Chiun, Chou, Hsoung-Wei, Ferng, Yuh-Ming, Kang, Chin-Hsiang
Published in The International journal of pressure vessels and piping (01.01.2020)

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Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

by Chou, Hsoung-Wei, Huang, Chin-Cheng
Published in Nuclear engineering and design (15.04.2016)

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$Comparison of ASME pressure–temperature limits on the fracture probability for a pressurized water reactor pressure vessel$

Comparison of ASME pressure–temperature limits on the fracture probability for a pressurized water reactor pressure vessel

by Chou, Hsoung-Wei, Huang, Chin-Cheng
Published in Annals of nuclear energy (01.10.2017)

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Demonstration of Structural Integrity of Boiling Water Reactor Pressure Vessels Under Ultimate Response Guideline Operation

by Chou, Hsoung-Wei, Huang, Chin-Cheng, Huang, Pin-Chiun, Ferng, Yuh-Ming
Published in Nuclear technology (01.12.2020)

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Development and application of safety assessment system in PFM analysis for RPV to study the influence of the neutron fluence estimate

by Chang, Ching-Hsin, Chou, Hsoung-Wei, Huang, Chin-Cheng
Published in The International journal of pressure vessels and piping (01.11.2018)

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$The influence of chemistry concentration on the fracture risk of a reactor pressure vessel subjected to pressurized thermal shocks$

The influence of chemistry concentration on the fracture risk of a reactor pressure vessel subjected to pressurized thermal shocks

by Huang, Pin-Chiun, Chou, Hsoung-Wei, Ferng, Yuh-Ming
Published in Nuclear engineering and design (01.02.2016)

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Probabilistic ageing and risk analysis tools for nuclear piping

by Qian, Guian, Chou, Hsoung-Wei, Niffenegger, Markus, Huang, Chin-Cheng
Published in Nuclear engineering and design (15.04.2016)

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Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

by Chou, Hsoung-Wei, Huang, Chin-Cheng
Published in International Journal of Nuclear Energy (25.11.2015)

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