Author

Hsoung-Wei Chou

Observed names: H. W. Chou, Hsoung‐Wei Chou

Institution: National Atomic Research Institute

Observed institutions: Institute of Nuclear Energy Research, National Atomic Research Institute, National Cheng Kung University

ORCID: 0000-0003-3086-3467

Observed ORCID: 0000-0003-3086-3467

OpenAlex ID: A5006205162

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

2006 · Hsoung‐Wei Chou, Jong‐Shin Huang, et al. · Journal of Applied Polymer Science · 59

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

2007 · Hsoung‐Wei Chou, Jong‐Shin Huang · Journal of Applied Polymer Science · 23

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

2014 · Hsoung-Wei Chou, Chin-Cheng Huang · Nuclear Engineering and Design · 20

Probabilistic fracture analysis for boiling water reactor pressure vessels subjected to low temperature over-pressure event

2012 · Chin-Cheng Huang, Hsoung-Wei Chou, et al. · Annals of Nuclear Energy · 14

Stress intensity factor assessment for reactor pressure vessel nozzles containing postulated corner cracks

2022 · Chih-Hsuan Lee, Hsoung-Wei Chou · Engineering Fracture Mechanics · 13

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

2008 · Hsoung‐Wei Chou, Jong‐Shin Huang · Journal of Applied Polymer Science · 12

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

2016 · Hsoung-Wei Chou, Chin-Cheng Huang · Nuclear Engineering and Design · 11

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Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events

2014 · Hsoung-Wei Chou, Chin-Cheng Huang · ASME Pressure Vessels and Piping Conference (PVP) · 11

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

2019 · Pin-Chiun Huang, Hsoung-Wei Chou, et al. · International Journal of Pressure Vessels and Piping · 9

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

2019 · Hsoung-Wei Chou, Yu-Yu Shen, et al. · International Journal of Pressure Vessels and Piping · 9

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

2017 · Hsoung-Wei Chou, Chin-Cheng Huang · Annals of Nuclear Energy · 8

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Probabilistic ageing and risk analysis tools for nuclear piping

2016 · Guian Qian, Hsoung-Wei Chou, et al. · Nuclear Engineering and Design · 7

Fatigue life prediction for circular rubber bearings subjected to cyclic compression

2011 · Hsoung‐Wei Chou, Jong‐Shin Huang · Journal of Applied Polymer Science · 7

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

2022 · Sue-Ray Lin, Hsoung-Wei Chou, et al. · International Journal of Pressure Vessels and Piping · 6

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

2015 · Hsoung-Wei Chou, Chin-Cheng Huang · International Journal of Nuclear Energy · 6

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

2011 · Hsoung‐Wei Chou, Jong‐Shin Huang · Journal of Applied Polymer Science · 6

Reactor pressure vessel integrity assessment by probabilistic fracture mechanics – A plant specific analysis

2013 · Bo-Yi Chen, Chin-Cheng Huang, et al. · International Journal of Pressure Vessels and Piping · 5

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

2024 · Hsoung-Wei Chou, Jiun-Shya Yu, et al. · International Journal of Pressure Vessels and Piping · 4

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

2023 · Yu-Yu Shen, Kuei-Jen Cheng, et al. · Nuclear Engineering and Technology · 4

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

2015 · Pin-Chiun Huang, Hsoung-Wei Chou, et al. · Nuclear Engineering and Design · 3

Nonlinear structural analysis for structural integrity re-evaluation of existing spent fuel pool in nuclear power plant under beyond design basis earthquake

2025 · Ting-Yu Fan, Chin-Yu Lin, et al. · Nuclear Engineering and Technology · 2

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

2021 · Hsoung-Wei Chou, Ching-Hsin Chang · Annals of Nuclear Energy · 2

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

2020 · Hsoung-Wei Chou, Chin-Cheng Huang, et al. · Nuclear Technology · 2

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

2018 · Ching-Hsin Chang, Hsoung-Wei Chou, et al. · International Journal of Pressure Vessels and Piping · 2

Probabilistic Fracture Analysis for Boiling Water Reactor Pressure Vessels Subjected to Low Temperature Over-Pressure Event

2010 · Hsoung-Wei Chou, Chin-Cheng Huang, et al. · ASME Pressure Vessels and Piping Conference (PVP) · 2

Works count: 42

Citation count: 255

H-index: 9

I10-index: 8