Haozheng J. Qu
Materials Science · Purdue University West Lafayette
Publications
38
Citations
259
Est. group size
~2
Recurring co-author estimate
Active years
9
Publishing since 2018
This researcher studies how metal alloys used in nuclear reactors degrade over time, focusing especially on FeCrAl alloys (iron-chromium-aluminum materials being developed as improved fuel cladding for nuclear reactors) and stainless steels used in spent fuel storage. Their work examines processes like oxidation, corrosion, and stress-related cracking under conditions that mimic reactor and storage environments, aiming to understand how alloy composition and manufacturing methods affect material durability and safety.
Publication output has grown substantially over the last decade, rising from occasional single papers in 2017-2021 to a sustained pace of 7-11 papers per year in 2022-2024, followed by a slight decline in 2025-2026.
Generated by claude-sonnet-5 from public bibliographic data · Jul 20, 2026
- Effect of Nickel on Long-term Hydrothermal Corrosion of FeCrAl Alloys in Boiling Water Reactor Environments
SSRN Electronic Journal · 2026
- Chloride-induced stress corrosion crack propagation mechanisms in austenitic stainless steel are mechanically driven
Scripta Materialia · 2025
- Synergistic roles of Cr and Mo in low temperature steam oxidation of FeCrAl alloys
Corrosion Science · 2025
- Sequential oxidation and hydrothermal corrosion of FeCrAl alloys at BWR top-of-core conditions
Corrosion Science · 2025
- Oxidation behaviour of PM-C26M FeCrAl alloy in low-temperature steam 400 – 900 °C
Nuclear Materials and Energy · 2025
- Chloride-induced stress corrosion cracking in Austenitic steels for SNF storage canisters - Recent understanding and advances in mitigation and repair
Journal of Nuclear Materials · 2024
- Effect of molybdenum on the oxidation resistance of FeCrAl alloy in lower temperature (400 °C) and higher temperature (1200 °C) steam environments
Corrosion Science · 2024
- Effect of Al content on steam oxidation behavior for ferritic Fe-21Cr-xAl alloys
Journal of Nuclear Materials · 2024
- Hydrothermal Corrosion of Latest Generation of FeCrAl Alloys for Nuclear Fuel Cladding
Materials · 2024
- Corrosion behavior of additively manufactured FeCrAl in out-of-pile light water reactor environments
npj Materials Degradation · 2024
- New insights on the effects of chemistry and temperature on α’ precipitation during aging of FeCrAl alloys
Journal of Nuclear Materials · 2024
- Effect of Al Content on Steam Oxidation Behavior for Ferritic Fe-21cr-Xal Alloys
SSRN Electronic Journal · 2024
- Corrosion behavior of additively manufactured FeCrAl in out-of-pile light water reactor environments
Research Square · 2024
- New Insights on the Effects of Chemistry and Temperature on Α’ Precipitation During Aging of Fecral Alloys
SSRN Electronic Journal · 2024
- Nanoindentation Orientation and Solute Influenced Deformation Mechanisms in Irradiated Bcc Fe Binary Alloys
SSRN Electronic Journal · 2024
- SSRN Electronic Journal×7
- Journal of Nuclear Materials×6
- Corrosion Science×4
- npj Materials Degradation×2
- Microscopy and Microanalysis×2
- Maria A. Okuniewski
Materials Science · Purdue University West Lafayette
- Logan Joyce
Materials Science · Purdue University West Lafayette
- Jian Gan
Materials Science · Purdue University West Lafayette
- Brandon Miller
Materials Science · Purdue University West Lafayette
- Laura Hawkins
Materials Science · Purdue University West Lafayette
This profile was generated automatically from public scholarly data (OpenAlex). Group size and activity levels are estimates derived from co-authorship patterns.
Last updated Jul 20, 2026.
Claim or correct this profile