Guiqiu (Tony) Zheng

Post-Irradiation Examination (PIE) of Nuclear Materials Following High-Temperature, In-Core Corrosion Testing in Molten Flibe
Experimental Design for the Investigation of Tritium Behavior in Molten Flibe Under Neutron Irradiation
Tritium Retention in Graphite for the Application of Molten Fluoride Salt as Coolant in Both Fission and Fusion Nuclear Reactors
Light-Water Reactor Fuel and Cladding Development Including Modeling and Irradiation Studies
Materials Microstructural Characterization, Especially Radiation Damage and Corrosion
Molten Salts Electrochemical Analysis