Bio
Dr. Yong Yan is a distinguished R & D staff scientist of the Nuclear Fuel Development Section at Oak Ridge National Laboratory. He received his Ph.D. in Material Physics from the University of Could Bernard Lyon 1, France.
Dr. Yan’s research interests are mainly focused on nuclear fuel materials characterizations, with an emphasis on Loss of Coolant Accident (LOCA) tests with irradiated nuclear fueled rods and post quench ductility study of nuclear cladding material for light water reactors. This work provides technical assistance to the US NRC for the rulemaking of the Code of Federal Regulations. He is actively in a systematical study to assess the effect of hydrogen concentration & morphology and oxygen pickup under normal operation and accident conditions on fuel cladding embrittlement, by addressing (1) the alloy-related effects (including traditional Zirconium alloys Zicolay-2 and Zircolay-4; advanced alloys M5, ZIRLO, E110 and E110M; as well as accident tolerant fuel cladding materials FeCrAl and coated Zr tubing); (2) the burnup-related effects (impacts of corrosion layer, hydrogen pickup, fuel –cladding bond, fuel fragmentation). He is also active in development of testing techniques for zirconium hydriding, high temperature stem oxidation, ring compression testing of irradiated Zr cladding. Dr. Yan has authored or co-authored over eighty peer-reviewed publications in the areas of nuclear fuel and cladding materials, high superconductors, and high temperature materials.