Coating Development for Improved Performance Nuclear Reactor Fuel Cladding (26/07/22)
Speaker and Affliation:
Prof. Kumar Sridharan
Grainger Professor
Departments of Engineering Physics and Materials Science & Engineering
University of Wisconsin, Madison, U.S
Biography
Dr. Kumar Sridharan received his B. Tech in Metallurgy from BHU-IT, and his PhD in Materials Science & Engineering from the University of Wisconsin, Madison. His expertise spans a broad spectrum of areas in materials science, including, surface modification and coatings, cold spray materials deposition technology, corrosion, nuclear reactor materials, irradiation effects in materials, and intersections of materials science and manufacturing. He is author of over 350 publications in these areas including seven invited book chapters, journal articles, conference proceedings, and governmental technical reports. He serves on the editorial committee of the journal Advanced Materials and Processes, and previously served on the editorial committees of the journals, International Materials Reviews and Journal of Materials Engineering and Performance. Prof. Sridharan has served on expert panels of the U.S. Department of Energy, U.S. Nuclear Regulatory Commission, and the International Atomic Energy Agency. He has provided research mentorship to well over 100 undergraduate and graduate students, post-doctoral researchers, and staff scientists. Prof. Sridharan is an elected Fellow of American Society for Materials, Fellow of Institute of Materials, UK, and Fellow of American Nuclear Society, in recognition of his contributions to the areas of surface engineering, corrosion, nuclear reactor materials, and education.
When?
26th July, 2022 (Wednesday), 12:01 PM (India Standard Time)
Location
Department of Materials Engineering, KPA auditorium
Abstract
Zirconium-alloys have been used successfully as materials for light water reactor (LWR) fuel cladding (tubes that contain uranium-dioxide fuel pellets) for many decades because of their excellent neutron transparency, and good strength and hydrothermal corrosion resistance under normal operating conditions. However, in the rare event of loss of active cooling, decay heat can rapidly increase the temperature of the Zr-alloy cladding. This can result in severe oxidation of Zr-alloy in reaction with steam and loss of mechanical integrity of the cladding. This was unfortunately the case in the Fukushima-Daiichi accident in Japan in 2011. A near-term approach to increase coping time in such events is to coat the Zr-alloy cladding with an oxidation-resistant material. In this presentation I will discuss our research on the development of the cold spray process for the deposition of oxidation-resistant materials on Zr-alloy cladding, with emphasis on Cr coatings. In cold spray, powder of the coating material is propelled at supersonic velocities on to the surface of a substrate to form a coating. The particle temperature is low and coating forms in solid-phase due to high strain-rate deformation of the particles upon impact with the substrate, and an associated adiabatic shear process. Results of this developmental study, including the performance of these coatings in steam environments up to 1300 oC will be presented. In situ examination of radiation damage in a TEM showed lower defect evolution kinetics in the cold spray Cr coatings compared to a reference bulk annealed Cr. It is noted that other materials and/or coating technologies are being investigated for this application and the ultimate implementation of the coated cladding concept will be decided by various stakeholders in industry and governmental agencies.