Contact Information robbkr@ornl.gov 865.576.4730 orcid.org/0000-0002-3988-438X Key Links Google Scholar Profile Related Organizations Fusion and Fission Energy and Science Directorate Kevin R Robb Group Leader, Energy Systems Development Group, Nuclear Energy and Fuel Cycle Division All Publications Facility to Alleviate Salt Technology Risks (FASTR): Preliminary Design Report with Failure Modes and Effects Analysis... Evaluation of Power Fluidic Pumping Technology for Molten Salt Reactor Applications... ALLOY SELECTION AND C-276 CODE DESIGN VALUE EXTENSION FOR ADVANCED MOLTEN SALT TECHNOLOGY TEST FACILITIES EXPERIMENTATION... Simplified High-Temperature Molten Salt CSP Plant Preconceptual Design... Chloride Salt Purification by Reaction With Thionyl Chloride Vapors to Remove Oxygen, Oxygenated Compounds, and Hydroxides... Experimental investigation on the coolability of nuclear reactor debris beds using seawater... Measuring the solubility of xenon in molten chloride salt... Corrosion of 316H Stainless Steel Specimens in Two FLiBe (LiF-BeF2) Salt Batches... Validating modern methods for impurity analysis in fluoride salts... Tribological behavior of ceramic-alloy bearing contacts in molten salt lubrication for concentrating solar power... Effects of Particle Size and Concentration of Magnesium Oxide on the Lubricating Performance of a Chloride Molten Salt for Co... Molten Salt Air-Cooled Heat Exchanger Fluid Dynamics... Design Overview of the Facility to Alleviate Salt Technology Risks (FASTR)... Assessment on the Practicality of Off-the-Shelf Valves for Use in Molten Salt... Impact of FeCrAl ATF Concept on BWR Upper Internal Structures During Station Blackouts... Thermal Analysis Capability of UNF-ST&DARDS Development of Streamlined Nuclear Safety Analyses for Used Nuclear Fuel Applications Sensitivity Analysis for Best-Estimate Thermal Models of Vertical Dry Cask Storage Systems... Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: motivation and overview Heat Up and Failure of BWR Upper Internals During a Severe Accident Preconceptual design of a fluoride high termperature salt-cooled engineering demonstration reactor: Motivation and overview Design of a Universal Canister System for US High-Level Waste FUKUSHIMA DAIICHI UNIT 1 EX-VESSEL PREDICTION: CORE–CONCRETE INTERACTION UNF-ST&DARDS: A Unique Tool for Automated Characterization of Spent Nuclear Fuel and Related Systems Design of a Universal Canister System for U.S. High-Level Waste Pagination Current page 1 Page 2 Page 3 Next page ›› Last page Last »
