Kevin Robb Group Leader, Energy Systems Development Group, Nuclear Energy and Fuel Cycle Division Contact ROBBKR@ORNL.GOV All Publications Molten Salt Air-Cooled Heat Exchanger Fluid Dynamics Design Overview of the Facility to Alleviate Salt Technology Risks (FASTR) ALLOY SELECTION AND C-276 CODE DESIGN VALUE EXTENSION FOR ADVANCED MOLTEN SALT TECHNOLOGY TEST FACILITIES EXPERIMENTATION Evaluation of Power Fluidic Pumping Technology for Molten Salt Reactor Applications Facility to Alleviate Salt Technology Risks (FASTR): Preliminary Design Report with Failure Modes and Effects Analysis 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 Development of Streamlined Nuclear Safety Analyses for Used Nuclear Fuel Applications Thermal Analysis Capability of UNF-ST&DARDS 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... Design of a Universal Canister System for US High-Level Waste Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview 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 Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks Analysis of Operational and Safety Performance for Candidate Accident Tolerant Fuel and Cladding Concepts Core Design Characteristics of the Fluoride Salt-Cooled High Temperature Demonstration Reactor Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core-Concrete Interaction Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview Thermal Modeling Sensitivities with COBRA-SFS for Vertical Dry Casks with Limited Internal Convection... Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core–Concrete Interaction... Pagination First page « First Previous page ‹‹ Page 1 Current page 2 Page 3 Next page ›› Last page Last » Key Links ORCID Google Scholar Profile Organizations Fusion and Fission Energy and Science Directorate Nuclear Energy and Fuel Cycle Division Advanced Reactor Engineering and Development Section Energy Systems Development Group
