Kevin R Robb Group Leader, Energy Systems Development Group, Nuclear Energy and Fuel Cycle Division Contact robbkr@ornl.gov | 865.576.4730 All Publications 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... Preconceptual design of a fluoride high temperature 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... Design of a Universal Canister System for U.S. High-Level Waste... UNF-ST&DARDS: A Unique Tool for Automated Characterization of Spent Nuclear Fuel and Related Systems... 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... 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 Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core-Concrete Interaction Thermal Modeling Sensitivities with COBRA-SFS for Vertical Dry Casks with Limited Internal Convection... Heat up and potential failure of BWR upper internals during a severe accident... Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents... Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core–Concrete Interaction... A Unified Spent Nuclear Fuel Database and Analysis System... COBRA-SFS DRY CASK MODELING SENSITIVITIES IN HIGH-CAPACITY CANISTERS... Streamlining Analysis Capabilities for SNF Management... 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