Kevin R Robb

Kevin R Robb

Energy Systems Engineering and Testing Team Lead

Bio

Kevin Robb is the lead of the Energy Systems Engineering and Testing Team within the Advanced Reactor Engineering Group at Oak Ridge National Laboratory. His primary research interests are in thermal-hydraulics and nuclear reactor safety. He is supporting the development of fluoride salt-cooled reactor technology through a number of experiments including the liquid salt test loop, salt flow calibration stand, and FLiBe purification facility. He supported the domestic post-accident analyses of the 2011 accidents at Fukushima Daiichi and is the lead at ORNL for the ongoing data collection and forensics work. He is working on the development of accident tolerant fuel (ATF) concepts to provide enhanced safety performance for existing nuclear reactors. Finally, he is the investigating the thermal performance of as-loaded spent nuclear fuel casks. He received his B.S. degrees in Mechanical Engineering and Nuclear Engineering from Pennsylvania State University. He received his M.S. and Ph.D. degrees in Nuclear Engineering and Engineering Physics from the University of Wisconsin-Madison.

Projects

Fluoride Salt-Cooled Reactor Technology
Reactor Safety Technology
Accident Tolerant Fuel
Used Nuclear Fuel

Publications

K. R. Robb, “Heat Up and Failure of BWR Upper Internals during a Severe Accident,” Nuclear Engineering and Design, 314, pp 293-306, April 2017.

K. R. Robb, M. T. Farmer, M. W. Francis, “Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core-Concrete Interaction,” Nuclear Technology, 196, No. 3, Dec. 2016.

M. T. Farmer, K. R. Robb, M. W. Francis, “Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Melt Spreading,” Nuclear Technology, 196, No. 3, Dec. 2016.

K. Banerjee, et. al., “Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks,” Nuclear Technology, 195, August 2016.

K. R. Robb, J. W. McMurray, K. A. Terrani, “M2FT-16OR020205042: Severe Accident Analysis of BWR Core Fueled with UO2/FeCrAl with Updated Materials and Melt Properties from Experiments,” ORNL/TM-2016/237, June 2016.

K. R. Robb, P. K. Jain, T. J. Hazelwood, “High-Temperature Salt Pump Review and Guidelines – Phase I Report,” ORNL/TM-2016/199, May 2016.

N. R. Brown, et al., “Core Design Characteristics of the Fluoride Salt-Cooled High Temperature Demonstration Reactor,” ICAPP 2016, San Francisco, CA, USA, April 17-20, 2016.

A. L. Qualls, et al., “Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design,” ORNL/TM-2016/85, February 2016.

K. R. Robb, “Thermal Modeling Sensitivities with COBRA-SFS for Vertical Dry Casks with Limited Internal Convection,” Proc. of American Nuclear Society (ANS) 2015 Winter Meeting, Washington D. C., Nov. 08-12, 2015.

K. R. Robb, “Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits During BWR Station Blackout Accidents,” Proc. of NURETH-16, Chicago, IL, USA, August 30-September 4, 2015.

J. Rempe, et. al., “US Efforts in Support of Examinations at Fukushima Daiichi,” ANL/LWRS-15/2, August 2015.

M. Snead, et. al., “Technology Implementation Plan ATF FeCrAl Cladding for LWR Application,” ORNL/TM-2014/353, May 2015.

L. L. Snead, et. al., “Technology Implementation Plan Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application,” ORNL/TM-2015/220, April 2015.

R. Devoe, K. R. Robb, “COBRA-SFS Dry Cask Modeling Sensitivities in High-Capacity Canisters,” 2015 International High Level Radioactive Waste Management (IHLRWM), Charleston, SC, USA, April 12–16, 2015.

M. T. Farmer, L. Leibowitz, K. A. Terrani, K. R. Robb, “Scoping Assessments of ATF Impact on Late Stage Accident Progression Including Molten Core-Concrete Interaction,” J. of Nuclear Materials, 448, pp. 534–540, 2014.

L. J. Ott, K. R. Robb, D. Wang, “Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions,” J. of Nuclear Materials, 448, pp. 520–533, 2014.

K. R. Robb, M. W. Francis, L. J. Ott, “Insight from Fukushima Daiichi Unit 3 Investigations using MELCOR,” Nuclear Technology, vol. 186, iss. 2, May 2014.

G.L. Yoder, Jr., et al., “An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor,” Annals of Nuclear Energy, 64, pp. 511–517, February 2014.

D. E. Holcomb, et. al., “Fluoride Salt-Cooled High-Temperature Reactor Research Development and Demonstration Roadmap,” ORNL/TM-2013/401, November 2013.

R.O. Gauntt, et. al., “Fukushima Daiichi Accident Study (Status as of April 2012),” SAND2012-6173, June 2012.

Facilities Used

Thermal Hydraulics High Bay