Robert K Salko
Robert K Salko
Staff Research Scientist
Dr. Robert Salko graduated from the Pennsylvania State University in 2006 with bachelor’s degrees in both mechanical and nuclear engineering, in 2009 with a master’s degree in nuclear engineering, and in 2012 with a PhD in nuclear engineering. His research topics included thermal-hydraulics code development and modeling and simulation, with emphasis on subchannel tools. In 2012, he began supporting the COBRA-TF (CTF) subchannel code, which was adopted by the Consortium for Advanced Simulation of Light Water Reactors (CASL) for use as the thermal-hydraulic subchannel component in the core simulator software the consortium was developing. Some of Robert’s significant contributions involved: 1) implementation of software quality assurance measures in the maintenance and development of CTF, 1) performance improvements, including full domain-decomposition based parallelization of the code, 3) implementation of improved models, 4) supporting coupling to neutronics, fuel performance, and Crud chemistry packages, and 5) code verification, validation, and benchmarking activities. Robert has also been involved in numerous modeling and simulation activities with CTF, and also provides support to the CTF User Group.
Co-recipient of an ORNL Signicant Event Award for contribution to CASL milestone, "Qualify VERA-CS for Multi-Cycle (with Fuel Reloading) PWR Core Simulations Capability", October, 2015
Chosen as the CASL 2014 Technical Contributor of the Year (CASL "Knight" award)
Graduated with honors in Nuclear Engineering, December, 2006
Obtained Engineer in Training certicate, April, 2006
Received scholarships based on academic achievement from Dominion (fall, 2004), Exelon Nuclear (summer, 2005), and the American Nuclear Society (summer, 2005)
Consortium for Advanced Simulation of Light Water Reactors
V. Kucukboyaci, Y. Sung, and R. Salko, "COBRA-TF Parallelization and Application to PWR Reactor Core Subchannel DNB Analysis," ANS MC2015-Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method, 2015.
R. Salko, T. Lange, V. Kucukboyaci, Y. Sung, J. Gehin, and M. Avramova, "Development of COBRA-TF for Modeling Full-Core, Reactor Operating Cycles," Advances in Nuclear Fuel Management V (ANFM 2015), Hilton Head Island, South Carolina, USA, March 29-April 1, 2015.
R. Pawlowski, K. Clarno, R. Montgomery, R. Salko, T. Evans, J. Turner, and D. Gaston, "Design of a High Fidelity Core Simulator for Analysis of Pellet Clad Interaction," Proceedings of the ANS Joint International Conference on Mathematics and Computation (M&C 2015), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method, Nashville, TN, USA, 2015.
R.K. Salko, R.C. Schmidt, and M.N. Avramova, "Optimization and Parallelization of the Thermal-Hydraulic Subchannel Code for High-Fidelity Multi-Physics Applications", Annals of Nuclear Energy (2014), doi:10.1016/j.anucene.2014.11.005.
S. Palmtag, K. Clarno, G. Davidson, R. Salko, T. Evans, J. Turner, and R. Schmidt, "Coupled Neutronics and Thermal-Hydraulic Solution of a Full-Core PWR using VERA-CS," Proceedings of International Topical Meeting on Advances in Reactor Physics (PHYSOR), Kyoto, Japan, 2014.
R.K. Salko, et al, "Improvements, Enhancements, and Optimization of COBRA-TF," International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2013), Sun Valley, Idaho, USA, May 5-9, 2013.
R.K. Salko and M.N. Avramova, "Uncertainty Analysis of Sub-channel Code Calculated ONB Wall Superheat in Rod-Bundle Experiments using the GRS Methodology," Progress in Nuclear Energy, 65, pp. 42--49, May, 2013.
P. Peturaud, R.K. Salko, A. Bergeron, S. Yagnik, and M.N. Avramova, "Analyses of Single-Phase Heat Transfer and Onset of Nucleate Boiling in a Rod Bundle with Mixing Vane Grids," International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-14, Toronto, Ontario, Canada, September 25-30, 2011.
R.K. Salko, M.N. Avramova, and A. Ohnuki, "Modification of COBRA-TF for Improved Vessel-Wide Transient Analysis," American Nuclear Society Meeting, Las Vegas, NV, Winter, 2010.
P. Peturaud, R.K Salko, A. Bergeron, and M.N. Avramova, "Analyses of Single-Phase Heat Transfer and Onset of Nucleate Boiling in Rod Bundles," International Conference On Nuclear Engineering 18, Xi'an, China, May 17-21, 2010.
Oak Ridge Leadership Computing Facility