Associate R&D Staff, Thermal Hydraulics Group, NEFCD
Dr. Vineet Kumar is an Associate R&D staff in the Thermal Hydraulics group within NEFCD at Oak Ridge National Laboratory (ORNL). Vineet received his PhD (2019) in nuclear engineering at the University of Illinois Urbana-Champaign, and his MS (2012) in mechanical engineering at Purdue University. He has over 5 years academic and industrial experience in thermal-hydraulics system modeling, multi-phase flows experimentation and modeling, and computational fluid dynamics. At ORNL, Vineet has supported various projects such as the Proton Power Upgrade (PPU) project at the Spallation Neutron Source, the NEAMS project on improving two-phase closure modeling in COBRA-TF - the thermal hydraulics code in the VERA suite, as well as NEUP, and HPC for manufacturing projects. Vineet is currently working on supporting the Used Fuel Systems Group on CFD modeling and validation projects, as well as coupling system codes in RAVEN for the Integrated Energy Systems (IES) Program.
- PhD, Nuclear Engineering, University of Illinois, Urbana-Champaign (2019)
- MS, Mechanical Engineering, Purdue University, West Lafayette (2012)
- Bachelors, Mechanical Engineering, Anna University, Chennai, India (2009)
- Experienced in thermal hydraulic system codes and high-fidelity modeling and simulation of various engineering systems.
- Experienced in commercial computational fluid dynamics tools including STAR-CCM+, FLUENT and ANSYS CFX.
- Experience with reactor safety tools and packages, including COBRA-TF, and RELAP5.
- Experienced in programming and scripting languages, including Python and MATLAB.
Awards and Honors
- Barclay Jones Fellowship (2016)
- Felix T Adler Fellowship (2018)
2. Ooi, Z. J., Kumar, V., Brooks, C.S., “Validation of the Interfacial Area Transport Equation Coupled with the Void Transport Equation for Prediction of Flashing Flows,” Nuclear Science and Engineering, 1-22 (2020).
3. Kumar, V., Ooi, Z. J., Brooks, C.S., “Forced convection steam-water experimental database in a vertical annulus with local measurements,” Int. J. Heat Mass Transf., 137, pp. 216-228 (2019).
4. Ooi, Z. J., Kumar, V., Brooks, C.S., “Experimental Database of Two-Phase Natural Circulation with Local Measurements,” Progress in Nuc. Ener., 116, 124-136 (2019).
5. Bottini, J. L., Kumar, V., Hammouti, S., Ruzic, D., Brooks, C.S., “Influence of wettability due to laser-texturing on critical heat flux in vertical flow boiling,” Int. J. Heat Mass Transf. 127, pp. 806-817 (2018).
6. Kumar, V., Brooks, C.S., “Inter-group mass transfer modeling in the two-group two-fluid model with interfacial area transport equation in condensing flow,” Int. J. Heat Mass Transf., 119, pp. 688-703 (2018).
7. Ooi, Z. J., Kumar, V., Bottini, J. L., Brooks, C.S., “Experimental investigation of variability in bubble departure characteristics between nucleation sites in subcooled boiling flow,” Int. J. Heat Mass Transf., 118, pp. 327-339 (2018).
8. Kumar, V., Brooks, C.S., “Validation and model sensitivity of the interfacial area transport equation in condensing flows,” Int. J. Heat Mass Transf., 113, pp. 647-661 (2017).
9. Fullmer, W.D., Kumar, V., Brooks, C.S., “Validation of RELAP5/MOD3.3 for subcooled boiling, flashing and condensation in a vertical annulus,” Progress in Nuclear Energy, 93, pp. 205-217 (2016).