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Caleb Brooks
Caleb Brooks

Caleb Brooks

Assistant Professor
(217) 265-0519
111C Talbot Laboratory

For more information

Education

  • Purdue University, Ph.D., School of Nuclear Engineering, 2014
  • Purdue University, M.S., School of Nuclear Engineering, 2012
  • Purdue University, B.S., School of Nuclear Engineering, 2008

Academic Positions

  • Assistant Professor, University of Illinois, Department of Nuclear, Plasma, and Radiological Engineering, 2014 - present

Resident Instruction

  • NPRE/ME 598: Computational Multi-phase Flow
  • NPRE 448: Nuclear System Engineering and Design
  • NPRE 511: Nuclear Reactor Heat Transfer

Research Areas

  • Advanced Reactor Designs
  • Interfacial Area Transport
  • Reactor Physics
  • Reactor Thermal-Hydraulics
  • Reactor Transient and Accident Analysis
  • Two-Phase Flow and Heat Transfer

Selected Articles in Journals

Articles in Conference Proceedings

  • Chen, J., Brooks, C.S., CFD Simulation of Xenon removal by Helium sparging in molten salt, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Portland, Oregon, USA, August 18-23, 2019.
  • Borowiec, K., Kozlowski, T., Brooks, C.S., Validation and uncertainty quantification for two-phase natural circulation flows using TRACE code, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH), Portland, Oregon, USA, August 18-23, 2019.
  • Zhu, L., Ooi, Z.J., Kumar, V., Brooks, C.S., Current Intergroup Mass Transfer Limitations in the Multi-group Two-fluid Model, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Portland, Oregon, USA, August 18-23, 2019.
  • Zhu, L., Kumar, V., Ooi, Z.J., Brooks, C.S., Current Capability of Interfacial Area Transport Equation in Subcooled Boiling, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Portland, Oregon, USA, August 18-23, 2019.
  • 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, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Portland, Oregon, USA, August 18-23, 2019.
  • Bottini, J.L., Zhu, L., Ooi, Z.J., Zhang, T., Brooks, C.S., A New Dataset with Local Measurement and Visualization of Subcooled Boiling in an Internally Heated Annulus Channel, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Portland, Oregon, USA, August 18-23, 2019.
  • Zhang, T., Ooi, Z.J., Brooks, C.S., Transient local thermal hydraulics data for two-phase flow instability in natural circulation, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Portland, Oregon, USA, August 18-23, 2019.
  • Wang, L., Brooks, C.S., Analysis of wall nucleation modeling for flow boiling in Fluent, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Portland, Oregon, USA, August 18-23, 2019.
  • Kumar, V., Brooks, C.S., Validation of interfacial area concentration approaches for prediction of gas-dispersed condensing flows, 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Portland, Oregon, USA, August 18-23, 2019.
  • Bottini, J.L., Hammouti, S., Ruzic, D., Brooks, C.S., Boiling and Critical heat flux on surfaces of modified wettability and roughness, International Topical Meeting on Advances in Thermal Hydraulics, ANS Winter Meeting, Orlando, FL, November 11-15, 2018.
  • Ooi, Z.J., Kumar, V., Brooks, C.S., Measurement of two-phase natural circulation in a vertical annulus, International Topical Meeting on Advances in Thermal Hydraulics, ANS Winter Meeting, Orlando, FL, November 11-15, 2018.
  • Kumar, V., Ooi, Z.J., Brooks, C.S., Measurement of steam-water flow in a vertical annulus, International Topical Meeting on Advances in Thermal Hydraulics, ANS Winter Meeting, Orlando, FL, November 11-15, 2018.
  • Colgan, N., Bottini, J.L., Brooks, C.S., Flow boiling at subatmospheric pressure. American Nuclear Society - Nuclear and Emerging Technologies for Space, Las Vegas, NV, February 26 - March 1, 2018.
  • Bottini, J.L., Kumar, V., Hammouti, S., Ruzic, D., Brooks, C.S., Critical heat flux on laser-textured surface in flow boiling, 3rd Thermal and Fluids Engineering Conference (TFEC), Fort Lauderdale, FL, USA, March 4-7, 2018.
  • Borowiec, K., Wang, C., Kozlowski, T., Brooks, C.S., Uncertainty Quantification for Steady-Steady PSBT Benchmark using Surrogate Models, 2017 ANS Winter Meeting, Washington, DC, October 29 - November 2, 2017
  • Ooi, Z.J., Kumar, V., Bottini, J., Brooks, C.S., Experimental Investigation of Variability in Bubble Departure Diameter and Bubble Departure Frequency Between Nucleation Sites, The 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), Xian, China,September 3 - 8, 2017
  • Kumar, V., Brooks, C.S., Validation of the Interfacial Area Transport Equation Coupled with Mass Continuity for Prediction of Condensing Flows, The 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), Xian, China,September 3 - 8, 2017
  • Bottini, J., Kumar, V., Brooks, C.S., Critical Heat Flux Experiments Under Low Flow Conditions in a Vertical Rectangular Channel, The 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), Xian, China,September 3 - 8, 2017
  • Zou, L., Zhao, H., Zhang, H., Brooks, C.S., A Revisit to the Hicks Hyperbolic Two-pressure Two-phase Flow Model, The 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), Xi’an, China, September 3-8, 2017.
  • Kumar, V., & Brooks, C.S., The two-group two-fluid model with interfacial area transport equation in condensing flow, 2017 Japan-U.S. seminar on Two-Phase flow Dynamics, Hokkaido, Japan, June 22-24, 2017.
  • Guojun, H., Kozlowski, T., Brooks, C.S., Uncertainty quantification of TRACE sucbooled boiling model using BFBT experiments, The 16th International Topical Meeting of Reactor Thermal Hydraulics (NURETH-16), Chicago, IL, USA, August 30-September 4, 2015.
  • Brooks, C.S., Lietwiler, C.D., Fullmer, W.D., Assessment of RELAP5/MOD3.3 for subcooled boiling, flashing and condensation in a vertical annulus, The 16th International Topical Meeting of Reactor Thermal Hydraulics (NURETH-16), Chicago, IL, USA, August 30-September 4, 2015.
  • Sharma, S.L., Brooks, C.S., Schlegel, J., Hibiki, T., Ishii, M., Buchanan, J.R., Turbulent induced bubble collision force model development and assessment for adiabatic dispersed air-water two-phase flow with two-fluid model, The 16th International Topical Meeting of Reactor Thermal Hydraulics (NURETH-16), Chicago, IL, USA, August 30-September 4, 2015.
  • Brooks, C.S., Hibiki, T., Ishii, M., The Interfacial Area Transport Equation and Boundary Condition Sensitivity in Subcooled Boiling Flow, Japan-U.S. Seminar on Two-phase Flow Dynamics, West Lafayette, IN, May 10-15, 2015.
  • Shi, S., Brooks, C.S., Eoh, J., Ishii, M., Pressurized startup transient analyses for the BWR-type NMR-50, American Nuclear Society Winter Meeting, Anaheim, CA, USA, November 9-13, 2014.
  • Shi, S., Brooks, C.S., Lin, Y.-C., Schlegel, J.P., Hibiki, T., Ishii, M., Experimental investigation of natural circulation instability in a BWR-type small modular reactor, ASME 2014 Small Modular Reactors Symposium, Washington, DC, USA, April 15-17, 2014.
  • Brooks, C.S., Silin, N., Hibiki, T., Ishii, M., Experimental investigation of bubble departure diameter and bubble departure frequency in sub-cooled flow boiling in a vertical annulus, Proceedings of the ASME Summer Heat Transfer Conference, Minneapolis, MN, USA, July 14-19, 2013.
  • Brooks, C.S., Ozar, B., Hibiki, T., Ishii, M., Two-group relative velocity in boiling two-phase flow, Proceedings of the 20th International Conference on Nuclear Engineering, Anaheim, CA, USA, July 30-August 3, 2012.
  • Brooks, C.S., Liu, Y., Hibiki, T., Ishii, M., Void fraction covariance in two-phase flows, Proceedings of the 20th International Conference on Nuclear Engineering, Anaheim, CA, USA, July 30-August 3, 2012.
  • Chen, S.W., Brooks, C.S., Hibiki, T., Ishii, M., Mori, M., Macke, C., Experiment of adiabatic two-phase flow in an annulus under low-frequency vibration, Proceedings of the 20th International Conference on Nuclear Engineering, Anaheim, CA, USA, July 30-August 3, 2012.
  • Ishii, M., Brooks, C.S., Ozar, B., Hibiki, T., Interfacial area transport of subcooled boiling flow in a vertical annulus, Japan-U.S. Seminar on Two-phase Flow Dynamics, Tokyo, Japan, June 7-12, 2012.
  • Ozar, B., Brooks, C.S., Hibiki, T., Ishii, M. Interfacial area transport of steam-water two-phase flow in a vertical annulus at elevated pressures during sub-cooled boiling, The 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Toronto, Ontario, Canada, September 25-29, 2011.
  • Hibiki, T., Ishii, M., Brooks, C.S., Overview of interfacial area transport data development, American Nuclear Society Winter Meeting, Las Vegas, NV, USA, November 7-11, 2010.
  • Hibiki, T., Ishii, M., Liu, Y., Brooks, C.S., Overview of interfacial area transport equation development, American Nuclear Society Winter Meeting, Las Vegas, NV, USA, November 7-11, 2010.

Professional Societies

  • American Society of Thermal and Fluids Engineers (ASTFE) Member
  • American Nuclear Society (ANS) Member
  • Atomic Energy Society of Japan (AESJ)
  • American Society of Mechanical Engineers (ASME) Member
  • Alpha Nu Sigma National Honor Society Member

Honors

  • AESJ Thermal Hydraulics Division Shorei-sho Award (young member research award) (March 2018)
  • AESJ Society-wide Shorei-sho Award (young member research award) (March 2017)
  • Purdue University College of Engineering Outstanding Research Award (April 2013)
  • Purdue University College of Engineering Outstanding Service Award (April 2013)

Courses Taught

  • ME 598 - Computational Multi-Phase Flow
  • NPRE 448 - Nuclear Syst Engrg & Design
  • NPRE 511 - Nuclear Reactor Heat Transfer
  • NPRE 598 - Computational Multi-Phase Flow