The Grainger College of Engineering
Nuclear, Plasma & Radiological Engineering

Caleb Brooks

Caleb Brooks
Caleb Brooks
Associate Professor and Donald Biggar Willett Faculty Scholar
(217) 265-0519
111C Talbot Laboratory

For More Information

Education

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

Academic Positions

  • 2024-Present, WPI Associate Professor, International Institute for Carbon-Neutral Energy Research, Kyushu University
  • 2020-Present, Associate Professor, Department of Nuclear, Plasma, and Radiological Engineering, University of Illinois
  • 2014-2020, Assistant Professor, Department of Nuclear, Plasma, and Radiological Engineering, University of Illinois at Urbana-Champaign

Resident Instruction

  • NPRE/ME 598: Computational Multi-phase Flow
  • NPRE 511: Nuclear Reactor Heat Transfer
  • NPRE 449: Nuclear System Engineering and Design
  • NPRE 448: Nuclear System Engineering and Design
  • NPRE 349: Introduction to NPRE Heat Transfer
  • NPRE 247: Modeling Nuclear Energy System

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

  • Chaube, A., Novak, A., Shaver, D., Brooks, C.S., Natural Convection CFD Modeling of a Microreactor Air Jacket, American Nuclear Society Winter Meeting, Washington DC, November 13-14, 2024
  • Chandra, N., Kalinichenko, D., Ross, M., Brooks, C.S., Bindra, H., Optimal Sizing Toolbox for Energy Generation and Storage for a Nuclear Hybrid Microgrid, American Nuclear Society Winter Meeting, Washington DC, November 13-14, 2024
  • OBrien, K., Brooks, C.S., Gioja, L., Grunloh, T.P., Baroi, C., Novak, A.J., He, J., Integrating Direct Air Capture (DAC) with Microreactors: Pathways to Net Negative, 17th International Conference on Greenhouse Gas Control Technologies GHGT-17, October 20-24, 2024, Calgary Canada
  • Shao, H., Zhang, T., Brooks, C.S., Zhu, L., Demonstrated potential of machine learning in driving post-processing of four-sensor conductivity probes, Specialist Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal Hydraulics and Severe Accidents SWINTH-2024, Dresden, Germany, June 17-20, 2024
  • Aziz, M.A.B., Callaway, C., Brown, N., Brooks, C.S., Integrated Energy System Model in OpenModelica for Producing Energy, Powering a Desalination Plant and for District Heating, 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20), Washington D.C., United States, August 20-25, 2023
  • Chen, J., Brooks, C.S., Experimental Method for Measurement of Density and Viscosity of High Temperature Heat Transfer Fluid, 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20), Washington D.C., United States, August 20-25, 2023
  • Malik, M.S., Chen, J., Di Fulvio, A., Brooks, C.S., Grunloh, T.P., “Uncoupled DEM Simulation to Investigate the Impact of Coolant flow on Pebble flow in Pebble Bed Reactor” 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20), Washington D.C., United States, August 20-25, 2023
  • Zhang, T., Hua, T., Ooi, Z.J., Zou, L., Brooks, C.S., Modeling the Prismatic HTGR Core in SAM by Representative Channels, 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20), Washington D.C., United States, August 20-25, 2023
  • Zhang, T., Brooks, C.S., Simulation of Oscillatory Flow Induced by Flashing Instability Using the ASYST System Analysis Code, 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20), Washington D.C., United States, August 20-25, 2023
  • A. Chaube, A.J. Novak, C.S. Brooks, Demonstration of Cardinal for Modelling Core-Radial Expansion in Sodium-Cooled Fast Reactor Assemblies, The International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C), Ontario Canada, August 13-17, 2023
  • A. Chaube, A.J. Novak, D.R. Shaver, C.S. Brooks, Heat Transfer Performance Assessment of a Microreactor Air Jacket using CFD, The International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C), Ontario Canada, August 13-17, 2023
  • Malik, M.S., Chen, J., Brooks, C.S., CFD Simulation for Evaluating Pressure Drop and Heat Transfer Correlations for Pebble Bed Reactor, Computational Fluid Dynamics for Nuclear Reactor Safety (CFD4NRS-9) OECD NEA Workshop, College Station, Texas A&M University, United States, February 20-22, 2023
  • Novak, A.J., Chaube, A., Shaver, D.R., Brooks, C.S. Validation of NekRS-MOOSE Conjugate Heat Transfer Coupling for a 7-Pin Bare Bundle, Proceedings of ANS 2022
  • McCord, M., Bottini, J. L., and Brooks, C. S., Experimental Comparison of Transient and Stable Film Boiling Minimum Rewet Conditions at Low Pressure and Low Flow Rate, International Topical Meeting on Advances in Thermal Hydraulics, Anaheim, CA, June 12-16, 2022
  • Bottini, J.L., Ooi, Z.J., Brooks, C.S., Trends in vapor generation and intergroup mass transfer in the two-group two fluid model in flows with phase change, Japan-U.S. Seminar on Two-phase Flow Dynamics, Ann Arbor, MI, USA, May 8-11, 2022 (invited)
  • Zhang, T., Brooks, C.S., Stability Tests and Analysis on a Low-pressure Natural Circulation Loop with Flashing Instability, 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH), Brussels, Belgium, March 6-11, 2022 (Best Paper Award)
  • Zhu, L., Zhang, T., Ooi, Z.J., Brooks, C.S., Shan, J., Pan, L., Validation of Interfacial Mass Transfer Closures in Drift Flux Model for Condensation Flow, 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH), Brussels, Belgium, March 6-11, 2022
  • Zhu, L., Ooi, Z.J., Zhang, T., Brooks, C.S., Pan, L., Flow Regime Identification for Boiling Flow and Study of Feed-in Information in Unsupervised Machine Learning, 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH), Brussels, Belgium, March 6-11, 2022
  • Bottini, J. L., Brooks, C. S., Modeling Vapor Mass Generation in Vertical Boiling Flow Using the Two-Group Two-Fluid Model, 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH), Brussels, Belgium, March 6-11, 2022
  • Chen, J., Brooks, C.S., Simulation and Analysis of a Prototypical Bubble Separator for Molten Salt Reactor, ANS Winter Meeting, Washington, D.C., November 31-December 3, 2021
  • Lee, A. J. H., Wodrich, L., Brooks, C.S., Kozlowski, T., Modeling Micro-Reactor Benefits to an Existing Campus Microgrid, ANS Winter Meeting, Washington, D.C., November 31-December 3, 2021
  • Zhu, L., Bottini, J.L., Brooks, C.S., Comparison of intergroup mass transfer coefficient correlations in two-group IATE for subcooled boiling flow, ANS Winter Meeting, Chicago, IL, November 16-19, 2020
  • Zhang, T., Brooks, C.S., Linear Stability Analysis of Flashing Instability with Homogeneous Equilibrium Model, ANS Winter Meeting, Chicago, IL, November 16-19, 2020
  • McCord, M.L., Bottini, J.L., Brooks, C.S., Transient measurement of film boiling during cooldown from CHF in vertical flow, ANS Winter Meeting, Chicago, IL, November 16-19, 2020
  • Chen, J., Brooks, C.S., Simulation and analysis of a prototypical xenon removal system, ANS Winter Meeting, Chicago, IL, November 16-19, 2020
  • Chen, J., Brooks, C.S., Experimental study of fluid motion and mass transfer in a cylindrical bubble column, International Topical Meeting on Advances in Thermal Hydraulics, Paris, France, March 31-April 3, 2020
  • Ooi, Z.J., Brooks, C.S., Beyond time-averaged measurements using conductivity probes, International Topical Meeting on Advances in Thermal Hydraulics, Paris, France, March 31-April 3, 2020
  • Chen, J., Brooks, C.S., Experimental study of fluid motion and mass transfer in a cylindrical bubble column, International Topical Meeting on Advances in Thermal Hydraulics, Paris, France, March 31-April 3, 2020
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • 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
  • Guojun, H., Kozlowski, T., Brooks, C.S., Uncertainty quantification of TRACE subcooled 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
  • 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
  • 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
  • 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
  • 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., 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
  • 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

Journal Editorships

  • Editorial Board, Experimental and Computational Multiphase Flow, Springer

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

Recent Courses Taught

  • NPRE 247 - Modeling Nuclear Energy System
  • NPRE 349 - Intro to NPRE Heat Transfer
  • NPRE 449 (NPRE 448) - Nuclear Syst Engrg & Design
  • NPRE 511 - Nuclear Reactor Heat Transfer

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