Selected Articles in Journals

• Miernicki, E.A, Heald, A.L., Huff, K.D., Brooks, C.S., Margenot, A.J., Nuclear waste heat use in agriculture: history and opportunities, Journal of Cleaner Production, 267, 121918, DOI: 10.1016/j.jclepro.2020.121918
• Bottini, J.L., Zhu, L., Ooi, Z.J., Zhang, T., Brooks, C.S., 2020 Experimental study of boiling flow in a vertical Heated Annulus with local two-phase measurements and visualization, International Journal of Heat and Mass Transfer, 155, 119712, DOI: 10.1016/j.ijheatmasstransfer.2020.119712
• Kumar, V., Brooks, C.S., 2020, Benchmark of interfacial area concentration approaches for the two-fluid model in gas-dispersed condensing flow, Progress in Nuclear Energy, 124, 103329, DOI: 10.1016/j.pnucene.2020.103329
• Borowiec, K., Kozlowski, T., Brooks, C.S., 2020. Validation and uncertainty quantification for two-phase natural circulation flows using TRACE code, Nuclear Science and Engineering (in-press), DOU:10.1080/00295639.2020.1713671
• Ooi, Z.J., Kumar, V., Brooks, C.S., 2020. Validation of the Interfacial Area Transport Equation Coupled with the Void Transport Equation for Prediction of Flashing Flows, Nuclear Science and Engineering, (in-press), DOI: 10.1080/00295639.2020.1732123
• Zhang, T., Ooi, Z.J., Skrzpek, M., Brooks, C.S., 2020. A multi-dimensional dataset for two-phase instability in low pressure natural circulation based on direct transient local measurement, International Journal of Heat and Mass Transfer, Vol 151, 119447. DOI:10.1016/j.ijheatmasstransfer.2020.119447
• Ooi, Z.J., Kumar, V., Brooks, C.S., 2019. Experimental database of two-phase natural circulation with local measurements, Progress in Nuclear Energy, Vol. 116, 124-136. DOI:10.1016/j.pnucene.2019.04.003
• Kumar, V., Ooi, Z.J., Brooks, C.S., 2019. Forced convective steam-water experimental database in a vertical annulus with local measurements, International Journal of Heat and Mass Transfer, Vol 137, 216-228. DOI:10.1016/j.ijheatmasstransfer.2019.03.099
• Colgan, N., Bottini, J.L., Ooi, Z.J., Brooks, C.S., 2019. Experimental study of wall nucleation characteristics in flow boiling under subatmospheric pressures in a vertical square channel, International Journal of Heat and Mass Transfer, Vol. 134, 58-68. DOI:10.1016/j.ijheatmasstransfer.2018.12.153
• Ju, P., Liu, Y., Brooks, C.S., Ishii, M., 2019. Prediction of Interfacial Shear Stress of Vertical Upward Adiabatic Annular Flow in Pipes, International Journal of Heat and Mass Transfer, Vol 133, 500-509. DOI:10.1016/j.ijheatmasstransfer.2018.12.057
• Colgan, N., Bottini, J.L., Martinez-Cuenca, R., Brooks, C.S., Experimental study of critical heat flux in flow boiling under subatmospheric pressure in a vertical square channel, International Journal of Heat and Mass Transfer, Vol 130, 514-522. DOI:10.1016/j.ijheatmasstransfer.2018.10.082
• Bottini, J.L., Kumar, V., Hammouti, S., Ruzic, D., Brooks, C.S., 2018. Influence of wettability due to laser-texturing on critical heat flux in vertical flow boiling, International Journal of Heat and Mass Transfer, Vol 127, 806-817. DOI:10.1016/j.ijheatmasstransfer.2018.06.113
• Kumar, V., Brooks, C.S., 2018. Inter-group mass transfer modeling in the two-group two-fluid model with interfacial area transport equation in condensing flow, International Journal of Heat and Mass Transfer, Vol 119, 688-703. DOI:10.1016/j.ijheatmasstransfer.2017.11.087.
• Ooi, Z.J., Kumar, V., Bottini, J.L., Brooks, C.S., 2018. Experimental investigation of variability in bubble departure characteristics between nucleation sites in subcooled boiling flow, International Journal of Heat and Mass Transfer, Vol 118, 327-339. DOI:10.1016/j.ijheatmasstransfer.2017.10.116
• Kumar, V., Brooks, C.S., 2017. Validation and model sensitivity of the interfacial area transport equation in condensing flows, International Journal of Heat and Mass Transfer, Vol 113, 647–661. DOI:10.1016/j.ijheatmasstransfer.2017.05.118
• Sharma, S.L., Hibiki, T., Ishii, M., Brooks, C.S., Schlegel, J.P., Liu, Y., Buchanan Jr., J.R., 2017 Turbulence-induced Bubble Collision Modeling and Validation in Adiabatic Two-phase Flow, Nuclear Engineering and Design, Vol 312, 399-409. DOI:10.1016/j.nucengdes.2016.05.006
• Fullmer, W.D., Kumar, V., Brooks, C.S., 2016. Validation of RELAP5/MOD3.3 for Subcooled Boiling, Flashing and Condensation in a Vertical Annulus, Progress in Nuclear Energy, Vol 93, 205-217 DOI: 10.1016/j.pnucene.2016.08.013
• Dandekar, A.V., Brooks, C.S., 2016. Modeling of Void Fraction Covariance in Two-phase Flows with Phase Change, International Journal of Heat and Mass Transfer, Vol 100, 231-242. DOI:10.1016/j.ijheatmasstransfer.2016.04.066
• Brooks, C.S. and Hibiki, T., 2016. Modeling and validation of interfacial area transport equation in subcooled boiling flow, Journal of Nuclear Science and Technology, Vol 53 (8), 1192-1204. DOI: 10.1080/00223131.2015.1096852
• Shi, S., Schlegel, J.P., Brooks, C.S., Lin, Y-C., Eoh, J., Liu, Z., Zhu, Q., Liu, Y., Hibiki, T., Ishii, M., 2015. Experimental investigation of natural circulation instability in a BWR-type small modular reactor, Progress in Nuclear Energy, Vol 85, 96-107. DOI:10.1016/j.pnucene.2015.06.014
• Ju, P., Brooks, C.S., Ishii, M., Liu, Y., Hibiki, T., 2015. Film thickness of vertical upward co-current adiabatic flow in pipes, International Journal of Heat and Mass Transfer, Vol 89, 985-995. DOI:10.1016/j.ijheatmasstransfer.2015.06.002
• Shi, S., Wu, Z., Liu, Z., Schlegel, J.P., Brooks, C.S., Eoh, J., Yan, Y., Liu, Y., Yang, W.S., Ishii, M., 2015. Experimental study of natural circulation instability with void reactivity feedback during startup transients for a BWR-type SMR, Progress in Nuclear Energy, 83, 73-81. DOI:10.1016/j.pnucene.2015.03.003
• Brooks, C.S. and Hibiki, T., 2015. Wall Nucleation Modeling in Subcooled Boiling Flow, International Journal of Heat and Mass Transfer, Vol 86., 183-196, DOI: 10.1016/j.ijheatmasstransfer.2015.03.005.
• Brooks, C.S., Silin, N., Hibiki, T., Ishii, M., 2015. Experimental investigation of wall nucleation characteristics in flow boiling, Journal of Heat Transfer, Vol 137 (5), DOI: 10.1115/1.4029593.
• Martinez, R., Brooks, C.S., Julia, J.E., Hibiki, T., Ishii, M., 2015. Stochastic nature of wall nucleation and its impact on the time averaged boundary condition, Journal of Heat Transfer, Vol 137 (2), DOI:10.1115/1.4028975.
• Brooks, C.S., Ozar, B., Hibiki, T., Ishii, M., 2014. Interfacial area transport of subcooled boiling flow in a vertical annulus, Nuclear Engineering and Design, Vol 269, 152-163, DOI:10.1016/j.nucengdes.2013.04.041.
• Brooks, C.S., Liu, Y., Hibiki, T., Ishii, M., 2014. Effect of void fraction covariance on relative velocity in gas-dispersed two-phase flow, Progress in Nuclear Energy, Vol 70, 209-220, DOI:10.1016/j.pnucene.2013.10.006.
• Ozar, B, Brooks, C.S., Euh, D.J, Hibiki, T., Ishii, M., 2013. Investigation of one-dimensional interfacial area transport for vertical upward air-water two-phase flow in an annular channel at elevated pressures, Nuclear Engineering and Design, Vol 263, 362-379, DOI:10.1016/j.nucengdes.2013.05.018.
• Ozar, B, Brooks, C.S., Hibiki, T., Ishii, M., 2013. Interfacial area transport of vertical upward steam-water two-phase flow in an annular channel at elevated pressures, International Journal of Heat and Mass Transfer, Vol 57, 504-518, DOI:10.1016/j.ijheatmasstransfer.2012.10.059.
• Brooks, C.S., Hibiki, T., Ishii, M., 2012. Interfacial Drag Force in One-dimensional Two-fluid Model, Progress in Nuclear Energy, Vol 61, 57-68, DOI:10.1016/j.pnucene.2012.07.001.
• Brooks, C.S., Ozar, B., Hibiki, T., Ishii, M., 2012. Two-group Drift-flux Model in Boiling Flow, International Journal of Heat and Mass Transfer, Vol 55, 6121-6129, DOI:10.1016/j.pnucene.2012.07.001.
• Brooks, C.S., Paranjape, S.S., Ozar, B., Hibiki, T., Ishii, M., 2012. Two-group Drift-flux Model for Closure of the Modified Two-fluid Model, International Journal of Heat and Fluid Flow, Vol 37, 196-208, DOI:10.1016/j.ijheatfluidflow.2012.04.002.

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), Xi’an, 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), Xi’an, 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), Xi’an, 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., 2015. 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.
• Brooks, C.S., Lietwiler, C.D., Fullmer, W.D., 2015. 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.
• Sharma, S.L., Brooks, C.S., Schlegel, J., Hibiki, T., Ishii, M., Buchanan, J.R., 2015. 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.
• Brooks, C.S., Hibiki, T., Ishii, M., 2015. 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.
• Shi, S., Brooks, C.S., Eoh, J., Ishii, M., 2014. Pressurized startup transient analyses for the BWR-type NMR-50, American Nuclear Society Winter Meeting, Anaheim, CA, USA, November 9-13.
• Brooks, C.S., Silin, N., Hibiki, T., Ishii, M., 2013. 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.
• Brooks, C.S., Ozar, B., Hibiki, T., Ishii, M., 2012. 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.
• Brooks, C.S., Liu, Y., Hibiki, T., Ishii, M., 2012. Void fraction covariance in two-phase flows”, Proceedings of the 20th International Conference on Nuclear Engineering, Anaheim, CA, USA, July 30-August 3.
• Chen, S.W., Brooks, C.S., Hibiki, T., Ishii, M., Mori, M., Macke, C., 2012. 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.
• Ishii, M., Brooks, C.S., Ozar, B., Hibiki, T., 2012. 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.
• Ozar, B., Brooks, C.S., Hibiki, T., Ishii, M., 2011. 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.
• Hibiki, T., Ishii, M., Brooks, C.S., 2010. Overview of interfacial area transport data development, American Nuclear Society Winter Meeting, Las Vegas, NV, USA, November 7-11.
• Hibiki, T., Ishii, M., Liu, Y., Brooks, C.S., 2010. Overview of interfacial area transport equation development, American Nuclear Society Winter Meeting, Las Vegas, NV, USA, November 7-11.

Reports

• Brooks, C.S., Kozlowski, T., Zou, L., H. Zhang, Golchert, B.M., Ooi, Z.J., Wang, C., Borowiec, K., Kumar, V., “Validation of RELAP-7 for Forced Convective and Natural Circulation Reactor Flows: Final Project Report” UIUC Technical Report, Project ID: DOE-16-10630, 2019.
• Kozlowski, T., Borowiec, K., Wang, C., Brooks, C.S., Zou, L., Golchert, B., 2018, “Validation of RELAP-7 for Forced Convective and Natural Circulation Reactor Flows: Task 2.2 Validation under natural circulation flow” UIUC Technical Report, Project ID: DOE-16-10630, Milestone ID: M2NU-16-IL-UIUC-030401-066.
• Heald, A.L., Miernicki, E.A., Fairhurst, R.E., Margenot, A.J., Huff, K.D., Brooks, C.S., 2018. Investigation of Agricultural Uses of Nuclear Waste Heat. UIUC Technical Report. October, 2018.
• Brooks, C.S., 2018, Identification of experiments to support seven-equation two-phase flow model assessment, Idaho National Laboratory, Technical Report INL/EXT-18-51203.
• Brooks, C.S., Ooi, Z.J., Kumar, V., Kozlowski, T., Zou, L., Golchert, 2018, “Validation of RELAP-7 for Forced Convective and Natural Circulation Reactor Flows: Task 2.1 Two-phase natural circulation data for validation of system analysis codes” UIUC Technical Report, Project ID: DOE-16-10630, Milestone ID: M2NU-16-IL-UIUC-030401-065.
• Kozlowski, T., Borowiec, K., Wang, C., Brooks, C.S., Zou, L., Golchert, B., 2018, “Validation of RELAP-7 for Forced Convective and Natural Circulation Reactor Flows: Task 1.2 Validation under forced convective flow” UIUC Technical Report, Project ID: DOE-16-10630, Milestone ID: M2NU-16-IL-UIUC-030401-063.
• Brooks, C.S., Kozlowski, T., Zou, L., Golchert, B., Borowiec, K., Wang, C., Ooi, Z.J., Kumar, V., 2017, “Validation of RELAP-7 for Forced Convective and Natural Circulation Reactor Flows: Task 1.1 Synthesis of existing forced convective two-phase flow datasets for validation of RELAP7” Technical Report, Project ID: DOE-16-10630, Milestone ID: M2NU-16-IL-UIUC-030401-062. 
• Liang, X., Bottini, J.L., Ooi, Z.J., Brooks, C.S., 2017. “Natural Circulation Demonstration Facility”. Technical report, UIUC/NPRE/MTCL-2017-01.
• Ishii, M., Hibiki, T., Yang, W.S., Schlegel, J.P., Wu, Z., Shi, S., Y.C., Lin, Eoh, J.H., Brooks, C.S., Clark, C. “Investigation of natural circulation instability and transients in passively safe modular reactors” Purdue Technical Report PU/NE-13-10, 2013
• Brooks, C.S., Hibiki, T., Ishii, M. “Foundation and Treatment of the Four-field Two-fluid Model in Annular Flows” Purdue Technical Report PU/NE-12-26, 2012
• Brooks, C.S., Hibiki, T., Ishii, M. “Two-group Drift Flux Model” Purdue Technical Report PU/NE-12-04, 2012
• Ishii, M., Hibiki, T., Liu, Y., Sharma, S.L., Lee, D.Y., Brooks, C.S. “Implementation of interfacial area transport equation into 3-D CFD code, task 1: spatial benchmarking of interfacial area transport equation, subtask 1f: Systematic Benchmarking of the Two-group IATE with Two Gas Momentum Equations for Uniform and Non-Uniform Inlet Boundary Conditions.” Purdue Technical Report PU/NE-12-07, 2012
• Ishii, M., Hibiki, T., Liu, Y., Sharma, S.L., Lee, D.Y., Brooks, C.S. “Implementation of interfacial area transport equation into 3-D CFD code, task 1: spatial benchmarking of interfacial area transport equation, subtask 1f: Additional Diffusion Force Model Development and Near Wall Treatment of Two-Fluid Model.” Purdue Technical Report PU/NE-12-20, 2012
• Brooks, C.S., Hibiki, T., Ishii, M. “Interfacial drag force in one-dimensional two-fluid model” Purdue Technical Report PU/NE-11-07, 2011
• Ishii, M., Hibiki, T., Ozar, B., Brooks, C.S. “Interfacial Area Transport with Boiling and Condensation in an Annulus” Purdue Technical Report PU/NE-09-62, 2009
• Ishii, M., Hibiki, T., Ozar, B., Brooks, C.S. “Data Report for Steam-Water Boiling Flow in a Vertical Annulus” Purdue Technical Report PU/NE-09-57, 2009
• Ishii, M., Liu, Y., Brooks, C.S. “Scaling Analysis and Model Development of Air Entrainment before the ECCS Pump Suction Section” Purdue Technical Report PU/NE-09-52, 2009
• Ishii, M., Brooks, C.S., Paranjape, S.S., Ozar, B. “PWROG Air Entrainment Tests Data Report for 12 Inch Test Section” Purdue Technical Report PU/NE-09-53, 2009
• Ishii, M., Brooks, C.S., Paranjape, S.S. “Air Entrainment in Nuclear Safety System Suction Piping: Design Drawings, Equipment and Instrument Specifications, and Test Plan for 4 inch Tests” Purdue Technical Report PU/NE-09-54, 2009
• Ishii, M., Brooks, C.S., Schlegel, J.P. “Air Entrainment in Nuclear Safety System Suction Piping: Test Procedures and Shakedown Test Planning for 4 inch Tests” Purdue Technical Report PU/NE-09-55, 2009
• Ishii, M., Brooks, C.S., Schlegel, J.P., Chen, S.W. “PWROG Air Entrainment Tests Data Report for 4 Inch Test Section” Purdue Technical Report PU/NE-09-56, 2009
• Ishii, M., Brooks, C.S., Schlegel, J.P. “PWROG Air Entrainment Tests Data Report for 4 Inch Heated Test Section” Purdue Technical Report PU/NE-09-63, 2009
• Ishii, M., Liu, Y., Brooks, C.S. “Scaling Analysis of Air Entrainment Tests” Purdue Technical Report PU/NE-09-66, 2009
• Ishii, M., Brooks, C.S., Sawant, P.H., Paranjape, S.S. “Air entrainment in the flow of an emergency core cooling suction: Design Drawings, Equipment and Instrumentation Specifications, and Test Plan” Purdue Technical Report PU/NE-08-31, 2008
• Ishii, M., Brooks, C.S., Sawant, P.H., Paranjape, S.S. “Air entrainment in the flow of an emergency core cooling suction: Test Procedures and Shake-down Test Planning” Purdue Technical Report PU/NE-08-30, 2008