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Juliana Pacheco Duarte

Juliana Pacheco Duarte

Associate Professor

Prof. Juliana Pacheco Duarte is an Associate Professor in the Department of Nuclear Engineering & Engineering Physics at the University of Wisconsin-Madison. Her research focuses on safety analysis, thermal-hydraulics, and risk assessment of advanced nuclear systems. Dr. Duarte’s expertise encompasses safety analysis, experimental design for studying two-phase heat transfer phenomena at high-pressure conditions, and computational thermal-hydraulic analysis using subchannel codes (COBRA, CTF) and system safety codes (TRACE, MELCOR). Her expertise encompasses several machine learning and statistical methods in safety, fire data, and reliability analysis. Before coming to the U.S., Dr. Duarte worked as a graduate student at the Brazilian Navy’s thermal-hydraulics division, designing critical heat flux experiments for nuclear propulsion reactors. She currently applies her expertise to the design of fusion machines.

Department

Nuclear Engineering & Engineering Physics

Contact

439, Engineering Research Building
1500 Engineering Dr
Madison, WI

  • PhD 2018, University of Wisconsin-Madison
  • BS 2016, State University of Campinas (UNICAMP)
  • MS 2014, University of São Paulo (USP)
  • BS 2013, Federal University of Rio de Janeiro (UFRJ)

  • Safety analysis
  • Thermal-hydraulics
  • Post-critical heat flux
  • Two-phase flow

Affiliated Departments

  • 2023 DOE Nuclear Energy University Program, Distinguished Early Carrer Program Award
  • 2022 American Nuclear Society, ATH’22 Best Paper Award
  • 2019 US Nuclear Regulatory Commission Nuclear Education Program, Faculty Development award
  • 2014 Regional Council of Engineering and Agronomy of Rio de Janeiro State (CREA/RJ), IV OSCAR NIEMEYER Award for Scientific and Technological Projects
  • 2014 CAPES, Science without Borders Fellowship for Ph.D. studies
  • 2013 Navy Technological Center in São Paulo (CTMSP), Fellowship
  • 2012 Foundation for Research Support of the State of Rio de Janeiro (FAPERJ), Scholarship
  • 2010 National Institute of Science and Technology of Innovative Nuclear Reactors – INCT/CNPq, Scholarship
  • 2008 National Council for Scientific and Technological Development (PIBIC/CNPq) Scholarship, Scholarship

  • Erickson, C., Allaf, M. A., Corradini, M., & Duarte, J. P. (2027). Analysis of severe accident progression in a BWR considering power uprates and ATF cladding materials. Annals of Nuclear Energy, 240, 112717.
  • Armstrong, R., Vermaak, J., Folsom, C., Seo, S., Pacheco Duarte, J., & Jensen, C. (2026). Advanced Multiphysics Code Coupling for Cladding Surface Thermocouples During Two-Phase Heat Transfer from Nuclear Fuel. Nuclear Technology, 1--15.
  • Dunbar, C., Hwang, D., & Duarte, J. P. (2026). Critical heat flux prediction for internally heated annuli with uniform power. Nuclear Engineering and Design, 448, 114685.
  • Na, U., Serrao, B. P., Seo, J., Fitrony, M. A., Suh, Y., Kwon, S., Duarte, J. P., Won, Y., & Jo, H. (2026). Extracting Bubble Information from High-Pressure Nucleate Boiling Using a Deep Learning Approach. Nuclear Engineering and Technology, 104400.
  • Allaf, M. A., Ejnik, G., Zaccomer, E., Jung, W. H., Corradini, M., & Duarte, J. P. (2026). Investigating the Role of Chromium-Coated Clad During Accident Progression for an SMR Using MELCOR. Nuclear Technology, 212(2), 328--346.
  • Serrao, B. P., Na, U., Jo, H., & Pacheco Duarte, J. (2026). Model-Free Solution Estimation of Critical Heat Flux Using Genetic Programming. Nuclear Technology, 1--19.
  • Armstrong, R., Pacheco Duarte, J., Corradini, M., & Jensen, C. (2026). Review of ECCS Acceptance Criteria and Experimental Basis Evolution Toward Fuel Fragmentation, Relocation, and Dispersal Studies. Nuclear Science and Engineering, 1--18.
  • Garcia, A., Hwang, D., Diem, S., Bier, V., Sontag, A., & Pacheco Duarte, J. (2026). Review of Tokamak Fusion Machine Safety Systems. Fusion Science and Technology, 1--20.
  • Allaf, M. A., Zaccomer, E., Ejnik, G., Jung, W. H., Corradini, M., & Duarte, J. P. (2026). Transient Behavior of the NuScale 250-MW (thermal) Power Module with Cr-Coated Zr Cladding During a Severe Accident. Nuclear Technology, 212(6), 1595--1608.
  • Moon, J. H., Allaf, M. A., & Duarte, J. P. (2026). Transition boiling heat transfer model under dryout and rewet cycles. Annals of Nuclear Energy, 236, 112403.

  • INTEGSCI 299 - Independent Study (Summer 2026)
  • N E 699 - Advanced Independent Study (Summer 2026)
  • N E 790 - Master's Research and Thesis (Summer 2026)
  • N E 890 - Pre-Dissertator's Research (Summer 2026)
  • N E 990 - Research and Thesis (Summer 2026)
  • N E 602 - Special Topics in Reactor Engineering (Spring 2026)
  • N E 790 - Master's Research and Thesis (Spring 2026)
  • N E 890 - Pre-Dissertator's Research (Spring 2026)
  • N E 990 - Research and Thesis (Spring 2026)
  • N E 999 - Advanced Independent Study (Spring 2026)
  • N E 411 - Nuclear Reactor Engineering (Fall 2025)
  • N E 790 - Master's Research and Thesis (Fall 2025)
  • N E 890 - Pre-Dissertator's Research (Fall 2025)
  • N E 990 - Research and Thesis (Fall 2025)
  • N E 699 - Advanced Independent Study (Summer 2025)
  • N E 790 - Master's Research and Thesis (Summer 2025)
  • N E 890 - Pre-Dissertator's Research (Summer 2025)
  • N E 990 - Research and Thesis (Summer 2025)
  • ISYE 574 - Methods for Probabilistic Risk Analysis of Nuclear Power Plants (Spring 2025)
  • N E 574 - Methods for Probabilistic Risk Analysis of Nuclear Power Plants (Spring 2025)
  • N E 790 - Master's Research and Thesis (Spring 2025)
  • N E 890 - Pre-Dissertator's Research (Spring 2025)
  • N E 990 - Research and Thesis (Spring 2025)
  • N E 790 - Master's Research and Thesis (Fall 2024)
  • N E 890 - Pre-Dissertator's Research (Fall 2024)
  • N E 990 - Research and Thesis (Fall 2024)
  • N E 890 - Pre-Dissertator's Research (Summer 2024)
  • N E 990 - Research and Thesis (Summer 2024)