Courses:  297A:  Introduction to Nuclear engineering I

                 497A:  Introduction to Nuclear engineering II

                 430  :  Design of Reactor Systems

                 403  :  Advanced Reactor design

                 470  :  Power Plant Simulation

                 450  :  Radiation Detection and Measurements

                 515  :  Two-Phase Heat Transfer

                 312  :  Information technology, Technical Presentation and Speaking

                 505  :  Reactor Control Theory

                 501  :  Reactor Engineering 

                 512  :  Reactor Fuel Management

                 597B:  Advanced Reactor engineering

                 515  :  Audit - Two-Phase Flow and Heat Transfer

                 597D:  Reactor Safety Engineering

                 412  :  Turbulent Flow

                 590  :  Engineering Colloquium  

                 600  :  Thesis Research  

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Research and Teaching:

• Large break Loss Of Coolant Accident (LOCA) analysis (TRAC user group)

  • Served as a member of the TRAC maintenance and development group supporting the TRAC code maintenance work

  • Identification and ranking of phenomena leading to fuel peak cladding temperatures from a probabilistic approach in reactor systems, using integral and separate effect tests

  • Performed plant performance analysis under postulated accident scenarios in reactor systems

• Developed phenomena identification and ranking tables (PIRT) for peak cladding temperatures (BWR) during large break loss of coolant accidents (for the U.S.NRC)

• Conducted theoretical research in two phase flow leading to the explanation of liquid film dryout in boiling water reactors (BWR)

• Designed and conducted extensive laboratory tests modeling BWR core flow across spacer grids to obtain unique quantitative and visual data of liquid film entrainment and deposition at spacer grids affecting the departure from nucleate boiling (DNB)

• Derived and incorporated the fuel assembly spacer grid model in the COBRA-TF thermal hydraulic code (Framatome-ANP), which became the first in its class to be able to mechanistically model effects of spacer grids in accident and operating transient analysis

• Addressed and resolved code issues supporting the Framatome COBRA-TF group

• Performed micro scale physical modeling of thermal hydraulic and heat transfer phenomena in fouled and chemically cleaned steam generator tubes (FERMI grants)

• Proposed and implemented modifications to the steam generator tube test loop, solving numerous technical issues

• Performed computational fluid dynamic (CFD) modeling of flow past PWR spacer grids

• Used MACCS code (MELCOR Accident Consequence Code System for the calculation of the health and economic consequences of accidental atmospheric radiological releases) to assess environmental impact from radionuclide dispersion (E.g., Consejo de Seguridad Nuclear, CSN)

• Student evaluation in undergraduate courses: Introduction to nuclear engineering, (thermodynamic and heat transfer design principles in reactor systems)

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