About

The work focuses on testing and modeling of thermal-hydraulic phenomena to gain a better physical understanding of TH margins. The testing is performed using a high temperature (<1500 C) quench facility along with multitude of surface characterization capability including wettability, roughness, emmisivity and porosity. The heat transfer surface is exposed to neutron/gamma/steam at different water chemistries to characterize the most representative heated surface relevant to nuclear reactors. The model developement and simulations are perforemd using system-level and subchannel codes such as TRACE/RELAP, VIPRE and CTF. STARCCM+ CFD tool is also used for particular aspects of TH analysis.

Non-MIT collaborators include Sandia National Laboratory, University of Wisconsin.

Impact of Nuclear Environment on Two-Phase Heat Transfer

Student: Arunkumar Seshadri (PhD)

Utilizing unique facilities at MIT, the heat transfer surface wettability in characterized after exposre to prototypic nuclear environment (Neuron/Gamma/Oxidation/CRUD/ etc.)

Journal Publication: Seshadri A., Philips B., Shirvan K., "Towards Understanding the Effects of Irradiation on Quenching Heat Transfer," Journal of Heat and Mass Transfer, Accepted, 2018.

Quench Heat Transfer of ATF Materials

Postdoc: Wei Li Student: Arunkumar Seshadri

Quenching heat transfer of accident tolerant fuel materials is investigated throuhg testing and modeling and simulation. The ATF candidates include Cr, FeCrAl, Mo, SiC.

Journal Publication: Seshadri A., Shirvan K., “Quenching Heat Transfer Analysis of Accident Tolerant Coated Fuel Cladding,” Nuclear Engineering and Design Accepted, 2018.

Mechanistic Critical Heat Flux Modeling

Student: Xingang Zhao (PhD)

Mechanistic Critical Heat Flux(CHF) model based on a data driven (machine learning) and phenomenological driven (mirco-scale) appraoches applicable to both steady state and transient conditions is currently being developed for implementation in CTF.

Journal Publication: Zhao X., Salko R., Wysocki A., Shirvan K., "Validation and Benchmarking of CTF for Single- and Two-Phase Flow," Nuclear Technology, Accepted, 2018.