2021 Annual Meeting

Presentation information

Oral presentation

III. Fission Energy Engineering » 302-1 Advanced Reactor System

[1C01-06] Optimizatioin/Code Development on Advanced Reactor

Wed. Mar 17, 2021 10:15 AM - 12:00 PM Room C (Zoom room 3)

Chair: Motoyasu Kinoshita (MOSTECH)

11:15 AM - 11:30 AM

[1C05] Verification of Neutronics-Thermalhydraulics Coupling UDF for FLUENT Code using MSRE Experiment

Analysisi of Pump Startup and Trip Experiment

*Hiroyasu MOCHIZUKI1 (1. Tokyo Institute of Technology)

Keywords:Molten Salt Reactor, MSRE, Pump startup and coastdown, Neutronics-thermalhydraulics coupling, FLUENT, PID control

Using the fuel pump startup and trip experiments of the thermal neutron molten salt reactor experiment, MSRE, which was in operation for four years since 1965 at Oak Ridge National Laboratory, the computation function of the User Defined Function (UDF) incorporated in the FLUENT code has been verified. A PID controller model of the control rod has been added to the UDF which had a function to calculate neutronics and thermal-hydraulics coupling.