Important thermal-hydraulic parameters of the 3-MW TRIGA MARK-II research reactor operating under both steady-state and transient conditions are reported. Neutronic analyses were performed by using the CITATION diffusion code and the MCNP4B2 Monte Carlo code. The output of CITATION and MCNP4B2 were input to the PARET thermal-hydraulic code to study the steady-state and transient thermal-hydraulic behavior of the reactor. To benchmark the PARET model, data were obtained from different measurements performed by thermocouples in the instrumented fuel (IF) rod during the steady-state operation both under forced- and natural-convection mode and compared with the calculation. The mass flow rates needed for input to PARET were taken from the Final Safety Analysis Report for a downward forced coolant flow equivalent to 3500 gal/min. For natural convection cooling of the reactor, the mass flow rate was generated using the NCTRIGA code. Peak fuel temperatures measured by the thermocouples in the IF rods at different power levels of the TRIGA core were compared with the values calculated by PARET. The axial distribution of the temperatures of the fuel centerline, fuel surface, and the cladding surface in the hot channel were calculated for the reactor operating at the full-power level. Fuel surface heat flux and heat transfer coefficients for the hot channel were also calculated for the reactor operating at the full-power level. The investigated results were found to be in good agreement with the experimental and operational values. The testing of the PARET model calculations through benchmarking the available TRIGA experimental and operational data for pulse-mode operations showed that PARET can successfully be used to analyze the transient behavior of the reactor. Major transient parameters, such as peak power and prompt energy released after pulse, full-width at half-maximum of pulse peak, and maximum fuel centerline temperatures for different fuel elements at different pulses, were computed and compared with the experimental and operational values. It was observed that pulsing of the reactor inserting an excess reactivity of 1.996 $ shoots the reactor power level to 873 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 512°C. The investigation on maximum available reactivity insertion at full power (2.24 $) by the transient rod raises the reactor power to 1629 MW, and the fuel centerline temperature from the calculations is found to be 937°C.