Best-estimate codes, which couple neutronic and thermal-hydraulic solvers, are mainly used for safety analyses of nuclear power plants. During the past decade, the application of these codes to research reactors gained considerable interest and many improvements were presented to them. The increasing interest in the application of best-estimate codes to safety analyses of research reactor is largely driven by advancements in this field concerning power reactors and the diffusion of knowledge and capabilities to smaller, more diverse systems. The current study is a continuous effort in this framework and presents the coupled neutronic and thermal-hydraulic code development for the analysis of protected and unprotected transient behavior of research reactors. The coupling between neutronic and thermal-hydraulic processes is realized by considering the mutual feedbacks between them; the fuel and coolant properties (temperatures and density) variation affect the core's reactivity and hence the neutronic fission chain reaction, which in turn affects the fuel and coolant properties via a heat generation model for the reactor's power. More specifically, this study deals with the extension of the thermal-hydraulic model to the two-phase flow regime of the THERMO-T code. The extended THERMO-T model is validated against experimental results from the SPERT-IV, which was driven mainly by the coolant density reactivity feedback. This allows a more accurate evaluation of the adequacy of available and relevant two-phase flow models and correlations, which are selected from the domain of large power reactors. This is done in order to encourage and ensure standardization of modeling procedure of all types of reactors as part of the international community's continuous efforts towards this goal.