[en] Evaluation of thermal-hydraulic conditions, fuel behavior, and reactor kinetic during various operating and postulated accident conditions results in conclusions that support decision-making process, the review of license application, and the resolution of other technical issues related to nuclear safety. Also these activities increase the understanding and involvement into new technical developments. Thermal-hydraulic research activities at TAEA focus on the application of computer codes that simulate the behavior of the reactor system. The computer codes are used to analyzed loss of coolant accidents, and system transients in light water nuclear reactors and to assess the consequences if imbalance occurs and to determine the effectiveness of mitigating actions. TAEA has used nuclear reactor system codes (RELAP5/Mod3.2 and higher versions, PARCS) and nuclear plant visual analyzer codes (NPA, SNAP, and XNGR5) obtained by in the framework of the CAMP Agreement signed between TAEA and the United States Nuclear Regulatory Commission (US NRC). TAEA performs and documents the code assessments including improvements and error corrections. Moreover, research activities concerning the passive cooling application and simulations of advanced nuclear power plant have been carried out by both experimental and theoretical means. For example, the experiment test facility, which was designed to investigate the effect of noncondensable gases on condensation, was conducted in cooperation with the Mechanical Engineering Department of the Middle East Technical University in Ankara (Turkey) and was finished. The text matrix obtained from this research was also submitted to US NRC data bank. Application of RELAP5 code f system transients include International Standard Problem (ISP) studies (ISP 33, 38, 42, 45, 46), accident analysis for different reactors and special topics in nuclear heat transfer problems (mid-loop operation). Research studies of severe accidents assess the detailed behavior of reactor and containment systems, including the means by which these accidents may be prevented or mitigated. Research studies address fuel damage, progression of accident scenarios, ability to maintain damaged fuel within the reactor pressure vessel, and fission product transport. Two ISP's were carried out in the context of core degradation