The High-Temperature Gas-cooled Reactor (HTGR) has been selected as one of the next-generation nuclear power plants because of its passive safety features. The Korea Atomic Energy Research Institute (KAERI) has been studying how to utilize HTGR efficiently and safely. The HTGR uses graphite as a moderator and helium as a coolant. Once tritium is produced, it is released into the coolant; once released from the core, tritium travels within the primary loop. Because the coolant is gas phase, it is easy to transport to other systems. While it circulates in the primary loop, tritium is involved in processes that include leakage, purification, and permeation. KAERI has been developing a tritium behavior analysis code named TRitium Overall Phenomena analYsis (TROPY) to analyze tritium transport and predict the amount of tritium in the loop in the HTGR core. In this paper, the functions of the TROPY code are introduced, and the amount of tritium in each loop and the amount released into the product hydrogen from the MHTGR 350-MW(thermal) core are explained.