Irıs Küçük Modüler Reaktörünün (Smr) Relap5/Scdapsım Sistem Kodu İle Modellenmesi

Abstract

Due to many reasons such as its compact design, low initial investment costs, shorter installation times compared to conventional nuclear power plants, compatibility with nuclear disarmament treaty, compatibility of developing countries electrical transmission lines, modular installation, and improved passive safety systems, intensive studies have been made on small modular reactors since 2000. In this thesis, the IRIS reactor and a small break loss of coolant accident that may occur at this reactor were analyzed. IRIS has been designed under the sponsorship of the US Department of Energy and Nuclear Energy Research Initiative (NERI) with the lead of the Westinghouse Company, and it was designed by an international consortium of nuclear industry, research laboratories and universities. IRIS has 1000 MWth and 335 MWe power, it is a pressurized light water reactor. IRIS reactor has integral design consists of main reactor power unit such as main coolant pumps, steam generators and pressurizer in the pressure vessel and it is modeled with RELAP5/SCDAPSIM system code. By virtue of the integral structure of the IRIS SMR, there is no pipe connection larger than 4 inches (10.16cm) (Chemical and Volume Control System connection) to primary loop inside the reactor pressure vessel. The aim of this thesis is to model the IRIS reactor core, coolant system and safety system using RELAP5/SCDAPSIM system code and to analyze the behavior of the reactor during small break loos of coolant accident that can occur at the reactor Chemical and Volume Control System connection. In this model, reactor primary and secondary loop components, pressure vessel and passive safety systems except containment cooling system are modeled. According to the results of the analysis of small break loss of coolant accident, the mass flow started from pressure vessel to containment and reactor pressure started to decrease when the break occurred at 2000th second. The pressurizer water level and reactor pressure dropped their critical level at 2017th second and 2021th second, respectively. With the decrease of the reactor pressure below the critical level, the reactor was shutdown, Emergency Heat Removal System (EHRS) was activated. With the activation of EHRS, the main feed water isolation valve and main steam isolation valve were closed. Automatic Depressurization System was activated in order to assist EHRS to reducing the reactor pressure at 2036th second. Long Term Gravity Driven Makeup System was activated when the pressure difference between reactor pressure vessel and containment reached below the 13,79 kPa. In this study, containment cooling system is not modeled and EHRS design data was not found that's why the pressure of the containment and the pressure vessel was become equal more than expected level; however this level is still 0.3 MPa lower than to containment design pressure (1.4 MPa).