NEUTRONIC STUDY OF HTGR (High Temperatue Gas-Cooled) NUCLEAR REACTOR USING THORIUM NUCLEAR FUEL (Th-U233)O2 and (Th-U233)N

• Main Author: AZZAHWA (NIM : 10212019), SARAH
• Format: Final Project
• Language: Indonesia
• Online Access: https://digilib.itb.ac.id/gdl/view/24213
• Institution: Institut Teknologi Bandung

The development of design in nuclear reactor is aimed to make nuclear reactor capable to produce energy with the best safety feature. HTGR nuclear reactor (High Temperature Gas-Cooled Reactor) is one of nuclear reactor in Generation IV. Beside the development of reactor’s design, research in developing nuclear fuel is also increasing in the world. One of nuclear fuel material that being developed is Thorium. <br /> <br /> Thorium is a fertile element that more abundant in earth if we compare it to Uranium. Thorium nuclide could be used as nuclear fuel by absorpe neutron so it will transmute to Uranium-233, whice is fissile. In this research, we use Thorium mixed with Uranium-233 as TRISO fuel in HTGR reactor. This research aimed to determine composition between Uranium-233 and Thorium so HTGR reactor could reach criticallity wheter it using (Th-U233)O2 or (Th-233U)N. <br /> <br /> In this research , we calcuate neutronic calculation for period about 10 years where each interval is 180 days (6 months) by using SRAC. Based on analysis in each neutronic parametric, especially multiplication factor and reactivity, we could see that HTGR reactor reach it criticality fully for 10 years period if we using fissile percentation about 7.6-8.8% for producing 50MWt and using fissile percentation about 8.0-9.0% for producing 100MWt.