NEUTRONIC CALCULATION FOR PB-BI COOLED CANDLE REACTOR USING MCNP CODE

• 主要作者: AFIFAH (NIM : 20216049), MARYAM
• 格式: Theses
• 語言: Indonesia
• 主題: Fisika
• 在線閱讀: https://digilib.itb.ac.id/gdl/view/28788

CANDLE reactor is one of an innovative fast reactor design that employed with specific burn-up strategy. An initial core should be prepared which is contracted a special reactor for the first several years. After initial core have burned, the new fuel for the remaining cores is generated which can be sustained to use only by natural uranium. The fuel breeding using U10Zr have been investigating to obtain better CANDLE burn-up performance and it proposed the ease of designing a long life reactor. In this study, we also used lead bismuth cooled as the survey parameter to develop the coolant effect for neutronic performance with thermal power output is 2000 MW. Reactor calculation have been performed with Monte Carlo method, MCNP code, to determined detail analysis calculation. This reactor design demonstrated the sustainability of uranium fuel to produce plutonium as breeding fuel concept. The result of the calculation showed that spent fuel burnup is about 400 MWd/tHM during 80 years operation, it means, about 40% of natural uranium burns up without enrichment. This value is competitive which presently planned by fast reactor systems.