DESIGN AND ANALYSIS OF 208PB-BI EUTECTIC-COOLED REACTOR USING MCANDLE BURNUP SCHEME WITH RADIAL FUEL SHUFFLING

• Main Author: Widiawati, Nina
• Format: Dissertations
• Language: Indonesia
• Online Access: https://digilib.itb.ac.id/gdl/view/61089
• Institution: Institut Teknologi Bandung

Currently, the dominant type of nuclear reactor in the world is the Light Water Reactor (LWR). However, the utilization of natural uranium at LWR is very low. The reactor employed enriched uranium-235 as fuel. However, only about 5% of enriched uranium fuel undergoes fission reactions. Fast Breeder Reactor (FBR) with an open fuel cycle also utilizes quite low natural uranium, as it uses higher enriched uranium as fuel. Meanwhile, FBR with a closed fuel cycle requires reprocessing technology of spent fuel. The uranium enrichment process will leave a lot of depleted uranium. The technology of fuel enrichment and reprocessing of spent fuel are sensitive issues and expensive. Reactor with a modified CANDLE (Constant Axial shape of Neutron flux, nuclides densities, and power shape During Life of Energy production) burnup scheme (MCANDLE) can increase the utilization of natural uranium as fuel. It is because after operating for one cycle, the reactor only requires natural uranium as fuel input without any enrichment and reprocessing of fuel. Fissile fuel is needed to start the reactor operation, for example, plutonium. However, the scenario of operating a fast reactor in 2050 encourage minimizing the loading of plutonium in the reactor. Therefore, in this study, an analysis was carried out on the use of 208Pb-Bi as a coolant because the neutron captured cross-section of the 208Pb is small so that it is expected to minimize fuel loading in the reactor. Optimization of the direction of refueling is carried out to obtain a design that has an average reactivity of around 1% ?k/k and a Power Peaking Factor (PPF) of less than two during operation. Then, an analysis of the initial core design was carried out for the optimization design with plutonium as ignition fuel. Neutronic calculations were performed using SRAC (Standard Reactor Analysis Code System) and JENDL (Japanese Evaluated Nuclear Data Library) 4.0 as nuclear data libraries. The SRAC calculation using the deterministic method has also been compared with the results of the MCNP6 (Monte Carlo N particle-6) calculation using the probabilistic method. The difference in the calculation results is still below 4%. Other calculation results show that the 208Pb-Bi coolant can minimize fuel loading in the reactor with the MCANDLE burnup scheme. By optimizing the refueling direction, a scheme that has an excess reactivity lower than 1% ?k/k and a PPF is lower than two during operation time is also obtained. The scheme also has spent fuel burnup level of ~48% FIMA (Fissions per Initial Metal Atom). It means that this design is able to increase the utilization of natural uranium because about 48% of natural uranium undergoes fission without any enrichment process or fuel reprocessing. Then, the initial reactor core design for the selected scheme has an average reactivity value of about 1% ?k/k during operation using the Advanced Pressurized Water Reactor (APWR) spent fuel.