THERMALHYDRAULIC STUDY ON THE CYLINDRICAL-FUELED NUCLEAR RESEARCH REACTOR

• Main Author: UMAR (NIM : 33102004); Tim Pembimbing : Prof. Dr. Ir. Aryadi Suwono; Dr. Ir. Nathanael P. Ta, EFRIZON
• Format: Dissertations
• Language: Indonesia
• Online Access: https://digilib.itb.ac.id/gdl/view/15493
• Institution: Institut Teknologi Bandung

Safety aspect plays an imperative role in the design process and operation of a nuclear reactor. In order to achieve the capability to predict various operating condition of the reactor, analytical methods deployed in the design and operation activities shall be well examined and reliable. In the case of the TRIGA 2000 reactor employing cylindrical nuclear fuel, the analyses of experiments show that the analytical methods applied in the reactor design yields less conservative results. Based on this reason, improved analytical and experimental methods shall be innovated and promoted. This study covers the thermal-hydraulic characteristics analysis for a nuclear research reactor with cylindrical fuel by conducting various improvements in the analytical methods used to predict reactor-operating conditions. Both innovative numerical and experimental methods are applied as the approach to solve the thermal-hydraulic problems occur in the core of nuclear research reactor with cylindrical fuel. Numerical approach for three-dimensional analysis is applied in the finite volume method. While for improving the analytical methods, empirical correlations derived from the experimental assessment are deployed. In addition, other aspects that caused the weakness in the present analytical methods are also considered. The results of this research shows that the coolant flow in the reactor core is not purely induced by the force due to the density difference of the coolant, but also affected by the flow coming in from the primary cooling system. On the other hand, the application of the enhanced analytical method is capable to predict the governing thermal-hydraulic parameters of this cylindrical-fueled core with improved accuracy. This is also the case for the three-dimensional analysis using CFD software that simulates almost the entire components of the reactor core, is also capable to predict the two-phase boiling phenomena taking place in the core. The results of this research will contribute to improving the safety performance in the design process as well as in the operation of the cylindrical-fueled nuclear research reactor. Meanwhile, the improved analytical method is expected to serve as back up resources in enhancing the theoretical thermal-hydraulic analysis. Finally, the results of this research can serve as a useful reference and benchmark in thermal-hydraulic analysis of the similar nuclear research reactor.