DESIGN ANALYSIS ON OPERATING PARAMETER OF OUTLET TEMPERATURE AND VOID FRACTION IN RDE STEAM GENERATOR.HTGR is one of the next generation reactor types. HTGR is currently considered as one of the leading reactors for the future nuclear power plant. The steam generator is one of the main components in HTGR as well as in RDE. In the steam generator, the heat is transferred by high temperature helium gas in the shell side to water in the tube side to generate the superheated steam. The purpose of this work is to design the operating parameter of outlet temperature and void fraction of steam based on feed water mass flow rate and inlet temperature variations in RDE steam generator. In this work, the ChemCAD program was used. Both inlet and outlet temperature of helium gas have been set up as boundary conditions. The result shows that using the mass flow rate of 4.3 kg/s -4.8 kg/s and water inlet temperature of 110 o C -160 o C, the superheated steam outlet temperature (void fraction = 1.0) is obtained in the range of 275.5 o C -600 o C. This analysis is beneficial to assess 10 MW RDE design especially in the steam generator system operating parameters.

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Steady-state and transient analysis of reactor core under Reactivity-Initiated Accident (RIA) conditions are important for reactor operation safety. The reactor dynamics are influenced by neutronic and thermal-hydraulic aspects of the core. In this study, steady-state and transient analysis under RIA conditions of the RSG-GAS multipurpose reactor was carried out using MTR-DYN and EUREKA-2/RR programs. Neutronic calculations were performed using a few group cross-sections generated by Serpent 2 with the latest cross-section data ENDF/B-VIII.0. Steady-state conditions were carried out with a nominal power of 30 MW, while transient under RIA conditions occurred because the control rod was pulled too quickly while the reactor operated. These transient RIA conditions were performed for two cases, during start-up with an initial power of 1 W, and within power range with an initial power of 1 MW. Thermal-hydraulic parameters considered in this study are reactor power, the temperature of the fuel, cladding, and coolant. The calculated maximum fuel temperature at a steady state is 126.02°C. Meanwhile, the calculated maximum fuel temperature during RIA conditions at the initial power of 1 W and 1 MW are 64.38°C and 137.14°C, respectively. There are no significant differences in thermal-hydraulic parameters between each used program. The thermal-hydraulic parameters such as the maximum temperature of the coolant, cladding, and fuel under this postulated RIA condition are within the acceptable reactor operation safety limits.

A conversion of the TRIGA reactor Bandung has been planned by replacing fuel from rod type to plate type. The fuel replacement causes the fuel arrangement or reactor core configuration should be changed as well as the cooling system. Currently, the reactor cooling system takes place by natural convection with the flow direction from the bottom to the top of the core. The replacement of plate-type fuel with a small distance between plates causes the core cooling process to decrease, therefore to make more effective the forced convection cooling system with the direction of upward flow is required. It is expected that the cooling process will increase. The ability of the cooling system depends on flow distribution in the core at the time of reactor in operating, therefore it is necessary to analyze the distribution of coolant flow and the velocity distribution of the cooling system from the top reactor pool to the reactor core to determine that the coolant flow from the inlet pipe mostly goes to the core. The purpose of this research is to analyze the distribution of coolant flow and flow velocity to the reactor core. The analysis was carried out in a simulation using computational fluid dynamics FLUENT software by creating reactor core modeling. The result shows that the cooling water flows from above along the reactor pool, then the stored energy is reduced so that the flowing slowdown. In this case, resulting in evenly flow distribution to the reactor core.