Abstract
In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.
Highlights
HTR-10 is a 10 MW pebble-bed high temperature gas-cooled experimental reactor with the characteristics of Generation IV nuclear reactors, which takes spherical elements with allceramic TRISO type coated particles as fuel, graphite as neutron moderator material and helium as cooling medium
E flow distribution in the reactor core is quite complex because the coolant is divided into multiple parts on the flow paths from the reactor pressure vessel (RPV) inlet to the outlet. e layout of the primary system of HTR-10 and flow direction of coolant in RPV are shown in Figure 1 [2]. e cold helium enters the RPV, flows down through the annular space between the RPV and the core vessel, and gathers in the helium cavity at the RPV bottom
A new scheme to simulate the coolant flow process is proposed based on the flow distribution scheme of the original core thermal hydraulic analysis model and combined with the actual position of the core bypass flow. e THERMIX code is adopted to solve the thermal hydraulic models of HTR-10 with bypass flow channels simulated at different positions, and the temperature distribution differences in the pebble bed and the side components in the reactor core are compared. e relationship between bypass flow ratio and temperature distribution is analyzed in different conditions of the coolant flow process
Summary
HTR-10 is a 10 MW pebble-bed high temperature gas-cooled experimental reactor with the characteristics of Generation IV nuclear reactors, which takes spherical elements with allceramic TRISO type coated particles as fuel, graphite as neutron moderator material and helium as cooling medium. A great number of graphite blocks are installed in the core vessel to shape the pebble bed for the fuel elements and the graphite balls, to moderate and reflect neutrons, and to form the flow path for the coolant. In order to cool the control rods effectively, another small part of the cold helium flows through the control rod channels into a small plenum in the bottom graphite reflectors and enters the hot helium plenum.
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