Validation of RELAP5/MOD3.3 for subcooled boiling, flashing and condensation in a vertical annulus

Continued development of system analysis codes has resulted in the recovery of conservatisms originally imposed on nuclear power reactors, allowing for an increase in the capacity of commercial nuclear reactors. These codes also play an instrumental role in the design and certification of new reactor systems. With the increased demand for passive natural circulation and gravity driven cooling options, these codes are met with the new challenge of simulating low pressure, low flow conditions. The objective of this work is to demonstrate the effectiveness of the widely used RELAP5/MOD3.3 code to simulate boiling, condensing and flashing flows under such conditions. Two-phase flow data in an internally heated vertical annulus with inner diameter of 19.1 mm and outer diameter of 38.1 mm is utilized for validation of the RELAP5/MOD3.3. The code calculation of pressure, temperature, void fraction, interfacial area concentration, and void weighted gas velocity along the 4.5 m test section is compared with data at five axial locations. In the 2.8 m heated section of the channel the code predictions compare favorably in general, although the error does increase at low system pressure. Beyond the heated length, code predictions of condensation and flashing show more noticeable disagreement along the 1.7 m unheated section. Condensation is consistently under-predicted. Flashing varies from relatively good agreement to complete failure, depending on the conditions at the exit of the heated section. User options related to boiling and condensation are also assessed, and shown to have marginal improvements in some conditions. In general the code consistently predicts the point of net vapor generation too soon along the heated length at low mass flux, over predicts the void fraction at the end of the heated length, and has large scatter in void fraction agreement at the end of the channel.