Thermal-hydraulic validation of two-phase models in THUNDER code against benchmark results and CFD codes

Autor: J. R. Maiorino, Duvan A. Castellanos-Gonzalez, João Manoel Losada Moreira, Deiglys Borges Monteiro
Rok vydání: 2020
Předmět:
Zdroj: Nuclear Engineering and Design. 369:110827
ISSN: 0029-5493
DOI: 10.1016/j.nucengdes.2020.110827
Popis: The study of two-phase flow behavior is a very important requirement in nuclear reactor systems, due to it can appear during normal operating conditions in a BWR or fast and slow transients for both PWR and BWR reactors. This work presents the evaluation at steady-state of two-phase flow prediction performed by THUNDER code (Thermal-Hydraulic Unit of Numerical Development for Elements with Rod-bundles), a new computational code that has been developed to perform a thermal-hydraulic analysis of LWR reactors with rod-type fuel assemblies. The physic and mathematical models used to develop the steady-state analysis of a nuclear reactor core are presented. The conservation equations of mass, momentum, and energy were established based on operating and geometric conditions, for single and two-phase flow using the subchannel and control volume methods, it applied to square arrays to the four sub-channel types (center-typical, center-thimble, edge, and corner) found in a PWR fuel assembly. THUNDER code performs a tridimensional analysis for the whole core solving simultaneously the conservation equations for each control volume. For two-phase flow treatment, the code uses the drift flux model. It provides detailed information of operational conditions e.g. coolant, cladding and center fuel temperatures, core pressure drop, coolant velocity, void fraction, coolant quality, onset of nuclear boiling (ONB) and departure from nucleate boiling ratio (DNBR). To validate the code, we consider experimental data from OECD/NRC benchmark based NUPEC PWR subchannel and bundle tests (PSBT) and CFD calculations from ANSYS-CFX®. It was evaluated the void fraction, average density and pressure drop for individual subchannel as well as for a rod bundle 5 × 5 fuel assembly. THUNDER code reproduced well the benchmark results presenting a mean absolute discrepancy of void fraction equal to 0.039 and density relative discrepancy equal to 0.11. This indicates that THUNDER code performs a two-phase flow evaluation within the acceptance criterion for reactors with rod-type fuel assemblies under steady-state conditions.
Databáze: OpenAIRE
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