Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT) | Zendy

Uwe Imke | Zendy; Víctor Hugo Sánchez-Espinoza | Zendy

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Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT)

Author(s) -

Uwe Imke,

Víctor Hugo Sánchez-Espinoza

Publication year - 2012

Publication title -

science and technology of nuclear installations

Language(s) - English

Resource type - Journals

SCImago Journal Rank - 0.417

H-Index - 24

eISSN - 1687-6083

pISSN - 1687-6075

DOI - 10.1155/2012/465059

Subject(s) - bundle , nuclear engineering , thermal hydraulics , coolant , pressurized water reactor , boiling water reactor , transient (computer programming) , work (physics) , engineering , mechanical engineering , materials science , computer science , heat transfer , mechanics , physics , composite material , operating system

SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT). The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy

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