Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT). (28th August 2012)

• Record Type: Journal Article
• Authors: Imke, Uwe; Sanchez, Victor Hugo
• Other Names: Novog David Academic Editor.
• Is Part Of: Science and technology of nuclear installations. Volume 2012(2012)
• Publication Date: 2012-08-28
• Subjects: Nuclear engineering -- Periodicals; Nuclear facilities -- Periodicals; Nuclear engineering; Nuclear facilities; Electronic journals; Periodicals; 621.48
• DOI: 10.1155/2012/465059
• Languages: English
• ISSNs: 1687-6075

Abstract

Abstract : 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.