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dc.contributor.authorZhao, Xingang
dc.contributor.authorWysocki, Aaron J
dc.contributor.authorShirvan, Koroush
dc.contributor.authorSalko, Robert K
dc.date.accessioned2021-10-27T20:30:56Z
dc.date.available2021-10-27T20:30:56Z
dc.date.issued2019
dc.identifier.urihttps://hdl.handle.net/1721.1/136127
dc.description.abstract© 2019, © 2019 American Nuclear Society. As part of the Consortium for Advanced Simulation of Light Water Reactors, the subchannel code CTF is being used for single-phase and two-phase flow analysis under light water reactor operating conditions. Accurate determination of flow distribution, pressure drop, and void content is crucial for predicting margins to thermal crisis and ensuring more efficient plant performance. In preparation for the intended applications, CTF has been validated against data from experimental facilities comprising the General Electric (GE) 3 × 3 bundle, the boiling water reactor full-size fine-mesh bundle tests (BFBTs), the Risø tube, and the pressurized water reactor subchannel and bundle tests (PSBTs). Meanwhile, the licensed, well-recognized subchannel code VIPRE-01 was used to generate a baseline set of simulations for the targeted tests and solution parameters were compared to the CTF results. The flow split verification problem and single-phase GE 3 × 3 results are essentially in perfect agreement between the two codes. For the two-phase GE 3 × 3 cases, flow and quality discrepancies arise in the annular-mist flow regime, yet significant improvement is observed in CTF when void drift and two-phase turbulent mixing enhancement are considered. The BFBT pressure drop benchmark shows close agreement between predicted and measured results in general, although considerable overprediction by CTF is observed at relatively high void locations of the facility. This overestimation tendency is confirmed by the Risø cases. While overall statistics are satisfactory, both BFBT and PSBT bubbly-to-churn flow void contents are markedly overpredicted by CTF. The issues with two-phase closures such as turbulent mixing, interfacial and wall friction, and subcooled boiling heat transfer need to be addressed. Preliminary sensitivity studies are presented herein, but more advanced models and code stability analysis require further investigation.en_US
dc.language.isoen
dc.publisherInforma UK Limiteden_US
dc.relation.isversionof10.1080/00295450.2018.1507221en_US
dc.rightsCreative Commons Attribution-Noncommercial-Share Alikeen_US
dc.rights.urihttp://creativecommons.org/licenses/by-nc-sa/4.0/en_US
dc.sourceDOE repositoryen_US
dc.titleAssessment of the Subchannel Code CTF for Single- and Two-Phase Flowsen_US
dc.typeArticleen_US
dc.contributor.departmentMassachusetts Institute of Technology. Department of Nuclear Science and Engineering
dc.relation.journalNuclear Technologyen_US
dc.eprint.versionAuthor's final manuscripten_US
dc.type.urihttp://purl.org/eprint/type/JournalArticleen_US
eprint.statushttp://purl.org/eprint/status/PeerRevieweden_US
dc.date.updated2021-08-10T17:39:46Z
dspace.orderedauthorsZhao, X; Wysocki, AJ; Shirvan, K; Salko, RKen_US
dspace.date.submission2021-08-10T17:39:48Z
mit.journal.volume205en_US
mit.journal.issue1-2en_US
mit.licenseOPEN_ACCESS_POLICY
mit.metadata.statusAuthority Work and Publication Information Needed


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