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Title: Assessment of the Subchannel Code CTF for Single- and Two-Phase Flows

Abstract

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,more » 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. As a result, preliminary sensitivity studies are presented herein, but more advanced models and code stability analysis require further investigation.« less

Authors:
 [1]; ORCiD logo [2];  [1]; ORCiD logo [2]
  1. Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
  2. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1545243
Grant/Contract Number:  
AC05-00OR22725
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Nuclear Technology
Additional Journal Information:
Journal Volume: 205; Journal Issue: 1-2; Journal ID: ISSN 0029-5450
Publisher:
Taylor & Francis - formerly American Nuclear Society (ANS)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; CTF; subchannel analysis; flow mixing; pressure drop; void fraction

Citation Formats

Zhao, Xingang, Wysocki, Aaron J., Shirvan, Koroush, and Salko, Jr., Robert K. Assessment of the Subchannel Code CTF for Single- and Two-Phase Flows. United States: N. p., 2018. Web. doi:10.1080/00295450.2018.1507221.
Zhao, Xingang, Wysocki, Aaron J., Shirvan, Koroush, & Salko, Jr., Robert K. Assessment of the Subchannel Code CTF for Single- and Two-Phase Flows. United States. doi:10.1080/00295450.2018.1507221.
Zhao, Xingang, Wysocki, Aaron J., Shirvan, Koroush, and Salko, Jr., Robert K. Fri . "Assessment of the Subchannel Code CTF for Single- and Two-Phase Flows". United States. doi:10.1080/00295450.2018.1507221. https://www.osti.gov/servlets/purl/1545243.
@article{osti_1545243,
title = {Assessment of the Subchannel Code CTF for Single- and Two-Phase Flows},
author = {Zhao, Xingang and Wysocki, Aaron J. and Shirvan, Koroush and Salko, Jr., Robert K.},
abstractNote = {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. As a result, preliminary sensitivity studies are presented herein, but more advanced models and code stability analysis require further investigation.},
doi = {10.1080/00295450.2018.1507221},
journal = {Nuclear Technology},
issn = {0029-5450},
number = 1-2,
volume = 205,
place = {United States},
year = {2018},
month = {9}
}

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Works referenced in this record:

A general technique for the prediction of void distributions in non-steady two-phase forced convection
journal, September 1971


Single- and two-phase turbulent mixing rate between adjacent subchannels in a vertical 2×3 rod array channel
journal, May 2004